NG-18-0121, Response to Second Round - Request for Additional Information (RAI) - Duane Arnold Energy Center (DAEC) - LAR TSCR-166, Adoption of EAL Scheme Pursuant to NEI 99-01

Response to Second Round - Request for Additional Information (RAI) - Duane Arnold Energy Center (DAEC) - LAR TSCR-166, Adoption of EAL Scheme Pursuant to NEI 99-01

ML18295A202
Person / Time
Site:	Duane Arnold
Issue date:	10/18/2018
From:	Dean Curtland NextEra Energy Duane Arnold
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
EPID L-2017-LLA-420, NG-18-0121
Download: ML18295A202 (796)

Text

{{#Wiki_filter:NEXTeraM ENERGY~ DUANE ARNOLD October 18, 2018 NG-18-0121 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Duane Arnold Energy Center Docket No. 50-331 Renewed Op. License No. DPR-49

Subject:

Response to Second Round - Request for Additional Information (RAI) - Duane Arnold Energy Center (DAEC) - LAR TSCR-166, Adoption of EAL Scheme Pursuant to NEI 99-01 -EPID L-2017-LLA-420

References:

1. NextEra Energy Duane Arnold, LLC letter NG-17-0235, License Amendment Request (TSCR-166), Adoption of EAL Scheme Pursuant to NEI 99-01, Revision 6, dated December 15, 2017 Q'v1L17363A069)

2. NextEra Energy Duane Arnold, LLC letter NG-18-0090, Response to Request for Additional Information Regarding License Amendment Request (TSCR-166), Adoption of Emergency Action Level Scheme Pursuant to NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors," July 26, 2018 (J'v1L18212A232)

3. NRC E-Mail: Final Second Round - Request for Additional Information (RAI) - Duane Arnold Energy Center (DAEC) -LAR TSCR-166, Adoption of EAL Scheme Pursuant to NEI 99 EPID L-2017-LLA-0420, October 9, 2018 In Reference 1, N extEra Energy Duane Arnold, LLC (hereafter N extEra Energy Duane Arnold) submitted a License Amendment Request for the Duane Arnold Energy Center (DAEC) pursuant to 10 CFR 50.90. In Reference 2, NextEra Energy Duane Arnold submitted additional information regarding that application in response to a NRC staff request for additional information.

Subsequent to that submittal, in Reference 3, NRC staff have identified an additional request for additional information. Enclosure 1 of this letter provides the NextEra Energy Duane Arnold response to the NRC staffs second round request for additional information. NextEra Energy Duane Arnold, LLC, 3277 DAEC Road, Palo, IA 52324 u

The following information is provided as attachments to the Enclosure to aid NRC review and approval and replaces the Attachments in their entirety from References 1 and 2:

• Attachment 1 - Updated Redline Markup of NEI 99-01 Revision 6
• Attachment 2 - Updated Clean Copy of the Proposed DAEC EAL Scheme
• Attachment 3 - Updated Deviations and Differences Matrix
• Attachment 4 - Updated Supporting Technical Information
• Attachment 5 - Updated DAEC EAL Scheme Wallboards This additional information does not impact the 10 CFR 50.92 evaluation of "No Significant Hazards Consideration" previously provided in the referenced application.

This letter makes no new commitments and does not change any existing commitments. If you have any questions regarding this matter, please contact Michael Davis, Licensing Manager at (319) 851-7032. I declare under penalty of perjury that the foregoing is true and correct. Executed on October 18, 2018 Dean Curtland Site Director N extEra Energy Duane Arnold, LLC Enclosure cc: Regional Administrator, USNRC, Region III, Project Manager, USNRC, Duane Arnold Energy Center Resident Inspector, USNRC, Duane Arnold Energy Center A. Leek (State of Iowa) fMHR Iii¥¢ &  ; &¥ a i&HWM !¥ & P@ au, iWlRW *&MW@ W NextEra Energy Duane Amold, LLC, 3277 DAEC Road, Palo, IA 52324

ENCLOSURE! DUANE ARNOLD ENERGY CENTER Response to Second Round - Request for Additional Information (RAI) Duane Arnold Energy Center (DAEC) LAR TSCR-166, Adoption of EAL Scheme Pursuant to NEI 99-01 EPID L-2017-LLA-420

Enclosure to NG 18-0121 RAI-DAEC-1 The proposed value for the Potential Loss of Fuel Clad Barrier threshold is based on the guidance provided by NEI 99-01, Revision 6, and is not based on site-specific conditions. Section 3.3, "NSSS Design Differences," of NEI 99-01, Revision 6, provides that, "Developers will need to consider the relevant aspects of their plant's design and operating characteristics when converting the generic guidance of this document into a site-specific classification scheme." Please verify that the Loss of the Fuel Clad Barrier threshold value for the Drywell and Torus radiation monitors, based on a loss of the reactor coolant system resulting in approximately 2% to 5% clad damage, considers the site- specific aspects of the DAEC design and operating characteristics. If the fuel clad barrier threshold value is not based on DAEC site-specific values, revise the proposed fuel clad barrier threshold value to align with site-specific values. DAEC Response After further clarifying discussion with the NRC staff during an October 3, 2018 telephone call, NextEra Energy Duane Arnold acknowledges that the previously proposed calculated values were based solely on a reactor coolant system activity of 300 uci/ gm Dose Equivalent Iodine (DEI) as provided by the applicable Developer Notes of NEI 99-01, Revision 6; and as such did not accurately reflect a site-specific range of fuel clad damage of approximately 2% -5%. NRC staff have clarified that the intent of this EAL is for the threshold value to provide a radiation monitor threshold value indicative of fuel clad damage of approximately 2% -5%. In the attachments to this letter, N extEra Energy Duane Arnold provides revised threshold values for the Drywell and Torus radiation monitors indicative of Loss of the Fuel Clad Barrier that are based on approximately 5% clad damage. NextEra Energy Duane Arnold intends to formally revise validation document V-23 (engineering calculation NEE-323-CALC-001) contained in Attachment 4 to reflect this new understanding of the intent of the EAL threshold prior to implementation of the revised EALs. The table on the following page provides the updated 5% clad release threshold values based on a scaling of the

• present 100% and 20% clad damage values provided in the current calculation.

1 of2

Enclosure to NG 18-0121 Development of Fission Product Barrier EAL Threshold Values from NEE-323-CALC-001 NOTE: Fuel Clad barrier LOSS 4.A(B) threshold values below are scaled from the 100% gap release instead of calculated based on 300uci/gm DEI as assumed in NEI 99-01 Revision 6 developer guidance. This variation from the NRC endorsed guidance is due to the calculated value not reflecting the intended 2-5% gap release threshold due to differences in plant design. The calculation will be formally revised to reflect this change in methodology. Drywell dose rate Torus dose rate Drywell dose rate Torus dose rate R/hr R/hr R/hr R/hr Values below are Values below are rounded for ease of rounded for ease of 1691 MWth 22700 2140 use, as well as to use, as well as to 100% Gap release provide a step-wise provide a step-wise After application of 0.2 scaling factor for 20% Gap release 5133 484 5000 500 CI'MNT banier LOSS 4A(B) After application of 0.05 scaling factor for 5% Gap release 1283 121 1250 , 125 Fuel Clad banier LOSS 4A(B) (' 2of2

ATTACHMENT 1 NEXTERA ENERGY DUANE ARNOLD, LLC DUANE ARNOLD ENERGY CENTER RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION RELATING TO LICENSE AMENDMENT REQUEST TSCR-166 UPDATED REDLINE MARKUP OF NEI 99-01 REVISION 6 311 pages follow

NEI 99 01. [Rewisien 6] Nuclear Energy Institute Duane Arnold Energy Center (DAEC) Emergency Action Levels Technical Bases Document TBD,2018 Nevember 2012

    }li1elf:!clr Energy !nstUute, 1776 I Street.\T. W., Suite 4()(), Washingte11 D.C. (2~. 7J9.8()()())

ACKNOWLEDGMENTS Tais doOO:R'lent was prepared by the }foelear &ergy Instirate (),TEI) Emergeaey Aetion Level (EAL) Task Foree. NEI ChoiFpeFsea: David Ymmg PFepaFRtiea Team Larry Baker Exeloa Nuclear/Corporate Craig BaftB:er PSEG Nuclear/Salem and Hope Creek Naelear Geueratiftg Stations/US}.. JohH. Egclorf DomiBioR Geaera-tion/KewaUBee Power gtatioa Jaek Lewis &tergy Nuclear/Corporate C. Kelly Walker Opera-tioas Support Services, Ine. Re:view Team Chris Boone Soutaern Nuclear/Corporate JohR Callahafl Xcel Eaergy/Corporate/USi\ Bill Chausse Enereon Serviees, Ine. KCH:t Croeker Progress &ergy/Bruns1.vick Nuclear Plant Don Crov,rl Duke ERergy/Corporate Roger Freeman Constellatioa ERergy }h1clear Group/Corporate Walt Lee TVA }raelear/Corporate Ken Meade FENOC/Corporate DoH Mothooa NextEra Energy/Corporate David Stobaugh EP Consulting,* LLC Niek Tomer Callaway Pla£1t/STARS Maureen Zw.valiek Diablo Carlyon Power Pla£1t1STARS NOTICE Neither NEI, nor a.fly of its employees, members, supporting organizations, contractors, or eonsultants make any warranty, eKpressed or implied, or assume any legal responsibility for the aecuraey or emnpletooess of, or assUiB:e aay liability for dam.ages restrlting from any use of, a.fly infonnation apparatus, methods, or process disclosed rn this report or that such may not infringe prr;ately o'.vB:ed rights.

           }h1e.'ear Energy Jnsfitu$e, 1776 I StreeLV W:, SuUe 400, Wcf!)hin-gten D.C. (20:2.739.8000)

NEI 99 QI (Re,,*isiea a) Nsvember 2QI2 TABLE OF CONTENTS 1 BASIS FOR EMERGENCY ACTION LEVELS ................................................................. 1 1.1 OPERATING REACTORS ************************************************************************************************** 1 1.2 INDEPENDENT SPENT FUEL STORAGE INSTALLATION SFSI ..................................... 2 1.3 NRC ORDER EA-12-051 .*..............*...*....*.........**.....*........*.*..**...*....*....*.*...**...*........**... 4 2 KEY TERMINOLOGY USED IN DAEC EAL SCHEME..................................................... 6 2.1 EMERGENCY CLASSIFICATION LEVEL {ECL) ............................................................... 6 2.2 INITIATING CONDITION IC ..........................................................................................8 2.3 EMERGENCY ACTION LEVEL EAL *....*....*.***............*..**..*..*...*.*.*.*****.....**.......**......*.. 8 2.4 FISSION PRODUCT BARRIER TIIRESHOLD .....................................................................8 3 DESIGN OF THE DAEC EMERGENCY CLASSIFICATION SCHEME ........................... 11 3.1 ASSIGNMENT OF EMERGENCY CLASSIFICATION LEVELS CLs ............................. 11 3 .2 TYPES OF INITIATING CONDITIONS AND EMERGENCY ACTION LEVELS .................... 17 3.3 DAEC-SPECIFIC ORGANIZATION AND PRESENTATION OF GENERIC INFORMATION18 3.4 IC AND E.AL MODE APPLICABILITY **************************************************************************** 20 4 DAEC SCHEME DEVELOPMENT .................................... ~ D ** * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * *

• 22 4.1 GENERAL DEVELOPMENT PROCESS ****************************************************************************22

 ,4.2 CRITICAL CHARACTERISTICS ................................... :..................................................23 4.3 INSTRUMENTATION USED FOR E.ALs .......................................................................... 25 4.4 EAL/THRESHOLD REFERENCES TO AOP AND EOP SETPOINTS/CRITERIA .............. 27 5  GUIDANCE ON USING THE DAEC EALS ..............:.......................                           11 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 5.1  GENERAL CONSIDERATIONS ******************************************** ~ ******************************************* 29 5.2 CLASSIFICATION METHODOLOGY *******************************************************************************31 5.3  CLASSIFICATION OF MULTIPLE EVENTS AND CONDITIONS ........................................31 5.4 CONSIDERATION OF MODE CHANGES DURING CLASSIFICATION .............................. 32 5.5 CLASSIFICATION OF IMMINENT CONDITIONS .............................................................33 5.6 EMERGENCY CLASSIFICATION LEVEL UPGRADING AND DOWNGRADING .................33 5.7 CLASSIFICATION OF SHORT-LIVED EVENTS ...............................................................35 5.8 CLASSIFICATION OF TRANSIENT CONDITIONS ............................................................ 35 5.9 AFTER-THE-FACT DISCOVERY OF AN EMERGENCY EVENT OR CONDITION ..............36 111

NEI 99 0 I (Revisiee 6) Nevem.ber 2012 5 .10 RETRACTION OF AN EMERGENCY DECLARATION .......................................................36 6 ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT ICS/EALS ........................ 37 7 COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTION ICS/EALS .................... 74 8 INDEPENDENT SPENT FUEL STORAGE INSTALLATION {ISFSll ICS/EALS ............ 117 9 FISSION PRODUCT BARRIER ICS EALS ................................................................120 10 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS ....... 179 11 SYSTEM MALFUNCTION ICS EALS ..........................................................................215 APPENDIX A - ACRONYMS AND ABBREVIATIONS ........................................................A-1 APPENDIX B - DEFINITIONS *****..******** D...................... 11**************11*********************11***********11****8*1 IV

NEI 99 Q1 (Revisioe e) November 2912 DE'/ElOMENT OF DUANE ARNOLD EMERGENCY ACTION LEVELS FOR NON PA&&l'.'E REACTORS TECHNICAL BASIS DOCUMENT 1 BASIS FOR EMERGENCY ACTION LEVELSREGULATORY BACKGROUND 1.1 OPERATING REACTORS Title 10, Code of Federal Regulations (CFR), Energy, contains the U.S. Nuclear Regulatory Commission (NRC) regulations that apply to nuclear power facilities. Several of these regulations govern various aspects of an emergency classification scheme. A review of the relevant sections listed below will aid the reader in understanding the key terminology provided in Section 3.0 of this document.

• 10 CFR § 50.47(a)(l)(i)
• 10 CFR § 50.47(b)(4)
• 10 CFR § 50.54(q)
• 10 CFR § 50.72(a)
• 10 CFR § 50, Appendix E, IV.B, Assessment Actions
• 10 CFR § 50, Appendix E, IV.C, Activation of Emergency Organization The above regulations are supplemented by various regulatory guidance documents.

Three documents of particular relevance to NEI 99-01 are: NUREG-0654/FEMA-REP-1, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support ofNuclear Power Plants, October 1980. [Refer to Appendix 1, Emergency Action Level Guidelines for Nuclear Power Plants] NUREG-1022, Event Reporting Guidelines 10 CFR § 50.72 and§ 50.73 Regulatory Guide 1.101, Emergency Response Planning and Preparedness for Nuclear Power Reactors 1

NEI 99 QI (Revisioa 6) November 2Ql2 The above list is not aU i-ncmsP.'e and it is strongly recommended that scheme de*,relopers consult with licensing/regulatory compliance personnel to identify and understand all applicable requirements and guidance. Questions may also be directed to the NEI Emergency Preparedness staff. 1.2 PEru.iMtENTLY DEFUELEB STATION J\IBI 99 0 I provides guidance for an emergency classification scheme applicable to a pennanently defueled station. This is a station that generated spent fuel under a IO CFR

            § 50 license, has pennanently ceased operations and will store the spent fuel onsite for an mcteaded period of time. The emergency classification levels applicable to this type of station are consistent .vith the requirements of IO CFR § 50 and the guidance in NUREG 1

0654/FEMA. REP I. In order to relffi, the e:ipergeacy plan requiremeB.ts applicable to an operatH'l:g station, the 0 1:mer ofa permaneB.tly defueled station must demonstrate that no credible e11ent can result HI: a significaB.t radiological release beyond the site boundary. It is m.pected that this verification 1vvill confirm that the source term and motive force available in the pennaneB.tly defueled condition are insufficieB.t to v;arrant classifications of a Site Area Emergency or General Etnergeacy. Therefore, the geaeric InitiatH1:g Conditions (ICs) and Emergency Action Levels (Ef..Ls) applicable to a permanently defueled station may result HI: either a Notification of Unusual Event (J\tOUE) or an Alert classification. The generic ICs and Ei"..Ls are preseB.ted in Appendi1c C, P-en-11aHe¥1;lfy Defi1e.led &alien

            !Cs/EALs.
 +.:-3-l.1_1NDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) .

Selt:cted guidance in NEI 99-01 is applicable to licensees electing to use their 10 CFR 50 emergency plan to fulfill the requirements of 10 CFR 72.32 for a stand-alone ISFSI. The emergency classification levels applicable to an ISFSI are consistent with the requirements of 10 CFRf-50 and the guidance in NUREG 0654/FEMA-REP-1. The initiating conditions germane to a 10 CFR--§- 72.32 emergency plan (as described in NUREG-1567) are subsumed within the classification scheme for a 10 CFRf-50.47 emergency plan. The generic ICs and EALs for an ISFSI are presented in Section 8, ISFSI ICs/EALs. IC E-HUl covers the spectrum of credible natural and man-made events included within the scope of an ISFSI design. This IC is not applicable to installations or facilities that may process and/or repackage spent fuel (e.g., a Monitored Retrievable Storage Facility or an ISFSI at a spent fuel processing facility). In addition, appropriate aspects ofIC HUI and IC HAI should also be included to address a HOSTILE ACTION directed against an ISFSI. The analysis of potential onsite and offsite consequences of accidental releases associated with the operation ofan ISFSI is contained in NUREG-1140, A Regulatory Analysis on Emergency Preparedness for Fuel Cycle and Other Radioactive Material Licensees. NUREG-1140 concluded that the postulated worst-case accident involving an ISFSI has insignificant consequences to public health and safety. This evaluation shows that the 2

NEI 99 01 (R-evisieH a) Nevemeer 2012 maximum offsite dose to a member of the public due to an accidental release of radioactive materials would not exceed I rem Effective Dose Equivalent. Regarding the above information, the expectations for an offsite response to an Alert classified under a IO CFR f-72.32 emergency plan are generally consistent with those for a Notification of Unusual Event in a 10 CFR f-50.47 emergency plan (e.g., to provide assistance ifrequested). Also, the licensee's Emergency Response Organization (ERO) required for 10 CFR f-72.32 emergency plan is different than that prescribed for a 10 CFR f-50.47 emergency plan (e.g., no emergency technical support function)~ 3

NEI 99 01 (Revisioa 6) Novemeer 2012 +:4.Ll.._NRC ORDER EA-12-051 The Fukushima Daiichi accident of March 11, 2012, was the result of a tsunami that exceeded the plant's design basis and flooded the site's emergency electrical power supplies and distribution systems. This caused an extended loss of power that severely compromised the key safety functions of core cooling and containment integrity, and ultimately led to core damage in three reactors. While the loss of power also* impaired the spent fuel pool cooling function, sufficient water inventory was maintained in the pools to preclude fuel damage from the loss of cooling. Following a review of the Fukushima Daiichi accident, the NRC concluded that several measures were necessary to ensure adequate protection of public health and safety under the provisions of the backfitrule, 10 CFR 50.109(a)(4)(ii). Among them was to provide each spent fuel pool with reliable level instrumentation to significantly enhance the ability of key decision-makers to allocate resources effectively following a beyond design basis event. To this end, the NRC issued Order EA-12-051, Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation, on March 12, 2012, to all US nuclear plants with an operating license, construction permit, or combined construction and operating license. NRC Order EA-12-051 states, in part, "All licensees ... shall have a reliable indication of the water level in associated spent fuel storage pools capable of supporting identification of the following pool water level conditions by trained personnel: (I) level that is adequate to support operation of the normal fuel pool cooling system, (2) level that is adequate to provide substantial radiation, shielding for a person standing on the spent fuel pool operating deck, and (3) level where fuel remains covered and actions to implement make-up water addition should no longer be deferred." To this end, all licensees must provide:

• A primary and back-up level instrument that will monitor water level from the normal level to the top of the used fuel rack in the pool;
• A display in an area accessible following a severe event; and
• Independent electrical power to each instrument channel and provide an alternate remote power connection capability.

NEI 12-02, Industry Guidance for Compliance with NRC Order EA-12-051, "To Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation, " provides guidance for complying with NRC Order EA-12-051. NEI 99-01, Revision 6, includes three EALs that reflect the availability of the enhanced spent fuel pool level instrumentation associated with NRC Order EA-12-051. These EALs are included within 01..isting IC2 ~RA2, aHd Hev,r ICs AS2 RS2, and ~RG2. Associated E,'\L Botos, bases aHd developer Botos are also provided. It is recommended that these EALs be implemeHted vA1eH the enhaHced spent fuel pool level instrumeHtatioH is available for use. 4

NEI 99 0 I (Re,,isiea a) Nevemeer 2012 2 KEY TERMINOLOGY USED IN NEI 99 OlDAEC EAL SCHEME There are several key terms that appear throughout the "NEI 99 0 lEAL methodology. These terms are introduced in this section to support understanding of subsequent material. As an aid to the reader, the following table is provided as an overview to illustrate the relationship of the terms to each other. Emergency Classification Level Unusual Event I Alert I SAE I GE I Initiating Condition I Initiating Condition I Initiating Condition I Initiating Condition Emergency Action Emergency Action Emergency Action Emergency Action Level (1) Level (1) Level (1) Level (1)

• Operating Mode
• Operating Mode
• Operating Mode
• Operating Mode Applicability Applicability Applicability Applicability
• Notes
• Notes
• Notes
• Notes
• Basis
• Basis
• Basis
• Basis (1) - When making an emergency classification, the Emergency Director must consider all information having a bearing on the proper assessment of an Initiating Condition. This includes the Emergency Action Level (EAL) plus the associated Operating Mode Applicability, Notes and the informing Basis information. In the Recognition Category F matrices, EALs are referred to as Fission Product Barrier Thresholds; the thresholds serve the same function as an EAL:--.:.

2.1 EMERGENCY CLASSIFICATION LEVEL (ECL) One of a set of names or titles established by the US Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions according to (1) potential or actual effects or consequences, and (2) resulting onsite and offsite response actions. The emergency classification levels, in ascending order of severity, are: Notification of Unusual Event (NOUE) Alert Site Area Emergency (SAE) General Emergency (GE) 2.1.1 Notification of Unusual Event (NOUE}1" Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases ofra:dioactive material requiring offsite response or monitoring are expected unless further degradation of safety syst0fB.sSAFETY SYSTEMS occurs.

 + This term is semetim.es shertened ~e Uft:1sual Eveat (UE) er other similar site Sfleeifie termiaelegy. The terms Notifieatiea efUHtlsHal Eveat, NOUE and Um1sHal Eveat are used iaterehangeably threHgheHt this deet1ffient 6

NEI 99 QI (R-evisiea a) NeYember 2QI2

Purpose:

The purpose of this classification is to assure that the first step in future response has been carried out, to bring the operations staff to a state of readiness, and to provide systematic handling of unusual event information and decision-making. 2.1.2 Alert Events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA PAG exposure levels.

Purpose:

The purpose of this classification is to assure that emergency personnel are readily available to respond if the situation becomes more serious or to perform confirmatory radiation monitoring if required, and provide offsite authorities current information on plant status and parameters. 2.1.3 Site Area Emergency Events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; I) toward site personnel or equipment that could lead to the likely failure of or; 2) that prevent effective access to, equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA P AG exposure levels beyond the site boundary. Purpose': The purpose of the Site Area Emergency declaration is'to* assure that emergency response centers are staffed, to assure that monitoring teams are dispatched, to assure that personnel required for evacuation of near-site areas are at duty stations if the situation becomes more serious, to provide consultation with offsite authorities, and to provide updates to the public through government authorities. 2.1.4 General Emergency (GE) Events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA PAG exposure levels* offsite for more than the immediate site area.

Purpose:

The purpose of the General Emergency declaration is to initiate predetermined protective actions for the public, to provide continuous assessment of information from the licensee and offsite organizational measurements, to initiate additional measures as indicated by actual or potential releases, to provide consultation with offsite authorities, and to provide updates for the public through government authorities. 7

NEI 99 QI (Revisien 6) Ne:vember 2Ql2 2.2 INITIATING CONDITION (IC) An event or condition that aligns with the definition of one of the four emergency classification levels by virtue of the potential or actual effects or consequences. Discussion: An IC describes an event or condition, the severity or consequences of which meets the definition of an emergency classification level. An IC can be expressed as a continuous, measurable parameter (e.g., RCS leakage), an event (e.g., an earthquake) or the status of one or more fission product barriers (e.g., loss of the RCS barrier). Appendix 1 of NUREG-0654 does not contain example Emergency Action Levels_ (EALs) for each ECL, but rather Initiating Conditions (i.e., plant conditions that indicate that a radiological emergency, or events that could lead to a radiological emergency, has occurred). NUREG-0654 states that the Initiating Conditions form the basis for establishment by a licensee of the specific plant instrumentation readings (as applicable) which, if exceeded, would initiate the emergency classification. Thus, it is the specific instrument readings that would be the EALs. Considerations fur the assignment of a particular Initiating Condition to an emergency classification le*;el are discassed in geetion 3. 2.3 EMERGENCY ACTION LEVEL {EAL) A pre-determined, site-specific, observable threshold for an Initiating Condition that, when met or exceeded, places the plant in a given emergency classification level. Discussion: EAL statements may utilize a variety of criteria including instrument

• readings and status indications; observable events; results of calculations and analyses; entry into particular procedures; and the occurrence of natural phenomena.

2.4 FISSION PRODUCT BARRIER THRESHOLD A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier. Discussion: Fission product barrier thresholds represent threats to the defense in depth design concept that precludes the release of radioactive fission products to the environment. This concept relies on multiple physical barriers, any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment. The primary fission product barriers are: Fuel Clad Reactor Coolant System (RCS) Containment

   ---Upon determination that one or more fission product barrier thresholds have been exceeded, the combination of barrier loss and/or potential loss thresholds is compared to the fission product barrier IC/EAL criteria to determine the appropriate ECL.

In some accident sequences, the ICs and EALs presented in the Abnormal Radiation Levels/ Radiological Effluent (AR) Recognition Category will be exceeded at the same 8

NEI 99 01 (Re¥isisa e) Nsvem.ber 2012 time, or shortly after, the loss of one or more fission product barriers. This redundancy is intentional as the former ICs address radioactivity releases that result in certain offsite doses from whatever cause, including events that might not be fully encompassed by fission product barriers (e.g., spent fuel pool accidents, design containment leakage following a LOCA, etc.). 9

      ,1 0

NEI 99 QI (Ro=visioa 6) November 2Ql2 3 DESIGN OF THE Nlil 99 01 DAEC EMERGENCY CLASSIFICATION SCHEME 3 .1 ASSIGNMENT OF EMERGENCY CLASSIFICATION LEVELS (ECLs) An effective emergency classification scheme must incorporate a realistic and accurate assessment of risk, both to plant workers and the public. There are obvious health and safety risks in underestimating the potential or actual threat from an event or condition; however, there are also risks in overestimating the threat as well (e.g., harm that may occur during an evacuation). The NEI 99 OlDAEC emergency classification scheme attempts to strike an appropriate balance between reasonably anticipated event or condition consequences, potential accident trajectories, and risk avoidance or minimization. There are a range of "non-emergency events" reported to the US Nuclear Regulatory Commission (NRC) staff in accordance with the requirements of 10 CFR ,§--50.72. Guidance concerning these reporting requirements, and example events, are provided in NUREG-1022. Certain events reportable under the provisions of 10 CFR ,§--50.72 may also require the declaration of an emergency.

• In order to align each Initiating Conditions (IC) with the appropriate ECL, it was necessary to determine the attributes of each ECL. The goal of this process is to answer the question, "What events or conditions should be placed under each ECL?" The following sources provided information and context for the development of ECL attributes.

Assessments_ of the effects and consequences of different types of even~~ and conditions *

• Typical DAEC abnormal and emergency operating procedure setpoints and transition criteria Typical DAEC Technical Specification limits and controls Radiological Effluent Technical Specifications (RETS)/Offsite Dose Calcttlation Assessment Manual (ODAM) radiological release limits Review of selected Updated Final Safety Analysis Report (UFSAR) accident analyses Environmental Protection Agency (EPA) Protective Action Guidelines (PAGs)

NUREG 0654, Appendix 1, Emergency Action Level Guidelines for Nuclear Power Plants Industry Operating Experience Input from industry DAEC subject matter experts and NRG staff members The following ECL attributes were-are created used by the Revision 6 Preparation Team to aid in the development of ICs and Emergency Action Levels (EALs). The team. decided to IDclttde the attribtttes ID this revision since theyThe attributes may be useful in briefing and training settings (e.g., helping an Emergency Director understand why a particular condition is classified as an Alert). It should be stressed that developers not attempt to redefme these attribtttes or apply them in any fashion that \vorud change the generic guidance contaIDed ID this doc1:lmeat\ + The use of EGL aUrilmtes is at the diseretioa ofa lieensee and is not a requirement of the }ffi.C. Ifa licensee chooses in incorperatc the EGL attriln1tes iate their scheme basis docu1Ee1J:t, it ll'Kist be very elear that the NR£ staff has not endorsed their acceptability or applicatien for BHY p:1Fpose. In particular, the staff does not consider the 11

N

NEI 99 QI (RsYisiea 6) Nevsmb@r 2912 3.1.1 Notification of Unusual Event (NOUE) A Notification of Unusual Event, as defined in section 2.1.1, includes but is not limited to an event or condition that involves: (A)A precursor to a more significant event or condition. (B) A minor loss of control of radioactive materials or the ability to control radiation levels within the plant. (C) A consequence otherwise significant enough to warrant notification to local, State and Federal authorities. 3.1.2 Alert An Alert, as defined in section 2.1.2, includes but is not limited to an event or condition that involves: (A)A loss or potential loss of either the fuel clad or Reactor Coolant System (RCS) fission product barrier. (B) An event or condition that significantly reduces the margin to a loss or potential loss of the fuel clad or RCS fission product barrier. (C) A significant loss of control ofradioactive materials resulting in an inability to control radiation levels within the plant, or a release of radioactive materials to the environment that co{!ld result in doses greater than 1% of an EPA P AG at or beyond the site boundary. (D)A HOSTILE ACTION occurring within the OWNER CONTROLLED AREA, including those directed at an Independent Spent Fuel Storage Installation (ISFSI). 3.1.3 Site Area Emergency (SAE) A Site Area Emergency, as defined in section 2.1.3, includes but is not limited to an event or condition that involves: (A)A loss or potential loss of any two fission product barriers - fuel clad, RCS and/or containment. (B) A precursor event or condition that may lead to the loss or potential loss of :multiple fission product barriers within a relatively short period of time. Precursor events and conditions of this type include those that challenge the monitoring and/or control of multiple safety systemsSAFETY SYSTEMS. (C)A release of radioactive materials to the environment that could result in doses greater than 10% of an EPA P AG at or beyond the site boundary. (D)A HOSTILE ACTION occurring within the plant PROTECTED AREA. 14

NEI 99 01 (R~visioa 13) November 2012 3.1.4 General Emergency (GE) A General Emergency, as defined in section 2.1.4, includes but is not limited to an event or condition that involves: (A)Loss of any two fission product barriers AND loss or potential loss of the third barrier

         - fuel clad, RCS and/or containment.

(B) A precursor event or condition that, unmitigated, may lead to a loss of all three fission product barriers. Precursor events and conditions of this type include those that lead directly to core damage and loss of containment integrity. (C) A release of radioactive materials to the environment that could result in doses greater than an EPA PAG at or beyond the site boundary. (D)A HOSTILE ACTION resulting in the loss of key safety functions (reactivity control, core cooling/RPV water level or RCS heat removal) or damage to spent fuel. 3.1.5 15

NEI 99 QI (Revisioa 6) November 2Ql2 Risk-Informed Insights Emergency preparedness is a defense-in-depth measure that is independent of the assessed risk from any particular accident sequence; however, the development of an effective emergency classification scheme can benefit from a review of risk-based assessment results. To that end, the development and assignment of certain ICs and EALs also considered insights from several site-specific probabilistic safety assessments (P8A also knovm as probabilistic risk assessment, PRA). Some generic insights from this review included:

1. Accident sequences involving a prolonged loss of all AC power are significant contributors to core damage frequency at many Pressurized Water Reactors (PVlRs) aE:El-Boiling Water Reactors (BWRs). For this reason, a loss of all AC power for greater than 15 minutes, with the plant at or above Hot Shutdown, was assigned an ECL of Site Area Emergency, Precursor events to a loss of all AC power were also included as an Unusual Event and an Alert.

A station blackout coping analyses performed in response to 10 CFR ,§--50.63 and Regulatory Guide 1.155, Station Blackout, may be used to determine a time-based criterion to demarcate between a Site Area Emergency and a General Emergency. The time dimension is critical to a properly anticipatory emergency declaration since the goal is to maximize the time available for State and local officials to develop and implement offsite protective actions.

2. For severe core damage events, uncertainties exist in phenomena important to accid~nt progressions leading to containment failure. _Because of.these uncertainties, predicting the status of containment integrity may be difficult under severe accident conditions. _This is why maintaining containment integrity alone following sequences leading to severe core damage is an insufficient basis for not escalating to a General Emergency.

3. PSAs indicated that leading contributors to latent fatalities were sequences involving a containment bypass, a large Loss of Coolant Accident (LOCA) with early containment failure, a Station Blackout lasting longer than the site speeifieDAEC coping period, and a reactor coolant pump seal failure. The generic EAL methodology needs to be sufficiently rigorous to address these sequences in a timely fashion.

16

NEI 99 01 (R-evisieH e) NeY@mber 2012 3.2 TYPES OF INITIATING CONDITIONS AND EMERGENCY ACTION LEVELS The NEI 99-01 methodology makes use of symptom-based, barrier-based and event-based ICs and EALs. Each type is discussed below. Symptom-based ICs and EALs are parameters or conditions that are measurable over some range using plant instrumentation (e.g., core temperature, reactor coolant level, radiological effluent, etc.). When one or more of these parameters or conditions are off-normal, reactor operators will implement procedures to identify the probable cause(s) and take corrective action. Fission product barrier-based ICs and EALs are the subset of symptom-based EALs that refer specifically to the level of challenge to the principal barriers against the release of radioactive material from the reactor core to the environment. These barriers are the fuel cladding, the reactor coolant system pressure boundary, and the containment. The barrier.:-based ICs and EALs consider the level of challenge to each individual barrier - potentially lost and lost - and the total number of barriers under challenge. Event-based ICs and EALs define a variety of specific occurrences that have potential or actual safety significance. These include the failure of an automatic reactor scra~ to shut down the reactor, natural phenomena (e.g., an earthquake), or man-made hazards such as a toxic gas release. 17

NEI 99 01 (Revisiea e)

P.Tevember 2012 3.3 N888 DESIGN DIFFERENCES The }JEI 99 01 emergency classification scheme accounts for the design differences benveen PWRs and BV/Rs by specifying E.t\Ls unique to each type of

       }raclear Steam. Supply 8ysteH1: (N888). There are also significant design differeaces am.o>>g PVlR }TS88s; therefore, gi:iidance is provided to aid ill the development of EA.Ls appropriate to differe>>t PV/R N888 types. Where necessary, development guidance also addresses unique co>>siderations for advanced non passive reactor designs such as the Advanced Boiling Water Reactor (AB::X:VR), the A,dvanced Pressl:lrized "Nater Reactor (APWR) and the Evolatiollflf)' Power Reactor (EPR).

Developers 1.vill need to consider the relevant aspects of their plam's design and operati>>g eharacteristics *.vhen converting the generic guidance of this document iH:to a site specific classification scheme. The goal is to mai.Btain as mace fidelity as possible to the iH:teHt of generic ICs ana &:'\Ls vtithin the constraints imposed by th:e plant design B:B:a operating characteristics. To this end, developers of a scheme for an advanced non passive reactor may need to add, modify or delete some information contained in this E:locaraent; th:ese ch:anges will be reviewea for acceptability by the }lRC as part of th:e sche1He approval process. The guidance in NEI 99 0 I is H:ot applicable to advanced passive light 1vvater reactor designs. fill Emergeacy ClassificatioH: Scheme for this type of plant should be E:le*;eloped iR accordance with NEI 07 01, 1\/ethed-elegyfor Develepment e.fEnwrgency* Aetien L!'\ 1els, AdvaneedP-assive Light W~wr Reaeters. 3-:43.3 DAEC-SPECIFIC ORGANIZATION ANQ PRESENTATION OF GENERIC INFORMATION

                                               'I The scheme's generic information is organized by Recognition Category in the following order.

A-R - Abnormal Radiation Levels / Radiological Effluent - Section 6 C - Cold Shutdown / Refueling System Malfunction - Section 7 E - Independent Spent Fuel Storage Installation (ISFSI) - Section 8 F - Fission Product Barrier - Section 9 H - Hazards and Other Conditions Affecting Plant Safety- Section 10 S - System Malfunction - Section 11 PD Pennanently Defueled Station i\:ppoodiJt C Each Recognition Category section contains a matrix showing the ICs and their associated emergency classification levels. The following information and guidance is provided for each IC: ECL - the assigned emergency classification level for the IC. Initiating Condition - provides a summary description of the emergency event or condition. Operating Mode Applicability - Lists the modes during which the IC and associated EAL(s) are applicable (i.e., are to be used to classify events or conditions). 18

NEI 99 01 (Revisien 6) Nevemeer 2012 Examf'le Emergency Action Level(s)-Provides examples ofreports and indications that are considered to meet the intent of the IC. Developers should address each mrnmple EAL. If the generic approach to the development of an eJ,Rfllple EAL caenot be used (e.g., an assIB11ed instrnm,entatioa range is not available at the plant), the developer should attempt to specify an altemate means for identifying entry into the IC. For Recognition Category F, the fission product barrier thresholds are presented in tables applicable to BWRs and PWRs, and arranged by fission product barrier and the degree of barrier challenge (i.e.,_-potential loss or loss). This presentation method shows the synergism among the thresholds, and supports accurate assessments. Basis - Provides background information that explains the intent and application of the IC and EALs. In some cases, the basis also includes relevant source information and references. 19

NEI 99 0 l (R-evisioR G) November 2012 Develape:r Nates Information that supports the development of the site specific ICs and EALs. This may include clarifications, references, ~rnmples, instrnctions for calcltlations, etc. De:veloper notes should not be inclttded in the site's emergency classification scheme basis document. Developers may elect to incluele information resulting from a de:r1eloper note action in a basis section. ECL i-t...ssignment f...ttributes Located *.vithin the De:r.zelop& Notes section, specifies the attrihate :aseel for assigning the IC to a given EGL.

  ~3.4 IC AND EAL MODE APPLICABILITY The NEI 99 01 DAEC emergency classification scheme was developed recognizing that the applicability of ICs and EALs will vary with plant mode. For example, some symptom-based ICs and EALs can be assessed only during the power operations, startup, or hot standby/shutdown modes of operation when all fission product barriers are in place, and plant instrumentation and safety systemsSAFETY SYSTEMS are fully operational. In the cold shutdown and refueling modes, different symptom-based ICs and I       EALs will come into play to reflect the opening of systems for routine maintenance, the 1.

unavailability of some safety systemSAFETY SYSTEM components and the use of alternate instrumentation. The following table shows which Recognition Categories are applicable in each plant mode. The ICs and EALs for a given Recognition Category are applicable in the indicated modes. MODE APPLICABILITY MATRIX Recognition Category Mode AR C E F H s Power Operations X X X X X Startup X X X X X Hot Staneley -' -' -' Hot Shutdown X X X X X Cold Shutdown X X X X Refueling X X X X ,-*, Defueled X X X X Permanently Defueleel 20

NEI 99 01 (RevisieH t3) Ne11smbsr 2012 Tyaieal B'\¥R DAEC Operating Modes Power Operations (1): Mode Switch in Run Startup (2): Mode Switch in Startup/Hot Standby or Refuel (with all vessel head closure bolts fully tensioned) Hot Shutdown (3): Mode Switch in Shutdown, Average Reactor Coolant Temperature >~212 °P (with all vessel head closure bolts fully tensioned) Cold Shutdown (4): Mode Switch in Shutdown, Average Reactor Coolant Temperature~ ~212 °P (with all vessel head closure bolts fully tensioned) Refueling (5): Mode Switch in Shutdown or Refuel;-ffllEI (with one or more vessel head closure bolts less than fully tensioned} Tv~ieal PWR OPERATING MODES Power Operations (1):Reactor Power> 5%, Keff> 0.99 Startup (2): Reactor Power 6 5%, Keff;:,_ 0.99 Hot Standby (3): RCS ;:,_ 350 °f, Keff < 0.99 Hot Shl:l-1:dov,'fl. (4): 200 °F < RtS < 350 °f, Keff< 0.99 Cold ShutdO'l,'fl. (5): RCS< 200 °f, Keff< 0.99 Refueling (6): One or more vessel head closure bolts less taan fully tensioned De1,elopers will need to incorporate ilie mode criteria from\H1it specific Technical Specifications into ilieir emergency classification scheme. In addition, the scheme must also include the following mode designation specific to NEI 99 01: Defueled (None): All fuel removed from the reactor vessel (i.e., full core offload during refueling or extended outage). 21

NEI 99 01 (RevisieH a) l!>reY@ffl98f 2012 4 SIT& SP&CIFIC SCM&M& D&\'&lOPM&NT GUIDANC&DEVELOPMENT OF THE DAEC EMERGENCY CLASSIFICATION SCHEME This section pro:vides detailed guidance for de veloping a site specific emergency classification 0 scheme. Conceptually, the approach discussed here mirrors the approach used to prepare emergency operating proceCRH"es generic material prepared by reactor vendor mvners groups is converted by each nuclear power plant into site specific emergency operating procedures. Likc,.vise, the emergency classification scheme developer will use the generic guidance in NEI 99 0 I to prepare a site specific emergency classification scheme and the associated basis docllfl'lent. It is important that the NEI 99 01 emergency classification scheme be implemented as an integrated package. Selected use of portions of this guidance is strongly discouraged as it will lead to an inconsistent or incomplete emergency classification scheme that will likely not receive the Hecessary regulatory approYal. 4.1 GENERAL IMPLEMENTATION CUIDANCEDEVELOPMENT PROCESS The guidance in NEI 99 01 is not intended to be applied to plants "as is"; how0'1er, developers should attempt to keep their site specific schemes as close to the generic guidance as possible. The goal is to meet the iB.teB.t of the geReric !B.itiating Conditions (ICs) and Emergency Action Levels (EA.Ls) within the conteJlt of site specific cliaracteristics locale, plant design, operating features, tenninology, etc. Meeting this goal vtill result in a sho1ter and less cumbersome NRt review and approval process, closer alignm.ent 1.vith. the schemes of other nuclear pmver plant sites and better positioning to adopt future industry wide scheme enhancements. V/hen properly developed, the The DAEC ICs and EALs should were developed to be unambiguous and readily assessable. As discussed in Seetion 3, the generic guidance inch:1des ICs and example EALs. It is the intent of this guidance th.at QQfil be included in site specific documents as each serves a specific purpose. The IC is the fundamental event or condition requiring a declaration. _The EAL(s) is the pre-determined threshold that defines when the IC is met. If some feature of the plant location or design is e:ot compatible with a generic IC or EAL, efforts should be made to identify an alternate IC or Ei\L. If an IC or EAL includes an explicit reference to a mode dependent technical specifieation limit that is not applicable to the plant, then th.at IC and/or Ef. .L need not be in.ch:l.ded in the site specific scheme. lB. these cases, devel0f)ers mU:St provide adeE}'aate documentation to justify v;hy the IC and/or EAL were not incorporated (i.e., sufficient detail to allov,r a third party to understand the decision not to ie:corporate the generic guidance). Useful acronyms and abbreviations associated with the }JEI 99 OlDAEC emergency classification scheme are presented in Appendix A, Acronyms and Abbreviations. -Site-specific entries may be added if necessary. 22

NEI 99 0 I (Rei:ision 6) November 2012 Many words or terms used in the NEI 99 OlDAEC emergency classification scheme have scheme-specific definitions. These words and terms are identified by being set in all capital letters (i.e., ALL CAPS). The definitions are presented in Appendix B, Definitions. Below are eRamples of aeceptable modifications to the generic gmdance. These may be incorporated depending upon site developer and user preferences. The ICs within a Recognition Category may be placed in reverse order for presentation purposes (e.g., start vlith a General Emergency at the left/top of a user aid, follov,red by Site Area Emergeacy, Alert and l'lOUE). The Initiating Condition BUill:bering may be changed. The first letter of a R-0cognition Category designation may be changed, as follows, provided the chaage is carried through. for all of the associated IC identifiers. e R may be used in lieu of A

• M may be used in lieu of 8 For example, the Abnormal Radiation Levels / Radiological Effluent category desigaator "A" (for Abnonnal) may be changed to "R" (for Radiation). This means that the associated ICs 'vvould be changed to RUl, RU2, RAJ, etc.

The ICs and EA.Ls from Recognition Categories 8 and C may be incorporated into a common presentation method (e.g., one table) provided that all related notes and mode applicability requirements are maintained. The ICs an:d EALs for EmergeHey Director judgment and security related events may be placed under separate Recognition Categories.

• The terms EAL and threshold may be used interchangeably.

The material in the Developer }fotes section: is included to assist developers with erafting correct IC and EAL statements. This material is not required to be in the final emergency classification scheme basis document. 4.2 CRITICAL CHARACTERISTICS As discussed above, developers are encouraged to keep their site specific schemes as dose to the generic guidan:ce as possible. When crafting the scheme, developers should satisfy themselvesDAEC ensured that certain critical characteristics have been met. These critical characteristics are listed below.

• The ICs, EALs, Operating Mode Applicability criteria, Notes and Basis information are consistent with industry guidance; while the actual wording may be different, the classification intent is maintained. With respect to Recognition Category F, a-sfte--

specific scheme mustDAEC include.§.1! some type of user-aid to facilitate timely and accurate classification of fission product barrier losses and/or potential losses. The user-aid logic must beis consistent with the classification logic presented in Section 9. 23

J:ITEI 99 0 I (Revisien 6) Nevemb@r 2012

• The ICs, EALs, Operating Mode Applicability criteria, Notes and Basis information are technically complete and accurate (i.e., they contain the information necessary to make a correct classification).
• EAL statements use objective criteria and observable values.
• ICs, EALs, Operating Mode Applicability and Note statements and formatting consider human factors and are user-friendly.
• The scheme facilitates upgrading and downgrading of the emergency classification where necessary.
• The scheme facilitates classification of multiple concurrent events or conditions.

24

NEI 99 0 I (Revision 6) November 2012 4.3 INSTRUMENTATION USED FOR EALs Instrumentation referenced in EAL statements should inclade that described in the emergency plan section which addresses 10 CFR 50.47(b)(8) and (9) and/or Chapter 7 of the FSAR. Instrumentation :esed fur Ei'\Ls need not be safety related, addressed by a Techaical Specificatioa or ODCM:/RETS coatrol requirement, nor powered Hom an emergency pmver so"Hrce; hov1ever, EAL developers sho:eld strive to DAEC incorporateg instrumentation that is reliable and routinely maintained in accordance with site programs and procedures. Alarms referenced in EAL statements should beare those that are the most operationally significant for the described event or condition. Scheme developers sho:eld enS"HFe that specified values :esed as EAL setpoints are within the calibrated range of the referenced instrumentation, and consider any automatic instrumentation functions that may impact accurate EAL assessment. In addition, EAL setpoint values should do not use terms such as "off-scale low" or "off-scale high" since that type ofreading may not be readily differentiated from an instrument failure. Fiadings and violations related to EAL instrumentation issues may be located on the

NRG 1.vebsite.

4.4 PRESENTATION OF SCHEME INFORMATION TO UsERS The US Nuclear Regulatory Commission (1-ffi.C) expects licensees to establish and .. maintain the capability to assess, classify and declare an emergency condition promptly vii.thin 15 minutes after the a-vailability of indications to plant* operators that an emergency action le¥el has beea, or may be, mweeded. ¥.'hen writing an emergency classification proced"Hfe and creating related user aids, the developer must determine the presentation method(s) that best supports the end users by facilitating acC"HFate and timely emergency classification, To this end, developers sho:ald consider the fullowing poi-ft.ts. The first users of an emergency classification procedure are the operators in the Control Room. During the allowable classification time period, they may have responsibility to per:fum1 other critical tasks, and \Vill likely have minimal assistance in making a classification assessment. As an emergency situation evolves, members of the Control Room staff are likely to be the first personnel to notice a change in plant conditions. They can assess the changed conditions and, ;vhen .varranted, recommend a different emergency 1 1 classification level to the TeclHrical Supp01t Center (TSC) and/or Emergency Operations Facility (EOF). . Emergency Directors in the TSC and/or EOF \Vill have more opportunity to focus on making an en1ergency classification, and will probably have advisors from Operations available to help them. Emergency classification scheme information for end asers should be presented in a manner with which licensed operators are most comfortable. Developers will need to work closely with representatives from the Operations and Operations Training Departmeats to develop readily usable and easily understood classification tools (e.g., a procedure and related mer aids). If necessary, an alternate method for presenting 25

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NEI 99 0 l (R-evisioR a)

                                                                                        }fovember 2012 4.6     B!.:SIS Docm.tENT A basis document is an integral part of an emergency classification scheme. The material in this document supports proper emergency classification decision making by providing informing background and developmeB-t information iB a readily accessible format. It can be referred to ia training situations and when making an actual emergency classification, if necessary. The document is also useful for establishing configuration management controls for EP related equipmeB-t and eKplaining an emergency classification to offsite authorities. The coB:tent of the basis document shoHld iB:clude, at a 1Hinimlffll, the following:

• A site specific Mode Applicability Matrix and description of operating modes, similar to that pre sooted in section 3 .5.
• A discussion of the emergency classification and declaratioa process reflecting the material preseH.ted in Section: 5. This material may be edited as needed to aliga ,.vith site specific emergency plan aad implementing procedure requirements.

D Each Initiating Coadition along with the associated EALs or fissioa product barrier thresholds, Operating Mode Applica-bility, Notes aad Basis informatioa.

• A listing of acroayms and defiaed tenns, si1Hilar to that preseB-ted in l ..ppendices /'..

and B, respectively. This material may be edited as needed to align. with site specific characteristics.

• l ..ny site specific backgroUB:d or teclmical appendices that the developers belie>,re wou:ld be useful in mr.plaining or usiflg elemeB-ts of the e1Hergency classification scheme.

A Basis sectioa should not coB-tain information that could modify the meaniag or iqteB-t of the associated IC or.EAL. Such information should be incorporated within the 'JC or EAL statements, or as an E,"..L Note. InfonHation in the Basis should only clarify and inform decision making for an emergency classification. Basis informatioa should be readily available to be referooced, if necessary, by the Emergency Director. For mcarn.ple, a copy of the basis document eo:ald be maiatained in the appropriate emergency response faciJities. Because the infonHation in a basis document can affect emergency classification decision making (e.g., the Emergency Director refers to it during an event), the NRG staff e1r.pects that changes to the basis document ,.vill be evalaated in accordance with the provisions of 10 CFR 50.54(q). 4:-14.4 EAL/THRESHOLD REFERENCES TO AOP AND EOP SETPOINTS/CRITERIA As reflected in the generic guidance,Some of the criteria/values used in several EALs and fission product barrier thresholds may be are drawn from a plaB-t'sDAEC AOPs and EOPs. This approach is intended to maintain good alignment between operational diagnoses and emergency classification assessments. Dcwelopers should verify that a-A,ppropriate administrative controls are in place to ensure that a subsequent change to an AOP orEOP is screened to determine ifan evaluation pursuant to 10 CFR 50.54(q) is required. 27

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NEI 99 Q1 (Revisioa a) No:i,1ember 2012 5 GUIDANCE ON MAKING EM&RG&NCY ClASSIFICATIONSUSING THE DAECEALS 5.1 GENERAL CONSIDERATIONS When making an emergency classification, the Emergency Director must consider all information having a bearing on the proper assessment of an Initiating Condition (IC). This includes the Emergency Action Level (EAL) plus the associated Operating Mode Applicability, Notes and the infonning Basis information. In the Recognition Category F matrices, EALs are referred to as Fission Product Barrier Thresholds; the thresholds serve the same function as an EAL. NRC regulations require the licensee to establish and maintain the capability to assess, classify, and declare an emergency condition within 15 minutes after the availability of indications to plant operators that an emergency action level has been exceeded and to promptly declare the emergency condition as soon as possible following identification of the appropriate emergency classification level. The NRC staff has provided guidance on implementing this requirement in NSIR/DPR-ISG-01, Interim Staff Guidance, Emergency Planning/or Nuclear Power Plants. All emergency classification assessments should be based upon valid indications, reports or conditions. A valid indication, report, or condition, is one that has been verified through appropriate means such that there* is no doubt regarding the indicator's operability, the condition's existence, or the report's accuracy. For example, validation could be accomplished through an instrument channel check, response on related or redundant indicators, or direct observation by plant personnel. The validation of indications should be completed in a manner that supports timely emergency declaration. For ICs and EALs that have a stipulated time duration (e.g., 15 minutes, 30 minutes, etc.), the Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time. If an ongoing radiological release is detected and the release start time is unknown, it should be assumed that the release duration specified in the IC/EAL has been exceeded, absent data to the contrary. A planned work activity that results in an expected event or condition which meets or exceeds an EAL does not warrant an emergency declaration provided that 1) the activity proceeds as planned and 2) the plant remains within the limits imposed by the operating license. Such activities include planned work to test, manipulate, repair, maintain or modify a system or component. In these cases, the controls associated with the planning, preparation and execution of the work will ensure that compliance is maintained with all aspects of the operating license provided that the activity proceeds and concludes as expected. Events or conditions of this type may be subject to the reporting requirements of 10 f-CFR 50.72. The assessment of some EALs is based on the results of analyses that are necessary to ascertain whether a specific EAL threshold has been exceeded (e.g., dose assessments, 29

NEI 99 QI (Revision 6) November 2Ql2 chemistry sampling, RCS leak rate calculation, etc.); the EAL and/or the associated basis discussion will identify the necessary analysis. In these cases, the 15-minute declaration period starts with the availability of the analysis results that show the threshold to be exceeded (i.e., this is the time that the EAL information is first available). The NRC expects licensees to establish the capability to initiate and complete EAL-related analyses within a reasonable period of time (e.g., maintain the necessary expertise on-shift). While the EALs have been developed to address a full spectrum of possible events and conditions which may warrant emergency classification, a provision for classification based on operator/management experience and judgment is still necessary. The }ffil 99 01-This scheme provides the Emergency Director with the ability to classify events and conditions based upon judgment using EALs that are consistent with the Emergency Classification Level (ECL) definitions (refer to Category H). The Emergency Director will need to determine if the effects or consequences of the event or condition reasonably meet or exceed a particular ECL defmition. A similar provision is incorporated into the Fission Product Barrier Tables; judgment may be used to determine the status of a fission product barrier. 5.2 30

NEI 99 Q1 (RevisieH e) Nevember 2Ql2 CLASSIFICATION METHODOLOGY To make an emergency classification, the user will compare an event or condition (i.e., the relevant plant indications and reports) to an EAL(s) and determine if the EAL has been met or exceeded. The evaluation of an EAL(s) must be consistent with the related Operating Mode Applicability and Notes. If an EAL has been met or exceeded, then the IC is considered met and the associated ECL is declared in accordance with plant procedures. When assessing an EAL that specifies a time duration for the off-normal condition, the "clock" for the EAL time duration runs concurrently with the emergency classification process "clock." For a full discussion of this timing requirement, refer to NSIR/DPR=- ISG-01. 5.3 CLASSIFICATION OF MULTIPLE EVENTS AND CONDITIONS When multiple emergency events or conditions are present, the user will identify all met or exceeded EALs. The highest applicable ECL identified during this review is declared. For mcample: If aa Alert EAL aecl a Site Area EmergeeeJ' EAL are n'let, y;heth.er at oBe UH.it or at two cliffereet 1:u1'its, a Site Area Emergeeey sho1:1lcl ae cleelared.

     +Additionally, there is no "additive" effect from multiple EALs meeting the same ECL.

For example: If two Alert EALs are met.,, whether at oee UH.it or at two cliffereE:t ueits, an Alert should be declared. Related guidance concerning classification of rapidly escalating events or conditions is provided in Regulatory Issue Summary (RIS) 2007-02, Clarification ofNRC Guidance for Emergency Notifications During Quickly Changing Events. 31

NEI 99 QI (Rsvisien 6)

                                                                                   ~Tewimbsr 2QI2 5 .4 CONSIDERATION OF MODE CHANGES DURING CLASSIFICATION The mode in effect at the time that an event or condition occurred, and prior to any plant or operator response, is the mode that determines whether or not an IC is applicable. If an event or condition occurs, and results in a mode change before the emergency is declared, the emergency classification level is still based on the mode that existed at the time that the event or condition was initiated (and Iiot when it was declared). Once a different mode is reached, any new event or condition, not related to the original event or condition, requiring emergency classification should be evaluated against the ICs and EALs applicable to the operating mode at the time of the new event or condition.

For events that occur in Cold Shutdown or Refueling, escalation is via EALs that are applicable in the Cold Shutdown or Refueling modes, even if Hot Shutdown (or a higher mode) is entered during the subsequent plant response. In particular, the fission product barrier EALs are applicable only to events that initiate in the Hot Shutdown mode or higher. 32

NEI 99 g I (Re11isioa a) No,,rember 2912 5.5 CLASSIFICATION OF IMMINENT CONDITIONS Although EALs provide specific thresholds, the Emergency Director must remain alert to events or conditions that could lead to meeting or exceeding an EAL within a relatively short period of time (i.e., a change in the ECL is IMMINENT). If, in the judgment of the Emergency Director, meeting an EAL is IMMINENT, the emergency classification should be made as if the EAL has been met. While applicable to all emergency classification levels, this approach is particularly important at the higher emergency classification levels since it provides additional time for implementation of protective measures. 5.6 EMERGENCY CLASSIFICATION LEVEL UPGRADING AND DOWNGRADING An ECL may be downgraded when the event or condition that meets the highest IC and EAL no longer exists, and other site-specific downgrading requirements are met. If downgrading the ECL is deemed appropriate, the new ECL would then be based on a lower applicable IC(s) and EAL(s). The ECL may also simply be terminated. 33

NEI 99 Q1 (R-evisiea i) I / Nevemb& 2Q12 The following approach to downgrading or terminating an ECL is recommended. ECL Action When Condition No Longer Exists Unusual Event Terminate the emergency in accordance with plant procedures. Alert Downgrade or terminate the emergency in accordance with plant procedures. Site Area Emergency with no Downgrade or terminate the emergency in 1ong-term plant damage accordance with plant procedures. Site Area Emergency with Terminate the emergency and enter recovery in 1ong-term plant damage accordance with plant procedures. General Emergency Terminate the emergency and enter recovery in accordance with plant procedures. A s noted above, guidance concerning classification of rapidly escalating events or Conditions is provided in RIS 2007-02. I 5.7 I 34

NEI 99 QI (Revisiea i)

                                                                                    }fovember 2Q 12 CLASSIFICATION OF SHORT-LIVED EVENTS As discussed in Section 3.2, event-based ICs and EALs define a variety of specific occurrences that have potential or actual safety significance. By their nature, some of these events may be short-lived and, thus, over before the emergency classification assessment can be completed. If an event occurs that meets or exceeds an EAL, the associated ECL must be declared regardless of its continued presence at the time of declaration. Examples of such events include a failure of the reactor protection system to automatically scra~ the reactor followed by a successful manual s c r ~ or an earthquake.

5.8 CLASSIFICATION OF TRANSIENT CONDITIONS Many of the ICs and/or EALs contained in this document employ time-based criteria. These criteria will require that the IC/EAL conditions be present for a defined period of time before an emergency declaration is warranted. In cases where no time-based criterion is specified, it is recognized that some transient conditions may cause an EAL to be met for a brief period of time (e.g., a few seconds to a few minutes). The following guidance should be applied to the classification of these conditions. EAL momentarily met during expected plant response - In instances where an EAL is briefly met during an expected (normal) plant response, an emergency declaration is not warranted provided that associated systems and components are operating as expected, and operator actions are performed in accordance with procedures. 35

NEI 99 QI (Revision 6) November 2Ql2 EAL momentarily met but the condition is corrected prior to an emergency declaration - If an operator takes prompt manual action to address a condition, and the action is successful in correcting the condition prior to the emergency declaration, then the applicable EAL is not considered met and the associated emergency declaration is not required. For illustrative purposes, consider the following example. An A TWS occurs and the auxiliary feedwater system fails to automatically start. Steam generator levels rapidly decrease and the plant enters an inadequate RCS heat removal condition (a potential loss of both the fuel clad and RCS barriers). If an operator manually starts the auxiliary feedwater system in accordance with an EOP step and clears the inadequate RCS heat removal condition prior to an emergency declaration, then the classification should be based on the ATWS only. It is important to stress that the 15-minute emergency classification assessment period is not a "grace period" during which a classification may be delayed to allow the performance of a corrective action that would obviate the need to classify the event; emergency classification assessments must be deliberate and timely, with no undue delays. The provision discussed above addresses only those rapidly evolving situations where an operator is able to take a successful corrective action prior to the Emergency Director completing the review and steps necessary to make the emergency declaration. This provision is included to ensure that any public protective actions resulting from the emergency classification are truly warranted by the plant conditions. 5.9 AFTER-THE-FACT DISCOVERY OF AN EMERGENCY EVENT OR CONDITION In some cases, an EAL may,be met but the emergency classification was not made at the time of the event or condition. This situation can occur when personnel discover that an event or condition existed which met an EAL, but no emergency was declared, and the event or condition no longer exists at the time of discovery. This may be due to the event or condition not being recognized at the time or an error that was made in the emergency classification process-. In these cases, no emergency declaration is warranted; however, the guidance contained in NUREG-1022 is applicable. Specifically, the event should be reported to the NRC in accordance with 10 CFR ,§--50.72 within one hour of the discovery of the undeclared event or condition. The licensee should also notify appropriate State and local agencies in accordance with the agreed upon arrangements. 5.10 RETRACTION OF AN EMERGENCY DECLARATION Guidance on the retraction of an emergency declaration reported to the NRC is discussed in NUREG-1022. 36

NEI 99 01 (R,wisiofl 6) December ?Q 10 6 ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT ICS/EALS Tahle AR 1: ReeegHitiee Categer,r "AR" IRitiatiRg C0Bd-iti0e MatFix UNUSUAL SITE ! ..REA GENER,A..L ALERT EVENT EMERGENCY EM£RGENCY A.Ul.B!I! ,A,.,AlM! ASlBfil AGl.Rfil Release of Release of Release of Release of gaseous or liquid gaseous or liquid gaseous radioacfr,ity gaseous radioacti11ity radioactivity greater radioactivity resultieg in resulting in offsite dose resulting in offsite than 2 times the (site offsite dose greater than greater than 100 mrem dose greater than specific effluem 10 rnrern TEDE or 50 TEDE or 500 rnrem 1,000 ffirern TEDE release controlling mrem thyroid CDE. thyroid CDE. or 5,000 rnrem document)ODCM Op. AwGl-es: A !l Op. Awdcs: All thyroid CDE. limits for 60 minutes Op. },lodes: or longer. Ali Op. Med-es: All AUl;R!ll AA2Bf4 ASl.Bfil. AGl.RQ illl'.PLANNED Significant Speet fuel pool Spent fuel loss of v,rater level lov/eriBg of water level le,*el at (site specific pool level can:0ot be above irradiated fuel. above, or damage to, Level 3 description)iQ restored to at least Op. }.lodes: All irradiated fuel. ft g in (Level 3}. (site specific Level 3 Op. AwGl-es: A!! ~ description)40 ft g in (Level 3) for 60 minutes or longer.

                                                                                 ~
                                  ,AnA.JBfil Radiation levels that impede access to equipment necessary for normal plant operatioas, All cooldovm or shutdown.

Op. Medcs: AJl 37

NEI 99 Ql (R@\'isieR 6) Nevemeer 2Ql2 AU1RU1 ECL: Notification of Unusual Event Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the fstte-specific effluent release controlling document)ODAM limits for 60 minutes or longer. Operating Mode Applicability: All Emergency Action Levels: Example Emergeney Aeti0n LeYels: ( l or 2 or 3) Notes:

• The Emergency Director should declare the Unusual Eventevent promptly upon determining that the applicable time60 minutes has been exceeded, or will likely be exceeded.
• If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit60 n1.inlites.
• If the effluent flow past an eflluent monitor is known to have stopped due to actions to isolate the release path, then the eflluent monitor reading is no longer valid for classification purposes.

+RU 1.1 Reading on ANY Table R-1 effluent radiation monitor greater than column "NOUE" for 2-times the (site specific effllient release controlling document) limits for 60 minutes or longer: Monitor NOU E Reactor Building ventilation rad 8.0E-04 uci/cc monitor (Kaman 3L4, 5L6, 7L8) V, Turbine Building ventilation rad

, 8.0E-04 uciLcc 0 monitor (Kaman 1L2)

(I) V, 113 Offgas Stack rad monitor (.9 2.0E-01 uci/cc (Kaman 9L10l LLRPSF rad monitor 1.2E-03 uci/cc (Kaman 12) GSW rad monitor 1.5E+03 cps (RIS-4767) 1 RHRSW & ESW rad monitor (RM -1997} RHRSW & ESW Rupture Disc rad B.4E+02 cps 1.0E+03 cps monitor (RM-4268) RUl .22- (site specific 1Ho0itor list aed threshold vallies corresponding to 2 tim:es the coetrolling documeet limits) Reading on AN¥ANY effluent radiation monitor greater than 2 times the alarm setpoint established by a current radioactivity discharge permit for 60 minutes or longer. 38

NEI 99 <H (Revisise G) Nsvember 2012 RUI.3 Sample analysis for a gaseous or liquid release indicates a concentration or release rate greater than 2 times the (site specific effluent release controlling document)ODAM limits for 60 minutes or longer. 39

l'ffiI 99 01 (Revisioa 6) Noveni.ber 2012 Definitions: Basis: This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared. Naelear pewer plafltsDAEC incorporates design features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. Classification based on effluent monitor readings assumes that a release path to the environment is established. _If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes. Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL. EAL RUl :1 - This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways. EAL RUI.2 - This EAL addresses radioactivity releases that cause effluent radiation monitor readings to exceed 2 times the limit established by a radioactivity discharge permit._ This EAL will typically be associated with planned batch releases from non-continuous release pathways (e.g., radwaste, waste gas). EAL RUl.3 -This EAL addresses uncontrolled gaseous or liquid releases that are detected by sample analyseis or environmental surveys, particularly on unmonitored pathways (e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.). Escalation of the emergency classification level would be via IC AA+RAl. Developer Notes: The "site specific efflueH:t release coH:trolling document" is the Radiological Effluent TeomHcal SpecifieatioH:s (RETS) or, for plaH:ts that have ffi'ljllellieH:ted GeH:eric Letter g9 01 \-the

+lmplemenlefien efPre-grammalie Centre.ls fer Radiele-gieal Ejfh1e19f Teehnieal Speeifleatie11s in the Ad111inislf*ative Cent-re ls Seefien efthe Teehnieal Speeiflealie11s and the Releeatien efPreeedura! Det-ai!s e>fRETS te the Ojfsite Dese Cakulatien A/anual er te #ie Preees-s Centre! Pt*eg,*lf11 40

NEI 99 01 (Revisioa 6) Novernbsr 2012 AU2RU2 ECL: Notification of Unusual Event Initiating Condition: UNPLANNED loss of water level above irradiated fuel. Operating Mode Applicability: All ExalH.f)le Emergency Action Levels: -l-RU2.l a. UNPLANNED water level drop in the REFUELING PATHWAY as indicated by ANY of the following:

• Report to control room (visual observation)
• Fuel pool level indication (LI-3413) Lm;g THANless than 36 feet and lowering
• WR GEMAC Floodup indication (LI-4541) coming on scale(site specific 107.rel iBdicatioBs).

AND

b. UNPLANNED rise in area radiation levels as indicated by ANY of the following radiation monitors.

                      *                     (site specific list of area radiatioB monitors) Spent Fuel Pool Area, RI-9178 e   North Refuel Floor, RI-9163 11  New Fuel Vault Area, RI-9153

• South Refuel Floor, RI-9164
• NW Drywell Area Hi Range Rad Monitor, RIM-9184A
• South Drywell Area Hi Range Rad Monitor, RIM-9184B Definitions:

UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. REFUELING PATHWAY: The reactor refueling cavity, spent fuel pool and fuel transfer canal. 43

NEI 99 Q1 (Revisiea G) Nevember 2012 Basis: This IC addresses a decrease in water level above irradiated fuel sufficient to cause elevated radiation levels. This condition could be a precursor to a more serious event and is also indicative of a minor loss in the ability to control radiation levels within the plant. It is therefore a potential degradation in the level of safety of the plant. A water level decrease will be primarily determined by indications from available level instrumentation. _Other sources of level indications may include reports from plant personnel (e.g., from a refueling crew) or video camera observations (if available). A significant drop in the water level may also cause an increase in the radiation levels of adjacent areas that can be detected by monitors in thqse locations. 44

l'IBI 99 Ql (Revisiea 6) l'foveme er 20 12 The effects of planned evolutions should be considered. For example, a refueling bridge area radiation monitor reading may increase due to planned evolutions such as lifting of the reactor vessel head or movement of a fuel assembly. Note that this EAL is applicable only in cases where the elevated reading is due to an UNPLANNED loss of water level. During preparation for reactor cavity flood up prior to entry into refuel mode, reactor vessel level instmment 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 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 instmrnent 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 when moving irradiated fuel into the secondary containment. 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 l 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 alam1. To eliminate minor level perturbations from concern, DAEC uses LI-3413 *indicated water level below 36 feet and lowering. Increased radiation levels can be detected by the local area radiation moRitors surrounding the spent ( fuel pool and refueling cavity areas. Applicable area radiation monitors are those listed in AOP 981. A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes. Escalation of the emergency classification level would be via IC AMRA2. DevelopeF Notes: The "site speeifie level iasieatioas" are those iasieatioas that may be 1:tses to tHOBitor water level in the various portions of the REFUELING PATHWAY. Specif:,* the mode applicability ofa particular indieatioa if it is aot available ia all moses . The "site speeifie list of area radi:ation moaitors" sho1:tld eoataia those area radi:atioa moaitors that wouls be expeeted to have iaereases readings following a seerease ia v,ater level in the site speeifie REFUBLil>tG PATHWAY. In eases wh,ere a rasiatioa moaitor(s) is aot a railable or 1

      '<¥01:1ld aot provise a useful iadieatioa, eoasiseratioe sb.ouls be given to iael1:1sing alternate indieatioas s*..1eb. as Ul'WLANNED ehanges in tank and/or sump levels.

Developmeet of the BALs sb.oula coasider the availability aad limitatioas ofmose dependeat, or other eoatrolles but temporary, radiation monitors. Specify the mose applicability of a particular monitor if it is Bot available in all moses. EGL Assigmneat Attributes: 3 .1.1.A and 3. l . l .B 45

                                                                                          }ffil 99 Q l (Re*,.isiea 6)

N0Y0Efl00f 2012 AA1RA1 ECL: Alert Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE. Operating Mode Applicability: All Emergency Action Levels: Example Emergeney Aetien Levels: ( l or 2 or 3 or 4) Notes:

• The Emergency Director should declare the Alert event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
• If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limitl5 minutes.
• If the effluent flow past an effiuent monitor is known to have stopped due to actions to isolate the release path, then the effiuent monitor reading is no longer valid for classification purposes.
• The pre-calculated effluent monitor values presented in EAL l.) should only be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

RA 1. 1 Reading on ANY of the following Table R-1 effluent radiation monitors greater than the reading shova1column "Alert" for 15 minutes or longer: Monitor Reactor Building ventilation rad monitor 1.lE-02 uci/cc {Kaman 3/4, 5/6, 7/8) V) Turbine Building ventilation rad monitor

, 1.4E-02 uci/cc 0 {Kaman 1/2}

Q) V) ro Offgas Stack rad monitor

                 <..!J                                                        4.SE+Ol uci/cc

{Kaman 9/10) LLRPSF rad monitor 1.4E-02 uci/cc {Kaman 12} GSW rad monitor 1.7E+04 cps {RIS-4767}

                 ]     RHRSW & ESW rad monitor (RM-1997}

RHRSW & ESW Ru12ture Disc rad monitor l.2E+04 cps 1.8E+04 cps (RM-4268} 46

1'1BI 99 01 (Revisiea i) Nevembsr 2012 RAl.2 (site speeifie monitor list afld threshold ralties) 1 Dose assessment using actual meteorology indicates doses greater than 10 mrem TEDE _ _ _ _or 50 mrem thyroid CDE at or beyond (site speeifie dose reeeptor point)SITE BOUNDARY. [Preferred] RA 1.3 Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond fsite-speeifie dose reeeptor point)the SITE BOUNDARY for one hour of exposure. RaAl.4 Field survey results indicate EITHER of the following at or beyond (site speeifie dose reeeptor poiat)the SITE BOUNDARY:

• Closed window dose rates greater than 10 mR/hr expected to continue for 60 minutes or longer.
• Analyses of field survey samples indicate thyroid CDE greater than 50 mrem for one hour of inhalation.

47

1-ffiI 99 01 (Revisiea b) Nevembsr 2012 Definitions: SITE BOUNDARY: That line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee. Basis: This IC addresses a release of g~seous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).

• This IC is modified by a note that EAL RAI. I is only assessed for emergency classification until a qualified dose assessor is performing assessments using dose projection software incorporating actual meteorological data and current radiological conditions.

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. _The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at I% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the I :5 ratio of the EPA PAG for TEDE and thyroid CDE. Classification based on effluent monitor readings assumes that a release path to the environment is established. _If the effluent flow past an effluent monitor is known to have* stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.

 ---Escalation of the emergency classification level would be via IC AS+RS I.

DeveletJeF Netes: While this IC may not be met absent ehallenges to one or more fission product barriers, it provides classification diversity and may be 1:1sed to classify events that wo1:1Id not reach the same EGL based on plant status or too fission prodl:lCt matrix alone. For many of the DBAs analyzed in the Updated Final Safety Analysis Report, the discriminator will not be the munber of fission prod1:1ct barriers ehallenged, bl:1t rather the amo1:1nt of radioactivity released to the enviromnent. The EPA PA,Gs are ~qiressed iH tern3.s of the sl:lffi of the effecti,.ze dose eq1:1ivalent (EDE) and the committed effective dose equivalent (CEDE), or as the thyroid committed dose eq1:1ivaleHt (CDE). For the pl:lFpose of these IC/EA.Ls, the dose q1:1antity total effective dose eqmvalent (TEDE), as defined in 10 CFR § 20, is 1:1sed in lie1:1 of" ... sum of EDE and CEDE.... ". The EPA P}cG guidance provides for the :ese of adult thyroid dose conversion factors; hov,rever, some states have decided to base protective actions OH child thyroid CDE. Nuclear povter 48

.j f Vt 0

NEI 99 01 (Revision 6) November 2012 AA2RA2 ECL: Alert Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel. Operating Mode Applicability: All Emergency Action Levels: Bcample Emergency Action Levels: (l or 2 or 3) RA2.l Uncoveryofirradiated fuel in the REFUELING PATHWAY. RA2.2 Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by ANY of the following radiation monitors::Hi Rad alann for ANY of the following ARMs:

• Spent Fuel Pool Area, RI-9178
• North Refuel Floor, RI-9163
• New Fuel Vault Area, RI-9153
• South Refuel Floor, RI-9164 Reading greater than 5 R/hr on AN¥ANY of the following radiation monitors (in Mode 5 only):
• NW Drywell Area Hi Range Rad Monitor, RIM-9I84A
• South Drywell Area Hi Range Rad Monitor, RIM-9I84B RA2.3 (site specific listing of radiation monitors, and the associatedrnadings, setpoints aadlor alarms)

Lowering of spent fuel pool level to (site specific Level 2 value). [See Developer }Votes]25 .17 feet. Definitions: REFUELING PATHWAY - The reactor refueling cavity, spent fuel pool and fuel transfer canal. Basis: This IC addresses events that have caused IMMINENT or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool (see Developer }l-elBB). 51

                                                                                 "NEI 99 Q1 (Revisiea a)

N0 1;0mber 2Ql2 These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant. Expected radiation monitor alarm(s) during preplanned transfer of highly radioactive material through the affected areas are not considered valid alarms for the purpose of comparison to these EALs. 52

NEI 99 01 (Re1.rision 6) November 2012 This IC applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with IC E-HUl. Escalation of the emergency would be based on either Recognition Category A-R or C ICs. EALRA2.l This EAL escalates from :M:R--RU2 in that the loss oflevel, in the affected portion of the REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation (e.g., reports from personnel or camera images), as well as significant changes in water and radiation levels, or other plant parameters. Computational aids may also be used (e.g., a boil off curve). Classification of an event using this EAL should be based on the totality of available indications, reports, and observations. 53

NEI 99 0 l (Revision 6) November 2012 While an area radiation monitor could detect an increase in a dose rate due to a lowering of water level in some portion of the REFUELING PATHWAY, the reading may not be a reliable indication of whether or not the fuel is actually uncovered. To the degree possible, readings should be considered in combination with other available indications of inventory loss. A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes. EALRA2.2 This EAL addresses a release of radioactive material caused by mechanical damage to irradiated fuel. Damaging events may include the dropping, bumping or binding of an assembly, or dropping a heavy load onto an assembly. ,t<._ rise in readingsAn alarm on these radiation monitors should be considered in conjunction with in-plant reports or observations of a potential fuel damaging event (e.g., a fuel handling accident). Threshold values for the Drywell monitors are only applicable in Mode 5 since the calculated radiation levels from damage to irradiated fuel would be masked by the typical background levels on these monitors during plant operation, and mechanical damage to a fuel assembly in the vessel can only happen with the reactor head removed. EALRA2.3 Spent fuel pool water level at this value is within the lower end of the level range necessary to prevent significant dose consequences from direct gamma radiation to personnel performing operations in the vicinity of the spent fuel pool. This condition reflects a significant loss of spent fuel pool water inventory and thus it is also a precursor"to a loss of the ability to adequately cool the irradiated fuel assembles stored in the pool. Escalation of the emergency classification level would be via ICs AS-I-RS 1 or ~RS2(see AS2 De1*eleper Aretes). Develeper Notes: ForEAL#l Depending 1:1pon the ava:ilability and range of mstrumeatation, this EAL may inel1:1de speeifie readm.gs mdieative of fuel uneovery; eonsider water and radiation level readmgs. Speeify the mode applicability of a particular indication ifit is not ava:ilable in all modes. For EAL #2 The "site speeific listing of radiation monitors, and the associated readings, setpeiats 8:Fld/or alarms" should eontain those radiation monitors that col:lld be 1:1Sed to identify damage to 8:Fl irradiated fuel assembly (e.g., confirmatory of a release of fission prodl:lct gases from irradiated melt-For EALs #1 and #2 De11elopers shol:lld research radiation monitor design documents or other infonnation sources to ensure that I) the EAL value being considered is .vithin the usable response and display 1 54

NEI 99 Ql (Rsvisioa 6) November 2Ql2 AA3RA3 ECL: Alert Initiating Condition: Radiation levels that impede access to equipment areas necessary for normal plant operations, eooldown or shtitdown. Operating Mode Applicability: All

 ---Emergencv Action Levels:

Exemple EmeFgeHey /*... etieH Levels: (1 or 2) Nate: If the equipment in the listed roorn or a-rea was akeady inoperable or out of serviee before the event occurred, then no emergency classification is warranted. RA3.l Dose rate greater than 15 mR/hr in ANY of the following areas:

• Control Room ARM-(RM-9162)
• Central Alarm Station (by survey)

(other site specific areas/rooms) 2 An illWL'\N}ffiD event results in radiation levels that prohibit or impede access to any of the following plant rooms or a-reas: (site SJ?eeific list of plant rooms or a-reas with entry related mode.applicability identified) Definitions: Basis: This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or impede personnel from performing actions necessary to maintain normal plant operation,* or to perform a normal plant cooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Director should consider the cause of the increased radiation levels and determine if another IC may be applicable. 56

                                                                                        })!BI 99 01 (Revisioa ~)

November 2012 For EAL 2, El:l'l Alert declaration is 1.varranted if eB:try mto the affected room/area is, or may be, procedurally required during the plant operating mode in effect at the time of the ele 1ated 1 radiation levels. The emergency classification is not contingent upon v,rhether entry is actually necessary at the time of the increased radiation le1;els. ,A..ccess shottld be considered as impeded if extraordiHary measlifes are necessary to facilitate entry of personnel into the affected room/area (e.g., installing temporary shielding, requiring use of non routine protective equipment, requesting an CJEtension in dose limits beyond normal adm.mistrative limits).

        ,'\n emergency declaration is not v1arranted if any of the followcing conditions apply.

The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required dttrin.g the operating mode in effect at the time of the elevated radiation levels). For mtample, the plant is in *Mode 1 v1hen the radiation H.'l:Crease occurs, and the procedures used for normal operation, cooldo1,vn a,ad shutdowH do not reqttire eHtry iHto the affected room until Mode 4. The increased radiation lev:els are a result of a planned activity that includes compensatory measrn*es 1.vhich address the temporary inaccessibility of a room or area (e.g., radiography, spent filter or resin transfur, etc.). The action for which room/area eHtry is required is of aH administrative or record keepiHg Hature (e.g., Hom1al rounds orroutine iHspections). The access control measlifes are of a conservative or precatttionary natlife, and would not actually pre11ent or impede a required action. Escalation of the emergency classification level would be via Recognition Category AR, C or F ICs. DevelepeF Netes: EAL#l The value of l 5mR/hr is derived from the GDC 19 value of 5 rem ifl: 30 days vnth adjustment for expected occupancy times. The "other site specific areas/roon1:s" should include any areas or rooms requiring continuotis occupancy to maintaiH aormal plant operatioH, or to perform a normal cooldown and shutdown. EAL#2 The "site specific list of plant rooms or areas vfith entry related mode applicability identified" should specify those rooms or areas that contain equipment which reqt1ire a manual./local actioa as specified ia operating procedures used for noffilal plant operation, cooldovm and shutdown. Do not include rooms or areas in vAuch actions ofa contingent or emergency natlife v1ould be performed. (e.g., an action to address a+1 off normal or emergency condition such as emergency repairs, corrective measures or emergency operations). ln addition, the list should specify the plant mode(s) during which entry .vould be required for each room or area. 1 57

NEI 99 01 (Re¥isioe 6) 1-fovember 2012 AS1RS1 ECL: Site Area Emergency Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE. Operating Mode Applicability: All Emergency Action Levels: Exemllle Emergeney Aetion Leyels: (1 or 2 or 3) Notes:

• The Emergency Director should declare the Site Area Emergeacyevent promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
• If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutesthe specified time limit.
• If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
• The pre-calculated effluent monitor values presented in EAL l.) should only be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

RS 1.1 Reading on ANY of the followiagTable R-1 effluent radiation monitors greater than column "SAE" the reading shown for 15 minutes or longer: Reaetor BHilcling 1,eatilation ma mot1itor (Komae 3/4, 5/6, 7/8 ) l.OE 01 ttCi/ee i T1:1reit1e Builaiflg ventilation raa J'flORiter (Kamafl l/2) l.OE 01 :-1Ci/ee

               ~    Qffgas Staek rad a1enitor (Kan'lan 9/ l 0)                              4.5E-t02 l:!Ci!ee LLRPSF      Faa 1'flonitor (KarnaH 12)                                   l. OE OI :-1Ci/ee Monitor Reactor Building ventilation rad monitor

{Kaman 3/4, 5/6, 7/8) V) Turbine Building ventilation rad monitor

l 0 {Kaman 1/2)

QJ V) Offgas Stack rad monitor l9 (Kaman 9/10) 59

0\ 0

NEI 99 Q1 (Revisiea (;) Nev@mblilf 2Q12 RS 1.--4(....siH,*te-spH-eRic>+ifi'H+;-c-'J,mFWO'Hil*it*O*'F-+h~*s+-taianFlfdrtHflr*'eA'Sn'flH'10fHldT"1'\AtaA-1l+1u-P.Jes'") 2 Dose assessment using actual meteorology indicates doses greater than 100 mrem TEDE or 500 mrem thyroid CDE at or beyond (site specific dese receptor peint)the SITE BOUNDARY. [Preferred] RSl.3 Field survey results indicate EITHER of the following at or beyond (site specific dose receptorpeint)the SITE BOUNDARY:

• Closed window dose rates greater than 100 mR/hr expected to continue for 60_-minutes or longer.
• Analyses of field survey samples indicate thyroid CDE greater than 500 mrem for one hour of inhalation.

61

NEI 99 Q1 (Revisioa a) Novembsr 201'.2 Definitions: SITE BOUNDARY: That line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee. Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public. This IC is modified by a note that EAL RS 1.1 is only assessed for emergency classification until a qualified dose assessor is perfonning assessments using dose projection software incorporating actual meteorological data and current radiological conditions. 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. Radiological effluent EALs are also included to provide a basis for classifying events and I. conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. Classification based on effluent monitor readings assumes that a release path to the environment is established._ If the effluent flow past an effluent monitor is known 'to have stopped due to actions to isolate the release path, then the effluent monitor reading is* no longer valid for classification purposes. If Kaman readings are not valid, field survey results may be utilized to assess this IC using EAL RS1.3. Escalation of the emergency classification level would be via IC AGlRG 1. Developer Notes: While this IC may not be met absent challenges to multiple fission product barriers, it provides classification diversity and may be used to classify events that would not reach the same EGL based on plaet status or the fissioe product matriJ( alone. For n1aey of the DBl.,s analyzed in the Updated Final Safety A.ualysis Report, the discriminator will not be the number of fission product ban:iers challenged, but rather the amount of radioactivity released to the enviromnent. The EPA PAGs are e1[pressed i:e. tenHs of the sum of the effective dose equi1,*alent (EDE) and the committed effective dose equivaleRt (CEDE), or as the thyi:oid eoffilmtted dose equivalent (CDE). For the purpose of these IC/EALs, the dose quantity total effective dose equivalent (TEDE), as defined in 10 CFR § 20, is used in lie..1 of" ... sum of EDE and CEDE .... ". The EPA PAG guidance provides for tl1e use of adult thyrnid dose cow;ersion factors; howe11er, some states have decided to base protective actions on child thyi:oid CDE. l'raclear power 63

l~ ,~ I

I' NEI 99 Q1 (Revisiea ~) Nw1embsr 2012 AS2RS2 [See Devek,per 1'.~les] ECL: Site Area Emergency Initiating Condition: Spent fuel pool level at (site speeifie Level 3 deseri,ption) 16.36 feet. Operating Mode Applicability: All Example Emergency Action Levels: RS2.l Lowering of spent fuel pool level to 16.36 feet.(site s13eeifie Level 3 va-lee). Definitions: Basis: This IC addresses a significant loss of spent fuel pool inventory control and makeup capability leading to IMMINENT fuel damage. This condition entails major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration. It is recognized that this IC would likely not be met until well after another Site Area Emergency IC was met; however, it is included to provide classification diversity. Escalation of the emergency classification level would be via IC AG+-RG- f or A:GQ.RG2. De-1elof)eF Notes: In aeeordanee *.vith the disffitssion in Seetion 1.4, }JP£ Order EA, 12 051, it is reeommended that this IC and EAL be implemented when the enhaneed spent fuel pool level instrumentation is available fur use. The "site speeifie Lcwel 3 value" is usually that spent fuel pool level where fuel remains eovered and aetions to implement make up *.vater addition should no longer be deferred. This site speeifie le11el is determiaed ia aeeordanee with NRG Order EA 12 051 and WEI 12 02, and applieable ewner's group guidanee. Develo13ers should modify the EAL and+or Basis seetion to reflect any site speeifie eonstraints or limitations asseeiated with the design or operation of instrumentation used to determiae the Level 3

  ¥alu&.-

EGL Assignment Attributes: 3.1.3.B 66

NEI 99 01 (Revisioa a)

                                                                                                                              ~+OY@EBe@f 2QJ?

ARG1 ECL: General Emergency - Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE. Operating Mode Applicability: All Emergency Action Levels: Example EmeFgeney A:etien Levels: (1 or 2 or 3) Notes:

• The Emergency Director should declare the General Ernergencyevent promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
• If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutesthe specified time limit.
• If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
• The pre-calculated effluent monitor values presented in EAL l.) should only be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

RGI. I Reading on A~Y of the follovlingTable R-1 effluent radiation monitors greater than the

                 ..]",.,..,,.... ..... 1.... ........
           * - - - - 1.a.5       ~     .. - ..... l'l' olumn "GE" for 15 minutes or longer:

Effluent MeniteF Glessifieetien '.fhFeshelds Monitol' GE R:eaetor Btii!Eliag ,'@fltilatioa raEI 1tto0iter fK:amaa 314, §la, +18) l.O&l-00 t-1Gi!ee j Tu*bine B~ttlEliag '<'@fltilatioa rad fllonitor (Ka111en 1,£2) Qf~as gtaek rnEI moaitor (Kamas 9,110)

l. OHOO \lCi/ee 4.§Ea;a03 :.:iGilee Monitor Reactor Building ventilation rad monitor l.lE+OO uci/cc (Kaman 3/4. 5/6. 7/8)

Turbine Building ventilation rad monitor l.4E+OO uci/cc (Kaman 1/2) 67

NEI 99 g l (ReYisi0B a) N0vea4er ?Ql2 RGl.2 Dose assessment using actual meteorology indicates doses greater than 1,000 mrem TEDE or 5,000 rnrem thyroid CDE at or beyond (site specific dose receptor point)the SITE BOUNDARY. [Preferred] RGl. 3 Field survey results indicate EITHER of the following at or beyond (site specific dose receptorpoint)the SITE BOUNDARY:

• Closed window dose rates greater than 1,000 mR/hr expected to continue for 60_

minutes or longer.

• Analyses of field survey samples indicate thyroid CDE greater than 5,000 rnrem for one hour of inhalation.

68

N1ll 99 Q1 (Revision 6) November 2012 Definitions: SITE BOUNDARY: That line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee. Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public. This IC is modified by a note that EAL RG [ 1 is only assessed for emergency classification until a qualified dose assessor is perfonning assessments using dose projection software incorporating actual meteorological data and current radiological conditions. 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. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological efiluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration ~fthe 1:5 ratio of the EPA PAG for TEDE and thyroid CD~. Classification based on effluent monitor readings assumes that a release path to the environment is established. _If the efiluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the efiluent monitor reading is no longer valid for classification purposes; If Kaman readings are not valid, field survey results may be utilized to assess this IC using EAL RG 1.3. Develaper Nates: The effla.eftt ICs/EALs a-re Hl:el-Hded to provide a basis for elassifyiHg eveats that eaHBot be readily classified 011 the basis ofplaat couditioHs alone. The iuel-Hsion of both types of ICs/E.ALs more fully addresses the spectrum of possible events and accidents. While this IC may not be met absent challenges to multiple fissioH product barriers, it provides classification diversity and may be 1:1sed*to classify eveftts that wo1:1ld not reach the same EGL based on plaat status or the fission product matrix alone. For maay of the DBAs analyzed in the Updated Final Safety Pd1alysis Report, the discriminator will not be the nlHllber of fission product earners challenged, eut rather the amount of radioactivity released to the environment. The EPA PA.Gs a-re e~(pressed IB tem1s of the sum of the effeetive dose equi:valeftt (EDE) and the co1lllllitted effective dose equivalent (CEDE), or as the thyroid committed dose equivaleftt (CDE). For the purpose of these IC/EA.Ls, the dose quantity total effective dose equivaleftt (TEDE), as defined in 10 CFR § 20, is used in lieu of" ... sum of EDE and CEDE.... ". 70

l>IBI 99 01 (Revisioa 6) Novemb0r 2012 AG2RG2 [See De*,eloper }Votes] ECL: General Emergency Initiating Condition: Spent fuel pool level cannot be restored to at least 16.36 feet:- (site specific Level 3 description) for 60_-minutes or longer. Operating Mode Applicability: All Exam13le Emergency Action Levels: Note: The Emergency Director should declare the Geaeral Emergency event promptly upon determining that the applicable time 60 millUtes has been exceeded, or will likely be exceeded. RG2. l Spent fuel pool level cannot be restored to at least 16.36 feet:- (site specific Level 3 val1::1e) for 60 minutes or longer. Definitions: Basis: This IG addresses a significant loss of spent fuel pool inventory control and makeup capability leading to a prolonged uncovery of spent fuel. This condition will lead to fuel damage and a radiological release to the environment. It is recognized that this IC would likely not be met until well after another General Emergency IC was met; however, it is included to provide classification diversity. Devel013er Nates: Ia accordaace v:ith the discussioa ia 8ectioa 1. 4, }m.C Order EA 12 051, it is recoffill'l.eaded that this IC aad E,'\L be implememed whea the eahaaced spOflt fuel pool le>1el iastnHH:eatatioa is available for use. The "site specific Level 3 value" is usually that speat fuel pool level *.vhere fuel remains covered aad aetioas to implement make up water additioa should ao longer be deferred. This site specific lcwel is determined ia accordance ,~vith NRG Order EA 12 051 and NEI 12 02, and applicable ovffl.er' s groap guidaace. Developers should modify the EAL aad/or Basis section to reflect aay site specific constraints or limitations associated .vith the design or operation of instrumematioa used to detennine the Level 3 1

 ¥al-ue-:

EGL A.ssigmn.ent A.ttribH.4:es: 3.1.4.C 73

NEI 99 QI (Revisie:H G) Ne:r,zember 2Q12 7 COLD SHUTDOWN/ REFUELING SYSTEM MALFUNCTION ICS/EALS Table C 1: Reeegnitiee Categeey "C" Ieitiatieg CeeElitiee 1\cfatFe UNUSUAL CENER,A. .L ALERT EVENT E"MERCENCY EM.ERG ENCY CUl UNPL<\}INED CAl Loss of CSl Loss of (reaetor CCl Loss of (reaetor loss of (reaetor (reaetor YesseVR.tS ¥essel/RC8 [PTJ'R] or vessel/RCS [PWR] or vessel/Rt8 [.PTJ'R] or [PWR] orRPV RPV [BWR]) RPV [BWR]) RPV [B ~]) inventory [BWR]) inYentory. inventory affeeting inventory affecting for 15 minates or Op. }.lodes: LJCeld eore deeay heat fuel elad .integrity with longer. Shutde,rn, Rcfuel,ing rerB.oval eapability. contaimnent Op. },lodes: Geld Op. }.lodes: LJCold challenged. 8huffl.ewn, Refueling§:.,_ Shutdewn, Refuel4ng Op. }Jedes: LJCeld

~                                                                                     ShuUltJJ,l'fl, Refeeling CUl Loss of all l:rat           Cf...2 Loss of all one AC p01.ver souree           offsite and all onsite to emergeney buses for          ACpov;erto 15 m.inutes or longer.         erB.ergeney buses for Op. },lodes: LJCol:d            15 minlttes or longer.

8h1:1ff!.em9,, Refueli11g, Op. },lodes: LJCold De.fueled Slmtdewn, Reji1eling, Defiwled cm rnwL1J-n,rnn C,A...J IE.ability to inerease in RCS maintain the plant in temperatm:e. eold SffiltdoYffl. Op. }.lodes: LJCeld Op. },1e6les: LJCold Shutdewn, Refueling Shutde.~*n, Rejiwling CU4 Loss of Vital DC po,.ver for 15 minutes or longer. Op. },1odes: LJCold Shutdewn, Refueling Clli Loss of all onsite or offaite communieations eapabili-ties. Op. 1'1edes: LJCeld 81.ta-tdov,'fl., Raji1eling, Defaeled

Ta-ele inteaeed for ase by  :

• E,'\L developers. 1 1

1IRe1US10R r

• lRleeRsee 1

I I d . not reqt11re

                                                                                                            . d. I 1oew.neRts     is                   1 L------------------~

74

I I I I I I I I I I I I I I I I I I I

--------.J

NEI 99 0 I (Rtwisiee ~) Nevember 2012 CU1 ECL: Notification of Unusual Event Initiating Condition: UNPLANNED loss of (reactor vessel/RCS [PWR] or RPV [BWR]) inventory for 15 minutes or longer. Operating Mode Applicability: Cold S:l:ratdown, Refuel~ Emergency Action Levels: Example EmeFgeuey Aetian Lenis: (1 or 2) Note: The Emergency Director should declare the UauSRal &,zefl-1:event promptly upon determining that the applicable time 15 minutes has been exceeded, or will likely be exceeded. CUI.I UNPLANNED loss ofreactor coolant results fin (reactor vessel/RO, [PWR] or RPV [BWR]) level less than a required lower limit for 15 minutes or longer. CUI.2 a. (Reactor vessel/RCS [PT¥R] or RPV_ [BWR]) leve~ cannot be monitored.

            --AND

_ _ _ _b. UNPLANNED level rise in Drywell/Reactor Building Equipment or Floor Drain sump, or Suppression Poolincrease in (site specific sump and/or tank) Suppression PooLor Dryv,rell and Reactor Building floor and eguipmefl-1: drain sump levels. Definitions: UNPLANNED: A paran1eter change or an event that is not I) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor (reactor vessel/RCS [PWR] er-RPV [B WR]) level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant. Refueling evolutions that decrease RCS water inventory are carefully planned and controlled. An UNPLANNED event that results in water level decreasing below a procedurally required limit warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the core covered. 76

NEI 99 Q1 (R-evisiea e) Nevsmber 2Q 12 EAL CUI .1 recognizes that the minimum required (reaetor vessel/RCS [PWcR] or RPV [B WR]) level can change several times during the course of a refueling outage as different plant configurations and system lineups are implemented. This EAL is met if the minimum level, specified for the current plant conditions, cannot be maintained for 15 minutes or longer. The minimum level is typically specified in the applicable-_operating procedure but may be specified in another controlling document. The 15-minute threshold duration allows sufficient time for prompt operator actions to restore and maintain the expected water level. This criterion excludes transient conditions causing a brief lowering of water level. 77

-.l 00

NEI 99 QI (Revisiea e) 1'Tevemaer 2Q 12 ---EAL CUl.2 addresses a condition where all means to determine (reaetor vessel/RCS [PWR] or RPV [BW'R]) level have been lost. In this eondition, operators may detennine that an inventory loss is oeeurring by observing ehanges in sump and/or tank levels. Sump and/or tame le>;*el ehanges must be evaluated against other potential sourees of ,vater flovl to ensure they are indieative ofleakage from the (reaetor vessel/Rt'.S [PWR] or RPV [BWR]). If all level indication were to be lost during a loss of RCS inventory event, the operators would need to determine that RSC inventory loss was occurring by observing sump and Suppression Pool 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 Suppression Pool water level may be indicative of valve misalignment or leakage in systems that discharge to the Torus. Sump and Suppression Pool 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. Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CAI or CA3. De¥elepeF Netes: EAL #1 k is reeognized that the minimum allo,vable reaetor vessel/RCS/RPV level may hav'e many values over the eourse ofa refueling outage. Developers should solicit input from licensed operators eoneerning the optimum vlOrding for this EAL statement. In partieular, determine if the generie wording is adequate to ensure aeeurate and tim.ely elassifieation, or if speeifie setpoiuts can be ineluded .vithout making the EAL statement unwieldy or potentially inconsistent 1 with aetions that may be taken during an. outage. If speeifie setpoints *are ineluded, these should be drav1H from applieable operating proeedures or other eoatrolliag documents. EAL #2.b Eater any "site specifie sump aRdlor tank" levels that eould be mlpected to inerease if there were a loss of inventory (i.e., the lost inventory .vould enter the listed sump or tank). 1 EGL ,A...ssignment A.ttributes: 3.1.1.A 79

NEI 99 QI (Revisioa G) Novembsr 2Ql2 CU2 ECL: Notification of Unusual Event Initiating Condition: Loss of all but one AC power source to emergency essential buses for 15 minutes or longer. Operating Mode Applicability: Cold Shutdovm, Refueling4, 5, Defueled Example Emergency Action Levels: Note: The Emergency Director should declare the Unusual Eventevent promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded. CU2.1 a. AC power capability to (site specific emergeacy bt1ses)IA3 and 1A4 buses is reduced to a single power source for 15 minutes or longer. AND

b. Any additional single power source failure will result in loss of all-ALL AC power to SAFETY SYSTEMS.

Definitions: SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related./'... system regt1ired for safe plaBt operation, cooling dowa the plant and/or placing it in the cold shutdovln condition, incmding the EGGS. Systems classified as safety related. Basis: This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety::-* related equipment. When in the cold shutdown, refueling, or defueled mode, this condition is not classified as an Alert because of the increased time available to restore another power source to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition is considered to be a potential degradation of the level of safety of the plant. An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.

• A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).
• A loss Qf all offsite power and loss of all emergency power sot1rees (e.g., onsite diesel generators) .vith a single train of emergency essential bt1ses being back fed from the l:lllit 1

main generator.

• A loss of emergency power sources (e.g., onsite diesel generators) with a single train of 80

NEI 99 01 (R-evisiea e) Neveffllier 2012 emergeHcy essential buses being ooclf-fed from an offsite power source. 81

NEI 99 QI (Revisiss (ci) Nw,rember 2912 Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power. The subsequent loss of the remaining single power source would escalate the event to an Alert in accordance with IC CA2. 82

C"l 00

NEI 99 QI (RsvisieR a) Nevembsr 2Ql2 CU3 ECL: Notification of Unusual Event Initiating Condition: UNPLANNED increase in RCS temperature. Operating Mode Applicability: Cold Shutdovm, RefuelingLl Emergency Action Levels: Example EmeFgeuey 1A... eti0u Le*;els: (1 or 2) Note: The Emergency Director should declare the UH.usual Eventevent promptly upon determining that the applicable time 15 miH:Utes has been exceeded, or will likely be exceeded. CU3.l UNPLANNED increase in RCS temperature to greater than (site specific TechH:ical 8pecificatioH: cold shutdo'.VH: temperature limit)2 l 2°F. CU3.2 Loss of ALL RCS temperature and (reactor vessel/RCS [PWR] or RPV [BWR]) level indication for 15 minutes or longer. Definitions: UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.* CONTAINMENT CLOSURE: Procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions. For DAEC, this is considered to be Secondary Containment as required by Technical Specifications. CONTl\,IN}.4fil>ff CLOSURE: The procedarally defi.H:ed eoH:di.tioH:s or actioH:s takea to secure eoH:tainmeH:t aad its associated structures, systems, and components as a fimctioaal barrier to fissioH: product release under shutdovm c011ditions. Basis: This IC addresses an UNPLANNED increase in RCS temperature above the Technical Specification cold shutdown temperature limit, or the inability to determine RCS temperature and level, represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Director should also refer to IC CA3. A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification. 85

NEI 99 Q1 (Rr,wisiea G) Nevember 2Ql2 EAL CU3. l involves a loss of decay heat removal capability, or an addition of heat to the RCS in excess of that which can currently be removed, such that reactor coolant temperature cannot be maintained below the cold shutdown temperature limit specified in Technical Specifications. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation. During an outage, the level in the reactor vessel will normally be maintained above the reactor vessel flange. Refueling evolutions that lower water level below the reactor vessel flange are carefully planned and controlled. A loss of forced decay heat removal at reduced inventory may result in a rapid increase in reactor coolant temperature depending on the time after shutdown. 86

NEI 99 QI (ReYisiea 6) Nevember 2Ql2 EAL CU3.2 reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor RCS conditions and operators would be unable to monitor key parameters necessary to assure core decay heat removal. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication. Escalation to Alert would be via IC CAI based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria. Develaper Nates: For EAL #l, enter the "site specific Technical Specification cold shutdovm temperature limit" 1.vliere indicated. Dr"T A

• A .L . 'l *1 1 A

ct:t:l nSSignment 1 itlnuUtes: :, . .1.n 87

NEI 99 91 (RsvisieR G) Nevember 2()12 CU4 ECL: Notification of Unusual Event Initiating Condition: Loss of Vital DC power for 15 minutes or longer. Operating Mode Applicability: Cold Shtttdoi.vn, Refueling1,,2 Example EmergeBeyEmergency Action Levels: Note: The Emergency Director should declare the Unusual E~vent promptly upon determining that the applicable time 15 minutes has been exceeded, or will likely be exceeded. CU4.l Indicated voltage is less than (site speeifie hus voltage value) I 05 VDC on BOTH Div 1 and Div 2 125 VDC busesrequired Vital DC buses for 15 minutes or longer. Definitions: SAFETY SYSTEM: A system required for safe plant operation. cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.A. syste1n required for safe plant operation, eooling dow,a the plant: frH:dlor plaeing it in the eold shutdown condition, including the EGGS. Systems classified as safety related. Basis: This IC addresses a loss of Vital DC power which compromises the ability to monitor and control operable SAFETY SYSTEMS when the plant is in the cold shutdown or refueling mode. In these modes, the core decay heat load has been significantly reduced, and coolant system temperatures and pressures are lower; these conditions increase the time available to restore a vital DC bus to service. Thus, this condition is considered to be a potential degradation of the level of safety of the plant. As used in this EAL, "required" means the Vital DC buses necessary to support operation of the in-service, or operable, train or trains of SAFETY SYSTEM equipment. For example, if Train A is out-of-service (inoperable) for scheduled outage maintenance work and Train B is in-service (operable), then a loss of Vital DC power affecting Train B would require the declaration of an Unusual Event. A loss of Vital DC power to Train A would not warrant an emergency classification. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Depending upon the event, escalation of the emergency classification level would be via IC CAI or CA3, or an IC in Recognition Category AR. DevelatJeF Netes: The "site speeifie b:as voltage val:ae" should be based on the minim:am be.s volta-ge neeessary for adequate operation of SAFETY SYSTEM equipment. This voltage value should incorporate a margin of at least 15 minutes of operation before the onset of inability to operate those loaes. This voltage is use.ally near th.e minimum *1oka-ge seleeted *..vhen battery siziag is perfom.1:ed. 88

°' 00

NEI 99 Q1 (RevisieH e) Nevember 2Ql2 cus ECL: Notification of Unusual Event Initiating Condition: Loss of all onsite or offsite communications capabilities. Operating Mode Applicability: Cold Shutdmvn, Refuelmg5, 64, 5, Defueled Emergency Action Levels: Example EmeFgeaeyEmeFgeney t ... etian LeYels: (1 or 2 or 3) CU5.l Loss of ALL of the following onsite communication methods: _*_(site specific list of commm1ications methods)Plant Operations Radio System

• In-Plant Phone System
• Plant Paging System (Gaitronics)

CU5.2 Loss of ALL of the following GRGoffsite response organization communications methods:

• DAEC All-Call phone
• All telephone lines (PBX and commercial)
• Cell Phones (including fixed cell phone system)
• Control Room fixed satellite phone system
• FTS Phone system (site specific list of comn11:1:nications methods)

CU5.3 Loss of ALL of the following NRC communications methods:

• FTS Phone system
• All telephone lines (PBX and commercial)
• Cell Phones (including fixed cell phone system)
• Control Room fixed satellite phone system

                *   (site specific list of communications methods)

Basis: This IC addresses a significant loss of on-site or offsite communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to G&Goffsite response organizations and the NRC. 90

NEI 99 g l (RevisieB. 6)

Nevemeer 2912 This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.).

91

NEI 99 QI (Revisiea 6) Nevemb@r 2Q 12 EAL CU5.1 addresses a total loss of the communications methods used in support of routine plant operations .. EAL CU5.2 addresses a total loss of the communications methods used to notify all GWoffsite response organizations of an emergency declaration. The offsite response organizations referred to here are the State of Iowa, Linn County, and Benton CountyThe OR-Os referred to here a-re (see Developer Notes). ---EAL CU5.3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration. De,:,rele~eF Netes: EAL #1 The "site specific list of communications methods" should include all communications methods used for routine plant communications (e.g., commercial or site telephones, page party systems, radios, etc.). This listing should include installed plant equipment and components, and Hot items O'Nfl.ed and maintaiHed by individuals. EAL #2 The "site specific list ofcommunicatioas methods" should iaclude all communications methods used to perfonn initial emergoocy aotifications to OROs as described in. the site Emergeacy Plan. The listiHg should include iastalled plant equipment and components, and not items ovt-B:ed and 1H&i.ntained by i-Bdividuals. ERample methods are riHg do'Nilfdedicated telephone lines, commercial telephoae lines, radios, satellite telephones and internet based communications techBology. In the Basis section, insert the site specific listiag of the OROs requiring notification of an emergency-*declaratioa from the Control Room in accordance with the' site Emergency Plan, and typically within 15 minutes. EAL #3 The "site specific list of cmnmunications methods" should include all conll11:UB:ications methods esed to perfonn iaitial emergeacy notifications to the NRG as described i-B the site Emergoocy Plan. The listing should inch1de installed plant equipment and components, and not items ov,'Hed and Il'laintained by i-Bdividuals. These methods are typically the dedicated Emergency Notificatioa System (ENS) telephone li-Be and commercial telephone lines. EGL i\ssigBll'leH.t Attributes: 3.1.1.C 92

NEI 99 QI (Revisies 0) Nevember 2Ql2 CA1 ECL: Alert Initiating Condition: Loss of (reactor vessel/RCS [P1¥R] or RPV [B1¥R]) inventory. Operating Mode Applicability: Cold Shutdown, Refueling1..2 Emergency Action Levels: Exam13le Eme.rgeneyEmergeney 1A... eti0n Levels: (1 or 2) Note: The Emergency Director should declare the Alei+event promptly upon determining that the applicable time 15_ minutes has been exceeded, or will likely be exceeded. CAI.I Loss of (reactor vessel/RCS [P1¥R] or RPV [B1¥R]) inventory as indicated by level less than (site specific level)l 19.5 inches.

a. (Reactor vessel/RCS [PWR] or RPV [B1¥R]) level cannot be monitored for 15 minutes or longer AND

b. UNPLANNED level rise in Drywell/Reactor Building Equipment or Floor Drain sump, or Suppression Pool UNPLANNED increase in (site specific sHmp and/or tank) levels due to a loss of (reactor vessel/RCS [P\VRJ or RPV [BV/RJ) inventory.

Definitions: UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier). This condition represents a potential substantial reduction in the level of plant safety. For EAL CA 1.1, a lowering of water level below (site specific level) 119 .5 inches indicates that operator actions have not been successful in restoring and maintaining (reactor vessel/R{:;S [PWR] or RPV [B14'R]) water level. The heat-up rate of the coolant will increase as the available water inventory is reduced. A continuing decrease in water level will lead to core uncovery. __/ Although related, EAL CAI .1 is concerned with the loss of RCS inventory and not the potential concurrent effects on systems needed for decay heat removal (e.g., loss of a Residual Heat Rem.oval suction point). An increase in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3. 93

NEI 99 QI (Revisi:ea a) Nevember 2Q 12 For EAL CA 1.2, the inability to monitor (reactor vessel/RCS [PWR] or RPV [B WR]) level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may detennine that an iBveB-1:ory loss is occurriBg by observiBg clianges iB surn.p ami/or tank levels. Sump aBd./or tank 19",rel chaBges Hi-list be evaluated agaiBst other poteHtial sources of water flo;v to eBsure they are indicative ofleakage from the (reactor vessel/RCS [PWR] or RPV [BWR]).the operators would need to detennine that ~RCS inventory loss was occurring by observing sump and Suppression Pool 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 Suppression Pool water level may be indicative of valve misalignment or leakage in systems that discharge to the Torus. Sump and Suppression Pool 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. 94

NEI 99 0 I (Revisiea '3) Nevel.'fl0er 2012 The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CS 1 If the (reaetor vessel/RCS [PJf'R] or RPV [BWR]) inventory level continues to lower, then escalation to Site Area Emergency would be via IC CS 1. Develeper Netes: For EAL #1 the "site specific level" should be based on either:

• [EWR] Lov,r Low EGGS aeteatiofl setpoint/Le:vel 2. This setpoint v.zas ehoseH: beeause it is a standard operationally significant setpoint at which some (typically high pressure EGGS) injection systems v.rould automatieally start and is a value significantly below the low RPV water level RPS actuation setpoint specified ifl IC CUI.
• [PWR] The minimum allmvable level that supports operation of normally used decay heat removal systems (e.g., Residual Heat Removal or Shutdovm Cooling). If multiple levels exist, specify each aloH:g with the appropriate mode or configuration dependency criteria.

For El*.L #2 The type and range of RCS level instramen-tation may vary duriflg an outage as the plant moves through various operating modes and refHoliBg ev.olutions, partieularly for a PWR. i\s appropriate to tho plant design, alternate means of determining RCS le*1el are installed to assure that the ability to monitor level \.vithin the range required by operating ** proeedures v,ri.11 not be iF1tem1pted. The instrumentation rafl:go neeossary to support implemootatioB of operating proeedures Hl the Cold Shlttdovlf.l and RefHeling modes may be differoflt (e.g., narrower) than that required durin.g modes higher than Cold Shlttdown. Enter any "site speeific sump and/or tank" levels that could be rncpected to increase if there v1ere a loss of iH:Veatory (i.e., the lost iH:Ve11-tory would oater the listed sum,p or tank). ** EGL A.ssignmeflt }Atriblttes: 3 .1.2.B 95

                                                                                    }I.EI 99 01 (Revisioa G)

November 2012 CA2 ECL: Alert Initiating Condition: Loss of all offsite and all onsite AC power to emergency essential buses for 15 minutes or longer. Operating Mode Applicability: Cold Shutdown, RefuelHJ:g1_,_j_, Defueled Emergency Action Levels: Example EmeFgeneyEmeFgeney ,A,._etian Levels: Note: The Emergency Director should declare the Alert event promptly upon determining that the applicable time 15_ miautes has been exceeded, or will likely be exceeded. CA2.l Loss of ALL offsite and ALL onsite AC Power to (site specific emergency buses) 1A3 and 1A4 buses for 15 minutes or longer. Definitions: SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.A system required for safe plant operation, cooliag dovm th.e plarn aad/or placing it in the cold shutdown condition, including the ECCS. Systems classified as safety related. Basis: This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink.

 ---When in the cold shutdown, refueling, or defueled mode, this condition is not classified as a Site Area Emergency because of the increased time available to restore an emergency bus to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition represents an actual or potential substantial degradation of the level of safety of the plant.

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

 ---Escalation of the emergency classification level would be via IC CS 1 or AS+RS 1.

DevelapeF Nates: For a pov;er sm.a=ce that has 11:1:altiple generators, the EAL and/or Basis section sh01:1ld reflect the minimum number of operating generators necessary for that source to provide adequate power to an ,\.C emergency bus. For mcample, if a backup pov;er source is comprised oftv;:o generators 96

NEI 99 01 (Rovisioa 6) November 2012 CA3 ECL: Alert Initiating Condition: Inability to maintain the plant in cold shutdown. Operating Mode Applicability: Cold Shutdovm, Refueling1.2 Emergency Action Levels: Exam13le EmeFgeney Aetion Levels: (1 or 2) Note: The Emergency Director should declare the 1AJert event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded. CA3.l UNPLANNED increase in RCS temperature to greater than (site specific Technical Specification cold shutdo*.vn temperature limit)212°F for greater than the duration specified in the followmg tableTable C-2. Table C-2t RCS Heat-up Duration Thresholds CONTAINMENT CLOSURE RCS Stateslntegritv Heat-up Duration Status Intact (am Bet at reaHced Not applicable 60 minutes* iB:¥efitOfj' [-PWR]) Not intact (er at reduced Established 20 minutes* iw1efitefj* [PWRB Not Established 0 minutes

• _If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable.

UNPLANNED RCS pressure increase greater than (site specific pressure reading) 10 psig due to a loss of RCS cooling.. (This EAL does Bet apply durmg 1Nater solid plaBt cenditieBs. [PJ¥R]) Definitions: UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. CONTAINMENT CLOSURE: Procedurally defined actions taken to secure containment and its associated structures, systems, and components as a ftmctional barrier to fission product release under existing plant conditions. For DAEC, this is considered to be Secondary Contaimnent as required by Technical Specifications. CONTAINMENT CLOSURE: The procedurally defined conditions or actions taken to secure ceBtaiBmefit a:B:d its associated stractures, s11steu:is, a:B:d cempenents as a functional aarrier to fission product release under shutdown conditions. 98

NEI 99 01 (R~wisioa 6) November 2012 Basis: This IC addresses conditions involving a loss of decay heat removal capability or an addition of heat to the RCS in excess of that which can currently be removed. Either condition represents an actual or potential substantial degradation of the level of safety of the plant. A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification. RCS integrity is intact when the RCS pressure boundary is in its nonnal condition for the cold shutdown mode of operation (e.g., no freeze seals or nozzle dams). 99

NEI 99 01 (R-svisien 6) Navsmb@r 2012 The RCS Heat-up Duration Thresholds table addresses an increase in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact, or RCS inventory is reduced (e.g., mid loop operation in PWRs) ..:. The 20-minute criterion was included to allow time for operator action to address the temperature increase. The RCS Heat-up Duration Thresholds table also addresses an increase in RCS temperature with the RCS intact. The status of CONTAINMENT CLOSURE is not crucial in this condition since the intact RCS is providing a high pressure barrier to a fission product release. The 60-minute time frame should allow sufficient time to address the temperature increase without a substantial degradation in plant safety. Finally, in the case where there is an increase in RCS temperature, the RCS is not intact or is at reduced inventory [PWR], and CONTAINMENT CLOSURE is not established, no heat-up duration is allowed (i.e., 0 minutes). This is because I) the evaporated reactor coolant may be released directly into the Containment atmosphere and subsequently to the environment, and

2) there is reduced reactor coolant inventory above the top of irradiated fuel.

EAL CA3.2 provides a pressure-based indication of RCS heat-up. Escalation of the emergency classification level would be via IC CS I or A&l-RS 1. Developer Notes: For El..L #1 Enter the "site specific Technical Specification cold shutdown temperature limit" \Vfl:ere indicated. The R{:;S should be considered intact or not i-R-tact in accordance with site specific criteria. For EAL #2 The "site specific pressure reading" should be the lov,est change in pressure that can be accurately determined using installed instrumentation, but not less than 10 psig. For PWRs, this IC and its associated EALs address the concerns raised by Generic Letter gg 17, Les.r; ofDecay Helt Remm'£ll. A. number of phenomena such as pressurization, vortm1.ing, steam generator U tube draining, RCS 1C¥el differences vken operating at a mid loop condition, decay heat removal system design, aed level instram.entation problems can lead to conditions :r.,vhere decay heat removal is lost and core uncovery can occur. NRG analyses show that there are sequences that can cause core uncovery in 15 to 20 minutes, and severe core damage within an hour after decay heat removal is lost. The allowed time frames are consistent *.vith the guidance provided liy Generic Letter gg 17 and believ,ed to be conservative given that a low presstHe Containment barrier to fission product release is established. EGL i\.ssigmnent ,A.1.ttributes: 3.1.2.B 100

NEI 99 01 (R-evisioa 6) November 2012 CA6 ECL: Alert Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode. Operating Mode Applicability: Cold Sh.utdo .vn, Refueling4, 5 1 Emergency Action Levels: Examf)le EmeFgeney f ... etisn Le,*els: .:... Notes:

• If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then this emergency classification is not warranted.

e -If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted. CA6.1 a. The occurrence of ANY of the Table C-3 hazardous events:The occurrence of ANY of the follo Ning haza£dous events: 1

• Seismic event (earthquake)
• Internal or external flooding event

                   ,

• High winds or tornado strike
• FIRE
• EXPLOSION
• Other events with similar hazard characteristics as determined by the Shift Manager or Emergency Director
• Seismic cwent (earthquake)
• futemal or mcternal floodiag e'lent
• High v1inds or tornado strike
• FIRE
• EXPWSION (site specific haza£ds)River level above 757 feet
• River 'Nater Supply (R~V8) pit Io,.v level alarm
• Other eveBts with simila£ hazard cha£acteristics as determined by the Shift

                        *Man:ager or Emergency Director
         ---AND

b. EITHER of the follmving:

101

                                                            ~!EI 99 01 (R-evisiea a)

Nevember 2012

---

1. Event damage has caused indications of degraded performance in-at;

     !efl5f one train of a SAFETY SYSTEM needed for the current operating mode.

2. 2EITHER of the following:.

• Event damage has caused indications of degraded performance to a second train of the SAFETY SYSTEM needed for the current operating mode,-BF,

_e_The event has caused resulted in VISIBLE DAMAGE to the second train of a SAFETY SYSTEM compoHe:et or structure needed for the current operating mode.::;: 102

                                                                                 }!El 99 01 (R-evision 6)

November 2012 Definitions: FIRE: Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed. EXPLOSION: A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction, or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events may require a post-event inspection to determine if the attributes of an explosion are present. SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.A system reguirnd for safe plant operation. cooling dov,rn the plant and/or placing it in the cold shutdown condition, including the EGGS. Systems classified as safety related. VISIBLE DAMAGE: Damage to a component or structure that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure. Damage resulting from an equipment failure and limited to the failed component (i.e., the failure did not cause damage to a stmcture or any other equipment) is not VISIBLE DAMAGE. Basis: This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the current operating mode. In order to provide the appropriate context for consideration of an ALERT classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train, and there must be either indications of perfonnance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance, and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE such that the potential exists for performance issues. Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has indications of degraded performance for criteria CA6.l.b.1 of this EAL: commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events. An event affecting a single-train SAFETY SYSTEM (i.e .. there are indications of degraded perfonnance and/or VISIBLE DAMAGE affecting the one train) would not be classified under CA6 because the two-train impact criteria that underlie the EALs and Bases would not be met. If an event affects a single-train SAFETY SYSTEM, then the emergency classification should be made based on plant parameters/symptoms meeting the EALs for another IC. Depending upon the circumstances, classification may also occur based on Shift Manager/Emergency Director judgement. 103

NEI 99 01 (R~visieR a) Nevsrabsr 2012 Indications of degraded perfonnance addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. VISIBLE DAMAGE addresses damage to a SAFETY SYSTEM train that is not in service/operation and that potentially could cause performance issues. Operators will make this determination based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. This IC addresses a hazardous evOHt that causes damage to a SAFETY SYSTEM:, or a structrn=e coBtainiBg SAFETY SYSTEM compoBOHts, Heeded for the etHTeBt operatiBg mode. This coBditioB sigaificaB-tly redtlces the margin to a loss or potefltial loss of a fission prodtlct barrier, and therefore represents aR actual or potOHtial substaB-tial degradatiofl of the level of safety of the plant:- EAL l.b. l addresses damage to a SAFETY SYSTEM traiB that is iB service/operatioa siBce iRdicatioas for i-t vlill be readily available. The iadicatioas of degraded performaRce shornd be significaat eaotlgh to catlse coRcem regarding the operability or reliability of the S.A..FETY SYSTEM tram. EAL l.a.2 addresses daB1age to a SAFETY SYST~'1 compoaeB-t that is not iB service/operation or readily appareB-t throHg-R iBdications aloae, or to a structl:lre contaiBing SAFETY SYSTEM compoROHts. Operators *.vill make this determiBation based OB the totality of available e'.'ent aRd damage report informatiofl. This is iB-tended to be a brief assessmeB-t not reqtliring lOHgthy aRalysis or qHantificatioH: of the damage. Escalation of the emergency classification level would be via IC CS 1 or AS lRS 1. Develef)eF Netes: For (site specific hazards), developers shotild consider inck1ding other significant, site specific hazards to tho btilleted list contaiaed iH: EAL 1.a (e.g., a seicho). }foeloar po,ver plant SAFETY SYSTEMS are comprised of t\vo or more separate and redtindaat traiBs of eqtlipmOHt iB accorda:H:ce vtith site specific design criteria. EGL ,A..ssigameB-t Attribtltos: 3.1.2.B 104

NEI 99 01 (R~Yisie a 6) Nev@mber 1012 CS1 ECL: Site Area Emergency Initiating Condition: Loss of (reactor vessel/RCS [PWR] or RPV [B WR]) inventory affecting core decay heat removal capability. Operating Mode Applicability: Cold Sln:1tdowR, Refuelmg4 5 Emergency Action Levels: Examlile Eme1*geney Aetien Le11els: (1 or 2 or 3) Note: The Emergency Director should declare the Site l\rea En:i:ergeacyevent promptly upon determining that the applicable time 30 miRutes has been exceeded, or will likely be exceeded. CS1 .1 a. CONTAINMENT CLOSURE not established. AND

b. (Reactor vesseb'RCS [PWR] or RPV [BWR]) level LESS THA,Nless than fs-ite-specific level)+64 inches:.?.

CSl.2 a. CONTAINMENT CLOSURE established. AND

b. (Reactor vessel/RCS [PWR] or RPV [BU<R]) level LESS THANless than +,-ire-specific level).+ 15:.?. inches CSl. 3 a. (Reactor vessel/RCS [PWR] or RPV [BWR]) level cannot be monitored for 30 minutes or longer.

AND

b. Core uncovery is indicated by AN¥-EITHER of the following:

                 *   (Site~ cific radiation monitor) Drywell Monitor (9184A/B) reading greater than *

• 5.0 R/hr
• Erratic source raRge monitor iedica-tion. [PWR]
• UNPLANNED level rise in D1ywell/Reactor Building Equipment or Floor Drain sump, or Suppression Pool UNPLAN1'IBD increase in (site specific SHR'lp aadlor tank)_levels of sufficient magnitude to indicate core uncovery

                 *   (Other site specific indications)

Definitions: 105

11-fEI 99 Q 1 (R-eYisiee a) Nev@me@r 2012 CONTAINMENT CLOSURE: Procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions. For DAEC, this is considered to be Secondary Containment as required by Technical Specifications. UNPLANNED: A parameter change or an event that is not I) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. 106

NEI 99 01 (Revisioa 6) November 2012 Basis: This IC addresses a significant and prolonged loss of (reactor vesseYRCS [PWR] or RPV [BWR]) inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration. 107

NEI 99 01 (R~visioa G) Novemeer 2012 Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If R:G&/'.reactor vessel level cannot be restored, fuel damage is probable. Outage/shutdown contingency plans typically provide for re-establishing or verifying CONTAINMENT CLOSURE following a loss of heat removal or RCS inventory control functions. The difference in the specified R:G&/'.reactor vessel levels of EALs CS 1.1.b and CS 1.2.b reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment.- .. In the Cold Shutdown and Refueling Modes, LT/LI-4559, 4560, and 4561 (RX VESSEL NARROW RANGE LEVEL) instruments read up to 22" high due to hot calibrations. LI-4541 (WR GEMAC, FLOODUP) should be used in these Modes for comparison to EAL thresholds since it is calibrated cold and reads accurately. If normal means of RPV level indication are not available due to plant evolutions, 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. In EAL CSl.3.a, the 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring. The inability to monitor (reactor ¥essel/R.-C8 [PWR] or RPV [.B WR]) level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative ofleakage from the (reactor ¥essel/RC8 [PWR] or RPV [.Bff'R]). These EALs address concerns raised by Generic Letter 88-17, Loss ofDecay 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. ---Escalation of the emergency classification level would be via IC CG 1 or AG+RG 1. Developer Notes: Accident B:Halyses suggest that fu.el damage may occ:ar w,ithia oae hour ofl:lfl:co*1ery depending upon the amo..mt of time since shutdown; refer to Generic Letter 88 17, SECY 91 283, 1-JUREG 1449 andNUMARC 91 06. The type aad range of RCS level iustramentatioa may vary during an outage as the plant moves thro..i-gh *;arious operating modes and refueling evolutions, particularly for a P\VR.. A,s 108

0

                                                                                        }TEI 99 01 (R-evisioa G)

November 2012 CG1 ECL: General Emergency Initiating Condition: Loss of (reactor vessel/RCS [PWR] or RPV [BWR]) inventory affecting fuel clad integrity with containment challenged. Operating Mode Applicability: Cold Shutdown, Refuelingi,__i Emergency Action Levels: Example EmergeneyEmergeaey

  ,\etiaa Levels: (1 or 2)

Note: The Emergency Director should declare the General Emergencyevent promptly upon determining that the applicable time 30 minutes has been exceeded, or will likely be exceeded. CG 1.1 a. (Reactor vessel/RCS [PWR] or RPV [BWR]) level LESS Tl-lA1'.J:less than fs-ite-specific lcwel)+ 15 "inches for 30 minutes or longer. AND b12:. ANY indication from the Secondary Containment Challenge Table (see belov,0C-l- CGl.2

a. (Reactor vessel/RCS [P1VR] or RPV [B1VR]) level cannot be monitored for 30 minutes or longer.

                  -----AND

b. Core uncovery is indicated by EIHERAN¥ of the following:

• Drywell Monitor (9184A/B) (Site specific radiation m.onit:or) reading GREA,TER THAJ'fareater than (site specific valee)5.0 R/hr.
• Erratic soru=ce ra:ftge monitor iedication [PWR]
• UNPLANNED level rise in Drywell/Reactor Building Equipment or Floor Drain sump, or Suppression Pool UNPL'\.}J:NED m.crease in (site specific Sl:lffi:p and/or tank) le0vels of sufficient magnitude to indicate core uncovery.:.

                  *   (Other site specific indications)

AND 111

Jl,ffil 99 01 (R-evisioe G) November 2012

c. ANY indication from the Secondary Containment Challenge Table (see below C-H.

Table C-1 Secondary Containment Challenge Table

• CONTAINMENT CLOSURE not established*
• Drywell Hydrogen or Torus Hydrogen GREi..TER THANgreater than 6% AND Drywell Oxygen or Torus Oxygen GREATER THANgreater than 5%

(EJrplosive ffiHEhire) eJEists iHside eoatainment

• UNPLANNED increase in containment pressure
• Secondary containment radiation monitors above max safe operating limits (MSOL) of EOP 3, Table 6radiation monitor reading above (site speeifie value) [BW~]
• If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-_minute time limit, then declaration of a General Emergency is not required.

112

NEI 99 01 (R-evisiea 6) Nevembsr 2012 Definitions: CONTAINMENT CLOSURE: Procedurally defined actions taken to secure containment and its associated strnctures, systems, and components as a functional barrier to fission product release under existing plant conditions. For DAEC, this is considered to be Secondary Containment as required by Technical Specifications.Cc>>JT.zt\Il{MEtJT CLOSURE: The proeedmally defi.H:ed eonditions or aetions taken to seeme eontainment and its assoeiated struetures, systems, and eomponents as a functional barrier to fission product release under shutdown conditions. UNPLANNED: A parameter change or an event that is not 1). the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area. Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If ~reactor vessel level cannot be restored, fuel damage is probable. With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required. The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a challenge to Containment integrity. 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 gas mixture in containment. If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged. In EAL CG 1.2.~ the 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrnmentation uncertainties). It also allows sufficient time 113

1'ffil 99 (H (R-evisiee 6) Nevemeer 2012 for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring. For EAL CG 1.2.b, the calculated radiation level on the Drywell Monitors (9184A/B) is without the reactor head in place. Calculated in radiation levels with the reactor head in place are below the normal variation in background readings of these monitors. 114

NEI 99 01 (R.avisioa 6) Novsmb@r 2012 The inability to monitor (roaster vesse1'Rt:S [PWR] or RPV [lUf'R]) level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative ofleakage from the (reaetor vessel/RCS [PWR] or RPV [BWR]). For the Containment Challenge Table, Secondary Containment max safe operating (MSOL) limits from EOP 3 are defined as the highest parameter value at which neither: 0) equipment necessary for the safe shutdown of the plant will fail nor (2) personnel access necessary for the safe shutdown of the plant will be precluded.

 +These EALs address concerns raised by Generic Letter 88-17, Loss ofDecay 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.

DevelepeF Netes: l ..eeideRt analyses sttggest that fael datHage may oeeur within oae hou:r ofuaeovery depeRdiag upoR the amouat of time siRce shutdowR; refer to Gooeric Letter 88 17, SECY 91 283, }JUREG 1449 aRdNUMA.RC 91 06. The type aRd range of RCS level instrumentatioa may vary during an outage as the plaRt moves through various operating modes and refueling evolutions, partieularly for a PWR. As appropriate to the plant design, alternate means of determining RCS level arc installed to assure that the ability to monitor level '.Vithin the range required by operating procedures will not be imerrupted. The iastrumentation raage neeessary to support implemeatation of operating proeedures in the Cold Shutdo1,vn and Refueling modes may be different (e.g., narrov,rer) thaa that required during modes higher than Cold Shutdown. For El..L #1.a The "site specific level" should be apprOJEimately the top of active fuel. If the availal3ility of OH scale level iHdieatioH is sueh that this level ,,ralt1e ean be determiHed duriHg some shutdown modes or coHditions, but not others, then specify the mode depeHdem aHdlor configumtioH states dmiRg which the level indicatioH is applicable. If the desiga and operatioH of

• vmter level iRstrumeatatioR is such that this level value eannot be determined at any time during Cold Shutdo'NB: or Refueling modes, then do not inelude EAL #1 (elassificatioa will be aceomplished iR aceordanee with EAL #2).

For EAL #2.b first bullet l .. s v;ater level in the reactor vessel Io,.vers, the dose rate above the core will iacrease. Enter a "site specific radiation monitor" that could be used to detect core uneovery and the assoeiated "site speeifie ,,ralue" indieative of eore uneovery. It is reeognized that the condition described by this IC may result in a radiation value beyond the operatiHg or display range of the installed radiation moRitor. IR those cases, El..L values should be determiRed

 ;vith a margin sufficieHt to ensure that an accurate monitor reading is available. For example, an E,A..L monitor readiHg might be set at 90% to 95% of the highest acffi:H"ate monitor readiRg. This provision not.vithstanding, if the estimatedlealet1lated monitor readiHg is greater than 115

NEI 99 01 (R~visisa e) Ns1.'@mb@r 2012 8 INDEPENDENT SPENT FUEL STORAGE INSTALLATION {ISFSI) ICS/EALS Table E 1: Reeegeitiee Categ01y "E" Ieitiating CeeEJ.itiee }.{at.-ix E IIUl Damage to a loaded cask CONFINEMENT BOUNDARY. Op. },{edes: All

Taele inteaded for use by :

1 EAL de*r.:elopers. 1 I ff l . r

                                                              . H11eensee 1He1:1s10H              .          I 1

doeu113.eBts is not rec:tuired. :

L------------------1 117

ISFSI 1\llALFUNCTION E-HU1 ECL: Notification of Unusual Event Initiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARY. Operating Mode Applicability: All Example Emergency Action Levels: E-HUl.l Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on contactf! radiation reading greater than the values shown belewon Table E-1 (2 times the site specific cask specific technica:l specification allov,able radiation le11el) on the surlace of the-spent fuel cask.

                        '...           Table E-1 Cask Dose Rates 61BTDSC 3 feet from RSM Surface                     800 mrem/hr Outside RSM Door - Centerline of DSC                 200mrem/hr End Shield Wall Exterior                     40mrem/hr Definition:

CONFINEMENT BOUNDARY: The barrier(s) between spent fuel and the enviromnent once the spent fuel is processed for dry storage. Basis: This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of a storage cask containing spent fuel. It applies to irradiated fuel that is licensed for dry storage beginning at the point that the loaded storage cask is sealed. The issues of concern are the creation of a potential or actual release path to the environment, degradation of one or more fuel assemblies due to environmental factors, and configuration changes which could cause challenges in removing the cask or fuel from storage. The existence of "damage" is determined by radiological survey. The technical specification multiple of "2 times", which is also used in Recognition Category A-RIC RAUl, is used here to distinguish between non-emergency and emergency conditions._ The emphasis for this classification is the degradation in the level of safety of the spent fuel cask and not the magnitude of the associated dose or dose rate. It is recognized that in the case of extreme damage to a loaded cask, the fact that the "on-contact" dose rate limit is exceeded may be determined based on measurement of a dose rate at some distance from the cask. Security-related events for ISFSls are covered under ICs HUI and HAL 118

                                                                                                                                       ~!EI 99 01 (Revisioa e)

November 2012 9 FISSION PRODUCT BARRIER ICS/EALS Table 9 F 1: Recognition Category "F" Initiating Condition Matrix ALERT A.ny Loss or any Potential Loss of either the Fuel Clad or RCS barrier. Loss or Potential Loss of any two barriers. Loss of any two baffiers and Loss or Potential Loss of the third ba1Tier. See Table 9 F 2 feF IH¥R EALs See Table 9 F 3 feF P'WR EALs Dl¥.rel0f)eF Nate: The adjacent logic flow diagram. is KJF use by developers and is not reEfUired KJF site specific implen1entation; hov,rever, a site specific sc1'te1He m.ust i11clude some type of user aid to facilita-te tim.ely and accurate classification of fission product bmTier losses and/or pote11tial losses. Such aids are typically comprised oflogic flow diagrams, "scoring" criteria or checkbox type m.atrices. The user aid logic must be consistent with that of the adjacent diagram. 120

NEI 99 01 (Revisiea 6) Nevember 2012 POTENTIAL POTENTIAL LOSS LOSS LOSS LOSS FUEL CLAD

                                                                      -  YES     EGl. - Loss of ANY Two Barriers AM! Loss or Potential Loss of Third Barrier
                                                        ' - - - - - - - - - - - - NO----~

POTENTIAL . POTENTIAL POTENTIAL LOSS LOSS LOSS LOSS LOSS LOSS FUEL CLAD CONTAINMENT ffil. - Loss or Potential Loss of ANY Two Barriers POTENTIAL POTENTIAL LOSS LOSS LOSS LOSS FUEL CLAD RCS

                   ' - - - - - - - - - - - - - - - - - - - - - - a . i f A l - A N Y Loss or ANY Potential Loss ofEII.UE.B Fuel Clad ill!. RCS 121

                                                                                                                                      }+EI 99 01 (Re=vision a)
                                                                                                                                             }J011ember 2012 De¥elepeF Netes

1. The logie used for these initiating eonditions refleets the follchving eonsiderations:

• The Fael Clad Barrier and the RCS Barrier are ,veighted more heavily than the Containment Barrier.

Unusual Event ICs assoeiated with fissioH: product barriers are addrnssed in RecognitioH: Category S.

2. For accident eonditions involving a radiological release, evaluation of the fission product barrier thresholds *.vill need to be performed in conjUHction with dose assess111ents to ensure coffeet and timely escalation of the emergency classification. For example, an evaluation of the fission prodaet barrier thresholds ffiay result in a Site Area Effiergeney classification 1.vhile a dose assessment may indicate that an BAL for General .Bmergeney IC AGl has been fficceeded.

3. The fission product barrier thresholds speeified *.vithin a scheffl:e are expected to reflect plant specific design and opcrnting characteristics.

This may req,Hire that developers ereate different thresholds than those provided in the gene1ic guidance.

4. A.lternittive presentation metsods for the Recognition Cittegoey F ICs and fission produet barrier thresholds are aeeeptable and ineh1de flmv charts, block diagrams, and checklist type tables. Developers must ensure that the site specific method addresses all possible threshold -

eombinations and classification outcomes shov1n in the BWR or PWR EAL fission product barrier tables. The NRG staff eonsiders the presentatioe ffiethod of the Recognition Category F infoffilation to be an important user aid and may reqt:1est a change to a partieular proposed 111ethod if, among other reasons, the esaE:ge is neeessary to promote eonsisteney aeross the industry.

5. As used ia this Recognition Category, the term RCS leakage encompasses not just those types defined in Teehnical Speeifications but also includes the loss of RCS mass to any location inside containment, a secondary side system (i.e., PWR steam geHerator tube leakage), an interfaciag system, or outside of containment. The release of liquid or steam mass from the RCS due to the as designed/expected operation of a relief valve is not considered to be RCS leakage.

6. At the Site Area Effiergeney le1!'el, classifioation decision malEers should maintaiH: cognizance of sow far present eonditioHs are fron:1 ffieeting a threshold that 1,vould require a General Emergency deelaration. For e1cainple, if the Fuel Clad and.R{;S fission produet barriers were both lost, then there should be frequeH:t assessments of containment radioacti*te inventory and integrity. 1\.ltemittively, if both the Fuel Clad and RCS fission product baiTiers v1ere potentially lost, the Emergency Director would have more assurance that there was no immediate need to escalate to a General Emergency.

7. The ability to escalate to a higher emergency elassifieittion level in response to degrading conditions should be maintained. For example, a steady increase in RCS leakage would represent aH: increasing risk to publie health and safety.

122

NBI 99 01 (RevisieR a) Nevemeer 2012 Table 9-IZF-14: B'\¥R DAEC EAL Fission Product Barrier Table Thresholds for LOSS or POTENTIAL LOSS of Barriers FAlALERT FS1 SITE AREA EMERGENCY FGlGENERALEMERGENCY ANY Loss OR ANY Potential Loss of Loss &F-OR Potential Loss of ffitY"'.ANY two Loss of ANY two barriers AND Loss OR EITHER the Fuel Clad OR RCS ban*ie~~ barriers. ANX_Loss ar .&NXany Potential Loss of either the Fuel Clad ar OR RC8 barrier. ?'*:> ..*'; .* **'"" :. .. *-; ,..'* ... * *,. , .,.. . *,1::; .::, F" : :+i*. *: .>'.*, ..

                                                                                                       ;?;!RCS'll~r,fie(t *'. *:f:*' *::l:    \?T *.,.  :/: corit;J~mert'.t Barri~r. *. . :.'<;' . ..,:r:.*
,*.**~ ,.              Jftiel Clatl Barrier * . , : . .: :: . :                                 '*                                    ' '"~

LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS

1. RCS Activity 1. Primary Containment 1. Primary Containment Conditions PressureConditions A. Coolant activity Not Applicable A. Primary Not Applicable A. UNPLANNED A. Primary greater than 300 containment rapid drop in containmentTorus uCi/ g!TI dose pressure greater primary pressure greater eguivalent I- than (site specific containmentDrywe than (site specifie 131.A. (Site ¥£lffiej1.Qfilg due to ti pressure ~53psig specific iHdicatioH:s RCS leakage. following primat)' OR that reactor coolant coHtaiameHtDrywe B. Drywell or Torus aetivity is greater ll pressure rise H2 cannot be than 300 µCi/gm OR determined to be dose equivalent I B. Primary L88 +HANless l+. cott:tainmeHtDrywe than 6% and ti pressure Drywell m:OR response not Torus 02 cannot consistent with be determined to LOCA conditions. be less than 5%

OR (site s11eeifie C. UNISOLABLE ex11Iash'e direct downstream mixture) exists pathway to the iHside 11rimery environment exists eaHteiHmeHt after primary OR 123

P.rnI 99 01 (RevisieH 6)

P.fovember 2012 t ,. ,, -;J - .*... ,,. *... .,.

                                   . :: . ;   **'t.*                                   ,.        *,*            .

.... . . E)uf Clad Barrier . ':': . Containment;.J.Jarr;"ier * -

LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS containment C. HC+L {Graph 4 of isolation signal EOP 2) exceeded. OR D. Intentional primary containment venting per EOPs 124

1>1EI 99 01 (Revisioa a) Woveffll3er 2012 0 Fu~l'Cla d B~hief

                                                       *c . :*.
                                                                                  ,,,.* . . *. J',.
                                                                                  'RCS<*Bafiier '* * *
                                                                                                                          \ ,*           ;
                                                                                                                                                    ..      'ff Contain'menf Ba'rrief ** *..       ,"-* '

LOSS I POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS I POTENTIAL LOSS

2. RPV Water Level 2. RPV Water Level 2. RPV Water Level A. SAG enti:y is A. RPV water level A. RPV water level Not Applicable Not Applicable A. SAG enti*y is requiredPrimary cannot be restored cannot be restored requiredPrimary eeat:aiH:meat: and maintained and maintained cemaiHffiCB'1:

floodiag requif"ed. above Esite s13eei:§e above f&iEe- fleediag

                            &P:\Z: 'vrate:1: le'f'el               s13eei§e RP:\l                                                                                                 required.

eeffespoaaiH:g to 'Natel' level the top ef active coffespoaaiag to

                            .fHelj+ 15 inches OR                   fhe to13 of aetive cannot be                             .fHelj+ 15 inches determined.                            OR cannot be determined.

3. Not Applicable 3. RCS Leak Rate 3. Primary Containment Isolation Failure Not Applicable Not Applicable A. UNISOLABLE A. UNISOLABLE A. UNISOLABLE Not Applicable break in 1\NY of primary system Qrimary system the foHowiag: leakage that leakage that Esite sp*eeiHc ' ' results in results in systeffis with exceeding the exceeding the potootial for high Max Normal Max Safe eHergy liH:e 0Qerating Limit OQerating Limit breaks)Main (MNOL) of EOP (MSOL) of EOP Steam, HPCI, 3, Table 6 for 3, Table 6 for Feedwater, EITHER of the EITHER of the RWCU, orRCIC following: following:

as indicated by the

• 1. Max
• Tem12erature failure of both ~tefffl:al OR isolation valves in Operating o Radiation Level anvANY one line Temperature to close AND --A:

_OR UNI80 EITHER:

• 2. Ma1( bABbB aireet 9 HighMSL Normal downstream flow or steam r, - A .... ,...."'

                                                                                                          .....,,r- - -  -*u 125

NEI 99 01 (Revision 6)

                                                                                                                                                                         }>lsvemaer 2012

• • '}

        "*   .':Fuetci~a BatHe/ ' ,*:

*.-;'.2'/ ,;..

                                                       ., . '~t: ..**/., ,,r* RCS'Barfiei!'f(* *. :':~:** ;;!~
                                                                  ' :i   ,*
                                                                                                               \1'; ,:,;;,;:
                                                                                                                                \'.}*       . . . toiitaiiifue11t:.Bafrfer .. o;i.   ?Ji LOSS          I POTENTIAL LOSS                               LOSS           POTENTIAL LOSS                                    LOSS                     POTENTIAL LOSS tunnel                 Radiation                                  pat1¥.-,.-e;, ta tee temgerature            be¥el-Level,*                              envirenment annunciators                                                      exists after OR                                                                13rimary centainment

• Direct regort iselation signal of steam release OR OR ---B-:

                                                         -B. Emergency                                                                              Intentie RPV                                                                    Hal flBmEtrj' Depressurization                                                       containment required.                                                              venting per
                                                                                                                                      ~

OR

                                                                                                                              -----G.-

mm;g LABLE 13rimary system leakage teat resHlts in e:Xeeeai:ng El'.JHER e:f the fuHovling:

l. Ma~1:

Safe G13eFating Temflerature. QR

2. MEH1: Safe GI3erating Area R:aeiation

                                                                                                                                     ~

126

t!EI 99 01 (RevisieH e) trevember 2012 1* . '"

                                          '.i} ' : .,J. ' '.\  ,**:<. :._ . .:,;  :;1,;'                      .'*<: .:., '*\:. :* ., '. '£:*         '" ;,, : ,';'* *.:,. *;.:,;.*,    :    ,'  ' *:,: . '

Fu~iCI[d'.BJrrief ' *.* ,* ,. * .'RCS Baffi~t. / *

                                                                                                                                  "'           *?,
                                                                                                                                                   . * "': Containmen't Barrier,*

LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS

4. Primary Containment Radiation 4. Primary Containment Radiation 4. Primary Containment Radiation A. D1~ell Monitor Not Applicable A. Dr~ell Monitor Not Applicable Not Applicable A. Dr~ell Monitor (9184A/B) (9184A/B) (9184A/B) reading greater reading greater reading greater than 1250 R/hr. than 5 R/hr after than 5000 R/hr.

OR reactor OR B. Torus Monitor shutdownA.- B. Torus Monitor (9185A/B) Primary (9185A/B) reading greater eeH:taiameat reading greater than 125 R/hr raeiatiea meaiter than 500 R/hrA.- reaeiag greater Prima:!';' than (site specifie ceHtaiHnieHt raeiatiea meaiter vah1e).

                                                                                            ~

reaaiflg greater thaa (site specific

                                                                                                                                                                                     *ralee).

5. Other Indications 5. Other Indications 5. Other Indications A. Fuel damage Not Ar;mlicableA:- Not Ar;mlicableA:- Not Am2licableA:- Not A:i;mlicableA:- A. Fuel damage assessment (site specifie as (site specific (site specific as (site specific as assessment indicates at least applicable) as applicable) applieable) applicable) (PA8AP 7.2) 5% fuel clad indicates at least damage.fsite-- 20% fuel clad specific as damage.fsite--

applieable) specific as apfJlicable)

6. Emergency Director Judgment 6. Emergency Director Judgment 6. Emergency Director Judgment A. ANY condition in A. ANY condition in A. ANY condition in A. ANY condition in A. ANY condition in A. ANY condition in the opinion of the the opinion of the the opinion of the the opinion of the the opinion of the the opinion of the Emergency Emergency Emergency Emergency Emergency Emergency Director that Director that Director that Director that Director that , Director that indicates Loss of indicates Potential indicates Loss of indicates Potential indicates Loss of indicates Potential 127 L__

NEI 99 Ol (RevisieH s)

                                                                                                           }fovember 2012
        . j1Ffrei
           .,. ClacfBarfief' ':

POTENTIAL LOSS POTENTIAL LOSS POTENTIAL LOSS the Fuel Clad Loss of the Fuel the RCS Barri'~r. , Loss of the RCS the Containment Loss of the Barrier. Clad Barrier. Barrier. Barrier. Containment Barrier. 128

6ZI zrnz Ja E[l:HaAeN (~ ae~S!Aa=tt) IO 66 IHN:

                                                                                                                                    ~j:fll 99 01 (ReYiSi0H 6)

Nevefflber 2012 Basis Information For llWR DAEC EAL Fission Product Barrier Table-9-F-F-1-4 BWR-DAEC FUEL CLAD BARRIER THRESHOLDS: The Fuel Clad barrier consists of the zircalloy or stainless steel fuel bundle tubes that contain the fuel pellets.

1. RCS Activity Loss I.A

                                      /

This threshold indicates that RCS radioactivity concentration is greater than 300 µCi/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and correspciBds to Em apf)roximate range of 2% to 5% fHel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier. It is recognized that sample collection and analysis of reactor coolant with highly elevated activity levels could require several hours to complete. Nonetheless, a sample-related threshold is included. as a backup to other indications. There is no Potential Loss threshold associated with RCS Activity. D0",relaf)eF Notes: Threshold values should be determifled assuming RCS radioactivity concentration equals 3 00 µCi/gm dose equivaleB.t I 131. Other site speciffo :mits IHay be used (e.g., µCi/cc). DepeB.ding apon site specific capabilities, this threshold may have a sample analysis component anc:Vor a radiation monitor reading component. Add this paragraph (or similar wording) to the Basis if the threshold iHcludes a sample analysis component, "It is recogHized that sample collection and aBalysis of reactor coolaB.t 1.vith highly elevated activity levels could requke several hours to complete. tJonetheless, a sample related threshold is included as a backup to other indications."

2. RPV Water Level 130

1'lEI 99 01 (Revision 6) NoYemeer 2012 Loss 2.A The Loss threshold represents any EOP requirement for entry into the Severe Accident Guidelines. This is identified in the BWROG EPGs/SAGs when adequate core cooling cannot be assured.The Loss threshold represeftts the EOP requirement for priffiary eontaimHent flooding. Tais is identified in tee B\VROG EPGs/SAGs when the pkrase, "Priffiary CoHtainffient FloodiRg Is Required," appears. Sinee a site speeific RPV water level is not specified kere, the Loss tkresaold pkrase, "Primary cofttaiffl'Heftt flooding required," also accommodates the EOP need.to flood the primary containmeftt wkeH RPV 1.vater level caHnot be detem1ined and core damage due to inadequate core cooling is believed to be occmTiHg. Potential Loss 2.A This water level corresponds to the top of the active fuel and is used in the EOPs to indicate a challenge to core cooling. 131

                                                                                                                                "NEI 99 01 (Revision 6)
                                                                                                                                       ~fovemeer 2012 B"'R FUEL CLAD BARRIER THRESHOLDS:

The RPV water level threshold is the same as RCS barrier Loss threshold 2.A. Thus, this threshold indicates a Potential Loss of the Fuel Clad barrier and a Loss of the RCS barrier that appropriately escalates the emergency classification level to a Site Area Emergency. 132

NEI 99 01 (RevisioR 6) November 2012 DAEC FUEL CLAD BARRIER THRESHOLDS {cont.): This threshold is considered to be exceeded when, as specified in the site specific EOPs, RPV water cannot be restored and maintained above the specified level following depressurization of the RPV (either manually, automatically or by failure of the RCS barrier) or when procedural guidance or a lack oflow pressure RPV injection sources preclude Emergency RPV depressurization. EOPs allow the operator a wide choice ofRPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurization of the RPV in order to facilitate RPV water level control with low-pressure injection sources. In some events, elevated RPV pressure may prevent restoration of RPV water level until pressure drops below the shutoff heads of available injection sources. Therefore, this Fuel Clad barrier Potential Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability oflow-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss ofRPV inventory. 133

NEI 99 01 (RevisioR 6)

                                                                                                                                    }fovember 2012 Df...EC FUEL CLAD Bf..RRIER THRESHOLDS feaet.)::

The term "cannot be restored and maintained above" means the value ofRPV water level is not able to be brought above the specified limit (top of active fuel). The determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the current value is below the top of active fuel, but does not permit extended operation below the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained. In high-power ATWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level to the top of active fuel in order to reduce reactor power. RPV water level is then controlled between the top of active fuel and the Minimum Steam Cooling RPV Water Level (MSCRWL). Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority. For such events, ICs SA§...SA6 or SS~ will dictate the need for emergency classification. Since the loss of ability to determine if adequate core cooling is being provided presents a significant challenge to the fuel clad barrier, a potential loss of the fuel clad barrier is specified. 134

Nel 99 0 l (RevisieR e) NeveHieer 2012 B~'R FUEL CLAD BARRIER THRESHOLDS: DeYeleJJer Notes: Loss 2.A The phrase, "Primary containment flooding required," should be modified to agree with the site specific EOP phrase indicating exit from all EOPs and entry to the SAGs (e.g., dryv,rell flooding required, etc.). Poteatial Loss 2.A The decision that "RPV water level eaHsot be determined" is directed by g1:1idance given in the RPV \Yater le'\1el eostrol seetiofls of the .60Ps.

3. Not Applicable (included for numbering consistency between barrier tables) 135

                                                                                                                                     !>lei 99 OI (Re,,*isieA 6)
                                                                                                                                             !>10'>'0A900F 2012 DAEC FUEL CLAD BARRIER THRESHOLDS (cont.) :

4. Primary Containment Radiation Loss 4.A and Loss 4.B The Drywell and Torus radiation monitor readin~ corresponds to an instantaneous release of all reactor coolant mass into the Drywell or primary Toruscontainment, assuming that reactor coolant activity equals 300 µCi/gm dose equiYalent I 131. Reactor coolant activity aboYe this ]eye) is greater thctn that eJtpeoted for iodiee spilces aed corresponds to ftfl-approximately range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.

The radiation monitor readin~ in this threshold tS-are higher than that specified for RCS Barrier Loss threshold 4.A since it indicates a loss of both the Fuel Clad Barrier and the RCS Barrier. Note that a combination of the two monitor readings appropriately escalates the emergency classification level to a Site Area Emergency. There is no Potential Loss threshold associated with Primary Containment Radiation . ffhe readieg sl:lould be determieed assuming the instantaeeous release aed dispersal of tl:le reaotor ooolant noble gas and iodine inYentor~*, 136

                                                                                                                                            }/Bl 99 oI (ReYisieA a)

NeveFAeer 2012 BWR DAEC FUEL CLAD BARRIER THRESHOLDS (e0Rt.): ther Indications... 5.

1. Other IRdieetiees Loss and/or Potential Loss 5.A There is no Potential Loss threshold associated with Other Indications .

Develeper Netes: Loss and/or Potential Loss 5.A De:i.relopers should determine if other reliable indicators exist to evaluate the status of this fission product barrier (e.g., review accident analyses described in the site final Safety i'illalysis Report, as apdated). The goal is to identify any unique or site specific indications that 1

         .vill promote timely and accurate assessment of barrier status.

Aay added thresholds should represent apprmcimately the same relative threat to the barrier as the other thresholds in this colufl'lfl. Basis information for the other thresholds may be used to gauge the re lative barrier threat level.

 ~ ~Emergency Director Judgment Loss 6.A This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad Barrier is lost.

Potential Loss 6.A This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Fuel Clad Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored. 137

   ~!el 99 0 I (Re,*isieA e)
          ~Jevemeer 2012 138

NEI 99 Ol (RevisioH a) tfovef!'lber 2012 InVR DAEC RCS BARRIER THRESHOLDS: The RCS Barrier is the reactor coolant system pressure boundary and includes the RPV and all reactor coolant system piping up to and including the isolation valves.

1. Primary Containment PFessu:reConditions Loss I.A The (site speeifie v'alue) primary eontainmeH:4:2 psig pressure is the drywell high pressure scram setpoint which indicates a LOCA by automatically initiating the-ECCS or eqHivalent makeup system.!.

There is no Potential Loss threshold associated with Primary Containment Pressure. DevelapeF Nates:

2. RPV Water Level Loss 2.A This .vater l+ 15 inches e¥el--corresponds to the top of active fuel (TAF) and is used in the EOPs to indicate challenge to core cooling.

1 The RPV water level threshold is the same as Fuel Clad barrier Potential Loss threshold 2.A. Thus, this threshold indicates a Loss of the RCS barrier and Potential Loss of the Fuel Clad barrier and that appropriately escalates the emergency classification level to a Site Area Emergency. This threshold is considered to be exceeded when, as specified in the site speeifie EOPs, RPV water cannot be restored and maintained above the specified level following depressurization of the RPV (either manually, automatically or by failure of the RCS barrier) or when procedural guidance or a lack of low pressure RPV injection sources preclude Emergency RPV depressurization EOPs allow the operator a wide choice ofRPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurization of the RPV in order to facilitate RPV water level control with low-pressure injection sources. In some events, elevated RPV pressure may prevent restoration ofRPV water level until pressure drops below the shutoff heads of available injection sources. Therefore, this RCS barrier Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss ofRPV inventory. 139

l'mI 99 01 (Re,*ision 6)

                                                                                                                                          }fovember 2012 1l\¥R DAEC RCS BARRIER THRE8HOLD8:

The term, "cannot be restored and maintained above," means the value of RPV water level is not able to be brought above the specified limit (top of active fuel). The determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the current value is below the top of active fuel, butfuel but does not permit extended operation beyond the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained. 140

NEI 99 0 l (RevisioH 6) November 2012 DAEC RCS BARRIER THRESHOLDS {cont.):

                                                 \

In high-power ATWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level to the top of active fuel in order to reduce reactor power. RPV water level is then controlled between the top of active fuel and the Minimum Steam Cooling RPV Water Level (MSCRWL). Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority.-For such events, ICs SAS or SS5 will dictate the need for emergency classification. There is no RCS Potential Loss threshold associated with RPV Water Level.

3. RCS Leak Rate Loss Threshold 3.A Large high-energy lines that rupture outside primary containment can discharge significant amounts of inventory and jeopardize the pressure-retaining capability of the RCS until they are isolated. If it is'determined that the ruptured line cannot be promptly isolated from the Control Room, the RCS barrier Loss threshold is met.

Loss Threshold 3.B Emergency RPV Depressurization in accordance with the EOPs is indicative of a loss of the RCS barrier. If Emergency RPV Depressurization is performed, the plant operators are directed to open safety relief valves (SRVs) and keep them open. Even though the RCS is being vented into the suppression pool, a Loss of the RCS barrier exists due to the diminished effectiveness of the RCS to retain fission products within its boundary.

• Potential Loss Threshold 3.A Potential loss of RCS based on* primary system leakage outside the primary containment is determined from EOP temperature or radiation Max Normal Operating values in areas such as main steam line tunnel, RCIC, HPCI, etc., which indicate a direct path from the RCS to areas outside primary containment.

A Max Normal Operating Limit (MNOL) value is the highest value of the identified parameter expected to occur during normal plant operating conditions with all directly associated support and control systems functioning properly. 141

MEI 99 01 (RevisieH i)

                                                                                                                                              }l"evemeer 2012 B,¥R Th.4..EC RCS BAR&IER THRESHOLDS:

The indi~ators reaching the threshold barriers and confirmed to be caused by RCS leakage from a primary system warrant an Alert classification. A primary system is defined to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system. An UNISOLABLE leak which is indicated by Max Normal OperatingMNOL values escalates to a Site Area Emergency when combined with Containment Barrier Loss threshold 3.A (after a containment isolation) and a General Emergency when the Fuel Clad Barrier criteria is also exceeded. DAEC RCS BARRIER THRESHOLDS (cont.): Developer Notes: Loss Threshold 3 .A The list of systems inelt:1ded in this threshold sho1,ild be the higa e0ergy lines 1.vaica, ifru-ptured and remain unisolated, ean rapidly depressarize the RPV. These lines are typieally isolated by aetaation of the Leak Detection system. Large high energy line breaks sueh as Main Steam Line (MSL), High Pressure Coolant Injection (HPCI), Feedi,vater, Reactor Water Clearmp (RWCU), Isolation Condenser (IC) or Reactor Core Isolation Cooling (RCIC) that are UNISOL".BLE represent a significant loss of the RCS bafl"ier.

4. Primary Containment Radiation Loss4.A The Drywellradiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for Fuel Clad Barrier Loss threshold 4.A since it indicates a loss of the RCS Barrier only.

There is no Potential Loss threshold associated with Primary Containment Radiation. Develo13er Netes: The reading should be determined assuming tae instantaneous-release and dispersal of the reactor eoolant Hobie gas and iodine inventory, with RCS activity at Technical Specification allowable limits, into the primary containment atmosphere. Using RCS activity at Technical Specification alfo,.v,able limits aligns this threshold with IC SU3. Also, RCS activity at tais level will typically result in primary coHtainment 142

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                                                                                                                                             }Hll 99 01 (Revisiea 6)

Neveml:Jer 2012 B"'R R-CS B,. . . RRIER THRESHOLDS: In some cases, the site specific physical location and sensitivity of the primary containment radiation monitor(s) may be such that radiation

• from a cloud of released RCS gases cannot be distinguished from radiatioa emanating from piping aad components containing elei,rated reactor eoolaet activity. If so, refer to the Developer Guidance for Loss/Potential Loss 5.A and determine if an alternate indication is available.

5. Other Indications There are no Loss or Potential Loss thresholds associated with Other Indications.

Devela~e.- Nates: Loss and/or Potential Loss 5.A Developers should detennine if other reliable indicators e1dst to evaluate the status of this fission product barrier (e.g., reviev,r accident analyses described in the site Fiaal Safety Analysis Report, as updated). Tee goal is to ideatify any unique or site specific i-ndications that 1.vill proffiote timely and accurate assessffient of barrier status. Any added thresholds should represent apprmdmately the same relative threat to the barrier as the other thresholds in this cokimn. Basis infonnation for the other thresholds may be used to gauge the relative barrier threat level.

6. Emergency* Director Judgment Loss 6.A This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the RCS barrier is lost.

Potential Loss 6.A This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored. Devela~er Nates: 144 L_----

NEI 99 01 (RevisieH 6) Nevemeer 2012 llWR DAEC CONTAINMENT BARRIER THRESHOLDS: The Primary Containment Barrier includes the drywell, the wetwell, their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves. Containment Barrier thresholds are used as criteria for escalation of the ECL from Alert to a Site Area Emergency or a General Emergency.

1. Primary Containment Conditions Loss l .A and 1.B Rapid UNPLANNED loss of primary containmeat drywell pressure (i.e., not attributable to drywell spray or condensation effects) following an initial pressure increase indicates a loss of primary contaiameatdrywell integrity. Primary coatainmeatDrywell pressure should increase as a result of mass and energy release into the primary containment from a LOCA. Thus, primary coataiamentdrywell pressure not increasing under these conditions indicates a loss of primary containment integrity.

These thresholds rely on operator recognition of an unexpected response for the condition and therefore a specific value is not assigned. The unexpected (UNPLANNED) response is important because it is the indicator for a containment bypass condition. Loss l.C The use of the modifier "direct" in defining the release path discriminates against release paths through inte1facing liquid systems or minor release pathways, such as instrument lines, not protected by the Primary Containment Isolation System (PCIS). The existence of a filter is not considered in the threshold assessment. Filters do not remove fission product noble gases. In addition, a filter could become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream. Following the leakage of RCS mass into primary containment and a rise in primary containment pressure, there may be minor radiological releases associated with allowable primary containment leakage through various penetrations or system components. Minor releases may also occur if a primary containment isolation valve(s) fails to close but the primary containment atmosphere escapes to an enclosed system. These releases do not constitute a loss or potential loss of primary containment but should be evaluated using the Recognition Category R ICs. 145

1'mI 99 0 l (Revisiafl i)

                                                                                                                                         ~foyeffiber 2012 DA.EC CONTAINMENT BARRIER TIIRE8HOLD8:

Loss 1.D EOPs may direct primary containment isolation valve logic(s) to be intentionally bypassed, even if offsite radioactivity release rate limits will be exceeded. Under these conditions with a valid primary containment isolation signal, the containment should also be considered lost if primaiy containment venting is actually perf01med. Intentional venting of primary containment for primary containment pressure or combustible gas control to the secondary containment and/or the environment is a Loss of the Containment. Venting for primaiy containment pressure control when not in an accident situation (e.g., to control pressure below the drywell high pressure scram setpoint) does not meet the threshold condition. DAEC CONTAINMENT BARRIER THRESHOLDS (cont.): DA.EC CONTAIN~iENT BARRIER TIIRE8HOLD8: Potential Loss I.A The threshold pressure is the primary containmenfforus internal design pressure. Structural acceptance testing demonstrates the capability of the primary containment to resist pressures greater than the internal design pressure. A pressure of this magnitude is greater than those expected to result from any design basis accident and, thus, represent a Potential Loss of the Containment barrier. Potential Loss 1.B If hydrogen concentration reaches or exceeds the lower flammability limit, as defined in plant EOPs, in an oxygen rich environment, a potentially explosive mixture exists. If the combustible mixture ignites inside the primary containment, loss of the Containment barrier could occur. Potential Loss l .C The Heat Capacity Temperature Limit (HC+L) is the highest suppression pool temperature from which Emergency RPV Depressurization will not raise:

• Suppression chamber temperature above the maximum temperature capability of the suppression chamber and equipment within the suppression chamber which may be required to operate when the RPV is pressurized, OR 146

NEI 99 01 (RevisioH 6)

                                                                                                                                    },fo,,remeer 2012 B~ZR CONTAINMENT Bl'...RRIER THRESHOLDS:
    -Suppression chamber pressure above Primary Containment Pressure Limit A, while the rate of energy transfer from the RPV to the containment is greater than the capacity of the containment vent.

The HC+L is a function of RPV pressure, suppression pool temperature and suppression pool water level. It is utilized to preclude failure of the containment and equipment in the containment necessary for the safe shutdown of the plant and therefore, the inability to maintain plant parameters below the limit constitutes a potential loss of containment. 147

                                                                                                                                     }ffil 99 01 (Revision 6)

November 2012 DAEC CONTAINMENT BARRIER THRESHOLDS (cont.):t Develaf)eT Notes: Potential Loss 1.B BWR EPGs/8AGs s19ecifically defo=ie tke limi-ts associated ,vith eJrf)losive mbctl:H=es in terms of deflagration cofl:eeHtratioHs of kydrogefl: and mcygen. For lM.k I/II containments the deflagration limits are "6% hydrogen and 5% oxygen in the drywell or suppression ck.amber". For l\'lk HI eontainmCHts, the limi-t is the "Hydrogen Deflagration Overpressure Limit". The threshold term "CXfJlosive mixture" is syn:onymous with the EPG/SAG "deflagration limits". Potential Loss l.C 8ince the HCTL is defined assammg a range of SUfJpression pool water levels as low as the ele*1atioH of the downeomer openiHgs in Mk I/II eontainmenES, or 2 feet abo:ve the elevation of the horizontal vents in a Mk III containment, it is unnecessary to coHsider separate COfltai-Hment barrier Loss or Potefltial Loss thresholds for abHormal su19pression pool 1.vater level eonditioHs. If desfred, developers may iselude a separate ContaiHment Potestial Loss threshold based on the inability to maifltain su1919ressioH 19001 *.vater level above the downeom.er opeaiHgs iH Mk I/II eoatainmeats, or 2 feet above the elevation of the horizontal vents iH a l\'lk III eontainmeflt '.vith RPV pressure above the minimrnR decay heat rem.0*1al pressure, if it will sim.plify the assessm.eflt of the s-uppression 19001 level eompoHeflt of the HCTL.

2. RPV Water Level There is no Loss threshold associated with RPV Water Level.

Potential Loss 2.A The Potential Loss threshold is identical to the Fuel Clad Loss RPV Water Level threshold 2.A. The Potential Loss requirement for Primary Containment Flooding indicates adequate core cooling cannot be restored and maintained and that core damage is possible. BWR EPGs/SAGs specify the conditions that require primary containment flooding. When primary containment flooding is required, the EPGs are exited and SAGs are entered. Entry into SAGs is a logical escalation in response to the inability to restore and maintain adequate core cooling. 148

NEI 99 0 l (Revisioa 6) Novefflber 2012 BWR CONV..ImiENT BARRIER THRE8HOLD8: PRA studies indicate that the condition of this Potential Loss threshold could be a core melt sequence which, if not corrected, could lead to RPV failure and increased potential for primary containment failure. In conjunction with the RPV water level Loss thresholds in the Fuel Clad and RCS barrier columns, this threshold results in the declaration of a General Emergency. Develeper Netes: The phrase, "Primary containment flooding required," should be modified to agree with the site specific EOP phrase indicating eKit from all EOPs and entry to the SAGs (e.g., diy.vell flooding required:;.etc.).

3. Primary Containment Isolation Failure These thresholds address incomplete containment isolation that allows an UNISOLABLE direct release to the environment.

Loss 3.A The 1:1se of the modifier "direct" in defining the release path discriminates against release paths through m.terfacing liquid systefns or minor release patlnvays, such as instrument lines, not protected by the Primary Containment Isolation System (PCIS). The existence of a filter is not considered in the threshold assessment. Filters do not remove fission product noble gases. In addition, a filter could become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been e1weeded) or 1,vater satura-tion from steam/high hlHTiidity in the release stream. Follm:ving the leakage of RCS mass into primary containment and a rise m. primary containment pressure, there may be minor radiological releases associated with allowable primary containment leakage thrnagh various penetrations or system components. Minor releases may also occur if a primary containment isolation valve(s) fails to close but the primary containment atmosphere escapes to an enclosed system. These releases do not constitute a loss or potential loss of primary containment but saoald be evah1ated 1:1sing the Recognition Ca-tegory A ICs. Loss 3.B EOPs may direct primary containment isolation valve logic(s) to be intentionally bypassed, even if offsite radioactivity release rate limits vlill be eRceeded. Under these conditions vlith a valid primary containment isolation signal, the containment sho1:1ld also be considered lost if prime:Fy containment venting is actually performed. Intentional venting of primary containment for primary containment pressure or combustible gas control to the secondary containment and/or the environment is a Loss of the Containment. Venting for primary containment press..tre control ween not in an accident situation (e.g., to control press*dre belov1 the d1ywell high pressure scram setpoint) does not meet the threshold condition. 149

NEI 99 Ol (RevisioH e) November 2012 Loss 3.GA The Max Safe Operating Limit (MSOL) for Temperature and the Max Safe Operating Radiation Level are each the highest value of these parameters at which neither: (1) equipment necessary for the safe shutdown of the plant will fail, nor (2) personnel access necessary for the safe shutdown of the plant will be precluded. EOPs utilize these temperatures and radiation levels to establish conditions under which RPV depressurization is required. BWR CONL\INMENT B,A_..RRJER THRESHOLDS: The temperatures and radiation levels should be confirmed to*be caused by RCS leakage from a primary system. A primary system is defined to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system. In combination with RCS potential loss 3.A this threshold would result in a Site Area Emergency. There is no Potential Loss threshold associated with Primary Containment Isolation FailureRCS Leak Rate. 150

                                                                                                                                      }IEI 99 01 (Re*,rision 6)

November 2012 DAEC CONTAINMENT BARRIER THRESHOLDS (cont.):+ Develef)er Notes: Loss 3.B Consideration ma-y ae given to specifying the specific procedural step vlithin tee Primary Containment Control BOP that defines intentioHal venting of the Primary Containment regardless of offsite radioactivity release rate.

4. Primary Containment Radiation

   . There is no Loss threshold associated with Primary Containment Radiation.

Potential Loss 4.A The drywell radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary coatainn1eatdrywell, assuming that 20% of the fuel cladding has failed. The radiation monitor reading for the torus corresponds to an instantaneous release of all reactor coolant mass directly into the torus, assuming that 20% of the fuel cladding has failed. This level of fuel clad failure is well above that used to determine the analogous Fuel Clad Barrier Loss and RCS Barrier Loss thresholds. NUREG-1228, Source Estimations During/ncident Response to Severe Nuclear Power Plant Accidents, indicates the fuel clad failure must be greater than approximately 20% in order for there to be a major release of radioactivity requiring offsite protective actions. For this condition to exist, there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential loss of containment which would then escalate the emergency classification level to a General Emergency. D~releper Notes: NUREG 1228, Sour-cc Estimatiom Di:trb9g Incident Response to Severe }lucleer Power ,.12/:e,it Aeeitlcnts, provides the aasis for using the 20% fuel claddiHg failure value. Unless there is a site Sflecific analysis justifying a different value, the readiag should ae determined assuming the instantaHeous release and diSj3ersal of the reactor coolant noale gas and iodine inventory associated v,rith 20% fuel clad failure into the f)rimary containment atmosphere. BWR CONTAINl\<IENT BARRIER THRESHOLDS:

5. Other Indications There is no Loss threshold associated with Other Indications Loss and/or Potential Loss 5.A 151

NEI 99 01 (Revision 6)

                                                                                                                                   }>fo veftlfl er 2012 Results obtained from procedure PASAP 7.2, Fuel Damage Assessment, indicate at least 25% fuel clad damage. Tlris sU0eategory addresses other site specifie thresholds that may be ineluded to indicate loss or potential loss of the Containment barrier based on plant speeifie design ooaraeteristies not eonsidered in the generic guidance. P ASAP 7.2 only shows whether fuel damage is greater than or less than 25%, thus this indication is not likely to be declared before containment ban-ier potential loss 4.A which indicates 20% fuel damage. However, this potential loss threshold adds an additional layer of diversity to the scheme.

152

                                                                                                                                    "NEI 99 0 l (Revisien 6)

Nevember 2012 DeveletJer Netes: Loss and/or Potential Loss 5.A Developers should detennine if other relia-ble indicators exist to evaluate the status of this fission product barrier (e.g., rcYliew accident analyses described in the site FiHal Safety Ana-lysis Report, as updated). The goal is to identify aHy *tmique or site specific indicatioHs that will promote timely and accurate assessment of barrier status. Any added thresholds should represent approximately the same relative threat to the baiTier as the other thresholds in this column. Basis information for the other thresholds may be used to ga:Uge the relative barrier threat level.

6. Emergency Director Judgment Loss 6.A This threshold addresses any *other factors that are to be used by the Emergency Director in determining whether the Containment barrier is lost.

Potential Loss 6.A This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Containment Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored. 153

NEI 99 01 (Re*1isi0H 6)

                           }fovember 2012 Developer Notes:

NeHe 154

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                                                                                                                                                      'l Restorat uate RCS heat                                                                                                  ion proeedUi'e removal                                                                                                        not effeotive eapability via                                                                                                within 15 steam                                                                                                          minutes.

generators as inElieateEl 19:y (site speeifio inElieations).

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2) Incorporated along with parameter and value thresholds (e.g., a fuel clad loss 1.vould have 2 thresholds such as "CETs > 1200°P" a-Rd "Core Cooling Red entry conditions met".

3) Used in lie:u of paran1eters and values for all thresholds.

With Ofle mc.ception, if a decision is made to include the CSFST eased thresholds, then all sueh allo1.ved thresholds m:ust ae :usecl iH the table (e.g., it is not permissiale to use only the C OraRge tenninus as a potential loss of the fuel clad barrier threshold and disregard all other G8F8T based thresholds). The ofle exception is the RC8 IH.-tegrity (P) GSFST. Because of the complmdty of the P Red decision point that relies OH aa assessment a pressure temperature curve, a P Red condition may be used as an RG8 potential loss threshold 1.vithout the need to incorporate the other CSF8T basecl thresholds. 160

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NEI 99 01 (RevisieH 6) Ne,*el'Hber 2012 P'NR FUEL CLAD Bz\RRIER THRESHOLDS: Potential Loss a.A This threshold addresses aay otller faetors tllat may ee 1.c1sed ey tlle Ernergene:Y Direetor ia determiniag vlhetller tlle F1:1el Clad Barrier is potentially lost. The Emergeney Direetor should also eonsider whether or not to deelare the earrier potentially lost in tlle event that barrier stems eannot ee monitored. DeveloJJer Notes: NeHe 164

1-lBI 99 01 (RevisioH 6) 1-fovember 2012 ~'R RCS B1\RRIER THRESHOLDS: The RCS Ban-ier includes the RCS primary side and its connections up to and including the pressurizer safety and relief valves, and other connections up to and including the primary isolation val,;es. RCS 01 SC Tu.he Leah:age Loss 1.A This threshold is based on an U}USOL'\BLE RCS leak of sufficient size to require an automatic or manual actuation of the Emergency Core Cooling System (EGGS). This condition clearly represents a loss of the RCS Barrier. This threshold is applicable to unidentified and pressl:H'e bouedary leakage, as well as idefitified leakage. It is also applicable to UNISOL}..BLE RCS leakage through an interfacing system. The mass loss may be into any location inside containment, to the secondary side (i.e., steam generator tube leakage) or outside of coetainment. A steam generator with primary to secondary leakage of sufficient 1nageitude to require a safety injection is considered to be RUPTURED. If a RUPTURED steam generator is also FAULTED outside of contait'lment, the declaration escalates to a Site z'\.rea Emergency since the Containment Barrier Loss threshold I .A will also be met.

• Potential Loss 1.l..

This threshold is based on an UNISOLABLE RCS leak that results in the ieability to maintaiR pressurizer level 1.vithin specified limits by operation of a normally esed eharging (makeep) pmnp, but an EGGS (SI) acteation has not occurred. The threshold is met ween an 013erating procecklfe, or operating crev.r supervision, directs that a standby charging (makeup) pum.p be placed in service to restore and maintain pressmizer level. . This threshold is applicable to unidentified and pressure boendary leakage, as well as identified leakage. It is also applicable to U}HSOLABLE RCS leakage throaga an interfacing system. TJ:ie mass loss may be into any locatioa iaside containn'lent, to the secondary side (i.e., steam generator tube leakage) or outside of containment. If a leaking steam generator is also FAULTED outside of containment, the declaration escalates to a Site i\rea Emergency .since the Containment Barrier Loss threshold l .A will also be met. Poteatial Loss l.B This coadition iadicates an Emtreme challenge to the integrity of the RCS presslH'e boundary due to pressurized thennal shock a transient that causes rapid RCS cooldown \.Vhile the RCS is in Mode 3 or higher (i.e., hot and pressurized). 165

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                                                  . to the sa4"\0 reIa!Jve be ased                      . tli<eat lo Ike b v **                                    >llll 99 N,.,..,,.,

(Revisioa g1Basis 0) 2Q12 ga-uge the relative bmTier threat le:1Tel1er as the other thresholds iH this e o,.unn. I PWR RCS BARRIER Emergeeey Direeter J 8:RESHOLDS: Loss 6.A gmeet This threshold Po~ential addresses aa ) ' other faetors that may be l:lsed b , t Loss 6.A Tlus thresHO!d addresses "" ' y 4te F<Hergeeey DiFeotor iH deleF . . 168

0\ \0

                                                                                                                                }Ull 99 01 (RevisieH e)
                                                                                                                                        }foveff!eer 2012 P\¥R CONTAINl:\tlENT Bz\RRIER THRESHOLDS:

Following an SG tube leak or rupture, there may be minor radiological releases through a secondary side system component (e.g., air ejectors, glad seal e1Eha1:1sters, valve packing, etc.). These types of releases do not constitute a loss or potential loss of containment bl:lt shm1ld be eval1:1ated using the Recognition Cittegory A. IGs. The emergency classification levels res1:1king from primary to, secondary leakage, 1.vith or *.vithout a steam release from the FAULTED SG, are summarized below. i..i:ffeeted SG is FAYLTED Oetside of Ceetaiemeet~ P te S Leali' Rate ~ N& Less than Ol' eEI"Ual to ~§ Ne We classification gpm (or other val1:1e per classification C'TTA T"\

         ~~
                      --                   1 -
                                                * ~ -....

1'T

                                                               .L    -
                                                                       ,'\

Greater than ~§ gpm Uftl:lsual E1rent Un..1s1:1al Event (or other 1alue per SU4 1 per SU4 per SU4 T"\ *-1 *-.,.....,. '11..T .'\

         ---      ........ ....,.,...-.... l.    -    ... - .... ,

Reeyuires operetioH of a Site l~.crea staedby chargiHg EmergeHcy per Alert per FA.I (makeup) pump (RCS F&l-

          ..... ..., --*           .....          ......,a    _I T      .'\

ReEI"Uires an a\:ltomatic Site Area or maFn1al EGGS (SI) Emergency per Alert per FA.1

                                                                                                 ~

actuation (RCS Berrier F&l-be5tH 170

Ill

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                                                                                                                                                  }11.Bl 99 01 (RevisieH 6)
                                                                                                                                                           *Nevember 2012 PWR CONTAINME~ff BnA RRIER          ~     ' Tlr      l Ids*
                                               'lfeS10 Inadequate Heat R~moval

• There is no Loss Potential Loss 2.A threshold assoei t d ,..

a e n ith Inadequate Heat Ren1oval. This condition t

• represents

                       .       an IMMINENT
                                     * *c o ncore fef                                                    d imelt t isequen o c e ,w*hIC,

• h if
• not corrected, eould lead ,,

coiramment failure. For this

-lemOHtati'"' efa preeederefs) t H te ee~...-, there""" alrea<ly ba¥e beOH a less eft "' 'essel fad.,.***"" ffie,easa<I pe!en!ial lf<\jee!e,y will likely lea<I te .,;;.*..::i7~tore ade'l;'ate eere eeehng is eet effeotive fsueeessr!*-~: Barner arul tile l'ael Gloe BGffie,. If The res_tamtiea f)f<leedere is eeasitlered ~.:* a -~:':'.selj'dent el'_'tllCftge ef !lie Gentainlllent B.2;;;;._ IB 15 - s , it ,. QSSllffied that the event 1ncreas1Rg. *whether or not the roce ~.::.h !lie emerga,ey elassilieatioo le-I' I derefs) '.:'ll be effee!tve slaeuld be awarent ,..;!;" e if core_ ex.it thermocouple read.in s are deer . . , . 15 . eaS1ng nHd,er if reactor vessel level is s"""'e aee1dCllt

              .               'e as seea asOl11 b15..de!e aealy,;es fe.g., NURliG
                                                                        * <l h                      "
                                                                ~;;-~e t-et the prneederefs) wiH net be m!ftn!es The Effie effeeti..                 I)'

e4g<Bey1£eeter slaeul<l esealate ii"'"'.?."" ~ 113 ef eere damage seCBaries, aed the! the li~lil 0Bel.1ded the! !ilnetiea_ resterati0B preeederes .;.,.* aHCst . . . fl"' ',de 15 Hl!eutes beyeed tile req1i d . ~eed ef tlOfllallIBlent liulure is **ery Silla!! . eere degra<le!,ea '" a s,g,,illeent I)C¥eloper Notes: ,re eHtry p01Ht te detemttfle if f)f<leedural aeti.:.s em, re*. "' me """"'5. Givea !lais, it is Bjll""!'riate te Seffie site apaaiJie bOl's aru!<er BOP . . ,erse !lie eere Rlelt seqHCHee. th ' user gmdehnes ma

• t el* h *
• ermoeo:113le readings necessary to drive aetion7 es a IS ~ee1s10H: makiH:g criteria coneemin .

e0fe eeehag preaeclere). Te maift! . . s fe.g., 3 GIITs readiag greeter !ianH I 200 F . . .. g the """'""' er ether attril,ntes ef reading !lareskelds. a,e e0HS1steeey With bOPs, these deeisi0B fllnl<iflg 'eri..':ia :.:e'.i;:"~d l,efere lraesitiefHBg te !!fl ffia<lequate Petent,el Less 2.A.I ~ e ase<l ,e the eore e,ut !laermee0U]'le Enter site specific criteria requirin actions Fer A readin pl_;,;'!laet fl2 i.!ee;;.;;1:::::  :,e ~i;:r:. g entry mto a core

                                                        ""'Y   alsecooling   'restoration I'reoe-are
                                                                    !,e use<!."

d er proffiJll imp!efHeatatiea ef eere eeelmg resteratiea Core Cooling Red P a t h .

• es mg ouse Ov,rners Group Emergency Response Guidel.mes, enter the parameters and values used iH: the 172

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The s!alus of I . eontaifilBeHt bu! shoald b ,. ruHE!lont -OSflhere esea N,v-" 2812 Loss H! - He eonla!Hmont O Barri..- Btlfiag "" ..'.'e:::':.~~.~ asiHg !he Reeogniti-:~::g:~d These releases do aol eoa n, COH!aiHIBent SU"'f' te

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• eea<li!i e ***I  :,-. 11 erefere rejlresonts a petontial7.:':::  :"1c eould result ia eellateral e SlljlpOfl S)'Slems a,e desigeed le . a ~,ere eef!!oiHment re
• enteriea is iaeloded lo .:1:~.mat,eal!y aeleale, aad less J:.::::.~ " .

greater than the satpeiBI at '"hie e onla!HB!ont f!arrier erefine this tllreshold. E1r.peeted monitor a1arms or readings may also 175 L

1'mI 99 01 (Revisiea 6) 1'fovefflaer 2012 Potential Loss 4.A The site specific pressure is the containment design pressure. For 13laats that have implemented Westinghouse Owners Group Effiergency Res13onse G-uidelines, the pressure value in Potential Loss 4.A is that used for the ContaiBment Red Path. If the GoBtainment G8F8T contains more than one Red Pata due to other depeBdencies (e.g., status . of containment isolation), enter the highest containment press\ire value shovm on the tree. This is typically the containment design pressure. P\VR GONTSl}fMfil'ff B,'\R..~IER Thresholds: Potential Loss 4.B D01,relo13ers may enter the miBimum containment atmospheric hydrogen concentration necessary to support a hydrogen bum (i.e., the lower deflagration limit). A concurrent containmeBt oxygen concentration may be included if the plant has this indication available in the Control R-0om. Potential Loss 4.C Enter the site specific pressure setpoint value that ach1ates containment pressure control systems (e.g., containment spray). Also enter the site specific containment pressure control Sj'Stem,'equipment that should be operating per design if the eontainmeflt pressure setpoint is reaclled. If desired, specific coHdition indicatioBs sucll as parameter valses can also be eBtered (e.g., a containment spnty flov,r rate less than a certain value). This threshold is not 8.jJJ3licable to tlle U.S. Evolutionary Power Reactor (EPR) design. J}/estinghouse ERG Plants As a potential loss indication, developers should consider including a thi=eslwld the san1e as, or sin1ilar to, "Containment Red eHtry conditions met" in accordance with the guidance at the front of this section. Other Indications Loss and/or Potential Loss 5.A Tllis sl:lbcategory addresses other site specific thresholds that may be included to indicate loss or potential loss of the Containment barrier based on plant specific design characteristics not considered in the generic guidance. Developer Notes: Loss and/or Potential Loss 5.A If si!e emergeacy opeFatiag procedures provide for *;eatiag of the contaiB:FBent as a meaHs ofpreventiHg catastrophic failure, a Loss threshold should be incladed for the containrnent barrier. Tllis threshold v,zould be met as soon as such veHting is IMMINfil'rT. Gontaiement venting as part ofrecovery actioHs is classified iH accordaHce with the radiological eftlueHt ICs. Developers should determine if other reliable iHdicators exist to evaluate the status of this fissioH product barrier (e.g., r01;iew accideHt analyses described in the site Final Safety Aaalysis Re13ort, as updated). The goal is to identify any uaique or site speeific iedieations that

*.vill promote timely and ace:.1rate assessH1eat of earrier status.

P\l/R CONTAINMfil'H BARRIER Thresholds:

.Any added thresholds should represeHt a1313ro1dmately the same relative threat to the barrier as the other thresholds iH this columH. Basis informatioR for the other tlH=esholds may be used to gasge the relath*e barrier threat level.

176

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Jifoveffiber 2012 Figare 9 F 4: P\¥R Cantainmeat Integrity ar Bypass ExalBJ)les Inside Containment Auxiliary Building Open valve I I

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NEI 99 QI (RevisioR a) No=vember 2Ql2 10 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS 179

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NEI 99 Q1 (Revisiea a) Nevel3lber 2012 UNU8U1-\L SITE AREi"*... CENER\L 1-\LERT EVENT Kl\iERCENCY El\iERGENCY IIU7 Other H,-\7 Other HS7 Other HG7 Other conditions e1tist which coaditions exist conditioas e1cist conditioH:s e1cist v,rliich mthe jlidgment of the YJ:.rliich in the which in the in the jlidgment of the Emergency Director jlidgment of the jlidgment of the Emergeacy Director warrant declaration of Emergency Director Emergency Director \varrant declaration of a(NO)UE. warrant declaration v1a-rraat declaration a Geaeral Emergeacy. Op. },lodes: All ofaa Alert. of a gite Area Op. },Eades: All Op. }.ledes: AU Emergeacy. Op. Alofi.es: All HU1 ECL: Notification of Unusual Event Initiating Condition: Confirmed SECURITY CONDITION or threat. Operating Mode Applicability: All Emergency Action Levels: EJrnmple Emergeacy Actioa LC'.*els: (1 or 2 or 3) HUI.} ** A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the (site specific secmity shift supervisioa).DAEC Security Shift Supervision. HUl.2 Notification of a credible security threat directed at the siteDAEC. HUI .3 A validated notification from the NRC providing information of an aircraft threat. Definitions: SECURITY CONDITION: Any Security Event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A SECURITY CONDITION does not involve a HOSTILE ACTION. HOSTILE ACTION: An act toward DAEC or its persom1el that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land. or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the nuclear power plant. Non-

 -terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area).

181

NEI 99 QI (Revision 6) November 2Q12 SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including theincluding the ECCS. These systems are classified as safety-related. Basis: This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of IO CFR f-73. 71 or 10_-CFR-§-50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HAI, HSI and HGl. Timely and accurate communications between DAEC Security Shift Supervision and the Control Room is essential for proper classification of a security-related event. Classification of these events will initiate appropriate threat-related notifications to plant personnel and GR:Goffsite response organizations. 182

NEI 99 QI (RevisieR a) Ne:vember 2Q12 Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program]. 183

NEI 99 Ql (Revision 6) Novsmbsr 2012 EAL HUI .1 references (site speeifie seeurity shift supervisioH)DAEC Security Shift Supervision because these are the individuals trained to confirm that a security event is occurring or has occurred. Training on security event confirmation and classification is controlled due to the nature of Safeguards and 10 CFR § 2.39Q information. EAL HUI .2 addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with Abnormal Operating Procedure (AOP) 914, Security Events. :fsite-specific proeedure). EAL HUl.3 addresses the threat from the impact of an aircraft on the plant. The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may also be provided by NORAD through the NRC. Validation of the threat is performed in accordance with (site specific procedure) Abnormal Operating Procedure (AOP) 914, Security Events.-:- Emergency plans and implementing procedures are public documents; therefore, EALs should do not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information sholi:ld beis contained in non publie doewn.ents s1:1eh as the Security Plan. Escalation of the emergency classification level would be via IC HAI. Develaf)eF Nates: The (site speeifie se&H"ity shift supervisioaj is the title of the on slrift individual responsible for supervision of the on shift security force. The (site speeific procedure) is the proeedure(s) used by Control Room and/or Seeuri-ty personnel to detennine if a seeurity threat: is eredible, and to validate reeeipt of aireraft threat in.formation. , Emergeney plans and implementing procedures are public docwn.ents; therefore, EA.Ls sholi:ld not ineorpora-t:e Secmity sensitive information. This ineludes information that may be advantageous to a potential ad7;ersary, s1:1ch as the particulars concerning a specific threat: or threat location. Security sensitive information should be contained in non public documents such as the Security Plan. With due eonsideration given to the aboYe developer note, EfLLs may contain alpha or numbered references to seleeted events described in the Security Plan and associated implementing procedures. Such referooces should not contain a recognizable description of the event. For e1..0:1.nple, 0:1.1 EAL 1nay be v.,orded as "Security e7;ent #2, #5 or #9 is reported by the (site specifie security shift supervision)." EGL Assignment Attributes: *3 .1.1.A 184

NEI 99 QI (RevisioR 6) No:vember 2Ql2 HU2 ECL: Notification of Unusual Event Initiating Condition: Seismic event greater than OBE levels. Operating Mode Applicability: All Example Emergency Action Levels: HU2. l Seismic event greater than Operating Basis Earthquake (OBE) as indicated by-:- receipt of the Amber Operating Basis Earthquake Light and the wailing seismic alarm on Definitions: DESIGN BASIS EARTHQUAKE (DBE): A DBE is vibratory ground motion for which certain (generally, safety-related) structures, systems, and components must be designed to remain functional. OPERATING BASIS EARTHQUAKE (OBE): An QBE is vibratory ground motion for which those features of a nuclear power plant necessary for continued operation without undue risk to the health and safety of the public will remain functional. Basis: This IC addresses a seismic event that results in accelerations at the plant site greater than those specified for an Operating Basis Earthquake (OBE}1". An earthquake greater than an OBE but less than a Safe ShutdownDesign Basis Earthquake (.g.sgDBE)~ should have no significant impact on safety-related systems, structures and components; however, some time may be required for the plant staff to ascertain the actual post-event condition of the plant (e.g., performs walk-downs and post-event inspections). Given the time necessary to perform walk-downs and inspections, and fully understand any impacts, this event represents a potential degradation of the level of safety of the plant. Event verification with external sources should not be necessary during or following an OBE. Earthquakes of this magnitude should be readily felt by on-site personnel and recognized as a seismic event (e.g., typieal lateml aeeeleratioB:s are iB: fficeess of 0.08g). The Shift Manager or Emergency Director may seek external verification if deemed appropriate (e.g., a call to the

 + AR QBE is vibratory groYRd m.otioR for wh.ieh those features of a m1elsar pmver plaRt Reeessary for eoRtinued operatioR *N-ith.o:it U:Rdue risk to the health and safety of the pl:lblie *.vill rea1aiR fuaetioaal.
 ;; Aa SSE is vibratory gmund motioa for whieh eertain (geaerally, safety related) stmerures, systes1s, and eo.mpoaents nmst be desigaed to remaia fuaetioaal.

185

NEI 99 QI (RevisieH a) Ne1,zeH.10er 2012 USGS, check internet news sources, etc.); however, the verification action must not preclude a timely emergency declaration. OBE events are detected in accordance with AOP 901. The QBE is associated with a peak horizontal acceleration of+/- 0.06g. Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or -SMSAS. Develef)eF Netes: This "site specific indication that a seismic event met or exceeded QBE limits" should be based OH the indicatioHs, alanns and displays of site specific seismic monitoring equipment. Indicatioas described in the EAL should be limited to those that are immediately w;ailable to Coatrol Room personae! and vmich can be readily assessed. Indications available outside the Control Room aad/or vAtich require loogthy times to assess (e.g., processing of scratch plates or recorded data) shoald Bot be used. The goal is to specify iHdications that can be assessed w-ithin. 15 miB:utes of the actHal or S'l:1:spected seismic event. For sites that do not have readily assessable QBE indicatioas within the Control Room, dei,,elopers should use the followin.g alteraate EAL (or similar wordin.g). (1) a. Control Room personnel feel an actual or potootial seismic evoot.

b. The occurren.ce of a seismic e11en.t is coafirmed in H:u1:aner deemed BfJpropriate by the Shift Manager or Emergency Director.

The E,'\L l .b statemoot is in.eluded to en.sure that a deelaratioa does n.ot restllt from felt vibratioas eatlsed by a aon. seismic source (e.g., a dropped heavy load). The Shift Manager or Emergoocy Director may seek ffiEtemal verificatioa if deemed appropriate (e.g., a call to the USGS, check intern.et aev,zs sources, etc.); hmvever, the verification action must aot preclude a timely emergency declaratioa. It is recognized that this alternate EAL wording may cause a site to declare an UBusual Even.t while another site, similarly affected but *.vith readily assessable QBE mdications iB: the Control Room, may 110t. The above akemate wording may also be tlsed to develop a cmnpeasatory EAL for use during periods *.vhen. a seismic 1non.itoring system capable of detecting an QBE is Otlt of service for mamten.aiice or repair. EGL Assignmoot A.ttribHtes: 3 .1.1.A 186

NEI 99 QI (RevisieR ~) l>+0Y0FFIB0f 2012 HU3 ECL: Notification of Unusual Event Initiating Condition: Hazardous event~ Operating Mode Applicability: All Emergency Action Levels: Example Emergency f ...etion Levels: (1 or 2 or 3 or 4 or 5 or 6) Note: EAL HU3.4 does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents. HU3 .1 A tornado strike within the PROTECTED AREA. HU3 .2 Internal room or area flooding of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component needed for the current operating mode. HU3 .3 Movement of personnel within the PROTECTED AREA is impeded due to an offsite

         . event involving hazardous materials (e.g., an offsite chemjcal spill or toxic gas release).

HU3 .4 A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles . HU3.5 (Site specific list of natm=al or technological hazard events)6 Definitions: PROTECTED AREA: The area under continuous access monitoring and control, and armed protection as described in the site Security Plan. SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECC8. Systems classified as safety related. Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant. EAL HU3.l addresses a tornado striking (touching down) within the Protected Area. EAL HU3 .2 addresses flooding of a building room or area that results in operators isolating power to a SAFETY SYSTEM component due to water level or other wetting concerns. Classification is also required if the water level or related wetting causes an automatic isolation of a SAFETY SYSTEM component from its power source (e.g., a breaker or relay trip). To 187

NEI 99 0 I (RevisieR a) Nevemeer 2012 warrant classification, operability of the affected component must be required by Technical Specifications for the current operating mode. EAL HU3.3 addresses a hazardous materials event originating at an offsite location and of sufficient magnitude to impede the movement of personnel within the PROTECTED AREA. 188

NEI 99 0 I (Revision 6) Noyeml:Jer 2012 EAL HU3.4 addresses a hazardous everit that causes an on-site impediment to vehicle movement and significant enough to prohibit the plant staff from accessing the site using personal vehicles. Examples of such an event include site flooding caused by a hurricane, heavy rains, up-river water releases, dam failure, etc., or an on-site train derailment blocking the access road. This EAL is not intended apply to routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the flooding around the Cooper Station during the Midwest floods of 1993, or the flooding around Ft. Calhoun Station in 2011. EAL H!zl:.5 addresses (site specific description). . Escalation of the emergency classification level would be based on ICs in Recognition Categories AR, F, Sor C. DevelafJeF Netes: The "Site specific list of natural or technological hazard events" should include other C'1ents that may be a precursor to a more significant e vent or condition, and that are appropriate to the site 0 location and characteristics. Notwithstanding the events specifically incb~ded as EALs above, a "gite specific list of natural or technological hazard C'tents" need not include short lived events for which the mctent of the damage and the resulting consequences can be determined ;.vithin a relatively short time frame. In these cases, a damage assessment can be performed soon after the C'v'ent, and the plant staff will be able to identify potential or actual impacts to plant systen:1:s and struchlres. This 1,:vill ,,enable prompt definition and implementation of compensatory or corrective measures with no appreciable increase in risk to the public. To the extent that a short lived event does cause immediate and significant damage to plant systems and struchlres, it 1.vill be classifiable under the Recognition Category F, g and C ICs and EALs. Events oflesser impact v;ould be expected to cause only small and localized damage. The consequences from these types of events are adequately assessed and addressed in accordance 1vvith Technical Specifications. In addition, the occurrence or effects of the C'v'ent , may be reportable under the requirements of 10 CFR 50.72. EGL A.ssignment Attributes: 3.1.1.A and 3.1.1.C 189

NEI 99 Ql (Revisiea a) Nevember 2QI2 HU4 ECL: Notification of Unusual Event Initiating Condition: FIRE potentially degrading the level of safety of the plant. Operating Mode Applicability: All Emergency Action Levels: Exnmple EmeFgeney }*... etian Levels: (I or 2 or 3 or 4) Note1:

• The Emergency Director should declare the Unusual Eventevent promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

HU4.l a. A FIRE is NOT extinguished within 15-minutes of ANY of the following FIRE detection indications:

• Report from the field (i.e., visual observation)
• Receipt of multiple (more than 1) fire alarms or indications
• Field verification of a single fire alarm AND

b. The FIRE is located within ANY of the follov.4ngTable H-1 plant rooms or areas-:-

(site specific list of plant rooms or areas) HU4.2 a. ,.

• Receipt of a single fire alarm ~with no other indications of a FIREj.

              ----AND

b. The FIRE is located within ANY of the followingTable H-1 plant rooms or areas (site specific list of plant rooms or areas)

              ---AND

c. The existence of a FIRE is not verified within 30-minutes of alarm receipt.

HU4.3 A FIRE within the plant or ISFSI Fferpl,f{nfs with In IS.F'S! e1:1.t8ide thepmnt PffJtected A-roof-PROTECTED AREA not extinguished within 60-minutes of the initial report, alarm or indication. HU4.4 A FIRE within the plant or ISFSI Fferp/.ants with l1i !SFS! eutside theplemt Pretected A-roof-PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish. Table H-1 Fire Areas 190

NEI 99 Ql (Re1:isiea a) Ne:vember 2Ql2

• 1031 DO and Day Tank Rooms~
• I 021 DO and Day Tank Rooms,
• Battery Room~
• Essential Switchgear Rooms,
• Cable Spreading Room
• Torus Room
• ,Intake Structure;
• Pumphouse
• Drywelk
• Torus
• NE, NW. SE Comer.Rooms,
• HPCIRoom;:
• RCICRoom,
• RHR Valve Room,
• North CRD Area,
• South CRD Area~
• CSTs
• Control Building,
• Remote Shutdown Panel I C3 88 Area,
• Panel 1C55/56 Area,
• SBOTRoom 191

NEI 99 QI (ReYisisH e) Nsvel'l'H3& 2Ql2 Definitions: FIRE: Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed. PROTECTED AREA: The area under continuous access monitoring and control, and armed protection as described in the site Security Plan. Basis: This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant. EALHU4.l The intent of the 15-minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., smoldering waste paper basket). In addition to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation of a suppression system, etc. Upon receipt, operators will take prompt actions to confirm the validity of an initial fire alarm, indication, or report. For EAL assessment purposes, the emergency declaration clock starts at the time that the initial alarm, indication, or report was received, and not the time that a subsequent verification action was performed. Similarly, the fire duration clock also starts at the time of receipt of the initial alarm, indication or report. EALHU4.2 This EAL addresses receipt of a single fire alarm, and the existence of a FIRE is not verified (i.e., proved or disproved) within 30-minutes of the alarm. Upon receipt, operators will take prompt actions to confirm the validity of a single fire alarm. For EAL assessment purposes, the 30-minute clock starts at the time that the initial alarm was received, and not the time that a subsequent verification action was performed. A single fire alarm, absent other indication( s) of a FIRE, may be indicative of equipment failure or a spurious activation, and not an actual FIRE. For this reason, additional time is allowed to verify the validity of the alarm. The 30-minute period is a reasonable amount of time to determine if an actual FIRE exists; however, after that time, and absent information to the contrary, it is assumed that an actual FIRE is in progress. 192

NEI 99 QI (Revisiea ~) Nev@mber 2Ql2 If an actual FIRE is verified by a report from the field, then EAL HU4. l is immediately applicable, and the emergency must be declared if the FIRE is not extinguished within 15-minutes of the report. If the alarm is verified to be due to an equipment failure or a spurious activation, and this verification occurs within 30-minutes of the receipt of the alarm, then this EAL is not applicable and no emergency declaration is warranted. EALHU4.3 In addition to a FIRE addressed by EAL HU4. l or EAL HU4.2, a FIRE within the plant or ISFSI PROTECTED AREA not extinguished within 60-minutes may also potentially degrade the level of plant safety. This basis extends to a FIRE occurring within the PROTECTED AREA of an ISFSI located outside the plant PROTECTED AREA. [Sentence for plants with an I8F8I outside the plant Protected Area] EALHU4.4 If a FIRE within the plant or ISFSI [forplB:nts with en f,STS! eutsid-e theplemt PretecledAreB:] PROTECTED AREA is of sufficient size to require a response by an offsite firefighting agency (e.g., a local town Fire Department), then the level of plant safety is potentially degraded. The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish. Declaration is not necessary if the agency resources are placed on stand-by, or supporting post-extinguishment recovery or investigation actions. Basis-Related Requirements from Appendix Rand NFPA-805 Criterion 3 of Appendix A to 10 CFR 50 states in part that "strnctures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions." The Nuclear Safety Goal ("NSG") in NFPA 805, Section 1.3.1 states, "The nuclear safety goal is to provide reasonable assurance that a fire during any operational mode and plant configuration will not prevent the plant from achieving and maintaining the fuel in a safe and stable condition." When considering the effects of fire, those systems associated with achieving and maintaining safe shutdown conditions assume major importance because a safe shutdown success path, free of fire damage, must be available to meet the nuclear safety goals, objectives and performance criteria for a fire under any plant operational mode or configuration. Because fire may affect safe shutdown systems and because the loss of function of systems used to mitigate the consequences of design basis accidents under post-fire conditions does notper se impact public safety, the need to limit fire damage to systems required to achieve and maintain safe shutdown conditions is greater than the need to limit fire damage to those systems required to mitigate the consequences of design basis accidents. 193

NEI 99 g I (Re0visioa 6) No1;ember 2912 In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of 1-hour fire barriers for the enclosure of cable and equipment and associated non-safety circuits of one redundant train (G.2.c). Even though DAEC has adopted the alternate approach provided by NFPA-805 in lieu of the deterministic requirements of Appendix R, the 30-minutes to verify a single alarm as used in EAL HU4.2 is considered a reasonable amount of time to determine if an actual FIRE exists without presenting a challenge to the nuclear safety perfonnance criteria.Basis Related Reguirements from Aupendix R Appendi1, R to 10 CFR 50, states in part: Criterion 3 ofAppoodiR A to this part specifies that "Structures, systems, and componoots important to safety shall be designed and located to minimize, consistent with other safety re(ftliremeats, the probability and effect of fires and e1.plosions." Vlhen coasideriug the effects of fire, those systems associated with achieving and maintaining safe shutdown conditions assume major importance to safety because damage to them can lead to core damage resulting from loss of coolant through boil off. Because fire may affect safe shutdowa systems and because the loss of ftm.etion of systems used to mitigate the conse(ftlences of design basis accideats under post fire coaditions does not per se impact public safety, the need to limit fire damage to systems required to achi011e aad.maintain safe shutdov111 conditions is greater than the need to limit fire damage to those systems required to mitigate the consequences of design basis accidents. In addition, A.ppoodix R to 10 CFR 50, requires, among other considerations, the use of !"hour fire barriers for the enclosure of cable and equipmeat and associated non safety circuits of one redundant train (G.2.c). A.s used in EAL #2, the 30 IBifltltes to verify a single alarFH is 1.vell within this 1,vorst case 1 hour time period.. Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9SA8. Dev:ela11er Notes: The "site specific list of plant rooms or areas" should specify those rooms or areas that contain SAFETY SYSTEM e(ftlipment. As noted in the EA.Ls aH:d Basis section, include the term ISF8I if the site has an ISF8I outside the plant Protected Area. EGL Assignment Attributes: 3 .1.1.A 194

NEI 99 QI (RsvisieH G) Ne 1,rembsr 2912 HU76 ECL: Notification of Unusual Event - Initiating Condition: Other conditions exist which in the judgment of the Emergency Director warrant declaration of a fNU)UE. Operating Mode Applicability: All E:xemlile Emergency Action Levels: HU-76 .1 Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of SAFETY SYSTEMS safety systems occurs. Definitions: SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the EGGS. Systems classified as safety related. Basis: This IC addresses 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 classification level description for a NODE. 195

NEI 99 QI (RevisieB a) Nevember 2912 HA1 ECL: Alert Initiating Condition: HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes. Operating Mode Applicability: All Example Emergency Action Levels.:.: (1 or 2) HALI A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the (site speoifie seemity shift se,pervision)DAEC Security Shift Supervision. HAf.2 A validated notification from NRC of an aircraft attack threat within 30 minutes of the site. Definitions: HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station. HOSTILE ACTION: An act toward DAEC or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air. land, or water using guns, explosives, PROJECTILES, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the nuclear power plant. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). OWNER CONTROLLED AREA: The site property owned by or otherwise under the control of the licensee. PROJECTILE: An object directed toward a nuclear power plant that could cause concern for its continued operability, reliability, or personnel safety. Basis: This IC addresses the occurrence of a HOSTILE ACTION within the OWNER CONTROLLED AREA or notification of an aircraft attack threat. This event will require rapid response and assistance due to the possibility of the attack progressing to the PROTECTED AREA, or the need to prepare the plant and staff for a potential aircraft impact. Timely and accurate communications between DAEC Security Shift Supervision and the Control Room is essential for proper classification of a security-related event. Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program]. 196

NEI 99 QI (Rsvisien 6) Nevsrnber 2012 As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Alert declaration will also heighten the awareness of Offsite Response Organizations, allowing them to be better prepared should it be necessary to consider further actions. This IC does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFRf-73.71 or 10 CFRf-50.72. EAL HAI .1 is applicable for any HOSTILE ACTION occurring, or that has occurred, in the OWNER CONTROLLED AREA. This includes any action directed against ffll.-the ISFSI ~ which is located outside the plant PROTECTED AREA. EAL HA 1.2 addresses the threat from the impact of an aircraft on the plant, and the anticipated arrival time is within 30 minutes. The intent ofthis EAL is to ensure that threat-related notifications are made in a timely manner so that plant personnel and GR:Goffsite response organizations are in a heightened state of readiness. This EAL is met when the threat-related information has been validated in accordance with *(Abnormal Operating Procedure (AOP) 914, Security Events site specific procedure).s. The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may be provided by NORAD through the NRC. In some cases, it may not' be readily apparent if an aircraft impact within the O~'ER CONTROLLED AREA was intentional (i.e., a HOSTILE ACTION). It is expected, although not certain, that notification by an appropriate Federal agency to the site would clarify this point. In this case, the appropriate federal agency is intended to be NORAD, FBI, FAA or NRC. The emergency declaration, including one based on other ICs/EALs, should nof be unduly delayed while awaiting notification by a Federal agency. Emergency plans and implementing procedures are public documents; therefore, EALs should do not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be is contained in non public documents such as the Security Plan. Escalation of the emergency classification level would be via IC HS 1. Devel0peF Notes: The (site specific security shift supervision) is the title of the on shift individual responsible for supervision of the on shift security force. Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security sensitive infonnation. This includes infonnation that may be advantageous to a potential adversary, such as the particulars concerning a specific tlKeat or threat location. Security sensitive information should be contained in non public documents such as the Security Plan. 197

\0 00

NEI 99 Q1 (Revisiea 6) Nevember 2Ql2 HA6HA5 ECL: Alert Initiating Condition: Control Room evacuation resulting in transfer of plant control to alternate locations. Operating Mode Applicability: All Example Emergency Action Level§: HA65 .1 An event has resulted in plant control being transferred from the Control Room to fsite-SJ:Jeeifie rernote sB-1:HdoW'H: panels and loeal eontrol stations)the Remote Shutdown Panel (1C388). Definitions: Basis: This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations outside the Control Room. -The loss of the ability to control the plant from the Control Room is considered to be a potential substantial degradation in the level of plant safety. Following a Control Room evacuation, control of the plant will be transferred to alternate shutdown locations. The necessity to control a plant shutdown from outside the Control Room, irl addition to responding to the event that required the evacuation of the Control Room, will present challenges to plant operators and other on-shift personnel. Activation of the ERO and emergency response facilities will assist in responding to these challenges. Escalation of the emergency classification level would be via IC HS61. Developer Nates: The "site SJ:Jeeifie remote slHttdmvn J:Janels and loeal eontrol stations" are the panels and eontrol stations refereneed in J:Jlant J:Jroeed:eres ased to eooldo*.vn and shatdovli'l: the J:)lant from a loeation(s) oatside the Control Room. EGL A.ssignmentAttributes: 3.1.2.B 202

NEI 99 QI (RsvisieR e) Nevsm.bsr 2912 HA7HA6 ECL: Alert Initiating Condition: Other conditions exist which in the judgment of the Emergency Director warrant declaration of an Alert. Operating Mode Applicability: All Example EmeFgeBeyEmergency Action Levels: HA+6. I Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels. Definitions: HOSTILE ACTION: An act toward DAEC or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, *explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be constrned to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the nuclear power plant. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station. PROJECTILE: An object directed toward a nuclear power plant that could cause concern for its continued operability, reliability, or personnel safety. Basis: This IC addresses 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 classification level description for an Alert. 203

NEI 99 QI (Revisiea (i) NeYember 2Ql2 HS1 ECL: Site Area Emergency Initiating Condition: HOSTILE ACTION within the PROTECTED AREA. Operating Mode Applicability: All Exa1BJ3le EmeFgeReyEmergency Action Levels: HSl.l A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the (site specific security shift supervision)DAEC Security Shift Supervision. Definitions: HOSTILE ACTION: An act toward DAEC or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES. and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the nuclear power plant. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between indi'viduals in the owner controlled area). HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station. i' .* HOSTILE FORCE: One or more individuals wh~ are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction. PROJECTILE: An object directed toward a nuclear power plant that could cause concern for its continued operability, reliability, or personnel safety. PROTECTED AREA: The area under continuous access monitoring and control, and armed protection as described in the site Security Plan. Basis: This IC addresses the occurrence of a HOSTILE ACTION within the PROTECTED AREA. This event will require rapid response and assistance due to the possibility for damage to plant equipment. Timely and accurate communications between DAEC Security Shift Supervision and the Control Room is essential for proper classification of a security-related event. Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program]. 204

NEI 99 QI (Revisisa 0) Nsvember 2Ql2 As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Site Area Emergency declaration will mobilize GR-Ooffsite response organization resources and have them available to develop and implement public protective actions in the unlikely event that the attack is successful in impairing multiple safety functions. This IC does not apply to a HOSTILE ACTION directed at fill-the ISFSI PROTECTED AREA which is located outside the plant PROTECTED AREA; such an attack should be assessed using IC HAI. It also does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements oflO CFR,§-73.71 or 10 CFR,§-50.72. Emergency plans and implementing procedures are public documents; therefore, EALs sholild do not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information shuald beis contained in H:OH: publie docl.ffllents sueh as the Security Plan. Escalation of the emergency classification level would be via IC HG 1. De'lelapeF Netes: The (site Sj)eeifie seeurity shift supervision) is the title of the on shift individual respoH:sible for S1.TI3ervision of the OH: shift security force. " Emergency plans aad implementing proeedures are publie documents; therefore, EA.Ls shoald not ineorporate Seemity seH:sitive information. This ineludes iH:fonnatioH: that may be advantageous to a potential adversary, sueh as the particalars concerning a specific threat or threat loeation. Seearity sensitive informatioH: should be eontained in. ROH: pablie docmnents such as the Security Plan. With due eonsideration given to the above developer note, EALs may contain alpha or numbered referenees to selected events described in the Sec'Hfity Plan and associated implementing procedures. Sach references sholild not contain a recognizable deseription. of the. event. For example, aH: EAL may be worded as "Security event #2, #5 or #9 is reported by the (site specifie seeurity shift supervision)." See the related Developer Note iH: Appendi1E B, Definitions, for gaidance on the development of a scheme definitioH: for the PROTECTED ,A.REA. EGL Assignment Attributes: 3.1.3.D 205

NEI 99 01 (RevisiaR a) NaveFBl:ier 2012 HS6HS5 ECL: Site Area Emergency Initiating Condition: Inability to control a key safety function from outside the Control Room. Operating Mode Applicability: All Exafflf)le Emergency Action Levels: Note: The Emergency Director should declare the Site Area Ernergencyevent promptly upon determining that (site specific number the applicable timeo@rninutes) has been exceeded, or will likely be exceeded. HS5~.l a. An event has resulted in plant control being transferred from the Control Room to (site specific remote shutdovm panels and_control stations) the Remote Shutdown Panel (1C388). AND

b. Control of ANY of the following key safety functions is not reestablished within (site specific number of20 minutesj.

• Reactivity control
• Core coolmg [PWR] I RPV water level [BWR]
• RCS heat removal Definitions:

Basis: This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations, and the control of a key safety function cannot be reestablished in a timely manner. The failure to gain control of a key safety function following a transfer of plant control to alternate locations is a precursor to a challenge to one or more fission product barriers within a relatively short period of time. The determination of whether or not "control" is established at the remote safe shutdown location(s)Remote Shutdown Panel {1C388-isl.i§. based on Emergency Director judgment. The Emergency Director is expected to make a reasonable, informed judgment within (the site specific time for transfer) ~20 minutes whether or not the operating staff has control of key safety functions from the remote safe shutdown location(s). AOP 915, "Shutdown Outside Control Room" provides the following CAUTION - "For Control Room evacuation as the result o{a fire, transfer o{control at panels JC388, JC389, JC390, JC391, 1C392and JC392 is required to be completed within 20 minutes." Escalation of the emergency classification level would be via IC FG 1 or CG 1. 206

NEI 99 0 I (Revisiaa e) Navember 2012 HS7HS6 ECL: Site Area Emergency Initiating Condition: Other conditions exist which in the judgment of the Emergency Director warrant declaration of a Site Area Emergency. Operating Mode Applicability: All Example Emergency Action Level~: HS.'.76.1 Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts, (I) toward site personnel or equipment that could lead to the likely failure of or, (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the site boundary. Definitions: HOSTILE ACTION: An act toward DAEC or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes*attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious,acts that are not part of a concerted attack on the nuclear power plant. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station. PROJECTILE: An object directed toward a nuclear power plant that could cause concern for its continued operability, reliability, or personnel safety. Basis: This IC addresses 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 classification level description for a Site Area Emergency. 208

NEI 99 Ql (RevisieH G) Nevember 2012 HG1 ECL: General Emergency Initiating Condition: HOSTILE ACTION resulting in loss of physical control of the facility. Operating Mode Applicability: All Example Emergency Action Levels: HGl.l a. A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the (site specific security shift supervision)DAEC Security Shift Supervision. AND

b. EITHER of the following has occurred:

1. ANY of the following safety functions cannot be controlled or maintained.

• Reactivity control
• Core cooling [PWR] / RPV water level [BJ47R]
• RCS heat removal OR

2. Damage to spent fuel has occurred or is IMMINENT.

Definitions: HOSTILE ACTION: An act toward DAEC or its personnel that includes the use of violent force to destroy eguipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILES, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be constmed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the nuclear power plant. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). HOSTAGE: 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. IMMINENT: The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. 209

NEI 99 Q1 (ReYision 6) Nov@Hlb@r 2Ql2 PROJECTILE: An object directed toward a nuclear power plant that could cause concern for its continued operability, reliability, or personnel safety. PROTECTED AREA: The area under continuous access monitoring and control, and armed protection as described in the site Security Plan. 210

N

NEI 99 Ql (RevisiaH (ci)

                                                                                      }T0Ye1Bl:ier 2Ql2 This IC addresses an event in which a HOSTILE FORCE has taken physical control of the facility to the extent that the plant staff can no longer operate equipment necessary to maintain key safety functions. It also addresses a HOSTILE ACTION leading to a loss of physical control that results in actual or IMMINENT damage to spent fuel due to 1) damage to a spent fuel pool cooling system (e.g., pumps, heat exchangers, controls, etc.) or, 2) loss of spent fuel pool integrity such that sufficient water level cannot be maintained.

Timely and accurate communications between the DAEC Security Shift Supervision and the Control Room is essential for proper classification of a security-related event. Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program]. Emergency plans and implementing procedures are public documents; therefore, EALs should do not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information shmild beis contained in BOfl: pablie doetH:HeB-1:s such as the Security Plan. 212

NEI 99 Q1 (Revisioa 6) November 2012 HG7HG6 ECL: General Emergency Initiating Condition: Other conditions exist which in the judgment of the Emergency Director warrant declaration of a General Emergency. Operating Mode Applicability: All Exam~le Emergency Action Levels: HG-16.1 Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area. Definitions: HOSTILE ACTION: An act toward DAEC or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs,. vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the nuclear power plant. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). HOST AGE: A person(s) held as leverage against the station to ensure that demands will be met by the station. IMMINENT: The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. PROJECTILE: An object directed toward a nuclear power plant that could cause concern for its continued operability, reliability, or personnel safety. Basis: This IC addresses 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 classification level description for a General Emergency. 214

                                                                                                           }ffil 99 Ol (R-evisi0H 6)

Nevember 2012 11 SYSTEM MALFUNCTION ICS/EALS Tahle S 1: Reeegeitien CategeFY "S" Initiating Cenaitien MatFix UNUSUAL SITE AREA GENERAL ALERT EVENT EMERGENCY EMERGENCY SUl Loss of all offsite SA..1 Loss of all but oRe SS1 Loss of all offsite SC 1 Prolonged loss of AC power eapaaility to AC power sol:H'ee to and all onsite AC p01,ver to all offsite anEl all oesite AC emergeae:y bt1ses for 15 emergency b1:1ses for 15 emergeacy bt1ses for 15 130vrer to emergeeey b1:1ses. 1B:lat1tes or loRger. mi-Bt1tes or loRger. R'Hil.1:1tes or loRger. Op. },1edes: l, 2, 3, 4Power Op. },1edes: 1. 2, 3, 4Pewer Op. },le6fC5*: L...l.....l.. Op. ,~1edes: l. 2. 3. 4Pewer Operatien, St6lr.'up, Het Openrlien, Slart,,1p, He! 1:?ewer Opert1tie11, Operation, StarlilfJ, Hel Standby, He.' Shul6f.ew19 Slandh:)', He! f.l!ntl6/o ,vt1 1 Startup, Het Stand-hr, Het Standhy, Hat Shutdown Shu!d-ewn SU2 illlPLANNED loss SA..2 U}JPLMJ:NED of Control Rooa1 loss of CoBtrol Room iaElieations for 15 llliButes indieatioes fer 15 a1in1:1tes or Iaeger. or longer *.vith a significaat Op. ,~1edes: Pewer traBsieat ia 13rogress. Open1tie11, Startup, Hel Op. ,~1edes: 1. 2. 3. 4 Standby*, Het SJu,116/ownL...l.,_ Pewe1* Operahen, Starl'tlfJ, 3, 4 Het Standby*, Het Slmtdewn SUJ Reaetor eoolaRt activity greater thaB Teehaical Speeificatioa allowable limits. Op. },1edes: l. 2, 3, 4Pewer Operatien, Startup, &t S.ta,i1dhy, Het Shu!d-ewn SU4 RCS leakage for 15 miBt1tes or loRger. Op. },1edes: 1, 2, 3, 4Pewer Operahen, Slartup, Het Standby*, Het Shuldown SUS Al:ltmnatic or S1A...S A1:1tomatic or SSS Ina-bility to rn:an1:1al (trip [PWR] / maHual (trip [PWR] / shatdov,rn the reactor scram [BWR]) fails to scram [BWR]) fails to ca1:1sieg a challenge to shutdov,n the reactor. shutdo1>vn the reactor, (core cooling [PWR] I Op. },fed-es: and subsequeHt ma.Hua! RPV water level [E WR]) P-ev,*e,* Operl#enl actioas ta:lcen at the or RC£ heat remoYal. reactor control consoles are eot successful in Opera.'ion] Op. Medes: Power 1 Table inteRded for use by *I I shutting down the

• EAL developers. 1 reactor. I ff l . . l' I I B.0 l:IS10B l:Bl0ef1S00 1 I d . . d Op. },fed-es: Pewer , o Ol:lffi eRts 1s Bet reqm.re . I 1 OperBtienl_ L------------------*

215

t. NEI 99 01 (RevisioH fi) NoveHlber 2012 SU1 ECL: Notification of Unusual Event Initiating Condition: Loss of all-ALL offsite AC power capability to emergency essential buses for 15_-minutes or longer. Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdovm.L..LJ Example Emergency Action Levels: Note: The Emergency Director should declare the Unusual E~vent promptly upon determining that the applicable time 15 minutes has been exceeded, or will likely be exceeded. SUI.I Loss of ALL offsite AC power capability to (site specific emergency bl:lses) 1A3 AND 1A4 buses for 15 minutes or longer. Definitions: Basis: This IC addresses a prolonged loss of offsite power. The loss of offsite power sources renders the plant more vulnerable to a complete loss of power to AC emergency essential buses- . .,_This condition represents a potential reduction in the level of safety of the plant. The intent of this EAL is to declare an- Notification of Unusual Event when offsite power has been lost and both of the emergency diesel generators have successfully started and energized their respective 4kv essential bus. For emergency classification purposes, "capability" means that an offsite AC power source(s) is available to the emergency essential buses, whether or not the buses are powered from it. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of offsite power. Escalation of the emergency classification level would be via IC SAL De"irelapeF Netes: The "site specific emergency b.1ses" are the bl:lses fed by offsite or emergency AC povter sources that supply power to the electrical distribution system that pov;ers SAFETY SYSTEMS. There is typically I emergency bus per train of SAFETY SYSTEMS. At multi UBit stations, the EALs may credit compensatory measures that are proced-urali.2:ed and can be implemented *within 15 minutes. Consider capabilities such as povter source cross ties, "swing" generators, other power so1H:"ces described in abnormal or emergency operating procedl:tfes, ete. Plants that ha-ve a proeedl:tfali.2:ed capability to supply offsite AC power to an 217

N 00

                                                                                                           "NEI 99 01 (R-evision 6)

November 2012 SU2SU3 ECL: Notification of Unusual Event Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer. Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot ShutdovmLJ.,_]_ Example Emergency Action Levels: Note: The Emergency Director should declare the Unusual E~vent promptly upon determining that the applicable time 15 minutes has been exceeded, or will likely be exceeded. SU3.l  ;+a.--An UNPLANNED event results in the inability to monitor one or more of the Table S-1 parameters from within the Control Room for 15 minutes or longer.AH UNPb'\NNED event results in the inability to monitor one or more of the following parameters from vtithin the Control Room for 15 minutes or longer. Reactor Power RPV ¥later Level RPV Pressure Primary GontainmeHt Pressure Suppression Pool Level

• Suppression Pool Temperature Suppression Pool Temperature Reactor power RPV Water Level RPV Pressure
• Primary Containment Pressure t
• Suppression Pool Level

                   ~L*
                   ;}J Suppression Pool Temperature
                       ,~~~:""~~Rn~,,1:,'p , ~ - : ' * , ' . * , ~ * , *r** c ~* *,i,t,.;t~*.,.:,,{!i,,,,,

Definitions: SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are 219

NEI 99 0 l (Revisiea a) Ne 1,rember 2012 classified as safety-related.A system regaired for safe plant operation, eooling dovm the plant aad-/or plaeing it in the eold shutdovm eondition, inernding the ECC8. Systems elassified- as safety related. UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: This IC addresses the difficulty associated with monitoring normal plant conditions without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition is a precursor to a more significant event and represents a potential degradation in the level of safety of the plant. As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). _For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room. 220

NEI 99 01 (Re,,risioH '3) November 2012 An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making. This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling [PWR] I RPV level [B WR] and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for reactor vessel level [PWR1 / RPV water level [BVlR] cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication. Escalation of the emergency classification level would be via IC SA-2.J_. Devel0per Nates: In the PV.Z:R parameter list column, the "site specific number" should reflect the 1ninimum number of steam generators necessary for plant cooldovm and shutdmvn. This criterion may also specify whether the level value should be 1.vide range, narrov,r range or both, depending upon the monitormg requiremen-ts memergency operati-Bg procedures. ,, DEY,'elopers may specify Gither pressurizer or reactor vessel level in the PWR pararn9ter column entry for RCS Level. The IllHHber, type, location and layout of Control Room iadications, and tac range of possible failure modes, can challenge the ability of an operator to accurately determine, within the time period available for emergency classification assessments, ifa specific percentage of indications have been lost. The approach used in this EAL facilitates prompt and accurate emergency classification assessments by focusing on the indications for a selected subset of parameters. By focusing on the availability of the specified parameter values, instead of the sources of those values, the EAL recognizes and accommodates the vlide variety of indications in nuclear pmver plant Control Rooms. Indication types and sources may be analog or digital, safety related or not, primary or alternate, individual meter value or computer group display, etc. A loss of plant annunciators '.vill be evaluated for reportability in accordance ',vith 10 CFR 50.72 (and the associated guidance in NUREG 1022), and reported if it significantly impairs tho capability to perform emergency assessments. Compensatory measures for a loss of am1unciation can be readily implemented and may inch1de increased monitoring of main control boards and more freq-aent plant roands by non licensed operators. Their alerting function notwithstanding, annunciators do not provide the parameter valaes or specific component status information used to operate the plant, or process through A.OPs or EOPs. Based on these considerations, a loss of annunciation is considered to be adequately addressed by reportability criteria, and therefore not included in tais IC and EAL. 221

N N N

NEI 99 Q1 (R-evisiea <:i) N0vea1ber 2Ql2 SU3SU4 ECL: Notification of Unusual Event Initiating Condition: Reactor coolant activity greater than Technical Specification allowable limits. Operating Mode Applicability: Pov1er Operation, Startup, Hot Standby, Hot Shutdovm.L..1.,_]_ Example Emergency Action Levels: (1 or 2) SU4.l (Site specific rndiatioa HJ.on-itor) reading greater than (site specific vahle). Pretreatment Offgas System (RM-4104) Hi-Hi Radiation Alarm. SU4. 2 Sample analysis indicates that reactor coolant specific activity is greater than 2.0 µCi/gm dose equivalent I-131 for 12 hours or longerSample analysis indicates that a reactor coolant activity value is greater than an allowable limit specified in Technical Specifications.,::* Definitions: Basis: This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant. For EAL SU4. l, 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. For EAL SU4.2, coolant samples exceeding the 2.0 µCi/gm dose equivalent I-131concentration require prompt action by DAEC Technical Specifications and are representative of minor fuel cladding degradation. Escalation of the emergency classification level would be via I Cs FA 1 or the Recognition Category A-R ICs. DevelepeF Netes: For EAL #1 Enter the radiation ffionitor(s) that may be Bsed to readily identify when RCS activity le:r.*els exceed Technical Specification allowable limits. This El\L may be developed using different HJ.ethods and sites shoald Bse existing capabilities to address it (e.g., development of ne\.v capabilities is not required). E1camples of CJcisting H1:ethodslcapabilities incllide:

• An installed radiation monitor on the letdov,'fl Sj'Stem or air ejector.
• iA.. hand held ffionitor or deployed detector reading *.vith pre calculated conversion values or readily impleHJ.entable conversion calculation capability.

223

NEI 99 Q1 (R-evisieH a) Ne,,ember 2012 SU4SU5 ECL: Notification of Unusual Event Initiating Condition: RCS leakage for 15 minutes or longer. Operating Mode Applicability: Pov:er Operation, Startup, Hot Standby, Hot 8hutdownLl.,_1 Exemf)le Emergency Action Levels: (1 or 2 or 3) Note: The Emergency Director should declare the Unusual E~vent promptly upon determining that the applicable time 15 minutes has been exceeded, or will likely be exceeded. SU5.l RCS unidentified or pressure boundary leakage greater than (site specific value)lO gpm for 15 minutes or longer. SU5.2 RCS identified leakage greater than (site specific value)25 gpm for 15 minutes or longer. SU5.3 Leakage from the RCS to a location outside containment greater than 25 gpm for 15 minutes or longer. Definitions: UNISOLABLE: An open or breached system line that cannot be isolated, remotely or locally. Basis: This IC addresses RCS leakage which may be a precursor to a more significant event. In this case, RCS leakage has been detected and operators, following applica~le procedures, have been unable to promptly isolate the leak. This condition is considered to be a potential degradation of the level of safety of the plant. EAL SU5.l and EAL SU5.2 are focused on a loss of mass from the RCS due to "unidentified leakage", "pressure boundary leakage" or "identified leakage" (as these leakage types are defined in the plant Technical Specifications). EAL SU5.3 addresses a RCS mass loss caused by an UNISOLABLE leak through an interfacing system. These EALs thus apply to leakage into the containment, a secondary-side system-fe;g:, steam generator t1:tbe leakage in a P'NR) or a location outside of containment.

 ,The leak rate values for each EAL were selected because they are usually observable with normal Control Room indications. Lesser values typically require time-consuming calculations to determine (e.g., a mass balance calculation). EAL SU5.l uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage.

225

NEI 99 Q1 (R-evisiea {i) Nevember 2012 The release of mass from the RCS due to the as-designed/expected operation of a relief valve does not warrant an emergency classification. For PVi'Rs, an emergency classification *.vould be required if a mass loss is caused by a relief valve that is not functioning as designed/expected (e.g., a relief valve sticks open and the line flow cannot be isolated). For B'NRs, aA stuck-open Safety Relief Valve (SRV) or SRV leakage is not considered either identified or unidentified leakage by Technical Specifications and, therefore, is not applicable to this EAL. 226

NEI 99 01 (Revisisa <:i) Nsvember 2012 The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible. Escalation of the emergency classification level would be via !Cs of Recognition Category A-R orF. Develaper Nates: EAL #1 For the site specific leak rate value, enter the higher of 10 gpm or the value specified in the site's Technical Specifications for this type ofleakage. EAL #2 For the site specific leak rate value, enter the higher of 25 gpm or the value specified in the site's Technical Specifications for this type of leakage. For sites that ha-v:e Technical Specifications that do not specify a leakage type for steam generator tube leakage, de*;elopers should include aB: EAL for tube leakage greater than 25 gpm for 15 minu.-tes or longer. EGL Assignment A.ttributes: 3 .1.1./'.. 228

NEI 99 01 (R-evisiea 13) Nevsa10@r 2012 SU5SU6 ECL: Notification of Unusual Event Initiating Condition: Automatic or manual (trip [P:WR] / scram [B\VR]) fails to shutdown the reactor. Operating Mode Applicability: Power Operation.L_l Nate: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. Exam13Ie Emergency Action Levels: (1 or 2) Note: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. SU6.1

a. An automatic (trip [PVlR.] / scram [BWR.]) did not shutdown .the reactor.

AND

b. ANY of the following manual actions taken at 1C05 are successful in lowering reactor power below 5% power

• Manual Scram Pushbuttons

                  .

• Mode Switch to Shutdown
• Alternate Rod Insertion (ARI)A subsequent manual action taken at the reactor control consoles (1C05) is successful rn shutting dov,rn the reactor.

SU6.2 a. A manual trip ([P'.VRJ / scram [BWR]) did not shutdown the reactor. AND

b. EITHER of the following:

1. -ANY of the following subsequent manual actions taken at 1C05 are successful in lowering reactor power below 5% power

                           ~ Manual Scram Pushbuttons

• Mode Switch to Shutdown
• Alternate Rod Insertion (ARDA subseq1:1:ent man.ual action taken. at the reactor control console (1C05)s is successful in shutting dov.rn the reactor.

                    ---_OR 230

NEI 99 Q1 (RevisieB 13) Nevember 2Ql2

2. -A subsequent automatic (trip [PVlRJ / scram [EVlRJ) is successful in shutting down the reactor.

Definitions: Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor ftrip [PJiVR] I scram [BJf'R]) that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic (tri-p [PJiVR] I scram [.BWR]) is successful in shutting down the reactor. _This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant. Following the failure on an automatic reactor (trip [PJiVR] I scram [BWR]), operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor (tri-p [PWR] / scram [BWR])). If these manual actions are successful in shatting down the reactor, core heat generation ;villscram quickly fall to a level within the capabilities of the plant's decay heat removal systems. 231

NEI 99 Ql (Re:visieH <ii) Nevember 2012 If an initial manual reactor (trip [PWR] l scram [BWR]) is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor (trip [PWR] I scram [BWR])) using a different switch). Depending upon several factors, the initial or subsequent effort to manually (trip [PWR] I scram [BR"1R]) the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor~ [PTVR] I scram [B WR]) signal. If a subsequent manual or automatic (trip [PWcR] I scram [B WR]) is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems. A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor~ [PTf'R] / scram [BTf'R])). This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles". Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action. [BWR] The plant response to the failure of an automatic or manual reactor (trip [PWRJ / scram [BV/RJ) will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC SA.§{i. Depending upon the plant response, escalation is also possible via IC FAl. Absent the plant conditions needed to meet either IC SA.§{i or FAI, an Unusual Event declaration is appropriate for this event. The reactor should be considered shutdown when it is producing less heat than the maximum decay heat load for which the SAFETY SYSTEMS are designed (typically 3 to 5% power).A reactor shutdov;n is detennined in accordance with applicable Emergency Operating Procedure criteria. Should a reactor (trip [PWRJ / scram [B'.VR]) signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied.

• If the signal causes a plant transient that should have included an automatic reactor~

[PWR] I scram [B')/R.J) and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and should be evaluated.

• If the signal does not cause a plant transient and the (trip [PWR] / scram [BWRJ) failure is determined through other means (e.g., assessment oftest results), then this IC and the EALs are not applicable and no classification is warranted.

De¥elepeF Netes: This IC is applicable in any Mode in *.vhich the actual reactor pov1er 16',el could exceed the power level at v:hich the reactor is considered shutdown. A P'.VR with a shu-tdoYm reactor power le,,rel that is less than or eqeal to the reactor pov,rer le 1el which defines the lower b01.,1f'ld of 1 Power Operation (Mode 1) .vill need to in.elude Sertuj) (Mode 2) in the OperatiBg Mode 1 Applicability. For mrnmple, if the reactor is considered to be shutdov,,ra at 3% and Power Operation starts at >5%, then the IC is also applicable ia Startup Mode. Developers may iaclude site specific EOP criteria indicative of a successful reactor shutdown in. 232

NEI 99 01 (RevisioH 6) Nove1nber 2012 SU6SU7 ECL: Notification of Unusual Event Initiating Condition: Loss of-all-ALL onsite or offsite communications capabilities. Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Slnitdovm.L.1.,_1 ExamJJle Emergency Action Levels: (1 or 2 or 3) SU7.1 Loss of ALL of the following onsite communication methods: _*_(site specific list of communications methods) Plant Operations Radio System

• In-Plant Phone System
• Plant Paging System (Gaitronics)

SU7.2 Loss of ALL of the following GRGoffsite response organization communications methods:

              *    (site specific list of communications methods) DAEC All-Call phone

• All telephone lines (PBX and commercial)
• Cell Phones (including fixed cell phone system)
• Control Room fixed satellite phone system
• FTS Phone system SU7.3 Loss of ALL of the following NRC communications methods:

_*_(site specific list of commuR-ications methods) FTS Phone system

• All telephone lines (PBX and commercial)
• Cell Phones (including fixed cell phone system)
• Control Room fixed satellite phone system Basis:

This IC addresses a significant loss of on-site or offsite communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to GRGoffsite response organizations and the NRC. This IC should be assessed only when extraordinary means are being utilized to make communications possible ( e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.). 235 J

NEI 99 01 (Revision 6) Nove1nber 2012 EAL SU7 .1 addresses a total loss of the communications methods used in support of routine plant operations. EAL SU7.2 addresses a total loss of the communications methods used to notify all GR:Goffsite response organizations of an emergency declaration. The GR:Goffsite response organizations referred to here are- the State oflowa, Linn County, and Benton County (see De,,reloper Notes). ---EAL SU7.3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration. De1veI0peF Netes: EAL #1 The "site speeifie list of eomml:Hl-iea-tioas methods" shottld iaelttde all eommunieatioas methods ttsed for romine plant communieatioas (e.g., eommereial or site telephoaes, page party systems, radios, etc.). This listiag shottld inclttde installed plant eqttipmest and compoaeBts, and H:Ot items owHed asd maiBtaiaed by iad~viduals. EAL #2 The "site specific list of comm-ueicatioss methods" shottld isclude all COHHRl:Hlieatioas methods used to perfonn initial emergency notifications to OROs as described in the site Emergency Plan. The listing shottld inclttde isstalled plant equipment and components, and aot items O'llfled aad maintaraed by radividuals. E1<.ample methods are rieg do1,-r,<H/dedicated telephone lines, c011Hnereial telephose lises, radios, satellite telephones and internet based eommunieatioss teehH:ology. lH the Basis seetioH, insert the site speeifie listisg of the OROs reqttiring notifieatios of an emergeney declaration from the Control Room iR aeeordanee with the site Emergeney Plan, and typieally withiH 15 minutes. EAL #3 The "site speeifie list of eommttBieations methods" shottld iBeh1de all eoHHffiH1ieations methods ttsed to perform iBitial emergeney Botifieations to the NRG as described in the site Emergency Plan. The listing should inelude iBstalled plaBt eqttipment and componeHts, and Bot items ovmed m1d maintaiHed by individuals. These methods are typically the dedicated Emergeney }fotifieatioH System (EN£) telephone line and eoflltH.ereial telephoBe lraes. EGL Assignment A.ttribtttes: 3.1.1.C 237

I

NEI 99 01 (Revisioa (oi) November 2012 SA1 ECL: Alert Initiating Condition: Loss of-a-It-ALL but one AC power source to emergency essential buses for 15 minutes or longer. Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown.L._l,_J_ Example Emergency Action Levels: Note: The Emergency Director should declare the Alert event promptly upon determining that the applicable time 15 _ minutes has been exceeded, or will likely be exceeded. SAl.l a. AC power capability to (site speeific emergency h1:1ses)lA3 and IA4 buses is reduced to a single power source for 15 minutes or longer. AND

b. Any ANYANY additional single power source failure will result in a loss of all ALL AC power to SAFETY SYSTEMS.

Definitions: SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.A system roq1:1ired for safe plant operation, cooling down the plant andlor placing it in the cold shutd01m1 condition, including the EGGS. Systems classified as safety related. Basis: This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power Source may be powering one, or more than one, train of safety:.- related equipment. This IC provides an escalation path from IC SUI. An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.

• A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an on.site diesel generator).

 *    /'.. loss of all offsite pov1er and loss of all emergency pmver sources (e.g., onsite diesel generators) *.vith a single train of emergency buses being back fed from the unit main generator.

• A loss of emergency power sources (e.g., on.site diesel generators) with a single train of essentialernergency buses being :t3-aek--fed from an offsite power source.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power. Escalation of the emergency classification level would be via IC SS 1. DeveI013eF Netes: For a po,ver source that has multiple generators, the EAL and/or Basis section should reflect the minimum munber of operating generators necessary for that soarce to provide required pov;er to 243

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NEI 99 Ql (Revision 6) Noven:i.ber 2012 SA2SA3 ECL: Alert Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress. Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdov,rnL.LJ Example Emergency Action Levels: Note: The Emergency Director should declare the fJert event promptly upon determining that the applicable time 15_ minutes has been exceeded, or will likely be exceeded. SA3.l a. An UNPLANNED event results in the inability to monitor one or more Table S-1 parameters from within the Control Room for 15 minutes or longer-An UNPL'\1'JNED event results in the inability to monitor one or more of the following parameters from within the Control Room for 15 minutes or longer.

                                   , lable ~-1 Safety System Paramet~r~
                                        ,    ,          ,'             ,~ : "' -;\ ,.

Reactor power RPV Water Level RPV Pressure Primary Containment Pressure Suppression Pool Level Suppression Pool Temperature Pewer RPV \Vater Level RPV Pressure Primary Containment Pressure Suppression Pool Level Suppression Pool Temperature Suppression Pool Temperatra:e AND

         =b*;....___;ANY of the Table S-2 transient events are in progress.

245

NEI 99 01 (R1wisiea (i) Nevember 2012

• Automatic or manual runback greater than 25% thermal reactor power
• Electrical load rejection greater than 25% full electrical load
• Reactor scram
• ECCS actuation
• Thermal power oscillations greater than 10%

trffEl:sient C¥ents in progress. Automatic or mant1al nm-back greater than 25% thermal reactor pmver Electrical load rejectioa greater than 25% full electrical load Reactor scram [BV/R] / trip [PWR] EGGS (SI) actuation Thefffial pov,rer oscillations greater than 10% [B'NRJ 246

NEI 99 Q1 (R-evision 6) No1,zember 2Ql2 Definitions: SAFETY SYSTEM: A system reguired for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.A. system required for safe plant operation, cooling down the plant and/or placing it in the cold slHitdoWH: condition, includiH:g the ECCK 8ysterns classified as safety related. UNPLANNED: A parameter change or a11 event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: This IC addresses the difficulty associated with monitoring rapidly changing plant conditions during a transient without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. During this condition, the margin to a potential fission product barrier challenge is reduced. It thus represents a potential substantial degradation in the level of safety of the plant. 247

NEI 99 0 l (Revisioa <:i) November 2012 As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room. An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making. This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling [PWR] I RPV level [BTVR] and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for reactor vessel level [PT¥R] I RPV water level [B JYR] cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication. Escalation of the emergency classification level would be via ICs FS 1 or IC AS+RS 1. Develef)eF Netes: In the PWR parameter list column, the "site specific number" should reflect the minimum B:Umber of steam generators necessary for plant cooldovm and shutdo~;m. This criterion may also specify whether the }e>,rel value shoeld be *.vi.de range, naHow range or both, depending ,lf)on the monitoring requirements in emergeaey operating procedures. D011elopers may specify either pressHFizer or reactor vessel level in the PWR parameter column entry for RCS Le11el. De>,relopers should consider if the "transient events" list needs to be modified to better reflect site specific plant operating characteristics and mcpected responses. The B:liB'laer, tY13e, location anel. layout of Control Room inel.ieations, B:B:el. the raH:ge of possible failHFe modes, can challenge the ability of an op erffl:or to accurately determine, *.vithin the time period available for emergency classification assessments, if a specific percentage of indications have been lost. The approach used in this EAL facilitates prompt and accurate emergency classification assessments by focusing on the inel.ications for a selecteel. subset of parameters. By focusing on the availability of the specified parameter values, instead of the soHFces of those values, the EAL recognizes and accommodates the wide variety of indications in nuclear pmver plant Control Rooms. Indication types and sournes may be analog or digital, safety related or not, primary or alternate, individual meter value or computer group display, etc. 248

NEI 99 0 l (RevisioH l:i) November 2012 SA6SA6 ECL: Alert Initiating Condition: Automatic or manual (trip [P1,VR] I scram [BWR]) fails to shutdown the reactor, and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor. Operating Mode Applicability: Pov,rer Operationl.,_1 Note: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. Example EmergeH.eyEmergency Action Levels-: SA6.l a. An automatic or manual (trip [P\VRJ / scram [BWR]) did not shutdown the reactor. AND

b. ALL of the following manual actions taken at 1C05 are not successful in lowering reactor power below 5% power

• Manual Scram Pushbuttons
• Mode Switch to Shutdown
• Alternate Rod Insertion (ARI)MaH:aal aetions takea at the reaetor eontrol eonsoles (1 COS) are not s:1eeessful in shutting dovm the reaetor.

Definitions: Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor ftrip [PWR] / scram [BWR]) that results in a reactor shutdown, and subsequent operator manual actions taken at the reactor control consoles to shutdown the reactor are also unsuccessful. This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if the reactor is subsequently shutdown by an action taken away from the reactor control consoles since this event entails a significant failure of the RPS. A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core ( e.g., initiating a manual reactor ftrip [P1¥R] I scram [B1¥R])). This action does not include manually driving in control rods or implementation of boron injection strategies. If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control consoles (e.g., locally opening breakers). _Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles.:.".,. 251

NEI 99 01 (R-evision 6) Novsniber 2012 Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action. [BWR] 252

NEI 99 01 (ReYisisa {i) Ns1,<e111ber 2012 The plant response to the failure of an automatic or manual reactor (trip [PWR] I scram [.B WR]) will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If the failure to shutdown the reactor is prolonged enough to cause a challenge to the core cooliBg [:PWRJ / RPV water level [B\"VR] or RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC SS.§..§. Depending upon plant responses and symptoms, escalation is also possible via IC FS 1. Absent the plant conditions needed to meet either iC SS.§..§ or FSl, an Alert declaration is appropriate for this event. It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F ICs; however, this IC and EAL are included to ensure a timely emergency declaration. The reactor should be considered shutdown when it is producing less heat than the maximum decay heat load for which the SAFETY SYSTEMS are designed (jypically 3 to 5% power).A reactor shutdovm is determined in accordance *.vith applicable Emergency Operating :Procedure criteria. DevelBfleF Netes: This IC is applicable in any Mode in which the actual reactor pov,'er level could exceed the power level at *.vhich the reactor is coBsidered shutdovm. A :PWR vtith a shutdov,rn reactor power 101:el that is less than or equal to the reactor power le11el v.rliich defines the lov.zer bound of Power OperatioB (Mode 1) will Beed to H'lch1de Startup (Mode 2) in the OperatiBg Mode Applicability. For example, if the reactor is considered to be shutdovm at 3% and Power Operation struts at >5%, then the IC is also applicable in Startup Mode. Developers may inch1de site specific EO:P criteria iHCHcative of a successful reactor shutdovA1 iH an EAL statemeHt, the Basis or both (e.g., a reactor power level). The tenn "reactor coBtrol consoles" may be replaced *.vith the appropriate site specific term (e.g., main coBtrol boards).

• EGL Assigsment Attri-eu-tes: 3 .1.2.B 254

NEI 99 Ql (Revisioa (oi) November 2Q 12 SA9SA8 ECL: Alert Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode. Operating Mode Applicability: Power OperatioH, Startup, Hot Stasdby, Hot Shutdovm.L..LJ Examf)le Emergency Action Levels: Notes:

• If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then this emergency classification is not warranted.

       - If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of the SAFETY SYSTEM, then this emergency classification is not warranted .

SA8.l a. The occurrence of ANY of the Table S-3 hazardous events:

• Seismic event (earthquake)
• Internal or external flooding event
• High winds or tornado strike
• FIRE
• EXPLOSION
• Other events with similar hazard characteristics as determined by the Shift Manager or Emergency Director The oe6'Hffeaee oL\NY of the follmviag hazardoes eveats:

Seism.ie e0veat (earthquake) Internal or mctemal flooding eveat High 1+vinds or tornado strike ~ EXPLOSION (site specific hazards) Other events with similar hazard characteristics as determined by the Shift Masager or EmergCHcy Director AND

b. 1. Event damage has caused indications of degraded performance in one train of a SAFETY SYSTEM needed for the current operating mode.

AND

2. EITHER of the following:

256

NEI 99 01 (Revisioa ti) November 2012

• Event damage has caused indications of degraded performance to a second train of the SAFETY SYSTEM needed for the current operating mode, OR
• The event has resulted in VISIBLE DAMAGE to the second train of a SAFETY SYSTEM needed for the current operating mode.,

Event damage has caused indications of degraded performance to a second train of the SAFETY 8Y8TEM needed for the current operating mode, or The event has resulted in VISIBLE DAM:z'\GE to the second train of the SAFETY 8Y8TEM needed for the c:arrent operatiag mode. Loss of the safety fuactioa ofa smgle train SAFETY 8Y8TEM. 257

NEI 99 01 (RevisieB ~) Nevember 2012 Definitions: EXPLOSION: A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events may require a post-event inspection to determine if the attributes of an explosion are present. FIRE: Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed. SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdovm condition, including the EGGS. Systems classified as safety related. VISIBLE DAMAGE: Damage to a component or structure that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure. Damage resulting from an equipment failure and limited to the failed component (i.e., the failure did not cause damage to a structure or any other equipment) is not VISIBLE DAMAGE. EITHER of the followmg:

1. Event damage has caused indications of degraded perfonnance in at least one train of a SAFETY SYSTEM needed fur the current operating mode.

OR

2. The event has caused VISIBLE DAM.1\GE to a SAFETY SYSTEM component or structllre needed for the current operating mode.

Basis: This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the current operating mode. In order to provide the appropriate context for consideration of an ALERT classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train, and there must be either indications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance. and the second SAFETY SYSTEM train must have indications of degraded perfonnance or VISIBLE DAMAGE such that the potential exists for perfonnance issues. Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has indications of degraded performance for criteria SA98. l .b. l of this EAL; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events. An event affecting a single-train SAFETY SYSTEM (i.e., there are indications of degraded performance and/or VISIBLE DAMAGE affecting the one train) would not be classified under SAS because the two-train impact criteria that underlie the EALs and Bases would not be met. If 258

NEI 99 01 (Revision 6)

                                                                                      *November 2012 an event affects a single-train SAFETY SYSTEM, then the emergency classification should be made based on plant parameters/symptoms meeting the EALs for another IC. Depending upon the circumstances, classification may also occur based on Shift Manager/Emergency Director judgement.

Indications of degraded performance addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. 259

NEI 99 Q1 (RevisiaB i) Na¥ember 2012 VISIBLE DAMAGE addresses damage to a SAFETY SYSTEM train that is not in service/operation and that potentially could cause perfonnance issues. Operators will make this detennination based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. 261

NEI 99 Ql (Re11isi0H '3) Nevember 2Ql2 This IC addresses a hazardous e>rent that causes damage to a SAFETY SYSTEM, or a struet1:lre containing SAFETY SYSTEM components, needed for the current operating mode. This condition significantly reduces the margin to a loss or potential loss of a fission product barrier, and therefore represents an actual or potential substantial degradation of the level of safety of the ~ E,'\L l.b.l addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performaH:ce should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM: train. EAL l.b.2 addresses damage to a SAFETY SYSTEM component that is not in service/operation or readily apparent through indications alone, or to a structure containing SAFETY SYSTEM components. Operators will make this determination based on the totality of available event and damage report i:Hfora1ation. This is intended to be a brief assessment not reg-airing lengthy analysis or quantification of the damage. Escalation of the emergency classification level would be via IC FS 1 or A&l-RS 1. DenlaJJeF Nates: For (site specific hazards), developers should consider inclading other significant, site specific hazards to the bulleted list contained in EAL l.a (e.g., a seiche). }fuclear pov,er plant SAFETY SYSTEMS are comprised of two or more separate and redoodant traias of eq,aipment ill aeeordaRce *.vith site specific desigR criteria. EGL Assignment Attribates: 3 .1.2.B 262

NEI 99 0 l (Revisiea 6) November 2012 SS1 ECL: Site Area Emergency Initiating Condition: Loss of ALLa!l offsite and all-ALL onsite AC power to emergeHcy essential -buses for 15 minutes or longer. Operating Mode Applicability: Pov.,rer OperatioH, Startup, Hot Standby, Hot Shutdown.L..U Example Emergency Action Levels: Note: The Emergency Director should declare the Site Area En1ergencyevent promptly upon determining that the applicable time 15 minutes has been exceeded, or will likely be exceeded. SSl.l Loss of ALL offsite and ALL onsite AC power to (site speeific emergeney hases) 1A3 and 1A4 buses for 15 minutes or longer. Definitions: SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.}.. system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown eoridition, including the EGGS. Systems elassified as safety related. Basis: This IC addresses a total l~ss of AC power that compromises the performance of all .,SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. In addition, fission product barrier monitoring capabilities may be degraded under these conditions. This IC represents a condition that involves actual or likely major failures of plant functions needed for the protection of the public. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via ICs AG+RGl, FGI or SGI. DevelapeF Nates: For a pov,rer souree that has multiple generators, the EAL and/or Basis seetioH should refleet the minimum H1:Hr1.ber of operating generators HecessaI)' for that souree to provide adequate power to an l ..C emergeHcy bus. For rnmmple, if a backup pov,rer source is comprised of two generators (i.e., t\vo 50% capacity generators sized to feed 1 AC emergency bus), the EAL and Basis sectioH ffl:Ust specify that both geHerators for that smffce are operatiHg. The "site speeific emergeHey buses" are the buses fed by offsi-te or emergeHey AC povler sources that supply power to the electrical distribution system that po*.vers SAFETY SYSTEMS. There is typieally 1 emergency bus per train of SAFETY SYSTEMS. 263

NEI 99 Q1 (R-evisioH <'i) November 2012 SS8SS2 ECL: Site Area Emergency Initiating Condition: Loss of ALL Vital DC power for 15 minutes or longer. Operating Mode Applicability: 1, 2, 3 Emergency Action Levels: Note: The Emergency Director should declare the Site Area Emergencyevent promptly upon determining that the applicable time 15 minutes has been exceeded, or will likely be exceeded. =-SS=2=*~1___Indicated voltage is less than (site specific bus voltage value) 105 VDC on ALL(site specific :vital DC busses) BOTH Div I and Div 2 125 VDC buses for 15 minutes or longer. Definitions: SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.A system reguired for safe pla:at operatioa, cooliag dowB the pla:at aad/or placiBg it ia the cold shl:ttdowa eoaditioa, iBcludiBg the ECC8. Systems classified as safety related. Basis: This IC addresses a loss of Vital DC power which compromises the ability to monitor and control SAFETY SYSTEMS. In modes above Cold Shutdown, this condition involves a major failure of plant functions needed for the protection of the public. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via ICs AG+RG 1, FG 1 or SG2. 265

NEI 99 Ql (Re~*isioa (i) No1,,ember 2012 SS6SS6 ECL: Site Area Emergency Initiating Condition: Inability to shutdown the reactor causing a challenge to (eore eooling [PJiT'R] / RPV water level [BWR]) or RCS heat removal. Operating Mode Applicability: Po'.ver OperationU E:xemf)le Emergency Action Levels: SS6.l a. An automatic or manual (trip [PWR] / scram [BWR]) did not shutdown the reactor.

            ---_AND

b. ALL of the following manual actions taken at 1C05 are not successful in lowering reactor power below 5% power:

• Manual Scram Pushbuttons *
• Mode Switch to Shutdown
• Alternate Rod Insertion (ARI)All manual aetions to shutdov,'fl the reaetor. have been lillsueeessful.

AND

c. EITHER of the following conditions exist:

_*_(Site speeifie indieation of an inability to adequately remove heat from the eore) Reaetor vessel ,.vaterRPV level cannot be restored and maintained above

                       -25 inches.

OR

                   *   (Site specifie indication of an inability to adequately remo'le heat from the
                       ~HCL (Graph 4 of EOP 2) exceeded.

Definitions: Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor~ [PTf'.R] I scram [BTf'.R]) that results in a reactor shutdown, all subsequent operator actions to manually shutdown the reactor are unsuccessful, and continued power generation is challenging the capability to adequately remove heat from the core and/or the RCS. This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration of a Site Area Emergency. In some instances, the emergency classification resulting from this IC/EAL may be higher than that resulting from an assessment of the plant responses and symptoms against the Recognition 266

NEI 99 0 l (R-evisioR 6) November 2012 Category F ICs/EALs. This is appropriate in that the Recognition Category F ICs/EALs do not address the additional threat posed by a failure to shutdown the reactor. The inclusion of this IC and EAL ensures the timely declaration of a Site Area Emergency in response to prolonged failure to shutdown the reactor. The reactor should be considered shutdown when it is producing less heat than the maximum decay heat load for which the SAFETY SYSTEMS are designed (typically 3 to 5% power). A reactor Sftlltdo ,VR is determined in accordance with applicable Emergency Operating 1 Procedure criteria. Escalation of the emergency classification level would be via IC AGl-RG 1 or FG 1. DevelapeF Netes: This IC is applicable in any Mode in *.vhich the actual reactor po*,ver level could exceed the pov,zer level at *.v,hich the reactor is coH:sidered shu-tdown. A PVlR. :with a shutd01.vn reactor pov,rer level that is less thaB or equal to the reactor power le., el ,,vtHch defiH:es the lower bound of 0 Power Operation (Mode 1) will need to include Startup (Mode 2) in the Operating Mode Applicability. For example, if the reactor is considered to be shutdovm at 3% and Pov1cr Operation starts at >5%, thee the IC is also applicable in Sta:Ftup Mode. De>,celopers may ieclude site specific EOP criteria inElicative of a sHccessful reactor Sftlltaovm is an E},L statement, the Basis or both (e.g., a reactor pov1er level). Site specific indication of an inability to adequately remove heat from the core: [BW'R] Reactor vessel water level cannot be restored and maietaieed above Minlll'l-l:ffi'l Steam - Cooling RPV Water Level (as described in the EOP ba~es}. [PWR] Insert site specific valHes for an incore/core exit thermocouple temperature and/or reactor vessel water level that drives entry into a core cooling restoration procedure (or otherwise reqHires implementation of prompt restoration actions). Alternately, a site may use incore/core exit thennocouple temperatures greater than 1,200eF aBd/or a reactor vessel *.vater level that corresponds to appro1cimately the middle of active fuel. Plants with reactor vessel le:vel instrumentation that cannot measure down to approximately the middle of active fuel should use the lowest on scale reading that is aot above the top of active fael. If the lo*11est OH scale reading is above the top of active fuel, then a reactor vessel level value should not be included. For plants that have implemented ',Vcstinghouse Owners Group Emergency Response Guidelines, enter the parameters used in the Core Cooling Red Path. Site specific indication ofan inability to adequately remove heat from the R{:;S: [BWR] Use the Heat Capacity Temperature Limit. This addresses the inability to remove heat via the main condenser and the suppression pool due to high pool *.vater temperature. [PWR] Insert site specific parameters associated ;vith iBadeq1:1:ate RCS heat removal via the steam generators. These parameters should be identical to those used for the Inadeq1:1ate Heat Removal threshold Fuel Clad BaHier Potential Loss 2.B and threshold RCS Barrier Potential Loss 2.A in the PVlR. EAL Fission Product Barrier Table. EGL Assignment Attrib1:1tes: 3.1.3.B 267

NEI 99 0 l (Revisiea 6) Nevsmber 2012 SG1 ECL: General Emergency Initiating Condition: Prolonged loss of all-ALL offsite and ALL-all onsite AC power to emergency essential buses. Operating Mode Applicability: Power OperatioH, StartHfJ, Hot Standby, Hot Shutdown.L_LJ Example Emergency Action Levels: Note: The Emergency Director should declare the General Emergencyevent promptly upon determining that (site specific hours) the applicable time 4 hours has been exceeded, or will likely be exceeded. SGl.l a. Loss of ALL offsite and ALL onsite AC power to IA3 and 1A4 busesfsite-specific emergency buses). AND

b. EITHER of the following:

                  ~Restoration of at least one AC emergency essential bus in less than (site--

specific hours)4 hours is not likely. OR

                  * (Site specific iHdicatioH of an inability to adequately remove heat fi:om the eerejRPV level cannot be restored and maintained above -25 inches.

Definitions: SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.A system required for safe plant operation, cooling dowH the plant aHd/or placing it in the cold shutdown condition, including the EGGS. Systems classified as safety related. Basis: This IC addresses a prolonged loss of all power sources to AC emergency essential buses. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A prolonged loss of these buses will lead to a loss of one or more fission product barriers. In addition, fission product barrier monitoring capabilities may be degraded under these conditions. The EAL should require declaration of a General Emergency prior to meeting the thresholds for IC FG 1. This will allow additional time for implementation of offsite protective actions. Escalation of the emergency classification from Site Area Emergency will occur if it is projected that power cannot be restored to at least one AC emergency essential bus by the end of the analyzed 4 hour station blackout coping period. Beyond this time, plant responses and event trajectory are subject to greater uncertainty, and there is an increased likelihood of challenges to multiple fission product barriers. 269

NEI 99 0 l (R~P,risiaH {i) N0Ye1Rber 2012 The estimate for restoring at least one essentialem.erge13:ey bus should be based on a realistic appraisal of the situation. Mitigation actions with a low probability of success should not be used as a basis for delaying a classification upgrade. The goal is to maximize the time available to prepare for, and implement, protective actions for the public. 270

NEI 99 Q1 (Revision 6) November :Wl2 The EAL will also require a General Emergency declaration if the loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core. Developer Notes: Although this IC and EAL may be viewed as redundant to the Fission Product Barrier ICs, it is iecla.ded to provide for a more timely escalation of the emergeacy classification level. The "site specific emergeacy bases" are the buses fed by offsite or e1HergeHcy AC po\ver sources that supply power to the electrical distribution system that powers SA.FETY SYSTEMS. There is typically 1 emergency bes per train of SAFETY SYSTEMS. The "site specific hours" to restore AC pov.zer to an er.eergency bti:S shoeld be based on the sffl:tion blacko:at coping aHalysis perfofft1ed in accordance with 10 CFR § 50.G3 and Regalatory Guide 1.155, St1tien Blac-keut. Site specific indication ofan inability to adequately remove heat from the core: [BWR] Reacter vessel water level cannot be restore<l an<l maietaiHed abo*.re MiHiml:llH Steam Cooling RPV V.later Level (as desc1ibed in the EOP bases). [PTfR] 1Hsert site specific valees for an iacore/core OJ(i-t thefft1ecouple temperatere and/or reactor vessel water level that drive eetry into a core cooling restoration proceoore (or othenvise requires implemeatatioa of prompt restoration actions). } ..lternately, a site may use incore/core CKit thermocouple temperateres greater thaa l ,200eF and/or a reactor vessel water level that correspoHds to approximately the middle of aetive fuel. Plants with reactor vessel level instrHmentation that cannot measme down to approximately the middle of active fuel should use the lov.zest on scale reading that is not above the top of active fuel. If the 101.vest on scale reading is aoove the top of active fuel, theH a reactor vessel 101,el vala.e:'should not be included. for plants that have implemeB:te<l \llestieghouse Owners Group Emergency Res130Hse Guidelines, enter the parameters used in the Core Cooling Red Path. EGL Assignment A.ttributes: 3.1.4.B 272

NEI 99 Ql (Revision 6) November 2Ql2 SG8SG2 ECL: General Emergency Initiating Condition: Loss of all-ALL AC and Vital DC power sources for 15 minutes or longer. Operating Mode Applicability: Pov,z:er Operation, Startup, Hot Standby, Hot ShutdoWH.L_l,_l Example Emergency Action Levels: Note: The Emergency Director should declare the Gooeral Emergencyevent promptly upon determining that the applicable time 15 minutes has been exceeded, or will likely be exceeded. SG2.l a. Loss of ALL offsite and ALL onsite AC power to (site specific emergency

                  ---bl-Hu-<.ise,,.,s!-1--) 1A3 and 1A4 buses for 15_-minutes or longer.

AND

b. Indicated voltage is less than (site specific bus voltage 11alue)l05 VDC on AlJ,.

(site speeifie Vital DC husses) BOTH Div 1 and Div 2 125 VDC buses for 15 minutes or longer. Definitions: SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related:}.. system required for safe plant operation, cooling dov,'il the plant and/or placing it in the cold shutdown coHdition, including the EGGS. Systems classified as safety related. Basis: This IC addresses a concurrent and prolonged loss of both AC and Vital DC power. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A loss of Vital DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both AC and DC power will lead to multiple challen&es to fission product barriers. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. The 15-minute emergency declaration clock begins at the point when both EAL thresholds are met.

• Developer Nates:

The "site specific emergeHcy buses" are the buses fed by offsite or emergency f ..G pov,z:er sources that supply pov,z:er to the electrical distribution system that powers SAFETY SYSTEMS. There is typically 1 emergency has per train of SAFETY SYSTEMS. The "site specific bus voltage value" should be based on the minimum bus voltage necessary for adequate operation of SAFETY SYSTEM equipmOHt. 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 wheH battery sizing is performed. 273

NEI 99 0 I (RevisieH e)

                                                                                                                                  *Nevember 2012 APPENDIX A - ACRONYMS AND ABBREVIATIONS AC ...................................................................................................................... Alternating Current AOP ................................................................................................. Abnormal Operating Procedure
.......................................................................................................................................... i'..

PR14 .................................................................................................... A.verage Pmver Range l\4eter ATWS ...... :............................................................................ Anticipated Transient Without Scram

.......................................................................................................................................... B

&V*l ................................................................................................................... Babcock and ¥1'ilcox

................................................................................. ~ ........................................................ B IIT ....................................................................................... BoroB: Injection InitiatioR Temperatra:e BWR ............................................................................................................. Boiling Water Reactor CDE ..............................................................................................,. ....... Committed Dose Equivalent CFR ...................................................................................................... Code of Federal Regulations

~ M T ............................................................................................................... Containment

.....................................................................................................*..................................... C SF ................................................................................................................ Critical Safety Function
.......................................................................................................................................... C SFST ........................................................................................ Critical Safety FooctioB: Starns Tree
.......................................................................................................................................... D Bi'................................................................................................................... Design Basis Accidoot DC .............................................................................................................................. Direct Current 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
.......................................................................................................................................... E PIP ................................................................................... Emergency Plan Implementing Procedure
.......................................................................................................................................... E PR ......................................................................................................... v,rolutioH:ary Po\ver Reactor
.......................................................................................................................................... E PRI ............................................................................................... Electric Pmver Research Institute
.......................................................................................................................................... E RD .................................................................................................. Emergency RespoH:se Gl:ii.deliRe
.......................................................................................................................................... F EMA ................................................................................ Federal Emergency Management Agency FS,<\R ................................................................................................... Final Safety ,\J:ialysis Report GE ...................................................................................................................... General Emergency HCTL .......................................................................................... Heat Capacity Temperature Limit HPCI .............................................................................................. High Pressure Coolant Injection
.......................................................................................................................................... H SI ................................................................................................................ HlHllaH: Syste1n Interface IC ........................................................................................................................ Initiating Condition

NEI 99 01 (Revisiea e) 1-~evsmber 2Ql2

.......................................................................................................................................... I D ............................................................................................................................... Inside Diruneter IPEEE. ............................ IB.dividual Plant Examination of External Events (Generic Letter gg 20)

ISFSI ........................................................................... Independent Spent Fuel Storage Installation Keff .................................................................................... Effective Neutron Multiplication Factor LCO ............................................................................................... Limiting Condition of Operation

... .- ...................................................................................................................................... L OCA .......................................................................................................... Loss of Coolant Accident
.......................................................................................................................................... 14 CR ..................................................................................................................... l\4ain Control Room
.......................................................................................................................................... 14 81\l ........................................................................................................ l\iain Steam Isolation \lalve 14SL ....................................................................................................................... 14ain Steani Line m.R, m.Rem, mrem, m.REM ............ , ............................................... milli-Roentgen Equivalent Man MW .................................................................................................................................... Megawatt NEI .................................. ~ .......................................................................... Nuclear Energy Institute
..................... .,.................................................................................................................... ~+

PP .............................................. * ...................................................................... Nuclear Po1.ver Plant

.......................................................................................................................................... N RC ................................................................................................. Nuclear Regulatory Commission

~TSSS ................................................................................................. Nuclear 8teani Suwly System NORAD ................................................................. North American Aerospace Defense Command fNO)UE ......................................... '. ................................................ fNotification Of) Unusual Event NUMARC 1 ............................................................... Nuclear Management and Resources Council OBE ....................................................................................................... Operating Basis Earthquake OCA ............................................ , ... ;............................................................ Owner Controlled Area

.......................................................................................................................................... 0 DCM/ODAM ......................................................... Offsite Dose Calculation (Assessment) Manual ORO ................................................................................................ Off site Response OrgaH:ization PA .................................... :......................................................................................... Protected Area
.......................................................................................................................................... p A.CS ...................................................................................... Priority A.ctuation and Control System PAG....................................................................................................... Protective Action Guideline
.......................................................................................................................................... P ICS ................................................................................... Process Information and Control System PRA/PSA .................................... Probabilistic Risk Assessment / Probabilistic Safety Assessment PWR ........................................................................................................ Pressurized Water Reactor
.......................................................................................................................................... p S ........................................................................................................................... Protection System PSIG ................................................................................................. Pounds per Square Inch Gauge R ......................................................................................................................................... Roentgen
.......................................................................................................................................... R CC .............................................................................................................. Reactor Control Console RCIC ............................................................................................... Reactor Core Isolation Cooling 1

NUMARC was a predecessor organization of the Nuclear Energy Institute (NEI).

NEI 99 QI (Rs1i'isi0a e) 1'fo1,SB3.0Sf 2Q 12 RCS ............................................................................................................. Reactor Coolant System Rem, rem, REM ...................................................................................... Roentgen Equivalent Man

.......................................................................................................................................... R ETS ......................................................................... Radiological Efflaea.t Tech.rical 8pecificatioa.s RPS ......................................................................................................... Reactor Protection System RPV ...................................................................................................... :...... Reactor Pressure Vessel
.......................................................................................................................................... R VLI8 ......................................................................... Reactor Vessel Level fustrumea.tatioa. System RWCU .......................................................................................................... Reactor'Water Cleanup
****************************************************************************************************************************************** s

,\R ................................................................................................................ Safety l ..nalysis Report

..................................*........................................................................................................ s l .. 8 .......................................................................................................... Safety lrlttomation System
.......................................................................................................................................... s BO ........................................................................................................................... Station Blackout SCBA.. .. ... .. ...... .. .... .. ..... .. ... ... ... .... .. ........... ....... .. ... .......... ... .. .. Self-Contained Breathing Apparatus
.......................................................................................................................................... s G .............................................................................................................................. Steam Generator
.......................................................................................................................................... s I ................................................................................................................................ Safety Injection
.......................................................................................................................................... s ICS .................................... :................................................ Safety Information and Control System
.................................................................................................................. .-........................ S PDS .............................................................................................. Safety Parameter Display System SRO ............................................................................................................ Senior Reactor Operator TEDE ............................................................................................. Total Effective Dose Equivalent TQAF .................................................................................................................. Top of Active Fuel TSC .......................................................................................................... Technical Support Center
.......................................................................................................................................... U FSAR .................................................................................... Updated Final Safety Analysis Report WOG .................................................................................................. 'Nestinghouse Owners Group

,-~ NEI 99 Q1 (R:e¥isieB e) Nevember 2Ql2 APPENDIX B - DEFINITIONS The following definitions are taken from Title 10, Code of Federal Regulations, and related regulatory guidance documents. Alert: Events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA PAG exposure levels. General Emergency: Events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area. Notification of Unusual Event (NOUE}1": Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systemsSAFETY SYSTEMS occurs. Site Area Emergency: Events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; 1) toward site personnel or equipment that could lead to the likely failure of or; . 2) that prevent effective access to, equipment needed for~- the protection of the public. _Any releases are not expected to result in exposure levels which exceed EPA PAG exposure levels beyond the site boundary. The following are key terms necessary for overall understanding the NEI 99 OlDAEC emergency classification scheme. ** Emergency Action Level (EAL): A pre-determined, site-specific, observable threshold for an Initiating Condition that, when met or exceeded, places the plant in a given emergency classification level. Emergency Classification Level (ECL): One of a set of names or titles established by the US Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions according to (1) potential or actual effects or consequences, and (2) resulting onsite and offsite response actions. The emergency classification levels, in ascending order of severity, are: Notification of Unusual Event (NOUE) Alert Site Area Emergency (SAE) General Emergency (GE) + This term is semstin~es shertsnsEl te Uausual Eveat (UE) er ether similar site speoifo: terminelegy. B-1

NEI 99 Q1 (Revisisn 6) Nsvember 2Ql2 Fission Product Barrier Threshold: A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier. Initiating Condition (IC): An event or condition that aligns with the definition of one of the four emergency classification levels by virtue of the potential or actual effects or consequences. Selected terms used in Initiating Condition and Emergency Action Level statements are set in all capital letters ( e.g., ALL CAPS). These words are defined terms that have specific meanings as used in this document. The definitions of these terms are provided below. CONFINEMENT BOUNDARY: (Insert a site specific definition for this term.) Denlaper Nate -The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. This corresponds to the pressure boundary for the Dry Shielded Canister (DSC) shell (including the inner bottom cover plate) base metal and associated confinement boundary welds. CONTAINMENT CLOSURE: Procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions. For DAEC, this is considered to be Secondary Containment as required by Technical Specifications. CONTAINMENT CLOSURE: (Insert a site specific definition for this term.) Develeper Nate Site specific procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions. For D}.ECs, this is considered to be Secondary Containment as required by Technical Specifications.The procedurally defined"conditions or actions taken to secure containment (primary or secondary for BVlR) and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. DESIGN BASIS EARTHQUAKE (DBE): A DBE is vibratory ground motion for which certain (generally, safety-related) structures, systems, and components must be designed to remain functional. EXPLOSION: A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events may require a post-event inspection to determine if the attributes of an explosion are present. FAULTED: The term applied to a steam generator that has a steam. leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam. generator to become completely depressurized. Developer Nate This tenn is applicable to PWRs only. B-2

NEI 99 Q1 (Revisiea e) Ne:i,reffii:>er 2012 FIRE: Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed. HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station. HOSTILE ACTION: An act toward a NPP-nuclear power plant or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end._ This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. _Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPPnuclear power plant._ Non-terrorism-based EALs should be used to address such activities (i.e.,_-this may include violent acts between individuals in the owner controlled area). 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. IMMINENT: The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI): A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage.

• NORMAL LEVELS: As applied to radiological IC/EA.Ls, the highest reading in the past twenty four hours excluding the current peak value.

OPERATING BASIS EARTHOUAKE (OBE): An OBE is vibratory ground motion for which those features of a nuclear power plant necessary for continued operation without undue risk to the health and safety of the public will remain functional. OWNER CONTROLLED AREA: (Insert a site specific definition for this term.) Deyel0JJeF Nate This term is typically taken to mean the site property owned by, or otherwise under the control of, the licensee. In some cases, it may be appropriate for a licensee to define a smaller area with a perimeter closer to the plant Protected Area perimeter (e.g., a site \Vith a large OCA ,vhere some portions of the bom1dary may be a significant distance from the Protected Area). In these cases, dC'":elopers should consider Hsing the boundary defined by the Restricted or Secured 0 .vner Controlled Area (ROC1VSOCA). The area and boundary selected for scheme use must 1 be consistent with the description of the same area and boundary contained in the Secm*ity Plan. PROJECTILE: An object directed toward a NPPnuclear power plant that could cause concern for its continued operability, reliability, or personnel safety. B-3

NEI 99 QI (Revisioe 6) November 2Ql2 PROTECTED AREA:_ (IRsert a site specific defiBitioH. for this term.) DevelepeF Nate This term is typically takoo to mean (Ihe area under continuous access monitoring and control, and armed protection as described in the site Security Plan. REFUELING PATHWAY:_ (l-Hsert a site specific defffiitioB for this term.) De"lelepeF Nete This descl1f)tioH shoald iHcll:lde all the c&vities, tl:taes, caBals aBd pools throagh wliich irradiated fuel may be moved, bat Bot incladiag the reactor vessel. Includes all the cavities, tubes, canals and pools through which irradiated fuel may be moved, but not including the reactor vessel. RUPTURE(D): The coHditioR of a stecHB. geRerator iH which 13rireary to secoHdary leakage is of safficieBt FHageirude to reqaire a safety mjectioR. DeveleIJeF Nete This teffH is a1313licable to P\VRs only. SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems are classified as safety-related. DevelepeF Nete This terFH n1ay be IBodified to ieclade the attribl:ltes of "safety related" in aeeordaHce with 10 CFR 50.2 or other site specific terminology, if desired. SECURITY CONDITION: Any Security Event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. _A SECURITY CONDITION does not involve a HOSTILE ACTION. SITE BOUNDARY: That line beyond which the land is neither owned, nor leased, nor otherwise controlled by the Company. UFSAR Figure 1.2-1 identifies the DAEC SITE BOUNDARY..,. UNISOLABLE: An open or breached system line that cannot be isolated, remotely or locally. UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. VISIBLE DAMAGE: Damage to a component or structure that is readily observable without measurements, testing, or analysis. _The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure. Damage resulting from an equipment failure and limited to the failed component (i.e., the failure did not cause damage to a structure or any other equipment) is not VISIBLE DAMAGE. B-4

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NEI 99 QI (Revisien 6) Nevember 2QI2 PD HU2 ECL: }Tetifieation of Uin1sual Ev:ent Ieitiatieg Ceeditiee: Hazardous event affecting SA.FETY SYSTEM equipment necessary for spent fuel cooling. 0):leFatieg *Mede A):lf)lieahility: Not A.pplicable ExamJ:lle Emergeeey AetieR Levels: (1) a. The occurrence off.u.~Y of the following hazardous ev:ents:

• Seismic e1i*ent (earthq1:1ake)
• Iaternal or eRternal flootling ev:ent
• High v.tin.ds or tornado strike
• FIRE
• EXPLOSION

               *    (site specific hazartls)

• Other ev:ents ',vith similar hazard characteristics as determinetl by the Shift Manager

b. The e*,zent has damaged at least one train of a SAFETY SYSTEM neetletl for spent fuel cooling. *

e. The tlan1agetl SA.FETY SYSTEM train(s) cannot, or potentially cannot, perform its design function based on EITHER:

• Indications of degraded performance
• VISIBLE Di'..MAGE This IC addresses a hazardous ev:ent that causes damage to at least one train of a SAFETY SYSTEM needed for spent fuel cooling. The damage ml.lst be of sufficient magnitude that the system(s) train cannot, or potentially cannot, perform its design function. This condition reduces the margin to a loss or potential loss of the fuel clad barrier, aad therefore represents a 13otefi.tial degradation of the lei,rel of safety of the f)lant.

For EAL l.c, the first bullet addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it 1,vill be readily av:ailable. For EAL l .c, the secontl bullet addresses damage to a SA.FETY SYSTEM train that is not in service/operation or readily apparent through indications alone. Operators ',vill make this B-6

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NEI 99 Q1 (Revisisn 6) Nsvemb@r 2Ql2 PDAA2 ECL1 Alert Initiating C0nditi0n1 UNPL'\NNED rise in plant radiation levels that impedes plant access required to maintain spent fuel integrity. OpeFating *Mede ,A..pplieahility: }Tot ,ALpplicable Example Emergeney f..etiae Levels: (1 or 2) (I) UNPLA}il'ffiD dose rate greater than 15 mR/hr in ANY of the following areas requiring contiB-1:lous occupancy to maintain control of radioactP.J:e material or operation of system.s needed to maintain spent fuel integrity: (site specific area list) (2) UNPLAfil'JED ,A..rea Radiation Monitor readings or Sllfvey reSillts indicate a rise by 100 mR/hr 0 1rer }+QR}.4AL LEVELS that impedes access to A,.W . of the follOYr..ng areas needed to maintain control of radioactive 1Haterial or operation of systems needed to maintain spent fuel integrity. (site specific area list) This IC addresses increased radiation levels that impede necessary access to areas containing equipment that must be operated manually or that req1:1ires local monitoring, in order to maintain systems needed to maintain spent fuel integrity. As 1:1sed here, 'impede' includes hindering or interfering, provided that the interference or delay is sufficient to sigmficantly* threaten necessary plant access. It is this impaired access that results in the actual or potential substantial degradation of the lcw:el of safety of the plant. This IC does not apply to anticipated temporary increases due to plalliled events. DevelapeF Nates: The value of 15mR/hr is derived from the GDC 19 value of5 rem in 30 days with adjustment for expected occupancy times. Although Section III.D.3 of"NUREG 0737, CtGtrificmioo ojT}.{]Actitm Pkm Requirements, provides that the 15 mR/hr value can be averaged over the 30 days, the value is used here 1.vithout averagie.g, as a 30 day duratiofl: implies an e*;ent potentially more significant than an Alert. The specified value of 100 mR/hr may be set to another value for a specific application with appropriate justification. EGL Assignmeat Attributes: 3 .1.2.C B-12

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ATTACHMENT 2 NEXTERA ENERGY DUANE ARNOLD, LLC DUANE ARNOLD ENERGY CENTER RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION RELATING TO LICENSE AMENDMENT REQUEST TSCR-166 UPDATED CLEAN COPY OF THE PROPOSED DAEC EAL SCHEME 125 pages follow

Duane Arnold Energy Center (DAEC) Emergency Action Levels Technical Bases Document . TBD, 2018 I

TABLE OF CONTENTS 1 BASIS FOR EMERGENCY ACTION LEVELS ***************************************************************** 1 1.1 OPERATING REACTORS ************************************************************************************************** 1 1.2 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSl) .....................................2 1.3 NRC ORDER*EA-12-051 ************************************************************************************************3 2 KEY TERMINOLOGY USED IN DAEC EAL SCHEME..................................................... 4 2.1 EME-RGENCY CLASSIFICATION LEVEL (ECL) ................................................... ~...........4 2.2 INITIATING CONDITION (IC) ****************************************************************************************** 6 2.3 EMERGENCY ACTION LEVEL (EAL) *.**********************************.*****..*************************.******* 6 2.4 FISSION PRODUCT BARRIER THRESHOLD ......................................................................6 3 DESIGN OF THE DAEC EMERGENCY CLASSIFICATION SCHEME ............................. 7 3.1 ASSIGNMENT OF EMERGENCY CLASSIFICATION LEVELS (ECLs) ...............................7 3.2 TYPES OF INITIATING CONDITIONS AND EMERGENCY ACTION LEVELS.................... 10 3.3 DAEC-SPECIFIC ORGANIZATION AND PRESENTATION OF GENERIC INFORMATIONll 3.4 IC,AND EAL MODE APPLICABILITY**************************************************************************** 12 4 DEVELOPMENT OF THE DAEC EMERGENCY CLASSIFICATION SCHEME ............... 13

4. 1 GENERAL DEVELOPl\fENT PROCESS ..................:~.:......................................................13 4.2 CRITICAL CHARACTERISTICS ****************************************************************~*********************13 4 . 3 INSTRUMENTATION USED FOR EALs ................ ~~........................................................ 14 4.4 EAL/THRESHOLD REFERENCES TO AOP AND EOP SETPOINTS/CRITERIA .............. 14 5 GUIDANCE ON -,SING THE DAEC EALS ******************************************************************** 15 5 .1 GENERAL CONSIDERATIONS ************~***************************************************************************15 5.2 CµssIFICATION METHODOLOGY *****************************************************~*************************16 5.3 CLASSIFICATION OF MULTIPLE EVENTS AND CONDITIONS ........................................ 16 5.4 CONSIDERATION OF MODE CHANGES DURING CLASSIFICATION .............................. 17 5.5 CLASSIFICATION OF IMMINENT CONDITIONS ............................................................. 17 5.6 EMERGENCY CLASSIFICATION LEVEL UPGRADING AND DOWNGRADING ................. 17

5. 7 CLASSIFICATION OF SHORT-LIVED EVENTS ........................................................... 0 *** 18 5.8 CLASSIFICATION OF TRANSIENT CONDITIONS ............................................................ 18 5.9 AFTER-THE-FACT DISCOVERY OF AN EMERGENCY EVENT OR CONDITION .............. 19 5.10 RETRACTION OF ANEMERGENCYDECLARATION ....................................................... 19 11

6 ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT ICS/EALS ........................ 20 7 COLD SHUTDOWN/ REFUELING SYSTEM MALFUNCTION ICS/EALS .................... 36 8 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS .............. 58 9 FISSION PRODUCT BARRIER ICS/EALS ****************************************************************** 60 10 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS ......... 75 11 SYSTEM MALFUNCTION ICS/EALS*************************************************************************** 96 APPENDIX A - ACRONYMS AND ABBREVIATIONS ....................................................... A-1 APPENDIX B - DEFINITIONS ***11*******************~****************111................................................. B-1 lll

DUANE ARNOLD EMERGENCY ACTION LEVELS TECHNICAL BASIS DOCUMENT 1 BASIS FOR EMERGENCY ACTION LEVELS 1.1 OPERATING REACTORS Title 10, Code of Federal Regulations (CFR), Energy, contains the U.S. Nuclear Regulatory Commission (NRC) regulations that apply to nuclear power facilities. Several of these regulations govern various aspects of an emergency classification scheme. A review of the relevant sections listed below will aid the reader in understanding the key terminology provided in Section 3.0 of this document.

• 10 CFR § 50.47(a)(l)(i)
• 10 CFR § 50.47(b)(4)
• 10 CFR § 50.54(q)
• 10 CFR § 50.72(a)
• 10 CFR § 50, Appendix E, IV.B, Assessment Actions
• 10 CFR § 50, Appendix E, IV.C, Activation of Emergency Organization The above regulations are supplemented by various regulatory guidance documents.

Three documents of particular relevance to NEI 99-01 are: NUREG-0654/FEMA-REP-l, Criteria for Preparation and Evaluation of Radiologi,cal Emergency Response Plans and Preparedness in Support ofNuclear Power Plants, October 1980. [Refer to Appendix 1, Emergency Action Level Guidelines for Nuclear Power Plants] NUREG-1022, Event Reporting Guidelines 10 CFR § 50. 72 and§ 50. 73 Regulatory Guide 1.101, Emergency Response Planning and Preparedness for Nuclear Power Reactors 1

1.2 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) Selected guidance in NEI 99-01 is applicable to licensees electing to use their 10 CFR 50 emergency plan to fulfill the requirements of 10 CFR 72.32 for a stand-alone ISFSI. The emergency classification levels applicable to an ISFSI are consistent with the requirements of 10 CFR 50 and the guidance in NUREG 0654/FEMA-REP-l. The initiating conditions germane to a 10 CFR 72.32 emergency plan (as described in NUREG-1567) are subsumed within the classification scheme for a 10 CFR 50.47 emergency plan. The generic ICs and EALs for an ISFSI are presented in Section 8, ISFSI ICs/EALs. IC E-HUl covers the spectrum of credible natural and man-made events included within the scope of an ISFSI design. This IC is not applicable to installations or facilities that may process and/or repackage spent fuel (e.g., a Monitored Retrievable Storage Facility or an ISFSI at a spent fuel processing facility). In addition, appropriate aspects of IC HUI and IC HAI should also be included to address a HOSTILE ACTION directed against an ISFSI. The analysis of potential onsite and offsite consequences of accid~ntal releases associated with the operation of an ISFSI is contained in NUREG-1140, A Regulatory Analysis on Emergency Preparedness for Fuel Cycle and Other Radioactive Material Licensees. NUREG-1140 concluded that the postulated worst-case accident involving an ISFSI has insignificant consequences to public health and safety. This evaluation shows that the maximum offsite dose to a member of the public due to an accidental release of radioactive materials would not exceed 1 rem Effective Dose Equivalent. Regarding the above information, the expectations for an offsite response to an Alert classified under a 10 CFR 72.32 emergency plan are generally consistent with those for a Notification of Unusual Event in a 10 CFR 50.47 emergency plan (e.g., to provide assistance ifrequested). Also, the licensee's Emergency Response Organization (ERO) required for 10 CFR 72.32 emergency plan is different than that prescribed for a 10 CFR 50.47 emergency plan (e.g., no emergency technical support function). 2

1.3 NRC ORDER EA-12-051 The Fukushima Daiichi accident of March 11, 2012, was the result of a tsunami that exceeded the plant's design basis and flooded the site's emergency electrical power supplies and distribution systems. This caused an extended loss of power that severely compromised the key safety functions of core cooling and containment integrity, and ultimately led to core damage in three reactors. While the loss of power also impaired the spent fuel pool cooling function, sufficient water inventory was maintained in the pools to preclude fuel damage from the loss of cooling. Following a review of the Fukushima Daiichi accident, the NRC concluded that several measures were necessary to ensure adequate protection of public health and safety under the provisions of the backfit rule, 10 CFR 50.109(a)(4)(ii). Among them was to provide each spent fuel pool with reliable level instrumentation to significantly enhance the ability of key decision-makers to allocate resources effectively following a beyond design basis event. To this end, the NRC issued Order EA-12-051, Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation, on March 12, 2012, to all US nuclear plants with an operating license, construction permit, or combined construction and operating license. NRC Order EA-12-051 states, in part, "All licensees ... shall have a reliable indication of the water level in associated spent fuel storage pools capable of supporting identification of the following pool water level conditions by trained personnel: ( 1) level that is adequate to support operation of the normal fuel pool cooling system, (2) level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pooLoperating deck, and (3) level where fuel remains covered and actions to implement make-up water addition should no longer be deferred." To this end, all licensees must provide:

• A primary and back-up level instrument that will monitor water level from the normal level to the top of the used fuel rack in the pool;
• A display in an area accessible following a severe event; and
• Independent electrical power to each instrument channel and provide an alternate remote power connection capability.

NEI 12-02, Industry Guidance for Compliance with NRC Order EA-12-051, "To Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation, " provides guidance for complying with NRC Order EA-12-051. NEI 99-01, Revision 6, includes three EALs that reflect the availability of the enhanced spent fuel pool level instrumentation associated with NRC Order EA-12-051. These EALs are included within ICs RA2, RS2, and RG2. 3

2 KEY TERMINOLOGY USED IN DAEC EAL SCHEME There are several key terms that appear throughout the EAL methodology. These terms are introduced in this section to support understanding of subsequent material. As an aid to the reader, the following table is provided as an overview to illustrate the relationship of the terms to each other. Emergency Classification Level Unusual Event I Alert I SAE I GE I Initiating Condition I Initiating Condition I Initiating Condition I Initiating Condition Emergency Action Emergency Action Emergency Action Emergency Action Level (I) Level (1) Level (1) Level (1)

• Operating Mode
• Operating Mode
• Operating Mode
• Operating Mode Applicability Applicability Applicability Applicability
• Notes
• Notes
• Notes
• Notes
• Basis
• Basis
• Basis
• Basis (1) - When making an emergency classification, the Emergency Director must consider all information having a bearing on the proper assessment of an Initiating Condition. This includes the Emergency Action Level (EAL) plus the associated Operating Mode Applicability, Notes and the informing Basis information. In the Recognition Category F matrices, EALs are referred to as Fission Product Barrier Thresholds; the thresholds serve the same function as an EAL.

2.1 EMERGENCY CLASSIFICATION LEVEL (ECL) One of a set of names or titles established by the US Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions according to (1) potential or actual effects or consequences, and (2) resulting onsite and offsite response actions. The emergency classification levels, in ascending order of severity, are: Notification of Unusual Event (NOUE) Alert Site Area Emergency (SAE) General Emergency (GE) 2.1.1 Notification of Unusual Event (NOUE) Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases ofradioactive material requiring offsite response or monitoring are expected unless further degradation of SAFETY SYSTEMS occurs.

Purpose:

The purpose of this classification is to assure that the first step in future response has been carried out, to bring the operations staff to a state of readiness, and to provide systematic handling of unusual event information and decision-making. 4

2.1.2 Alert Events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA PAG exposure levels.

Purpose:

The purpose of this classification is to assure that emergency personnel are readily available to respond if the situation becomes more serious or to perform confirmatory radiation monitoring if required, and provide offsite authorities current information on plant status and parameters. 2.1.3 Site Area Emergency Events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; 1) toward site personnel or equipment that could lead to the likely failure of or; 2) that prevent effective access to, equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed BPA PAG exposure levels beyond the site boundary.

Purpose:

The purpose of the.Site Area Emergency declaration is to assure that emergency response centers are staffed, to assure that monitoring teams are dispatched, to assure that personnel required for evacuation of near-site areas are at duty stations if the situation becomes more serious, to provide consultation with offsite authorities, and to provide updates to the public through government authorities. 2.1.4 General Emergency (GE) Events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.

Purpose:

The purpose of the General Emergency declaration is to initiate predetermined protective actions for the public, to provide continuous assessment of information from the licensee and offsite organizational measurements, to initiate additional measures as indicated by actual or potential releases, to provide consultation with offsite authorities, and to provide updates for the public through government authorities. 5

2.2 INITIATING CONDITION {IC) An event or condition that aligns with the definition of one of the four emergency classification levels by virtue of the potential or actual effects or consequences .. Discussion: An IC describes an event or condition, the severity or consequences of which meets the definition of an emergency classification level. An IC can be expressed as a continuous, measurable parameter (e.g., RCS leakage), an event (e.g., an earthquake) or the status of one or more fission product barriers (e.g., loss of the RCS barrier). Appendix 1 of NUREG-0654 does not contain example Emergency Action Levels (EALs) for each ECL, but rather Initiating Conditions (i.e., plant conditions that indicate that a radiological emergency, or events that could lead to a radiological emergency, has occurred). NUREG-0654 states that the Initiating Conditions form the basis for establishment by a licensee of the specific plant instrumentation readings (as applicable) which, if exceeded, would initiate the emergency classific_ation. Thus, it is the specific instrument readings that would be the EALs. 2.3 EMERGENCY ACTION LEVEL {EAL) A pre-determined, site-specific, observable threshold for an Initiating Condition that, when met or exceeded, places the plant in a given emergency classification level. Discussion: EAL statements may utilize a variety of criteria including instrument readings and status indications; observable events; results of calculations and analyses; entry into particular procedures; and the occurrence of natural phenomena. 2.4 FISSION PRODUCT BARRIER THRESHOLD A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier. Discussion: Fission product banier thresholds represent threats to the defense in depth design concept that precludes the release of radioactive fission products to the environment. This concept relies on multiple physical barriers, any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment. The primary fission product barriers are: Fuel Clad Reactor Coolant System (RCS) Containment Upon determination that one or more fission product barrier thresholds have been exceeded, the combination of barrier loss and/or potential loss thresholds is compared to the fission product barrier IC/EAL criteria to determine the appropriate ECL. In some accident sequences, the ICs and EALs presented in the Abnormal Radiation Levels/ Radiological Effluent (R) Recognition Category will be exceeded at the same time, or shortly after, the loss of one or more fission product barriers. This redundancy is intentional as the former ICs address radioactivity releases that result in certain offsite doses from whatever cause, including events that might not be fully encompassed by fission product barriers (e.g., spent fuel pool accidents, design containment leakage following a LOCA, etc.). 6

3 DESIGN OF THE DAEC EMERGENCY CLASSIFICATION SCHEME 3.1 ASSIGNMENT OF EMERGENCY CLASSIFICATION LEVELS (ECLs) An effective emergency classification scheme must incorporate a realistic and accurate assessment of risk, both to plant workers and the public. There are obvious health and safety risks in underestimating the potential or actual threat from an event or condition; however, there are also risks in overestimating the threat as well (e.g., harm that may occur during an evacuation). The DAEC emergency classification scheme attempts to strike an appropriate balance between reasonably anticipated event or condition consequences, potential accident trajectories, and risk avoidance or minimization. There are a range of "non-emergency events" reported to the US Nuclear Regulatory Commission (NRC) staff in accordance with the requirements of 10 CFR 50.72. Guidance concerning these reporting requirements, and example events, are provided in NUREG-1022. Certain events reportable under the provisions of 10 CFR 50. 72 may also require the declaration of an emergency. In order to align each Initiating Conditions (IC) with the appropriate ECL, it was necessary to determine the attributes of each ECL. The goal of this process is to answer the question, "What events or conditions should be placed under each ECL?" The following sources provided information and context for the development of ECL attributes. Assessments of the effects and consequences of different types of events and conditions DAEC abnormal and emergency operating procedure setpoints and transition criteria DAEC Technical Specification limits and controls Offsite Dose Assessment Manual (ODAM) radiological release limits Review of selected Updated Final Safety Analysis Report (UFSAR) accident analyses Environmental Protection Agency (EPA) Protective Action Guidelines (PAGs) NUREG 0654, Appendix I, Emergen_cy Action Level Guidelines for Nuclear Power Plants Industry Operating Experience Input from DAEC subject matter experts The following ECL attributes are used to aid in the development ofICs and Emergency Action Levels (EALs). The attributes may be useful in briefing and training settings (e.g., helping an Emergency Director understand why a particular condition is classified as an Alert). 7

3.1.1 Notification of Unusual Event (NODE) A Notification of Unusual Event, as defined in section 2.1.1, includes but is not limited to an event or condition that involves: (A)A precursor to a more significant event or condition. (B) A minor loss of control of radioactive materials or the ability to control radiation levels within the plant. (C) A consequence otherwise significant enough to warrant notification to local, State and Federal authorities. 3.1.2 Alert An Alert, as defined in section 2.1.2, includes but is not limited to an event or condition that involves: (A)A loss or potential loss of either the fuel clad or Reactor Coolant System (RCS) fission product barrier. (B) An event or condition that significantly reduces the margin to a loss or potential loss of the fuel clad or RCS fission product barrier. (C) A significant loss of control of radioactive materials resulting in an inability to control radiation levels within the plant, or a release of radioactive materials to the environment that could result in doses greater than 1% of an EPA P AG at or beyond the site boundary. (D)A HOSTILE ACTION occurring within the OWNER CONTROLLED AREA, in?luding those directed at an Independent Spent Fuel Storage Installation (ISFSI). 3 .1.3 Site Area Emergency (SAE) A Site Area Emergency, as defined in section 2.1.3, includes but is not limited to an event or condition that involves: (A) A loss or potential loss of any two fission product barriers - fuel clad, RCS and/or containment. (B) A precursor event or condition that may lead to the loss or potential loss of multiple fission product barriers within a relatively short period of time. Precursor events and conditions of this type include those that challenge the monitoring and/or control of multiple SAFETY SYSTEMS. (C) A release of radioactive materials to the environment that could result in doses greater than 10% of an EPA PAG at or beyond the site boundary. (D)A HOSTILE ACTION occurring within the plant PROTECTED AREA. 8

3.1.4 General Emergency (GE) A General Emergency, as defmed in section 2.1.4, includes but is not limited to an event or condition that involves: (A)Loss of any two fission product barriers AND loss or potential loss of the third barrier

         - fuel clad, RCS and/or containment.

(B) A precursor event or condition that, unmitigated, may lead to a loss of all three fission product barriers. Precursor events and conditions of this type include those that lead directly to core damage and loss of containment integrity. (C) A release of radioactive materials to the environment that could result in doses greater than an EPA PAG at or beyond the site boundary. (D)A HOSTILE ACTION resulting in the loss of key safety functions (reactivity control, core cooling/RPV water level or RCS heat removal) or damage to spent fuel. 3.1.5 Risk-Informed Insights Emergency preparedness is a defense-in-depth measure that is independent of the assessed risk from any particular accident sequence; however, the development of an effective emergency classification scheme can benefit from a review of risk-based assessment results. To that end, the development and assignment of certain ICs and EALs also considered insights from several site-specific probabilistic safety assessments. Some generic insights from this review included:

1. Accident sequenc~s involving a prolonged loss of all AC power are signific~t contributors to core damage frequency at many Boiling Water Reactors (BWRs). For this reason, a loss of all AC power for greater than 15 minutes, with the plant at or above Hot Shutdown, was assigned an ECL of Site Area Emergency. Precursor events to a.loss of all AC power were also included as an Unusual Event and an Alert.

A station blackout coping analyses performed in response to 10 CFR 50.63 and Regulatory Guide 1.155, Station Blackout, may be used to determine a time-based criterion to demarcate between a Site Area Emergency and a General Emergency. The time dimension is critical to a properly anticipatory emergency declaration since the goal is to maximize the time available for State and local officials to develop and implement offsite protective actions.

2. For severe core damage events, uncertainties exist in phenomena important to accident progressions leading to containment failure. Because of these uncertainties, predicting the status of containment integrity may be difficult under severe accident conditions. This is why maintaining containment integrity alone following sequences leading to severe core damage is an insufficient basis for not escalating to a General Emergency.

3. PSAs indicated that leading contributors to latent fatalities were sequences involving a containment bypass, a large Loss of Coolant Accident (LOCA) with early containment failure, a Station Blackout lasting longer than the DAEC coping period, and a reactor coolant pump seal failure. The generic EAL methodology needs to be sufficiently rigorous to address these sequences in a timely fashion.

9

3.2 TYPES OF INITIATING CONDITIONS AND EMERGENCY ACTION LEVELS The NEI 99-01 methodology makes use of symptom-based, barrier-based and event-based ICs and EALs. Each type is discussed below. Symptom-based ICs and EALs are parameters or conditions that are measurable over some range using plant instrumentation ( e.g., core temperature, reactor coolant level, radiological effluent, etc.). When one or more of these parameters or conditions are off-normal, reactor operators will implement procedures to identify the probable cause(s) and take corrective action. Fission product barrier-based ICs and EALs are the subset of symptom-based EALs that refer specifically to the level of challenge to the principal barriers against the release of radioactive material from the reactor core to the environment. These barriers are the fuel cladding, the reactor coolant system pressure boundary, and the containment. The barrier-based ICs and EALs consider the level of challenge to each individual barrier - potentially lost and lost - and the total number of barriers under challenge. Event-based ICs and EALs define a variety of specific occurrences that have potential or actual safety significance. These include the failure of an automatic reactor scram to shut down the reactor, natural phenomena (e.g., an earthquake), or man-made hazards such as a toxic gas release. 10

3.3 DAEC-SPECIFIC ORGANIZATION AND PRESENTATION OF GENERIC INFORMATION The scheme's generic information is organized by Recognition Category in the following order. R - Abnormal Radiation Levels / Radiological Effluent - Section 6 C - Cold Shutdown / Refueling System Malfunction - Section 7 E - Independent Spent Fuel Storage Installation (ISFSI) - Section 8 F - Fission Product Barrier- Section 9 H - Hazards and Other Conditions Affecting Plant Safety - Section 10 S - System Malfunction - Section 11 Each Recognition Category section contains a matrix showing the ICs and their associated emergency classification levels. The following information and guidance is provided for each IC: ECL - the assigned emergency classification level for the IC. Initiating Condition - provides a summary description of the emergency event or condition. Operating Mode Applicability - Lists the modes during which the IC and associated EAL(s) are applicable (i.e., are to be used to classify events or conditions). Emergency Action Level(s)-Provides examples ofreports and indications that are considered to meet the intent of the IC. For Recognition Category F, the fission product barrier thresholds are presented in tables and arranged by fission product barrier and the degree of barrier challenge (i.e., potential loss or loss). This presentationmethod shows the synergism among the thresholds, and supports accurate assessments. Basis - Provides background information that explains the intent and application of the IC and EALs. In some cases, the basis also includes relevant source information and references. 11

3.4 IC AND EAL MODE APPLICABILITY The DAEC emergency classification scheme was developed recognizing that the applicability of ICs and EALs will vary with plant mode. For example, some symptom-based ICs and EALs can be assessed only during the power operations, startup, or hot standby/shutdown modes of operation when all fission product barriers are in place, and plant instrumentation and SAFETY SYSTEMS are fully operational. In the cold shutdown and refueling modes, different symptom-based ICs and EALs will come into play to reflect the opening of systems for routine maintenance, the unavailability of some SAFETY SYSTEM components and the use of alternate instrumentation. The following table shows which Recognition Categories are applicable in each plant mode. The ICs and EALs for a given Recognition Category are applicable in the indicated modes. MODE APPLICABILITY MATRIX Recognition Category Mode R C E F H s Power Operations X X X X X Startup X X X X X Hot Shutdown X X X X X Cold Shutdown X X X X Refueling X X X X Defueled X X .x X DAEC Operating Modes Power Operations (I): Mode Switch in Run Startup (2): Mode Switch in Startup/Hot Standby or Refuel (with all vessel head closure bolts fully tensioned) Hot Shutdown (3): Mode Switch in Shutdown, Average Reactor Coolant Temperature >212 °F (with all vessel head closure bolts fully tensioned) Cold Shutdown (4): Mode Switch in Shutdown, Average Reactor Coolant Temperature~ 212 °F (with all vessel head closure bolts fully tensioned) Refueling (5): Mode Switch in Shutdown or Refuel (with one or more vessel head closure bolts less than fully tensioned) 12

4 DEVELOPMENT OF THE DAEC EMERGENCY CLASSIFICATION SCHEME 4.1 GENERAL DEVELOPMENT PROCESS The DAEC ICs and EALs were developed to be unambiguous and readily assessable. The IC is the fundamental event or condition requiring a declaration. The EAL(s) is the pre-determined threshold that defines when the IC is met. Useful acronyms and abbreviations associated with the DAEC emergency classification scheme are presented in Appendix A, Acronyms and Abbreviations. Many words or terms used in the DAEC emergency classification scheme have scheme-specific definitions. These words and terms are identified by being set in all capital letters (i.e., ALL CAPS). The definitions are presented in Appendix B, Definitions. 4.2 CRITICAL CHARACTERISTICS When crafting the scheme, DAEC ensured that certain critical characteristics have been met. These critical characteristics are listed below.

• The ICs, EALs, Operating Mode Applicability criteria, Notes and Basis information are consistent with industry guidance; while the actual wording may be different, the classification intent is maintained. With respect to Recognition Category F, DAEC includes a user-aid to facilitate timely and accurate classification of fission product barrier losses and/or potential losses. The user-aid logic is consistent with the classification logic presented in Section 9.
• The ICs, EALs~ Operating Mode Applicability criteria, Notes and Basis information are technically complete and accurate (i.e., they contain the information necessary to make a correct classification).
• EAL statements use objective criteria and observable values.
• ICs, EALs, Operating Mode Applicability and Note statements and formatting consider human factors and are user-friendly.
• The scheme facilitates upgrading and downgrading of the emergency classification where necessary.
• The scheme facilitates classification of multiple concurrent events or conditions.

13

4.3 INSTRUMENTATIONUSEDFOREALS DAEC incorporated instrumentation that is reliable and routinely maintained in accordance with site programs and procedures. Alarms referenced in EAL statements are those that are the most operationally significant for the described event or condition. EAL setpoints are within the calibrated range of the referenced instrumentation, and consider any automatic instrumentation functions that may impact accurate EAL assessment. In addition, EAL setpoint values do not use terms such as "off-scale low" or "off-scale high" since that type of reading may not be readily differentiated from an instrument failure. 4.4 EAL/THRESHOLD REFERENCES TO AOP AND EOP SETPOINTS/CRITERIA Some of the criteria/values used in several EALs and fission product barrier thresholds are drawn from DAEC AOPs and EOPs. This approach is intended to maintain good alignment between operational diagnoses and emergency classification assessments. Appropriate administrative controls are in place to ensure that a subsequent change to an AOP or EOP is screened to determine if an evaluation pursuant to 10 CFR 50.54(q) is required. 14

5 GUIDANCE ON USING THE DAEC EALS 5.1 GENERAL CONSIDERATIONS. When making an emergency classification, the Emergency Director must consider all information having a bearing on the proper assessment of an Initiating Condition (IC). This includes the Emergency Action Level (EAL) plus the associated Operating Mode Applicability, Notes and the informing Basis information. In the Recognition Category F matrices, EALs are referred to as Fission Product Barrier Thresholds; the thresholds serve the same function as an EAL. NRC regulations require the licensee to establish and maintain the capability to assess, classify, and declare an emergency condition within 15 minutes after the availability of indications to plant operators that an emergency action level has been exceeded and to promptly declare the emergency condition as soon as possible following identification of the appropriate emergency classification level. The NRC staff has provided guidance on implementing this requirement in NSIR/DPR-ISG-01, Interim Staff Guidance, Emergency Planning/or Nuclear Power Plants. All emergency classification assessments should be based upon valid indications, reports or conditions. A valid indication, report, or condition, is one that has been verified through appropriate means such that there is no doubt regarding the indicator's operability, the condition's existence, or the report's accuracy. For example, validation could be accomplished through an instrument channel check, response on related or redundant indicators, or direct observation by plant personnel. The validation of indications should be completed in a manner that supports timely emergency declaration. For ICs auci EALs that have a stipulated time duration (e.g., 15 minutes, 30 minutes, etc.), the Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition has (;)Xceeded, or will likely exceed, the applicable time. If an ongoing radiological release is detected and the refease start time is unknown, it should be assumed that the release duration specified in the IC/EAL has been exceeded, absent data to the contrary. A planned work activity that results in an expected event or condition which meets or exceeds an EAL does not warrant an emergency declaration provided that 1) the activity proceeds as planned and 2) the plant remains within the limits imposed by the operating license. Such activities include planned work to test, manipulate, repair, maintain or modify a system or component. In these cases, the controls associated with the planning, preparation and execution of the work will ensure that compliance is maintained with all aspects of the operating license provided that the activity proceeds and concludes as expected. Events or conditions of this type may be subject to the reporting requirements of 10 CFR 50.72. The assessment of some EALs is based on the results of analyses that are necessary to ascertain whether a specific EAL threshold has been exceeded (e.g., dose assessments, chemistry sampling, RCS leak rate calculation, etc.); the EAL and/or the associated basis discussion will identify the necessary analysis. In these cases, the 15-minute declaration 15

period starts with the availability of the analysis results that show the threshold to be exceeded (i.e., this is the time that the EAL information is first available). The NRC expects licensees to establish the capability to initiate and complete EAL-related analyses within a reasonable period of time (e.g., maintain the necessary expertise on-shift). While the EALs have been developed to address a full spectrum of possible events and conditions which may warrant emergency classification, a provision for classification based on operator/management experience and judgment is still necessary. This scheme provides the Emergency Director with the ability to classify events and conditions based upon judgment using EALs that are consistent with the Emergency Classification Level (ECL) defmitions (refer to Category H). The Emergency Director will need to determine if the effects or consequences of the event or condition reasonably meet or exceed a particular ECL definition. A similar provision is incorporated into the Fission Product Barrier Tables; judgment may be used to determine the status of a fission product barrier. 5.2 CLASSIFICATION METHODOLOGY To make an emergency classification, the user will compare an event or condition (i.e., the relevant plant indications and reports) to an EAL(s) and determine if the EAL has been met or exceeded. The evaluation of an EAL(s) must be consistent with the related Operating Mode Applicability and Notes. If an EAL has been met or exceeded, then the IC is considered met and the associated ECL is declared in accordance with plant procedures. When assessing an EAL that specifies a time duration for the off-normal condition, the "clock" for the EAL time duration runs concurrently w:ith the emergency classification process "clock." For a full discussion ofthis timing requirement, refer to NSIR/DPR-ISG-01. 5.3 CLASSIFICATION OF MULTIPLE EVENTS AND CONDITIONS When multiple emergency events or conditions are present, the user will identify all met or exceeded EALs. The highest applicable ECL identified during this review is declared. Additionally, there is no "additive" effect from multiple EALs meeting the same ECL. For example: If two Alert EALs are met, an Alert should be declared. Related guidance concerning classification of rapidly escalating events or conditions is provided in Regulatory Issue Summary (RIS) 2007-02, Clarification ofNRC Guidance for Emergency Notifications During Quickly Changing Events. 16

5.4 CONSIDERATION OF MODE CHANGES DURING CLASSIFICATION The mode in effect at the time that an event or condition occurred, and prior to any plant or operator response, is the mode that determines whether or not an IC is applicable. If an event or condition occurs, and results in a mode change before the emergency is declared, the emergency classification level is still based on the mode that existed at the time that the event or condition was initiated (and not when it was declared). Once a different mode is reached, any new event or condition, not related to the original event or condition, requiring emergency classification should be evaluated against the ICs and EALs applicable to the operating mode at the time of the new event or condition. For events that occur in Cold Shutdown or Refueling, escalation is via EALs that are applicable in the Cold Shutdown or Refueling modes, even if Hot Shutdown (or a higher mode) is entered during the subsequent plant response. In particular, the fission product barrier EALs are applicable only to events that initiate in the Hot Shutdown mode or higher.

• 5.5 CLASSIFICATION OF IMMINENT CONDITIONS Although EALs provide specific thresholds, the Emergency Director must remain alert to events or conditions that could lead to meeting or exceeding an EAL within a relatively short period of time (i.e., a change in the ECL is IMMINENT). If, in the judgment of the Emergency Director, meeting an EAL is IMMINENT, the emergency classification should be made .as if the EAL has been met. While applicable to all emergency classification levels, this approach is particularly important at the higher emergency classification levels since it provides additional time for implementation of protective measures.

5.6 EMERGENCY CLASSIFICATION LEVEL UPGRADING AND DOWNGRADING An ECL may be downgraded when the event or condition that meets the highest IC and EAL no'longer exists, and other site-specific downgrading requirements are met. If downgrading the ECL is deemed appropriate, the new ECL would then be based on a lower applicable IC(s) and EAL(s). The ECL may also simply be terminated. 17

The following approach to downgrading or terminating an ECL is recommended. ECL Action When Condition No Longer Exists Unusual Event Terminate the emergency in accordance with plant procedures. Alert Downgrade or terminate the emergency in accordance with plant procedures. Site Area Emergency with no Downgrade or terminate the emergency in long-term plant damage accordance with plant procedures. Site Area Emergency with Terminate the emergency and enter recovery in long-term plant damage accordance with plant procedures. General Emergency Terminate the emergency and enter recovery in accordance with plant procedures. As noted above, guidance concerning classification of rapidly escalating events or conditions is provided in RIS 2007-02. 5.7 CLASSIFICATION OF SHORT-LIVED EVENTS As discussed in Section 3.2, event-based ICs and EALs define a variety of specific occurrences that have potential or actual safety significance. By their nature, some of these events may be short-lived and, thus, over before the emergency classification assessment can be completed. If an event occurs that meets or exceeds an EAL, the associated ECL must be declared regardless of its continued presence at the time of declaration. Examples of such events include a failure of the reactor protection system to automatically scram the reactor followed by a successful manual scram or an earthquake. 5.8 CLASSIFICATION OF TRANSIENT CONDITIONS

• Many of the ICs and/or EALs contained in this document employ time-based criteria.

These criteria will require that the IC/EAL conditions be present for a defined period of time before an emergency declaration is warranted. In cases where no time-based criterion is specified, it is recognized that some transient conditions may cause an EAL to be met for a brief period of time (e.g., a few seconds to a few minutes). The following guidance should be applied to the classification of these conditions. EAL momentarily met during expected plant response - In instances where an EAL is briefly met during an expected (normal) plant response, an emergency declaration is not warranted provided that associated systems and components are operating as expected, and operator actions are performed in accordance with procedures. 18

EAL momentarily met but the condition is corrected prior to an emergency declaration - If an operator takes prompt manual action to address a condition, and the action is successful in correcting the condition prior to the emergency declaration, then the applicable EAL is not considered met and the associated emergency declaration is not required. For illustrative purposes, consider the following example. An ATWS occurs and the auxiliary feedwater system fails to automatically start. Steam generator levels rapidly decrease and the plant enters an inadequate RCS heat removal condition (a potential loss of both the fuel clad and RCS barriers). If an operator manually starts the auxiliary feedwater system in accordance with an EOP step and clears the inadequate RCS heat removal condition prior to an emergency declaration, then the classification should be based on the ATWS only. It is important to stress that the 15-minute emergency classification assessment period is not a "grace period" during which a classification may be delayed to allow the performance of a corrective action that would obviate the need to classify the event; emergency classification assessments must be deliberate and timely, with no undue delays. The provision discussed above addresses only those rapidly evolving situations where an operator is able to take a successful corrective action prior to the Emergency Director completing the review and steps necessary to make the emergency declaration. This provision is included to ensure that any public protective actions resulting from the emergency classification are truly warranted by the plant conditions. 5.9 AFTER-THE-FACT DISCOVERY OF AN EMERGENCY EVENT OR CONDITION In some cases, an EAL may be met but the emergency classification w.as not made at the time of the event or condition. This situation can occur when personnel discover that an event or condition existed which met an EAL, but no emergency was declared, and the event or condition no longer exists at the time of discovery. This may be due to the event or .condition not being recognized at the time or an error that :was made in the emergency classification process. In these cases, no emergency declaration is warranted; however, the guidance contained in NUREG-1022 is applicable. Specifically, the event should be reported to the NRC in accordance with 10 CFR 50.72 within one hour of the discovery of the undeclared event or condition. The licensee should also notify appropriate State and local agencies in accordance with the agreed upon arrangements. 5.10 RETRACTION OF AN EMERGENCY DECLARATION Guidance on the retraction of an emergency declaration reported to the NRC is discussed in NUREG-1022. 19

6 ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT ICS/EALS 20

RU1 ECL: Notification of Unusual Event Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODAM limits for 60 minutes or longer. Operating Mode Applicability: All Emergency Action Levels: Notes:

• The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
• If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
• If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.

RUl.l Reading on ANY Table R-1 effluent radiation monitor greater than column "NOUE" for 60 minutes or longer: Table R Effluent Monitor Classification Thresholds Monitor NOUE Reactor Build ing ventilation rad 8.0E-04 uci/cc mon itor (Kaman 3/4, 5/6, 7/8) VI Turbine Building ventilation rad

, 8.0E-04 uci/cc 0 monitor (Kaman 1/2)

QJ V) 1------------- ro Offgas Stack rad monitor (!) 2.0E-01 uci/cc (Kaman 9/10) LLRPSF rad monit or 1.2E-03 uci/cc (Kaman 12) GSW rad monitor 1.SE+03 cps (RIS-4767) S RHRSW & ESW rad monitor 8.4 E+02 cps

3° (RM -1997)

RHRSW & ESW Rupture Disc rad l.OE+03 cps mon itor (RM-4268) RUl .2 Reading on ANY effluent radiation monitor greater than 2 times the alarm setpoint established by a current radioactivity discharge permit for 60 minutes or longer. RUl .3 Sample analysis for a gaseous or liquid release indicates a concentration or release rate greater than 2 times the ODAM limits for 60 minutes or longer. 21

Definitions: None Basis: This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g. , an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared. DAEC incorporates design features intended to control the release ofradioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes. Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL.

• EAL RUl .1 - This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways.

EAL RUl .2 - This EAL addresses radioactivity releases that cause effluent radiation monitor readings to exceed 2 times the limit established by a radioactivity discharge permit. This EAL will typically be associated with planned batch releases from non-continuous release pathways (e.g., radwaste, waste gas). EAL RUl.3 - This EAL addresses uncontrolled gaseous or liquid releases that are detected by sample analysis or environmental surveys, particularly on unmonitored pathways (e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.). Escalation of the emergency classification level would be via IC RA 1. 22

RU2 ECL: Notification of Unusual Event Initiating Condition: UNPLANNED loss of water level above irradiated fuel. Operating Mode Applicability: All Emergency Action Levels: RU2 .l a. UNPLANNED water level drop in the REFUELING PATHWAY as indicated by ANY of the following:

• Report to control room (visual observation)
• Fuel pool level indication (LI-3413) less than 36 feet and lowering
• WR GEMAC Floodup indication (LI-4541) coming on scale AND

b. UNPLANNED rise in area radiation levels as indicated by ANY of the following radiation monitors.

                     * *spent Fuel Pool Area, RI-9178

• North Refuel Floor, Rl-9163
• New Fuel Vault Area, Rl-9153 l
• South Refuel Floor, Rl-9164
• NW Drywell Area Hi Range Rad Monitor, R1M-9I84A
• South Drywell Area Hi Range Rad Monitor, R1M-9I84B Definitions:

UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. REFUELING PATHWAY: The reactor refueling cavity, spent fuel pool and fuel transfer canal. 23

Basis: This IC addresses a decrease in water level above irradiated fuel sufficient to cause elevated radiation levels. This condition could be a precursor to a more serious event and is also indicative of a minor loss in the ability to control radiation levels within the plant. It is therefore a potential degradation in the level of safety of the plant. A water level decrease will be primarily determined by indications from available level instrumentation. Other sources of level indications may include reports from plant personnel (e.g. , from a refueling crew) or video camera observations. A significant drop in the water level may also cause an increase in the radiation levels of adjacent areas that can be detected by monitors in those locations. The effects of planned evolutions should be considered. For example, a refueling bridge area radiation monitor reading may increase due to planned evolutions such as lifting of the reactor vessel head or movement of a fuel assembly. Note that this EAL is applicable only in cases where the elevated reading is due to an UNPLANNED loss of water level. 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 indication (e.g. , not due to loss of compensating air signal or other instrument channel failure) ofreactor 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 when moving irradiated fuel into the secondary containment. 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 l 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 Ll-3413 indicated water level below 36 feet and lowering. Increased radiation levels can be detected by the local area radiation monitors surrounding the spent fuel pool and refueling cavity areas. Applicable area radiation monitors are those listed in AOP 981 . A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes. Escalation of the emergency classification level would be via IC RA2. 24

RA1 ECL: Alert Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE. Operating Mode Applicability: All Emergency Action Levels: Notes:

• The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
• If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
• If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
• The pre-calculated effluent monitor values presented in EAL 1.1 should only be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

RAl.l Reading on ANY Table R-1 effluent radiation monitor greater than column "Alert" for 15 minutes or longer: Reactor Building ventilation rad monitor (Kaman 3/4, 5/6, 7/8) Turbine Building ventilation rad monitor 0 (Kaman 1/2) 1.4E-02 uci/cc aJ ro Offgas Stack rad monitor 4.SE+Ol uci/cc l9 (Kaman 9/10) LLRPSF rad monitor 1.4E-02 uci/cc (Kaman 12) GSW rad monitor 1.7E+04 cps (RIS-4767)

               ~

, RHRSW & ESW rad monitor C"

1.2E+04 cps

(RM-1997)

RHRSW & ESW Rupture Disc rad mon itor 1.8E+04 cps (RM-4268) RAI.2 Dose assessment using actual meteorology indicates doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond SITE BOUNDARY. [Preferred] RAI.3 Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY for one hour of exposure. RAl.4 Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:

• Closed window dose rates greater than 10 mR/hr expected to continue for 60 minutes or longer.
• Analyses of field survey samples indicate thyroid CDE greater than 50 mrem for one hour of inhalation.

25

Definitions: SITE BOUNDARY: That line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee. Basis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release). This IC is modified by a note that EAL RAl .1 is only assessed for emergency classification until a qualified dose assessor is performing assessments using dose projection software incorporating actual meteorological data and current radiological conditions. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 1% of the EPA PAGof 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then t4~ .effluent monitor reading is no longer valid for classification purposes. Escalation of the emergency classification level would be via IC RS 1. 26

RA2 ECL: Alert Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel. Operating Mode Applicability: All Emergency Action Levels: RA2.l Uncovery of irradiated fuel in the REFUELING PATHWAY. RA2.2 Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by Hi Rad alarm for ANY of the following ARMs:

• Spent Fuel Pool Area, RI-9178
• North Refuel Floor, RI-9163
• New Fuel Vault Area, RI-9153
• South Refuel Floor, RI-9164 OR Reading greater than 5 R/hr on ANY of the following radiation monitors {in Mode 5 only):
• NW Drywell Area Hi Range Rad Monitor, RIM-9 l 84A

           "

• South Drywell Area Hi Range Rad Monitor, RIM-9 l 84B' RA2.3 Lowering of spent fuel pool level to 25 .17 feet.

Definitions: REFUELING PATHWAY -The reactor refueling cavity, spent fuel pool and fuel transfer canal. Basis: This IC addresses events that have caused IMMINENT or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant.

• Expected radiation monitor alarm(s) during preplanned transfer of highly radioactive material through the affected areas are not considered valid alarms for the purpose of comparison to these EALs.

27

This IC applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with IC E-HUl. Escalation of the emergency would be based on either Recognition Category R or C ICs. EALRA2.l This EAL escalates from RU2 in that the loss oflevel, in the affected portion of the REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation (e.g., reports from personnel or camera images), as well as significant changes in water and radiation levels, or other plant parameters. Computational aids may also be used. Classification of an event using this EAL should be based on the totality of available indications, reports, and observations. While an area radiation monitor could detect an increase in a dose rate due to a lowering of water level in some portion of the REFUELING PATHWAY, the reading may not be a reliable indication of whether or not the fuel is actually uncovered. To the degree possible, readings should be considered in combination with other available indications of inventory loss. A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes. EALRA2.2 This EAL addresses a release of radioactive material caused by mechanical damage to irradiated fuel. Damaging events may include the dropping, bumping or binding of an assembly, or dropping a heavy load onto an assembly. An alarm on these radiation monitors should be considered in conjunction with in-plant reports'or observations of a potential fuel damaging event (e.g.; a fuel handling accident). Threshold values for the Drywell monitors are only applicable in Mode 5 since the calculated radiation levels from damage to irradiated fuel would be masked by the typical background levels on these monitors during plant operation, and mechanical damage to a fuel assembly in the vessel can only happen with the reactor head removed. EALRA2.3 Spent fuel pool water level at this value is within the lower end of the level range necessary to prevent significant dose consequences from direct gamma radiation to personnel performing operations in the vicinity of the spent fuel pool. This condition reflects a significant loss of spent fuel pool water inventory and thus it is also a precursor to a loss of the ability to adequately cool the irradiated fuel assembles stored in the pool. Escalation of the emergency classification level would be via ICs RS 1 or RS2. 28

RA3 ECL: Alert Initiating Condition: Radiation levels that impede access to areas necessary for normal plant operation. Operating Mode Applicability: All Emergency Action Levels: RA3.l Dose rate greater than 15 mR/hr in ANY of the following areas:

• Control Room (RM-9162)
• Central Alarm Station (by survey)

Definitions: None Basis: This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or impede personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Director should consider the cause of the increased radiation levels and determine if another IC may be applicable. .. Escalation of the emergency classification level would be via Recognition Category R, C or F ICs. 29

RS1 ECL: Site Area Emergency Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 rnrem TEDE or 500 mrem thyroid CDE. Operating Mode Applicability: All Emergency Action Levels: Notes:

• The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
• If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
• If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
• The pre-calculated effluent monitor values presented in EAL 1.1 should only be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

RSI.I Reading on ANY Table R-1 effluent radiation monitor greater than column "SAE" for 15 minutes or longer: Monitor Reactor Building ventilation rad monitor (Kaman 3/4, 5/6, 7/8) Vl Turbine Building ventilation rad monitor

J 0 (Kaman 1/2)

QJ Vl ro Offgas Stack rad monitor (!) (Kaman 9/10) RSl.2 Dose assessment using actual meteorology indicates doses greater than 100 mrem TEDE or 500 mrem thyroid CDE at or beyond the SITE BOUNDARY. [Preferred] RSl.3 Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:

• Closed window dose rates greater than 100 mR/hr expected to continue for 60 minutes or longer.
• Analyses of field survey samples indicate thyroid CDE greater than 500 mrem for one hour of inhalation.

30

Definitions: SITE BOUNDARY: That line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee. Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public. This IC is modified by a note that EAL RS 1.1 is only assessed for emergency classification until a qualified dose assessor is performing assessments using dose projection software incorporating actual meteorological data and current radiological conditions. 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. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological eflluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. Classification based on effluent monitor.readings assumes that a release path to the environment is. an established. If the eflluent flow past effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes. If Kaman readings are not valid, field survey results may be utilized to assess this IC using EAL RSI.3. Escalation of the emergency classification level would be via IC RG 1. 31

RS2 ECL: Site Area Emergency Initiating Condition: Spent fuel pool level at 16.36 feet. Operating Mode Applicability: All Emergency Action Levels: RS2.l Lowering of spent fuel pool level to 16.36 feet. Definitions: None Basis: This IC addresses a significant loss of spent fuel pool inventory control and makeup capability leading to IMMINENT fuel damage. This condition entails major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration. It is recognized that this IC would likely not be met until well after another Site Area Emergency IC was met; however, it is included to provide classification diversity. Escalation of the emergency classification level would be via IC RGI or RG2. 32

RG1 ECL: General Emergency Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE. Operating Mode Applicability: All Emergency Action Levels: Notes:

• The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
• If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.
• If the effluent flow past an efiluent monitor is known to have stopped due to actions to isolate the release path, then the efiluent monitor reading is no longer valid for classification purposes.
• The pre-calculated effluent monitor values presented in EAL 1.1 should only be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

RGI.I Reading on ANY Table R-1 effluent radiation monitor greater than column "GE" for 15 minutes or longer: Monitor GE React or Building ventilation rad monitor 1.lE+OO uci/cc {Kaman 3/4, 5/6, 7/8)

                 ~    Turbine Building ventilation rad monitor 1.4E+OO uci/cc
                 ~    {Kaman 1/2)

V,

                 !'CJ

(.9 RGl .2 Dose assessment using actual meteorology indicates doses greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE at or beyond the SITE BOUNDARY. [Preferred] RGl.3 Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:

• Closed window dose rates greater than 1,000 mR/hr expected to continue for 60 minutes or longer.
• Analyses of field survey samples indicate thyroid CDE greater than 5,000 mrem for one hour of inhalation.

33

Definitions: SITE BOUNDARY: That line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee. Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public. This IC is modified by a note that EAL RG 1.1 is only assessed for emergency classification until a qualified dose assessor is performing assessments using dose projection software incorporating actual meteorological data and current radiological conditions. 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. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG forTEDE and thyroid CDE. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is kiio~ to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes. If Kaman readings are not valid, field survey results may be utilized to assess this IC using EAL RG 1.3. 34

RG2 ECL: General Emergency Initiating Condition: Spent fuel pool level cannot be restored to at least 16.36 feet for 60 minutes or longer. Operating Mode Applicability: All Emergency Action Levels: Note:

• The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

RG2.l Spent fuel pool level cannot be restored to at least 16.36 feet for 60 minutes or longer. Definitions: None Basis: This IC addresses a significant loss of spent fuel pool inventory control and makeup capability leading to a prolonged uncovery of spent fuel. This condition will lead to fuel damage and a radiological release to the environment. It is recognized that this IC would likely not be met until well after another General Emergency IC was met; however, it is jncluded to provide classification diversity. 35

7 COLD SHUTDOWN/ REFUELING SYSTEM MALFUNCTION ICS/EALS 36

CU1 ECL: Notification of Unusual Event Initiating Condition: UNPLANNED loss ofRPV inventory for 15 minutes or longer. Operating Mode Applicability: 4, 5 Emergency Action Levels: Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded. CUI.I UNPLANNED loss of reactor coolant results in RPV level less than a required lower limit for 15 minutes or longer. CUl.2 a. RPV level cannot be monitored. AND

b. UNPLANNED level rise in Drywell/Reactor Building Equipment or Floor Drain sump, or Suppression Pool.

Definitions: UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor RPV level concurrent with indications of coolant leakage. Either of these conditions is ~onsidered to be a potential degradation of the level of safety of the plant. Refueling evolutions that decrease RCS water inventory are carefully planned and controlled. An UNPLANNED event that results in water level decreasing below a procedurally required limit warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the core covered. EAL CUI .1 recognizes that the minimum required RPV level can change several times during the course of a refueling outage as different plant configurations and system lineups are implemented. This EAL is met if the minimum level, specified for the current plant conditions, cannot be maintained for 15 minutes or longer. The minimum level is typically specified in the applicable operating procedure but may be specified in another controlling document. The 15-minute threshold duration allows sufficient time for prompt operator actions to restore and maintain the expected water level. This criterion excludes transient conditions causing a brief lowering of water level. 37

EAL CUI .2 addresses a condition where all means to determine RPV level have been lost. If all level indication were to be lost during a loss of RCS inventory event, the operators would need to determine that RSC inventory loss was occurring by observing sump and Suppression Pool 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 Suppression Pool water level may be indicative of valve misalignment or leakage in systems that discharge to the Torus. Sump and Suppression Pool 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. Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CAI or CA3. 38

CU2 ECL: Notification of Unusual Event Initiating Condition: Loss of all but one AC power source to essential buses for 15 minutes or longer. Operating Mode Applicability: 4, 5, Defueled Emergency Action Levels: Note: The Emergency Director should declare the event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded. CU2.l a. AC power capability to 1A3 and 1A4 buses is reduced to a single power source for 15 minutes orlonger. AND

b. Any additional single power source failure will result in loss of ALL AC power to SAFETY SYSTEMS.

Definitions: SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as

• safety-related.

Basis: This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one; or more than one, train of safety-related equipment. When in the cold shutdown, refueling, or defueled mode, this condition is not classified as an Alert because of the increased time available to restore another power source to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition is considered to be a potential degradation of the level of safety of the plant. An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.

• A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).
• A loss of emergency power sources (e.g., onsite diesel generators) with a single train of essential buses beirig fed from an offsite power source.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power. The subsequent loss of the remaining single power source would escalate the event to an Alert in accordance with IC CA2. 39

CU3 ECL: Notification of Unusual Event Initiating Condition: UNPLANNED increase mRCS temperature. Operating Mode Applicability: 4, 5 Emergency Action Levels: Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded. CU3.l UNPLANNED increase in RCS temperature to greater than 212°F. CU3.2 Loss of ALL RCS temperature and RPV level indication for 15 minutes or longer. Definitions: UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. CONTAINMENT CLOSURE: Procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release

 ** under existing plant conditions. For DAEC, this is considered to be Secondary Containment as required by Technical Specifications.
    .Basis:

This IC addresses an UNPLANNED increase in RCS temperature above the Technical Specification cold shutdown temperature limit, or the inability to determine RCS temperature and level, represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Director should also refer to IC CA3. A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification. EAL CU3.l involves a loss of decay heat removal capability, or an addition of heat to the RCS in excess of that which can currently be removed, such that reactor coolant temperature cannot be maintained below the cold shutdown temperature limit specified in Technical Specifications. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation. During an outage, the level in the reactor vessel will normally be maintained above the reactor vessel flange. Refueling evolutions that lower water level below the reactor vessel flange are carefully planned and controlled. A loss of forced decay heat removal at reduced inventory may result in a rapid increase in reactor coolant temperature depending on the time after shutdown. 40

EAL CU3.2 reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor RCS conditions and operators would be unable to monitor key parameters necessary to assure core decay heat removal. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication. Escalation to Alert would be via IC CAI based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria. 41

CU4 ECL: Notification of Unusual Event Initiating Condition: Loss of Vital DC power for 15 minutes or longer. Operating Mode Applicability: 4, 5 Emergency Action Levels: Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded. CU4.l Indicated voltage is less than 105 VDC on BOTH Div I and Div 2 125 VDC buses for 15 minutes or longer. Definitions: SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related. Basis: This IC addresses a loss of Vital DC power which compromises the ability to monitor and

  .. control operable SAFETY SYSTEMS when the plant is in the cold shutdown or refueling mode.

In these modes, the core decay heat load has been significantly reduced, and coolant system temperatures and pressures are lower; these conditions increase the time available to restore a v:ital DC bus to service. Thus, this condition is considered to be a potential degradation of the

 *, level of safety of the plant.                               '
.. As used in this EAL, "required" means the Vital DC buses necessary to support operation of the in-service, or operable, train or trains of SAFETY SYSTEM equipment. For example, if Train A is out-of-service (inoperable) for scheduled outage maintenance work and Train B is in-service (operable), then a loss of Vital DC power affecting Train B would require the declaration of an Unusual Event. A loss of Vital DC power to Train A would not warrant an emergency classification.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Depending upon the event, escalation of the emergency classification level would be via IC CAI or CA3, or an IC in Recognition Category R. 42

CU5 ECL: Notification of Unusual Event Initiating Condition: Loss of all onsite or offsite communications capabilities. Operating Mode Applicability: 4, 5, Defueled Emergency Action Levels: CU5.l Loss of ALL of the following onsite communication methods:

• Plant Operations Radio System
• In-Plant Phone System
• Plant Paging System (Gaitronics)

CU5.2 Loss of ALL of the following offsite response organization communications methods:

• DAEC All-Call phone
• All telephone lines (PBX and commercial)

              * . Cell Phones (including fixed cell phone system)

• Control Room fixed satellite phone system
• FTS Phone system CU5.3 : 'Loss of ALL of the following NRC communications methods:
• FTS Phone system
• All telephone lines (PBX and commercial)
• Cell Phones (including fixed cell phone system)
• Control Room fixed satellite phone system Basis:

This IC addresses a significant loss of on-site or offsite communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to offsite response organizations and the NRC. This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.). EAL CU5.l addresses a total loss of the communications methods used in support of routine plant operations. 43

EAL CU5.2 addresses a total loss of the communications methods used to notify all offsite response organizations of an emergency declaration. The offsite response organizations referred to here are the State of Iowa, Linn County, and Benton County. EAL CU5 .3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration. 44

CA1 ECL: Alert Initiating Condition: Loss ofRPV inventory. Operating Mode Applicability: 4, 5 Emergency Action Levels: Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded. CALI Loss ofRPV inventory as indicated by level less than 119.5 inches. CAl.2 a. RPV level cannot be monitored for 15 minutes or longer AND

b. UNPLANNED level rise in Drywell/Reactor Building Equipment or Floor Drain sump, or Suppression Pool due to a loss of RPV inventory.

Definitions: UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier). This condition represents a potential substantial reduction in the level of plant safety. For EAL CAI .1, a lowering of water level below 119.5 inches indicates that operator actions have not been successful in restoring and maintaining RPV water level. The heat-up rate of the coolant will increase as the available water inventory is reduced. A continuing decrease in water level will lead to core uncovery. Although related, EAL CAI .1 is concerned with the loss of RCS inventory and not the potential concurrent effects on systems needed for decay heat removal (e.g., loss of a Residual Heat Removal suction point). An increase in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3. For EAL CAl.2, the inability to monitor RPV level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, the operators would need to determine that RCS inventory loss was occurring by observing sump and Suppression Pool 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 Suppression Pool water level may be indicative of valve misalignment or leakage in systems that discharge to the Torus. Sump and Suppression Pool level increases must be 45

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 oflevel indication was chosen because it is half of the EAL duration specified in IC CS 1 If the RPV inventory level continues to lower, then escalation to Site Area Emergency would be via IC CSL 46

CA2 ECL: Alert Initiating Condition: Loss of all offsite and all onsite AC power to essential buses for 15 minutes or longer. Operating Mode Applicability: 4, 5, Defueled Emergency Action Levels: Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded. CA2.l Loss of ALL offsite and ALL onsite AC Power to 1A3 and 1A4 buses for 15 minutes or longer. Definitions: SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related. Basis: This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. When in the cold shutdown, refueling, or defueled mode, this condition is not classified as a Sit~ Area Emergency because of the increased time available to restore an emergency bus to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition represents an actual or potential substantial degradation of the level of safety of the plant. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via IC CS 1 or RS 1. 47

CA3 ECL: Alert Initiating Condition: Inability to maintain the plant in cold shutdown. Operating Mode Applicability: 4, 5 Emergency Action Levels: Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded. CA3.l UNPLANNED increase in RCS temperature to greater than 212°F for greater than the duration specified in Table C-2. Table C-2 RCS Heat-up Duration Thresholds CONTAINMENT CLOSURE RCS Integrity Heat-up Duration Status Intact Not applicable 60 minutes* Established 20 minutes* Not intact Not Established 0 minutes

• If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable.

CA3.2 UNPLANNED RCS pressure increase greater than IO psig due to a loss of RCS cooling. , , Definitions: UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. CONTAINMENT CLOSURE: Procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions. For DAEC, this is considered to be Secondary Containment as required by Technical Specifications. Basis: This IC addresses conditions involving a loss of decay heat removal capability or an addition of heat to the RCS in excess of that which can currently be removed. Either condition represents an actual or potential substantial degradation of the level of safety of the plant. A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification. RCS integrity is intact 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). 48

The RCS Heat-up Duration Thresholds table addresses an increase in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact. The 20-minute criterion was included to allow time for operator action to address the temperature increase. The RCS Heat-up Duration Thresholds table also addresses an increase in RCS temperature with the RCS intact. The status of CONTAINMENT CLOSURE is not crucial in this condition since the intact RCS is providing a high pressure barrier to a fission product release. The 60-minute time frame should allow sufficient time to address the temperature increase without a substantial degradation in plant safety. Finally, in the case where there is an increase in RCS temperature, the RCS is not intact, and CONTAINMENT CLOSURE is not established, no heat-up duration is allowed (i.e., 0 minutes). This is because

1) the evaporated reactor coolant may be released directly into the Containment atmosphere and subsequently to the environment, and

2) there is reduced reactor coolant inventory above the top of irradiated fuel.

EAL CA3.2 provides a pressure-based indication of RCS heat-up. Escalation of the emergency classification level would be via IC CS 1 or RS 1. 49

CA6 ECL: Alert Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode. Operating Mode Applicability: 4, 5 Emergency Action Levels: Notes:

• If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then this emergency classification is not warranted.
• If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted.

CA6.1 a. The occurrence of ANY of the Table C-3 hazardous events: Seismic event (earthquake) Internal or external flooding event High winds or tornado strike FIRE EXPLOSION Other events with sir;nilar hazard characteristics as determined by the Shift Manager or Emergency Director AND

b. 1. Event damage has caused indications of degraded performance in one train of a SAFETY SYSTEM needed for the current operating mode.

AND

2. EITHER of the following:

• Event damage has caused indications of degraded performance to a second train of the SAFETY SYSTEM needed for the current operating mode, OR
• The event has resulted in VISIBLE DAMAGE to the second train of a SAFETY SYSTEM needed for the current operating mode.

50

Definitions: FIRE: Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed. EXPLOSION: A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction, or ove:rpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events may require a post-event inspection to determine if the attributes of an explosion are present. SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related. VISIBLE DAMAGE: Damage to a component or structure that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure. Damage resulting from an equipment failure and limited to the failed component (i.e., the failure did not cause damage to a structure or any other equipment) is not VISIBLE DAMAGE. Basis: This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the current operating mode. In order to provide the appropriate context for consideration of an ALERT classification, the hazardous event must have caused indications of degraded SAFETY ,, SYSTEM performance in one train, and there µmst be either indications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance, and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE such that the potential exists for performance issues. Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has indications of degraded performance for criteria CA6. l .b. l of this EAL; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events. An event affecting a single-train SAFETY SYSTEM (i.e., there are indications of degraded performance and/or VISIBLE DAMAGE affecting the one train) would not be classified under CA6 because the two-train impact criteria that underlie the EALs and Bases would not be met. If an event affects a single-train SAFETY SYSTEM, then the emergency classification should be made based on plant parameters/symptoms meeting the EALs for another IC. Depending upon the circumstances, classification may also occur based on Shift Manager/Emergency Director judgement. Indications of degraded performance addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. 51

VISIBLE DAMAGE addresses damage to a SAFETY SYSTEM train that is not in service/operation and that potentially could cause performance issues. Operators will make this determination based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. Escalation of the emergency classification level would be via IC RS 1. 52

CS1 ECL: Site Area Emergency Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability. Operating Mode Applicability: 4, 5 Emergency Action Levels: Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded. CSl.l a. CONTAINMENT CLOSURE not established. AND

b. RPV level less than +64 inches CS1.2 a. CONTAINMENT CLOSURE established.

AND

b. RPV level less than +15 inches CS1.3 a. RPV level cannot be monitored for 30 minutes or longer.

AND ,, . ,,

b. Core uncovery is indicated by EITHER of the following:

• Drywell Monitor (9184.AIB) reading greater than 5.0 R/hr
• UNPLANNED level rise in Drywell/Reactor Building Equipment or Floor Drain sump, or Suppression Pool of sufficient magnitude to indicate core uncovery Definitions:

CONTAINMENT CLOSURE: Procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions. For DAEC, this is considered to be Secondary Containment as required by Technical Specifications. UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. 53

Basis: This IC addresses a significant and prolonged loss of RPV inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration. Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If reactor vessel level cannot be restored, fuel damage is probable. Outage/shutdown contingency plans typically provide for re-establishing or verifying CONTAINMENT CLOSURE following a loss of heat removal or RCS inventory control functions. The difference in the specified reactor vessel levels ofEALs CSI.1.b and CSl.2.b reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment. In the Cold Shutdown and Refueling Modes, LT/LI-4559, 4560, and 4561 (RX VESSEL NARROW RANGE LEVEL) instruments read up to 22" high due to hot calibrations. LI-4541 (WR GEMAC, FLOODUP) should be used in these Modes for comparison to EAL thresholds since it is calibrated cold and reads accurately. If normal means ofRPV level indication are not available due to plant evolutions, redundant means ofRPV 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. In EAL CSI.3.a, the 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring. The inability to monitor RPV level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative ofleakage from the RPV. These EALs address concerns raised by Generic Letter 88-17, Loss ofDecay 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. Escalation of the emergency classification level would be via IC CG I or RGI. 54

CG1 ECL: General Emergency Initiating Condition: Loss of RPV inventory affecting fuel clad integrity with containment challenged. Operating Mode Applicability: 4, 5 Emergency Action Levels: Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded. CGI.1 a. RPV level less than +15 inches for 30 minutes or longer. AND

b. ANY indication from the Secondary Containment Challenge Table C-1.

CGI.2 a. RPV level cannot be monitored for 30 minutes or longer. AND

b. Core uncovery is indicated by EIHER of the following:

• Drywell Monitor (9184A/B) reading greater than 5.0 R/hr.
• UNPLANNED level rise in Drywell/Reactor Building Equipment or Floor Drain sump, or Suppression Pool of sufficient magnitude to indicate core uncovery.

AND

c. ANY indication from the Secondary Containment Challenge Table C-1.

Table C-1 Secondary Containment Challenge

• CONTAINMENT CLOSURE not established*
• Drywell Hydrogen or Torus Hydrogen greater than 6% AND Drywell Oxygen or Torus Oxygen greater than 5%
• UNPLANNED increase in containment pressure
• Secondary containment radiation monitors above max safe operating limits (MSOL) ofEOP 3, Table 6
• If CONTAINMENT CLOSURE is re-established prior to exceeding the 30 minute time limit, then declaration of a General Emergency is not required.

55

Definitions: CONTAINMENT CLOSURE: Procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions. For DAEC, this is considered to be Secondary Containment as required by Technical Specifications. UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA P AG exposure levels offsite for more than the immediate site area. Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If reactor vessel level cannot be restored, fuel damage is probable. With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required. The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a challenge to Containment integrity. 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 gas mixture in containment. If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged. In EAL CG 1.2.a, the 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring. For EAL CGl.2.b, the calculated radiation level on the Drywell Monitors (9184A/B) is without the reactor head in place. Calculated in radiation levels with the reactor head in place are below the normal variation in background readings of these monitors. 56

The inability to monitor RPV level may be caused by instrumentation and/or power failures or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV. For the Containment Challenge Table, Secondary Containment max safe operating (MSOL) limits from EOP 3 are defined as the highest parameter value at which neither: (1) equipment necessary for the safe shutdown of the plant will fail nor (2) personnel access necessary for the safe shutdown of the plant will be precluded. These EALs address concerns raised by Generic Letter 88-17, Loss ofDecay Heat Removal; SECY 91-283, Evaluation ofShutdown 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. 57

8 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS 58

E-HU1 ECL: Notification of Unusual Event Initiating Condition: Damage to a loaded cask CONFJNEMENT BOUNDARY. Operating Mode Applicability: All Emergency Action Levels: E-HUl.l Damage to a loaded cask CONFJNEMENT BOUNDARY as indicated by a radiation reading greater than the values shown on Table E-1 on the spent fuel cask. Table E-1 Cask Dose Rates 61BTDSC 3 feet from HSM Surface 800mrem/hr Outside HSM Door- Centerline of DSC 200mrem/hr End Shield Wall Exterior 40mrem/hr Definition: CONFJNEMENT BOUNDARY: The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. Basis: This IC addresses an event that results in damage to the CONFJNEMENT BOUNDARY of a storage cask containing spent fuef It applies to irradiated fuel that is licensed for dry storage beginning at the point that the loaded storage cask is sealed. The issues of concern are the creation of a potential or actual release path to the environment, degradation of one or more fuel assemblies due to environmental factors, and configuration changes which could cause challenges in removing the cask or fuel from storage. The existence of "damage" is determined by radiological survey. The technical specification multiple of"2 times", which is also used in Recognition Category RIC RUl, is used here to distinguish between-non-emergency and emergency conditions. The emphasis for this classification is the degradation in the level of safety of the spent fuel cask and not the magnitude of the associated dose or dose rate. It is recognized that in the case of extreme damage to a loaded cask, the fact that the "on-contact" dose rate limit is exceeded may be determined based on measurement of a dose rate at some distance from the cask. Security-related events for ISFSis are covered under ICs HUl and HAL 59

9 FISSION PRODUCT BARRIER ICS/EALS FA1' No

                                          .*FG1 60

Table F-1: DAEC EAL Fission Product Barrier Table Thresholds for LOSS or POTENTIAL LOSS of Barriers FAlALERT FSl SITE AREA EMERGENCY FGlGENERALEMERGENCY ANY Loss OR ANY Potential Loss of Loss OR Potential Loss of ANY two barriers. EITHER the Fuel Clad OR RCS barrier. LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS

1. RCS Activity 1. Primary Containment Conditions 1. Primary Containment Conditions A. Coolant activity Not Applicable A. Primary

• N~t Applicable A. UNPLANNED A. Torus pressure greater than 300 containment rapid drop in greater than 53

  µCi/gm dose                               pressure greater                             Drywell pressure      psig equivalent 1-131.                         than 2 psig due to                           following Drywell     OR RCS leakage.                                pressure rise       B. Drywell or Torus OR                    H2 cannot be B. Drywell pressure       determined to be response not           less than 6% and consistent with       Drywell OR Torus LOCA conditions.       02 cannot be OR                     determined to be C. UNISOLABLE             less than 5%

direct downstream OR pathway to the C. HCL (Graph 4 of environment exists EOP 2) exceeded. after primary containment isolation signal OR D. Intentional primary containment venting per EOPs 61

LOSS POTENTIAL LOSS . LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS

2. RPV Water Level 2. RPV Water Level 2. RPV Water Level A. SAG entry is A. RPV water level A. RPV water level Not Applicable Not Applicable A. SAG entry is required cannot be restored cannot be restored required and maintained and maintained above +15 inches above +15 inches OR cannot be OR cannot be determined. determined.

3. Not Applicable 3. RCS Leak Rate 3. Primary Containment Isolation Failure Not Applicable Not Applicable A. UNISOLABLE A. UNISOLABLE A. UNISOLABLE Not Applicable break in Main

• primary system primary system Steam, HPCI, leakage that leakage that Feedwater, results in results iri RWCU, or RCIC exceeding the exceeding the as indicated by the Max Normal Max Safe failure of both Operating Limit Operating Limit isolation valves in (MNOL) of EOP (MSOL) ofEOP ANY one line to 3, Table 6 for 3, Table 6 for close AND EITHER of the EITHER of the EITHER: following: following:
• HighMSL
• Temperature
• Temperature flow or steam OR OR tunnel
• Radiation
• Radiation Level temperature Level annunciators*

OR

• Direct report of steam release OR B. Emergency RPV Depressurization required. -

62

LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS

4. Primary Containment Radiation 4. Primary Containment Radiation 4. Primary Containment Radiation A. Drywell Monitor Not Applicable A. Drywell Monitor Not Applicable Not Applicable A. Drywell Monitor (9184NB) (9184A/B) (9184A/B) reading greater reading greater reading greater than 1250 R/hr. than 5 R/hr after than 5000 R/hr.

OR reactor shutdown OR B. Torus Monitor B. Torus Monitor (9185NB) (9185A/B) reading greater reading greater than 125 R/hr than 500 R/hr

5. Other Indications 5. Other Indications 5. Other Indications A. Fuel damage Not Applicable Not Applicable Not Applicable Not Applicable A. Fuel damage assessment assessment indicates at least indicates at least 5% fuel clad 20% fuel clad damage. damage.

6. Emergency Director Judgment 6. Emergency Director Judgment 6. Emergency Director Judgment A. ANY condition in A. ANY condition in A. ANY*condition in A. ANY condition in A. ANY condition in A. ANY condition in the opinion of the the opinion of the the opinion of the the opinion of the the opinion of the the opinion of the Emergency Emergency Emergency Emergency Emergency Emergency Director that Director that Director that Director that Director that Director that indicates Loss of indicates Potential indicates Loss of indicates Potential indicates Loss of indicates Potential the Fuel Clad Loss of the Fuel the RCS Barrier. Loss of the RCS the Containment Loss of the Barrier. Clad Barrier. Barrier. Barrier. Containment Barrier.

63

Basis Information For DAEC EAL Fission Product Barrier Table F-1 DAEC FUEL CLAD BARRIER THRESHOLDS: The Fuel Clad barrier consists of the zircalloy or stainless steel fuel bundle tubes that contain the fuel pellets.

1. RCS Activity Loss I.A This threshold indicates that RCS radioactivity concentration is greater than 300 µCi/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.

It is recognized that sample collection and analysis ofreactor coolant with highly elevated activity levels could require several hours to complete. Nonetheless, a sample-related threshold is included' as a backup to other indications.' There is no Potential Loss threshold associated with RCS Activity.

2. RPV Water Level Loss2.A The Loss threshold represents any EOP requirement for entry into the Severe Accident Guidelines.

This is identified in the BWROG EPGs/SAGs when adequate core cooling cannot be assured. Potential Loss 2.A This water level corresponds to the top of the active fuel and is used in the EOPs to indicate a challenge to core cooling. The RPV water level threshold is the same as RCS barrier Loss threshold 2.A. Thus, this threshold indicates a Potential Loss of the Fuel Clad barrier and a Loss of the RCS barrier that appropriately escalates the emergency classification level to a Site Area Emergency. 64

DAEC FUEL CLAD BARRIER THRESHOLDS (cont.): This threshold is considered to be exceeded when, as specified in the EOPs, RPV water cannot be restored and maintained above the specified level following depressurization of the RPV (either manually, automatically or by failure of the RCS barrier) or when procedural guidance or a lack oflow pressure RPV injection sources preclude Emergency RPV depressurization. EOPs allow the operator a wide choice ofRPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurization of the RPV in order to facilitate RPV water level control with low-pressure injection sources. In some events, elevated RPV pressure may prevent restoration of RPV water level until pressure drops below the shutoff heads of available injection sources. Therefore, this Fuel Clad barrier Potential Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss ofRPV inventory. The term "cannot be restored and maintained above" means the value ofRPV water level is not able to be brought above the specified limit (top of active fuel). The determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the current value is below the top of active fuel, but does not permit extended operation below the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained; In high-power ATWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level to the top of active fuel in order to reduce reactor power. RPV water level is then contrqlled between the top of active fuel and the Minimum Steam Cooling RPV Water Level (MSCRWL). Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority. For such events, ICs SA6 or SS6 will dictate the need for emergency classification. Since the loss of ability to determine if adequate core cooling is being provided presents a significant challenge to the fuel clad barrier, a potential loss of the fuel clad barrier is specified.

3. Not Applicable (included for numbering consistency between barrier tables) 65

DAEC FUEL CLAD BARRIER THRESHOLDS (cont.):

4. Primary Containment Radiation Loss 4.A and Loss 4.B The Drywell and Torus radiation monitor readings correspond to an instantaneous release of all reactor coolant mass into the Drywell or Torus, assuming that reactor coolant activity corresponds to approximately 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.

The radiation monitor readings in this threshold are higher than that specified for RCS Barrier Loss threshold 4.A since it indicates a loss of both the Fuel Clad Barrier and the RCS Barrier. Note that a combination of the two monitor readings appropriately escalates the emergency classification level to a Site Area Emergency. There is no Potential Loss threshold associated with Primary Containment Radiation.

5. Other Indications Loss 5.A Results obtained from procedure PASAP 7.2, Fuel Damage Assessment, indicate at least 5% fuel clad damage.

There is no Potential Loss threshold associated with Other Indications.

6. Emergency Director Judgment Loss 6.A This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad Barrier is lost.

Potential Loss 6.A This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Fuel Clad Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored. 66

DAEC RCS BARRIER THRESHOLDS: The RCS Barrier is the reactor coolant system pressure boundary and includes the RPV and all reactor coolant system piping up to and including the isolation valves.

1. Primary Containment Conditions Loss 1.A 2 psig is the drywell high pressure scram setpoint which indicates a LOCA by automatically initiating ECCS.

There is no Potential Loss threshold associated with Primary Containment Pressure.

2. RPV Water Level Loss2.A

       +15 inches corresponds to the top of active fuel (TAF) and is, used in the EOPs to indicate challenge to core cooling.

The RPV water level threshold is the same as Fuel Clad barrier Potential Loss threshold 2.A. Thus, this threshold indicates a Loss of the RCS barrier and Potential Loss of the Fuel Clad barrier and that appropriately escalates the emergency classification level to a Site Area Emergency. This threshold is considered to be exceeded when, as specified in the EOPs, RPV water cannot be restored and maintained above the specified level following depressurization of the RPV (either manually, automatically or by failure of the RCS barrier) or when procedural guidance or a lack oflow pressure RPV injection sources preclude Emergency RPV depressurization EOPs allow the operator a wide choice ofRPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurization ofthe RPV in order to facilitate RPV water level control with low-pressure injection sources. In some events, elevated RPV pressure may prevent restoration ofRPV water level until pressure drops below the shutoff heads of available injection sources. Therefore, this RCS barrier Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss ofRPV inventory. The term, "cannot be restored and maintained above," means the value of RPV water level is not able to be brought above the specified limit (top of active fuel). The determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the current value is below the top of active fuel but does not permit extended operation beyond the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained. 67

DAEC RCS BARRIER THRESHOLDS (cont.): In high-power ATWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level to the top of active fuel in order to reduce reactor power. RPV water level is then controlled between the top of active fuel and the Minimum Steam Cooling RPV Water Level (MSCRWL). Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority. For such events, ICs SAS or SS5 will dictate the need for emergency classification. There is no RCS Potential Loss threshold associated with RPV Water Level.

3. RCS Leak Rate Loss Threshold 3.A Large high-energy lines that rupture outside primary containment can discharge significant amounts of inventory and jeopardize the pressure-retaining capability of the RCS until they are isolated. If it is determined that the ruptured line cannot be promptly isolated from the Control Room, the RCS barrier Loss threshold is met.

Loss Threshold 3 .B Emergency RPV Depressurization maccordance with the EOPs is indicative of a loss of the RCS barrier. If Emergency RPV Depressurization is performed, the plant operators are directed to open safety relief valves (SRVs) and keep them open. Even though the RCS is being vented into the suppression pool, a Loss of the RCS barrier exists due to the diminished effectiveness of the RCS to retain fission products within its boundary. Potential Loss Threshold 3.A Potential loss of RCS based on primary system leakage outside the primary containment is determined from EOP temperature or radiation Max Normal Operating values in areas such as main steam line tunnel, RCIC, HPCI, etc., which indicate a direct path from the RCS to areas outside primary containment.

• A Max Normal Operating Limit (MNOL) value is'the highest value of the identified parameter expected to occur during normal plant operating conditions with all directly associated support and control systems functioning properly.

The indicators reaching the threshold barriers and confirmed-to be caused by RCS leakage from a primary system warrant an Alert classification. A primary system is defined to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system. An UNISOLABLE leak which is indicated by MNOL values escalates to a Site Area Emergency when combined with Containment Barrier Loss threshold 3.A (after a containment isolation) and a General Emergency when the Fuel Clad Barrier criteria is also exceeded. 68

DAEC RCS BARRIER THRESHOLDS (cont.):

4. Primary Containment Radiation Loss4.A The Drywell monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for Fuel Clad Barrier Loss threshold 4.A since it indicates a loss of the RCS Barrier only.

There is no Potential Loss threshold associated with Primary ~ontainment Radiation.

5. Other Indications There are no Loss or Potential Loss thresholds associated with Other Indications.

6. Emergency Director Judgment Loss 6.A This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the RCS barrier is lost.

Potential Loss 6.A This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored. 69

DAEC CONTAINMENT BARRIER THRESHOLDS: The Primary Containment Barrier includes the drywell, the wetwell, their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves. Containment Barrier thresholds are used as criteria for escalation of the ECL from Alert to a Site Area Emergency or a General Emergency.

1. Primary Containment Conditions Loss I .A and l .B Rapid UNPLANNED loss of drywell pressure (i.e., not attributable to drywell spray or condensation effects) following an initial pressure increase indicates a loss of drywell integrity. Drywell pressure should increase as a result of mass and energy release into the primary containment from a LOCA. Thus, drywell pressure not increasing under these conditions indicates a loss of primary containment integrity.

These thresholds rely on operator recognition of an unexpected response for the condition and therefore a specific value is not assigned. The unexpected (UNPLANNED) response is important because it is the indicator for a containment bypass condition. Loss l.C The use of the modifier "direct" in defining the reiease path discriminates against release paths through interfacing liquid systems or minor release pathways, such as instrument lines, not protected by the Primary Containment Isolation System (PCIS). The existence of a filter is not considered in the threshold assessment. Filters do not remove fission product noble gases. In addition, a filter could become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream. Following the leakage of RCS mass into primary containment and a rise in primary containment pressure, there may be minor radiological releases associated with allowable primary containment leakage through various penetrations or system components. Minor releases may also occur if a primary containment isolation valve(s) fails to close but the primary containment atmosphere escapes to an enclosed system. These releases do not constitute a loss or potential loss of primary containment but should be evaluated using the Recognition Category R ICs. Loss l.D EOPs may direct primary containment isolation valve logic(s) to be intentionally bypassed, even if offsite radioactivity release rate limits will be exceeded. Under these conditions with a valid primary containment isolation signal, the containment should also be considered lost if primary containment venting is actually performed. Intentional venting of primary containment for primary containment pressure or combustible gas control to the secondary containment and/or the environment is a Loss of the Containment. Venting for primary containment pressure control when not in an accident situation (e.g., to control pressure below the drywell high pressure scram setpoint) does not meet the threshold condition. 70

DAEC CONTAINMENT BARRIER THRESHOLDS (cont.): Potential Loss 1.A The threshold pressure is the Torus internal design pressure. Structural acceptance testing demonstrates the capability of the primary containment to resist pressures greater than the internal design pressure. A pressure of this magnitude is greater than those expected to result from any design basis accident and, thus, represent a Potential Loss of the Containment barrier. Potential Loss l .B If hydrogen concentration reaches or exceeds the lower flammability limit, as defined in plant EOPs, in an oxygen rich environment, a potentially explosive mixture exists. If the combustible mixture ignites inside the primary containment, loss of the Containment barrier could occur. Potential Loss 1.C The Heat Capacity Limit (HCL) is the highest suppression pool temperature from which Emergency RPV Depressurization will not raise:

• Suppression chamber temperature above the maximum temperature capability of the suppression chamber and equipment within the suppression chamber which may be required to operate when the RPV is pressurized, OR
• Suppression chamber pressure above Primary Containment Pressure Limit A, while the rate of energy transfer from the RPV to the containment is greater than the capacity of the containment vent.

The HCL is a function ofRPV pressure, suppression pool temperature and suppression pool water level. It is utilized to preclude failure of the containment and equipment in the containment necessary for the safe shutdown of the plant and therefore, the inability to maintain plant parameters below the limit constitutes a potential loss of containment. 71

DAEC CONTAINMENT BARRIER THRESHOLDS (cont.):

2. RPV Water Level There is no Loss threshold associated with RPV Water Level.

Potential Loss 2.A The Potential Loss threshold is identical to the Fuel Clad Loss RPV Water Level threshold 2.A. The Potential Loss requirement for Primary Containment Flooding indicates adequate core cooling cannot be restored and maintained and that core damage is possible. BWR EPGs/SAGs specify the conditions that require primary containment flooding. When primary containment flooding is required, the EPGs are exited and SAGs are entered. Entry into SAGs is a logical escalation in response to the inability to restore and maintain adequate core cooling. PRA studies indicate that the condition of this Potential Loss threshold could be a core melt sequence which, if not corrected, could lead to RPV failure and increased potential for primary containment failure. In conjunction with the RPV water level Loss thresholds in the Fuel Clad and RCS barrier columns, this threshold results in the declaration of a General Emergency.

3. Primary Containment Isolation Failure These thresholds address incomplete containment isolation that allows an UNISOLABLE direct release to the environment.

Loss 3.A The Max Safe Operating Limit (MSOL) for Temperature and Radiation Level are each the highest value of these parameters at which neither: (I) equipment necessary for the safe shutdown of the plant will fail, nor (2) personnel access necessary for the safe shutdown of the plant will be precluded. EOPs utilize these temperatures and radiation levels to establish conditions under which RPV depressurization is required. The temperatures and radiation levels should be confirmed to be caused by RCS leakage from a primary system. A primary system is defined to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an un:isolated break in the system. In combination with RCS potential loss 3.A this threshold would result in a Site Area Emergency. There is no Potential Loss threshold associated with RCS Leak Rate. 72

DAEC CONTAINMENT BARRIER THRESHOLDS (cont.):

4. Primary Containment Radiation There is no Loss threshold associated with Primary Containment Radiation.

Potential Loss 4.A The drywell radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the drywell, assuming that 20% of the fuel cladding has failed. The radiation monitor reading for the torus corresponds to an instantaneous release of all reactor coolant mass directly into the torus, assuming that 20% of the fuel cladding has failed. This level of fuel clad failure is well above that used to determine the analogous Fuel Clad Barrier Loss and RCS Barrier Loss thresholds. NUREG-1228, Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents, indicates the fuel clad failure must be greater than approximately 20% in order for there to be a major release of radioactivity requiring offsite protective actions. For this condition to exist, there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential loss of containment which would then escalate the emergency classification level to a General Emergency.

5. Other Indications There is no Loss threshold associated with Other Indications Potential Loss 5.A Results obtained from procedure P ASAP 7 .2, Fuel Damage A~sessment, indicate at least 25% fuel clad damage. PASAP 7 .2 only shows whether fuel damage is greater than or less than 25%, thus this indication is not likely to be declared before containment barrier potential loss 4.A which indicates 20% fuel damage. However, this potential loss threshold adds an additional layer of diversity to the scheme.

6. Emergency Director Judgment Loss 6.A This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Containment barrier is lost.

Potential Loss 6.A This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Containment Barrier is potentially lost. The Emergency Director should also consid~r whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored. 73

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10 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS 75

HU1 ECL: Notification of Unusual Event Initiating Condition: Confirmed SECURITY CONDITION or threat. Operating Mode Applicability: All Emergency Action Levels: HUI.I A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by DAEC Security Shift Supervision. HUl.2 Notification of a credible security threat directed at DAEC. HUI.3 A validated notification from the NRC providing information of an aircraft threat. Definitions: SECURITY CONDITION: Any Security Event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a

• potential degradation to the level of safety of the plant. A SECURITY CONDITION does not involve a HOSTILE ACTION.

HOSTILE ACTION: An act toward DAEC or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILES, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the* nuclear power plant. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related. Basis: This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR 73.71 or 10 CFR 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HAI, HSI and HGl. Timely and accurate communications between DAEC Security Shift Supervision and the Control Room is essential for proper classification of a security-related event. Classification of these events will initiate appropriate threat-related notifications to plant personnel and offsite response organizations. 76

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program]. EAL HUl.l references DAEC Security Shift Supervision because these are the individuals trained to confirm that a security event is occurring or has occurred. Training on security event confirmation and classification is controlled due to the nature of Safeguards and 10 CFR § 2.390 information. EAL HU 1.2 addresses the receipt of a credible security threat. The credibility of the threat is assessed iti accordance with Abnormal Operating Procedure (AOP) 914, Security Events. EAL HUl .3 addresses the threat from the impact of an aircraft on the plant. The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may also be provided by NORAD through the NRC. Validation of the threat is performed in accordance with Abnormal Operating Procedure (AOP) 914, Security Events. Emergency plans and implementing procedures are public documents;.therefore, EALs do not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information is contained in the Security Plan.

• Escalation of the emergency classification level would be*via IC HAL 77

HU2 ECL: Notification of Unusual Event Initiating Condition: Seismic event greater than OBE levels. Operating Mode Applicability: All Emergency Action Levels: HU2. l Seismic event greater than Operating Basis Earthquake (OBE) as indicated by receipt of the Amber Operating Basis Earthquake Light and the wailing seismic alarm on IC35. Definitions: DESIGN BASIS EARTHQUAKE (DBE): A DBE is vibratory ground motion for which certain (generally, safety-related) structures, systems, and components must be designed to remain functional. OPERATING BASIS EARTHQUAKE (OBE): An OBE is vibratory ground motion for which those features of a nuclear power plant necessary for continued operation without undue risk to the health and safety of the public will remain functional. Basis: This IC addresses a seismic event that results in accelerations at the plant site greater than those specified for an Operating Basis Earthquake (OBE). An earthquake greater than an OBE but less than a Design Basis Earthquak~ (DBE) should have no significant impact on safety-relate.d systems, structures and components; however, some time may be required for the plant staff to ascertain the actual post-event condition of the plant (e.g., performs walk-downs and post-event inspections). Given the time necessary to perform walk-downs and inspections, and fully understand any impacts, this event represents a potential degradation of the level of safety of the plant. Event verification with external sources should not be necessary during .or following an OBE. Earthquakes of this magnitude should be readily felt by on-site personnel and recognized as a seismic event. The Shift Manager or Emergency Director may seek external verification if deemed appropriate (e.g., a call to the USGS, check internet news sources, etc.); however, the verification action must not preclude a timely emergency declaration. OBE events are detected in accordance with AOP 901. The OBE is associated with a peak horizontal acceleration of+/- 0.06g. Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA8. 78

HU3 ECL: Notification of Unusual Event Initiating Condition: Hazardous events Operating Mode Applicability: All Emergency Action Levels: Note: EAL HU3.4 does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents. HU3.l A tornado strike within the PROTECTED AREA. HU3.2 Internal room or area flooding of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component needed for the current operating mode. HU3.3 Movement of personnel within the PROTECTED AREA is impeded due to an offsite event involving hazardous materials (e.g., an offsite chemical spill or toxic gas release). HU3.4 A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles. Definitions: PROTEC;rED AREA: The area under continuous access monitoring and control, and armed protection as described in the site Security Plan. SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related. Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant. , EAL HU3. l addresses a tornado striking (touching down) within the Protected Area. EAL HU3.2 addresses flooding of a building room or area that results in operators isolating power to a SAFETY SYSTEM component due to water level or other wetting concerns. Classification is also required if the water level or related wetting causes an automatic isolation of a SAFETY SYSTEM component from its power source (e.g., a breaker or relay trip). To warrant classification, operability of the affected component must be required by Technical Specifications for the current operating mode. EAL HU3.3 addresses a hazardous materials event originating at an offsite location and of sufficient magnitude to impede the movement of personnel within the PROTECTED AREA. 79

EAL HU3 .4 addresses a hazardous event that causes an on-site impediment to vehicle movement and significant enough to prohibit the plant staff from accessing the site using personal vehicles. Examples of such an event include site flooding caused by a hurricane, heavy rains, up-river water releases, dam failure, etc., or an on-site train derailment blocking the access road. This EAL is not intended apply to routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the flooding around the Cooper Station during the Midwest floods of 1993, or the flooding around Ft. Calhoun Station in 2011. Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, Sor C. 80

HU4 ECL: Notification of Unusual Event Initiating Condition: FIRE potentially degrading the level of safety of the plant. Operating Mode Applicability: All Emergency Action Levels: Notes:

• The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

HU4.l a. A FIRE is NOT extinguished within 15-minutes of ANY of the following FIRE detection indications:

• Report from the field (i.e., visual observation)
• Receipt of multiple (more than 1) fire alarms or indications
• Field verification of a single fire alarm AND

b. The FIRE is located within ANY Table H-1 plant rooms or areas HU4.2 a. Receipt of a single fire alarm with no other indications of a FIRE.

AND

b. The FIRE is located within ANY Table H-1 plant rooms or areas AND

c. The existence of a FIRE is not verified within 30-minutes of alarm receipt.

HU4.3 A FIRE within the plant or ISFSI PROTECTED AREA not extinguished within 60-minutes of the initial report, alarm or indication. HU4.4 A FIRE within the plant or ISFSI PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish.

• Table H-1 Fire Areas
• 1G3 l DG and Day Tank Rooms
• 1G21 DGandDayTankRooms
• Battery Rooms
• Essential Switchgear Rooms
• Cable Spreading Room
• TorusRoom
• Intake Structure
• Pumphouse
• Drywell
• Torus
• NE, NW, SE Corner Rooms
• HPCIRoom
• RCICRoom
• RHR Valve Room
• North CRD Area
• South CRD Area
• CSTs
• Control Building
• Remote Shutdown Panel 1C388 Area
• Panel 1C55/56 Area
• SBGTRoom 81

Definitions: FIRE: Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed. PROTECTED AREA: The area under continuous access monitoring and control, and armed protection as described in the site Security Plan. Basis: This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant. EALHU4.l The intent of the 15-minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., smoldering waste paper basket). In addition to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation of a suppression system, etc. Upon receipt, operators will take prompt actions to confirm the validity of an initial fire alarm, indication, or report. For EAL assessment purposes, the emergency declaration clock starts at the time that the initial alarm, indication, or report was received, and not the time that a subsequent verification action was performed. Similarly, the fire duration clock also starts at the time of receipt of the initial alarm, indication or report. EALHU4.2 This EAL addresses receipt of a single frre alarm, and the existence of a FIRE is not verified (i 7e., proved or disproved) within 30-minutes of the alarm. TJpon receipt, operators will take prompt actions to confrrm the validity of a single fire alarm. *For EAL assessment purposes, the 30-minute clock starts at the time that the initial alan:ii was received, and not the time that a subsequent verification action was performed. A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failure or a spurious activation, and not an actual FIRE. For this reason, additional time is allowed to verify the validity of the alarm. The 30-minute period is a reasonable amount of time to determine if an actual FIRE exists; however, after that time, and absent information to the contrary, it is assumed that an actual FIRE is in progress. If an actual FIRE is verified by a report from the field, then EAL HU4. l is immediately applicable, and the emergency must be declared if the FIRE is not extinguished within 15 minutes of the report. If the alarm is verified to be due to an equipment failure or a spurious activation, and this verification occurs within 30-minutes of the receipt of the alarm, then this EAL is not applicable and no emergency declaration is warranted. EALHU4.3 In addition to a FIRE addressed by EAL HU4.l or EAL HU4.2, a FIRE within the plant or ISFSI PROTECTED AREA not extinguished within 60-minutes may also potentially degrade the level of plant safety. This basis extends to a FIRE occurring within the PROTECTED AREA of an ISFSI located outside the plant PROTECTED AREA. 82

EALHU4.4 If a FIRE within the plant or ISFSI PROTECTED AREA is of sufficient size to require a response by an offsite firefighting agency (e.g., a local town Fire Department), then the level of plant safety is potentially degraded. The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish. Declaration is not necessary if the agency resources are placed on stand-by, or supporting post-extinguishment recovery or investigation actions. Basis-Related Reguirements from Appendix R and NFPA-805 Criterion 3 of Appendix A to IO CFR 50 states in part that "structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions." The Nuclear Safety Goal ("NSG") in NFPA 805, Section 1.3.1 states, "The nuclear safety goal is to provide reasonable assurance that a fire during any operational mode and plant configuration will not prevent the plant from achieving and maintaining the fuel in a safe and stable condition." When considering the effects of fire, those systeips associated with achieving and maintaining safe shutdown conditions assume major importance because a safe shutdown success path, free of fire damage, must be available to meet the nuclear safety goals, objectives and performance criteria for a fire under any plant operational mode or configuration. Because fire may affect safe shutd,own systems and because the loss of function of systems ., used to mitigate the consequences of design basis accidents under post-fire conditions does' not per se impact public safety, the need to limit fire damage to systems required to achieve and maintain safe shutdown conditions is greater than the need to limit fire damage to those systems required to mitigate the consequences of design basis accidents. In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of I-hour fire barriers for the enclosure of cable and equipment and associated non-safety circuits of one redundant train (G.2.c). Even though DAEC has adopted the alternate approach provided by NFPA-805 in lieu of the deterministic requirements of Appendix R, the 30-minutes to verify a single alarm as used in EAL HU4.2 is considered a reasonable amount of time to determine if an actual FIRE exists without presenting a challenge to the nuclear safety performance criteria. Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA8. 83

HU6 ECL: 'Notification of Unusual Event Initiating Condition: Other conditions exist which in the judgment of the Emergency Director warrant declaration of a NOUE. Operating Mode Applicability: All Emergency Action Levels: HU6.l Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of SAFETY SYSTEMS occurs. Definitions: SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related. Basis: This IC addresses 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 classification level description for a NOUE. 84

HA1 ECL: Alert Initiating Condition: HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes. Operating Mode Applicability: All Emergency Action Levels: HAI.I A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the DAEC Security Shift Supervision. HAI.2 A validated notification from NRC of an aircraft attack threat within 30 minutes of the site. Definitions: HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station. HOSTILE ACTION: An act toward DAEC or its personnelthat includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILES, vehicles, *or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the nuclear power plant. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). OWNER CONTROLLED AREA: The site property owned by or*otherwise under the control of the licensee. PROJECTILE: An object directed toward a nuclear power plant that could cause concern for its continued operability, reliability, or personnel safety. Basis: This IC addresses the occurrence of a HOSTILE ACTION within the OWNER CONTROLLED AREA or notification of an aircraft attack threat. This event will require rapid response and assistance due to the possibility of the attack progressing to the PROTECTED AREA, or the need to prepare the plant and staff for a potential aircraft impact. Timely and accurate communications between DAEC Security Shift Supervision and the Control Room is essential for proper classification of a security-related event. Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program}. As time and conditions allow, these events require a heightened state ofreadiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). 85

The Alert declaration will also heighten the awareness of Offsite Response Organizations, allowing them to be better prepared should it be necessary to consider further actions. This IC does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR 73.71 or 10 CFR 50.72. EAL HA 1.1 is applicable for any HOSTILE ACTION occurring, or that has occurred, in the OWNER CONTROLLED AREA. This includes any action directed against the ISFSI which is located outside the plant PROTECTED AREA. EAL HAI .2 addresses the threat from the impact of an aircraft on the plant, and the anticipated arrival time is within 30 minutes. The intent of this EAL is to ensure that threat-related notifications are made in a timely manner so that plant personnel and offsite response organizations are in a heightened state ofreadiness. This EAL is met when the threat-related information has been validated in accordance with Abnormal Operating Procedure (AOP) 914, Security Events. The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may be provided by NORAD through the NRC. In some cases, it may not be readily apparent if an aircraft impact within the OWNER CONTROLLED AREA was intentional (i.e., a HOSTILE ACTION). It is expected, although not certain, that notification by an appropriate Federal agency to the site would clarify this point. In this case, the appropriate federal,,agency is intended to be NORAD, FBI, FAA or NRC. The emergency declaration, including one based on other ICs/EALs, should not be unduly delayed while awaiting notification by a Federal agency. Emergency plans and implementing procedures are public documents; therefore, EALs do not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information is contained in the Security Plan. Escalation of the emergency classification level would be via IC HSI. 86

HAS ECL: Alert Initiating Condition: Control Room evacuation resulting in transfer of plant control to alternate locations. Operating Mode Applicability: All Emergency Action Level: HAS.I An event has resulted in plant control being transferred from the Control Room to the Remote Shutdown Panel (IC388). Definitions: None Basis: This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations outside the Control Room. The loss of the ability to control the plant from the Control Room is considered to be a potential substantial degradation in the level of plant safety. Following a Control Room evacuation, control of the plant will be transferred to alternate* shutdown locations. The necessity to control a plant shutdown from outside the Control Room, in addition to responding to the event that required the evacuation of the Control Room, will present challenges to plant operators and other on-shift personnel. Activation of the ERO and emergency response facilities will assist in re.sponding to these challenges. Escalation of the emergency classification level would be via IC HS5. 87

HA6 ECL: Alert Initiating Condition: Other conditions exist which in the judgment of the Emergency Director warrant declaration of an Alert. Operating Mode Applicability: All Emergency Action Level: HA6.1 Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels. Definitions: HOSTILE ACTION: An act toward DAEC or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on*the nuclear power plant. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station. PROJECTILE: An object directed toward a nuclear power plant.that could cause concern for its continued operability, reliability, or personnel safety. Basis: This IC addresses 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 classification level description for an Alert. 88

HS1 ECL: Site Area Emergency Initiating Condition: HOSTILE ACTION within the PROTECTED AREA. Operating Mode Applicability: All Emergency Action Levels: HS 1.1 A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the DAEC Security Shift Supervision. Definitions: HOSTILE ACTION: An act toward DAEC or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the nuclear power plant. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). HOSTAGE: 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.

PROJECTILE: An object directed toward a nuclear power plant that could cause concern for its continued operability, reliability, or personnel safety. PROTECTED AREA: The area under continuous access monitoring and control, and armed protection as described in the site Security Plan. Basis: This IC addresses the occurrence of a HOSTILE ACTION within the PROTECTED AREA. This event will require rapid response and assistance due to the possibility for damage to plant equipment. Timely and accurate communications between DAEC Security Shift Supervision and the Control Room is essential for proper classification of a security-related event. Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safegu.ards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program]. 89

As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Site Area Emergency declaration will mobilize offsite response organization resources and have them available to develop and implement public protective actions in the unlikely event that the attack is successful in impairing multiple safety functions. This IC does not apply to a HOSTILE ACTION directed at th.e ISFSI PROTECTED AREA which is located outside the plant PROTECTED AREA; such an attack should be assessed using IC HAL It also does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR 73.71 or 10 CFR 50.72. Emergency plans and implementing procedures are public documents; therefore, EALs do not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information is contained in the Security Plan. Escalation of the emergency classification level would be via IC HG 1. 90

HS5 ECL: Site Area Emergency Initiating Condition: Inability to control a key safety function from outside the Control Room. Operating Mode Applicability: All Emergency Action Levels: Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded. HS5.l a. An event has resulted in plant control being transferred from the Control Room to the Remote Shutdown Panel (1C388). AND

b. Control of ANY of the following key safety functions is not reestablished within 20minutes.

• Reactivity control
• RPV water level
• RCS heat removal Definitions:

None B;sis: This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations, and the control of a key safety function cannot be reestablished in a timely manner. The failure to gain control of a key safety function following a transfer of plant control to alternate locations is a precursor to a challenge to one or more fission product barriers within a relatively short period of time. The determination of whether or not "control" is established at the Remote Shutdown Panel (1C388) is based on Emergency Director judgment. The Emergency Director is expected to make a reasonable, informed judgment within 20 minutes whether or not the operating staff has control of key safety functions from the remote safe shutdown location(s). AOP 915, "Shutdown Outside Control Room" provides the following CAUTION - "For Control Room evacuation as the result ofa fire, transfer ofcontrol at panels 1 C388, 1 C389, 1 C390, 1 C391, and 1 C392 is required to be completed within 20 minutes." Escalation of the emergency classification level would be via IC FG 1 or CG 1. 91

HS6 ECL: Site Area Emergency Initiating Condition: Other conditions exist which in the judgment of the Emergency Director warrant declaration of a Site Area Emergency. Operating Mode Applicability: All Emergency Action Level: HS6.l Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts, (1) toward site personnel or equipment that could lead to the likely failure of or, (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the site boundary. Definitions: HOSTILE ACTION: An act toward DAEC or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the nuclear power plant. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station. PROJECTILE: An object directed toward a nuclear power plant that could cause concern for its continued operability, reliability, or personnel safety.

                                                 \

Basis: This IC addresses 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 classification level description for a Site Area Emergency. 92

HG1 ECL: General Emergency Initiating Condition: HOSTILE ACTION resulting in loss of physical control of the facility. Operating Mode Applicability: All Emergency Action Level: HGl.l a. A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the DAEC Security Shift Supervision. AND

b. EITHER of the following has occurred:

1. ANY of the following safety functions cannot be controlled or maintained.

• Reactivity control
• RPV water level
• RCS heat removal OR

2. Damage to spent fuel has occurred oris IMMINENT.

Definitions: HOSTILE ACTION: An act toward DAEC or its'personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosive*s, PROJECTILEs, vehicles, or other devices used to deliver destructive fqrce. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the nuclear power plant. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). HOSTAGE: 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. IMMINENT: The trajectory of events or conditions is such that an EAL will be niet within a relatively short period of time regardless of mitigation or corrective actions. PROJECTILE: An object directed toward a nuclear power plant that could cause concern for its continued operability, reliability, or personnel safety. PROTECTED AREA: The area under continuous access monitoring and control, and armed protection as described in the site Security Plan. 93

Basis: This IC addresses an event in which a HOSTILE FORCE has taken physical control of the facility to the extent that the plant staff can no longer operate equipment necessary to maintain key safety functions. It also addresses a HOSTILE ACTION leading to a loss of physical control that results in actual or IMMINENT damage to spent fuel due to 1) damage to a spent fuel pool cooling system (e.g., pumps, heat exchangers, controls, etc.) or, 2) loss of spent fuel pool integrity such that sufficient water level cannot be maintained. Timely and accurate communications between the DAEC Security Shift Supervision and the Control Room is essential for proper classification of a security-related event. Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program}. Emergency plans and implementing procedures are public documents; therefore, EALs do not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information is contained in the Security Plan. 94

HG6 ECL: General Emergency Initiating Condition: Other conditions exist which in the judgment of the Emergency Director warrant declaration of a General Emergency. Operating Mode Applicability: All Emergency Action Level: HG6.1 Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area. Definitions: HOSTILE ACTION: An act toward DAEC or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a"concerted attack on the nuclear power plant. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station. IMMINENT: The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. PROJECTILE: An object directed toward a nuclear power plant that could cause concern for its continued operability, reliability, or personnel safety. Basis: This IC addresses 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 classification level description for a General Emergency. 95

11 SYSTEM MALFUNCTION ICS/EALS 96

SU1 ECL: Notification of Unusual Event Initiating Condition: Loss of ALL offsite AC power capability to essential buses for 15 minutes or longer. Operating Mode Applicability: 1, 2, 3 Emergency Action Level: Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded. SUI.I Loss of ALL offsite AC power capability to 1A3 AND 1A4 buses for 15 minutes or longer. Definitions: None Basis: This IC addresses a prolonged loss of offsite power. The loss of offsite power sources renders the plant more vulnerable to a complete loss of power to AC essential buses. This condition represents a potential reduction in the level of safety of the plant. The intent of this EAL is to declare a Notification of Unusual Event when offsite power has been lost and both of the emergency diesel generators have successfully started and energized their respective 4kv essential bus.

• For emergency classification purposes, "capability" means that an offsite AC power source(s) is available to the essential buses, whether or not the buses are powered from it.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of offsite power. Escalation of the emergency classification level would be via IC SAl. 97

SU3 ECL: Notification of Unusual Event Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer. Operating Mode Applicability: 1, 2, 3 Emergency Action Level: Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded. SU3.l An UNPLANNED event results in the inability to monitor one or more of the Table S-1 parameters from within the Control Room for 15 minutes or longer.

• Reactor power
• RPV Water Level
• RPV Pressure Primary Containment Pressure Suppression Pool Level Suppression Pool Temperature Definitions:

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.

• UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

Basis: This IC addresses the difficulty associated with monitoring normal plant conditions without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition is a precursor to a more significant event and represents a potential degradation in the level of safety of the plant. As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room. 98

An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.

• This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, RPV level and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for RPV water level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication. Escalation of the emergency classification level would be via IC SA3. 99

SU4 ECL: Notification of Unusual Event Initiating Condition: Reactor coolant activity greater than Technical Specification allowable limits. Operating Mode Applicability: 1, 2, 3 Emergency Action Levels: SU4.l Pretreatment Offgas System (RM-4104) Hi-Hi Radiation Alarm. SU4.2 Sample analysis indicates that reactor coolant specific activity is greater than 2.0 µCi/gm

           . dose equivalent 1-131 for 12 hours orlonger.

Definitions: None Basis: This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant. For EAL SU4.l, 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-minut~ delay and decay of 1 Ci/sec. For EAL SU4.2, coolant samples exceeding the 2.0 µCi/gm dose equivalent I-13lconcentration require prompt action by DAEC Technical Specifications and are representative of minor fuel cladding degradation. Escalation of the emergency classification level would be via ICs FAl or the Recognition Category R ICs. 100

SUS ECL: Notification of Unusual Event Initiating Condition: RCS leakage for 15 minutes or longer. Operating Mode Applicability: 1, 2, 3 Emergency Action Levels: Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded. SU5.l RCS unidentified or pressure boundary leakage greater than 10 gpm for 15 minutes or longer. SU5.2 RCS identified leakage greater than 25 gpm for 15 minutes or longer. SU5.3 Leakage from the RCS to a location outside containment greater than 25 gpm for 15 minutes or longer. Definitions: UNISOLABLE: An open or breached system line that cannot be isolated, remotely or locally. Basis: This IC addresses RCS leakage which may be a precursor to a more significant event. In this case, RCS leakage has been detected and operators, following applicable procedures, have been unable to promptly isolate the leak. This condition is considered to be a potential degradation of the level of safety of the plant. EAL SU5.1 and EAL SU5 .2 are focused on a loss of mass from the RCS due to "unidentified leakage", "pressure boundary leakage" or "identified leakage" (as these leakage types are defined in the vlant Technical Specifications). EAL SU5.3 addresses a RCS mass loss caused by an UNISOLABLE leak through an interfacing system. These EALs thus apply to leakage into the containment, a secondary-side system) or a location outside of containment. The leak rate values for each EAL were selected because they are usually observable with normal Control Room indications. Lesser values typically require time-consuming calculations to determine (e.g., a mass balance calculation). EAL SU5.l uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage. The release of mass from the RCS due to the as-designed/expected operation of a relief valve does not warrant an emergency classification. A stuck-open Safety Relief Valve (SRV) or SRV leakage is not considered either identified or unidentified leakage by Technical Specifications and, therefore, is not applicable to this EAL. The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible. Escalation of the emergency classification level would be via ICs of Recognition Category R or F. 101

SU6 ECL: Notification of Unusual Event 1 Initiating Condition: Automatic or manual scram fails to shutdown the reactor. Operating Mode Applicability: 1, 2 Emergency Action Levels: Note: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. SU6.l a. An automatic scram did not shutdown the reactor. AND

b. ANY of the following manual actions taken at 1C05 are successful in lowering reactor power below 5% power

• Manual Scram Pushbuttons
• Mode Switch to Shutdown
• Alternate Rod Insertion (ARI)

SU6.2 a. A manual scram did not shutdown the reactor. AND

b. EITHER of the following:

1. ANY of the following subsequent manual actions taken at 1C05 are successful in lowering reactor power below 5% power.

• Manual Scram Pushbuttons *
• Mode Switch to Shutdown
• Alternate Rod Insertion (ARI)

OR

2. A subsequent automatic scram is successful in shutting down the reactor.

Definitions: None Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic scram is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant. Following the failure on an automatic reactor scram, operators will promptly' initiate manual actions at the reactor control console to shutdown the reactor (e.g., initiate a manual reactor scram quickly fall to a level within the capabilities of the plant's decay heat removal systems. 102

If an initial manual reactor scram is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control console to shutdown the reactor (e.g., initiate a manual reactor scram using a different switch). Depending upon several factors, the initial or subsequent effort to manually scram the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor scram signal. If a subsequent manual or automatic scram is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems. A manual action at the reactor control console is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram). This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control console". Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action. The plant response to the failure of an automatic or manual reactor scram will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC SA6. Depending upon the plant response, escalation is also possible via IC F Al. Absent the plant conditions needed to meet either IC SA6 or FAl, an Unusual Event declaration is appropriate for this event. The reactor should be considered shutdown when it is producing less heat than the maximum decay heat load for which the SAFETY SYSTEMS are designed (typically 3 to 5% power). Should a reactor scram signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied.

• If the signal causes a plant transient ~at should have included an automatic reactor scram and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and should be evaluated.
• If the signal does not cause a plant transient and the scram failure is determined through other m~ans (e.g., assessment oftest results), then this IC and the EALs are not applicable and no classification is warranted.

103

SU7 ECL: Notification of Unusual Event Initiating Condition: Loss of ALL onsite or offsite communications capabilities. Operating Mode Applicability: 1, 2, 3 Emergency Action Levels: SU7.l Loss of ALL of the following onsite communication methods:

• Plant Operations Radio System
• In-Plant Phone System
• Plant Paging System (Gaitronics)

SU7.2 Loss of ALL of the following offsite response organization communications methods:

• DAEC All-Call phone
• All telephone lines (PBX and commercial)
• Cell Phones (including fixed cell phone system)
• Control Room fixed satellite phone system
• FTS Phone system SU7.3 Loss of ALL of the following NRC communications methods:
• FTS Phone system *,
• All telephone lines (PBX and commercial)
• Cell Phones (including fixed cell phone system)
• Control Room fixed satellite phone system Basis:

This IC addresses a significant loss of on-site or offsite communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to offsite response organizations and the NRC. This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.). EAL SU7 .1 addresses a total loss of the communications methods used in support of routine plant operations. EAL SU7.2 addresses a total loss of the communications methods used to notify all offsite response organizations of an emergency declaration. The offsite response organizations referred to here are the State of Iowa, Linn County, and Benton County. EAL SU7.3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration. 104

SA1 ECL: Alert Initiating Condition: Loss of ALL but one AC power source to essential buses for 15 minutes or longer. Operating Mode Applicability: 1, 2, 3 Emergency Action Level: Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded. SAI.1 a. AC power capability to 1A3 and 1A4 buses is reduced to a single power source for 15 minutes or longer. AND

b. ANY additional single power source failure will result in a loss of ALL AC power to SAFETY SYSTEMS.

Definitions: SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related. Basis: This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment. This IC provides an escalation path from IC SUI. An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.

• A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).
• A loss of emergency power sources (e.g., onsite diesel generators) with a single train of essential buses being fed from an offsite power source.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power. Escalation of the emergency classification level would be via IC SSL 105

SA3 ECL: Alert Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress. Operating Mode Applicability: I, 2, 3 Emergency Action Level: Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded. SA3.l a. An UNPLANNED event results in the inability to monitor one or more Table S-1 parameters from within the Control Room for 15 minutes or longer.

                             ;L .,z,s *      . * :;::, *. *,:'. ,'                                          j . * ,..x:       .**~*'" ,. ~.1:, .*, ,~;,. :::':;,~~*'.: *1~

• r:

1 ;.:'.,. *  : .:~ ** *

• 1 * * * *
• Table;S-1.)) Safety System' Parameters
• i;,

&)~C~'.ff;"\'li$co;ti;~ii))i:;1 "'7!~!tf:")!1:Q',;\jll;/.'."\1[,!~\';:;f:::r;rf:'Z1"r;;f"~>>,;:,J'3:)t".'(';;,k)~!;ffl0fl'::~',11'1"~7');[

• Reactor power '
• RPV Water Level

!

• RPV Pressure
• Primary Containment Pressure
• Suppression Pool Level
• Suppression Pool Temperature AND

b. ANY of the Table S-2 transient events are in progress.

g;~;~ than 25% thermal reactor power i!

• Electrical load rejection greater than 25% full electrical load

                                                                                                                                                                  . ].*

• Reactor scram 1

r

• ECCS actuation
• Thermal power oscillations greater than 10%

106

Definitions: SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related. UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may

• be known or unknown.

Basis: This IC addresses the difficulty associated with monitoring rapidly changing plant conditions during a transient without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. During this condition, the margin to a potential fission product barrier challenge is reduced. It thus represents a potential substantial degradation in the level of safety of the plant. As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room. An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making. This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, RPV level and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for RPV water level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication. Escalation of the emergency classification level would be via ICs FSl or IC RSl. 107

SA6 ECL: Alert Initiating Condition: Automatic or manual scram fails to shutdown the reactor, and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor. Operating Mode Applicability: 1, 2 Note: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. Emergency Action Level: SA6.l a. An automatic or manual scram did not shutdown the reactor. AND

b. ALL of the following manual actions taken at 1C05 are not successful in lowering reactor power below 5% power

• Manual Scram Pushbuttons
• Mode Switch to Shutdown
• Alternate Rod Insertion (ARI)

Definitions: None Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results .in a reactor shutdown, and subsequent operator manual actions taken at the reactor control consoles to shutdown the reactor are also unsuccessful. This condition represents an actual oi- potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if the reactor is subsequently shutdown by an action taken away from the reactor control consoles since this event entails a significant failure of the RPS. A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram. This action does not include manually driving in control rods or implementation of boron injection strategies. If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control consoles (e.g., locally opening breakers). Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles." Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action. 108

The plant response to the failure of an automatic or manual reactor scram will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If the failure to shutdown the reactor is prolonged enough to cause a challenge to the RPV water level or RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC SS6. Depending upon plant responses and symptoms, escalation is also possible via IC FS1. Absent the plant conditions needed to meet either IC SS6 or FS1, an Alert declaration is appropriate for this event. It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F ICs; however, this IC and EAL are included to ensure a timely emergency declaration. The reactor should be considered shutdown when it is producing less heat than the maximum decay heat load for which the SAFETY SYSTEMS are designed (typically 3 to 5% power). 109

SAS ECL: Alert Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode. Operating Mode Applicability: 1, 2, 3 Emergency Action Level: Notes:

• If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then this emergency classification is not warranted.
• If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of the SAFETY SYSTEM, then this emergency classification is not warranted.

SAS.I a. The occurrence of ANY of the Table S-3 hazardous events:

                  ,:_r~,:. ~~1~~,;f::r~~.~~i~~t-~~.~:~;:~.;,},, ,/~~,:;,..w:w,+';.*.,t-i~¥:.:~;~--'l't': -J!~t~-~*;\~---,,:t}~irft,'¥'~~r-1:/:5
                    *j

• Table s:.3 Hazardous Events *; * *.

                         ,~,~-*'. --~'dr-, >* .~---~~~. ,,~ ,r*J~0r. '.,><,?,                                                           .,.J;

• Seismic event (earthquake)
• Internal or external flooding event
• High winds or tornado strike
• FIRE
• EXPLOSION
• Other events with similar hazard characteristics as determined by the*

Shift Manager or Emergency Director AND

b. 1. Event damage has caused indications of degraded performance in one train of a SAFETY SYSTEM needed for the current operating mode.

AND

2. EITHER of the following:

• Event damage has caused indications of degraded performance to a second train of the SAFETY SYSTEM needed for the current operating mode, OR
• The event has resulted in VISIBLE DAMAGE to the second train of a SAFETY SYSTEM needed for the current operating mode.

llO

Definitions: EXPLOSION: A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events may require a post-event inspection to determine if the attributes of an explosion are present. FIRE: Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed. SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related. VISIBLE DAMAGE: Damage to a component or structure that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure. Damage resulting from an equipment failure and limited to the failed component (i.e., the failure did not cause damage to a structure or any other equipment) is not VISIBLE DAMAGE. Basis: This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the current operating mode. *In order to provide the appropriate context for consideration of an ALERT classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train, and there must be either indications of performance issues with the second SAFETY,SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance, and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE such that the potential exists for performance issues. Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has indications of degraded performance for criteria SA8. l .b. l of this EAL; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events. An event affecting a single-train SAFETY SYSTEM (i.e., there are indications of degraded performance and/or VISIBLE DAMAGE affecting the one train) would not be classified under SA8 because the two-train impact criteria that underlie the EALs and Bases would not be met. If an event affects a single-train SAFETY SYSTEM, then the emergency classification should be made based on plant parameters/symptoms meeting the EALs for another IC. Depending upon the circumstances, classification may also occur based on Shift Manager/Emergency Director judgement. Indications of degraded performance addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. 111

VISIBLE DAMAGE addresses damage to a SAFETY SYSTEM train that is not in service/operation and that potentially could cause performance issues. Operators will make this determination based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. Escalation of the emergency classification level would be via IC FS 1 qr RS 1. 112

SS1 ECL: Site Area Emergency Initiating Condition: Loss of ALL offsite and ALL onsite AC power to essential buses for 15

 ,minutes or longer.

Operating Mode Applicability: 1, 2, 3 Emergency Action Level: Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded. SSl.l Loss of ALL offsite and ALL onsite AC power to 1A3 and IA4 buses for 15 minutes or longer. Definitions: SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related. Basis: This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. In addition, fi~~ion product barrier monitoring capabilities may be degraqed under these conditions. * - This IC represents a condition that involves actual or likely major failures of plant functions need(ld for the protection of the public. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via ICs RGI, FG 1 or SG1. 113

SS2 ECL: Site Area Emergency Initiating Condition: Loss of ALL Vital DC power for 15 minutes or longer. Operating Mode Applicability: 1, 2, 3 Emergency Action Level: Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded. SS2. l Indicated voltage is less than 105 VDC on BOTH Div 1 and Div 2 125 VDC buses for 15 minutes or longer. Definitions: SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related. Basis: Th1s IC addresses a loss of Vital DC power which compromises the ability to monitor and control SAFETY SYSTEMS. In modes above Cold Shutdown, this condition involves a major failure of plant functions needed for the protection of the public. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via I Cs RG 1, FG 1 or SG2. 114

SS6 ECL: Site Area Emergency Initiating Condition: Inability to shutdown the reactor causing a challenge to RPV water level or RCS heat removal. Operating Mode Applicability: I, 2 Emergency Action Levels: SS6.l a. An automatic or manual scram did not shutdown the reactor. AND

b. ALL of the following manual actions taken at IC05 are not successful in lowering reactor power below 5% power:

• Manual Scram Pushbuttons
• Mode Switch to Shutdown
• Alternate Rod Insertion (ARI)

AND

c. EITHER of the following conditions exist:

• RPV level cannot be restored and maintained above -25 inches.

OR

• HCL (Graph 4 of EOP 2) exceeded.

Definitions: None Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, all subsequent operator actions to manually shutdown the reactor are unsuccessful, and continued power generation is challenging the capability to adequately remove heat from the core and/or the RCS. This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration of a Site Area Emergency. In some instances, the emergency classification resulting from this IC/EAL may be higher than that resulting from an assessment of the plant respo:nses and symptoms against the Recognition Category F ICs/EALs. This is appropriate in that the Recognition Category F ICs/EALs do not address the additional threat posed by a failure to shutdown the reactor. The inclusion of this IC and EAL ensures the timely declaration of a Site Area Emergency in response to prolonged failure to shutdown the reactor. The reactor should be considered shutdown when it is producing less heat than the maximum decay heat load for which the SAFETY SYSTEMS are designed (typically 3 to 5% power). Escalation of the emergency classification level would be via IC RGI or FG I. 115

1 SG1 ECL: General Emergency Initiating Condition: Prolonged loss of ALL offsite and ALL onsite AC power to essential buses. Operating Mode Applicability: 1, 2, 3 Emergency Action Level: Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded. SGI.1 a. Loss of ALL offsite and ALL onsite AC power to 1A3 and 1A4 buses. AND

b. EITHER of the following:

• Restoration of at least one AC essential bus in less than 4 hours is not likely.

OR

• RPV level cannot be restored and maintained above -25 inches.

Definitions: SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related. Basis: This IC addresses a prolonged loss of all power sources to AC essential buses. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, *containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A prolonged loss of these buses will lead to a loss of one or more fission product barriers. In addition, fission product barrier , monitoring capabilities may be degraded under these conditions. The EAL should require declaration of a General Emergency prior to meeting the thresholds for IC FG 1. This will allow additional time for implementation of offsite protective actions. Escalation of the emergency classification from Site Area Emergency will occur if it is projected that power cannot be restored to at least one AC essential bus by the end of the 4 hour station blackout coping period. Beyond this time, plant responses and event trajectory are subject to greater uncertainty, and there is an increased likelihood of challenges to multiple fission product barriers. The estimate for restoring at least one essential bus should be based on a realistic appraisal of the situation. Mitigation actions with a low probability of success should not be used as a basis for delaying a classification upgrade. The goal is to maximize the time available to prepare for, and implement, protective actions for the public. The EAL will also require a General Emergency declaration if the loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core. 116

SG2 ECL: General Emergency Initiating Condition: Loss of ALL AC and Vital DC power sources for 15 minutes or longer. Operating Mode Applicability: 1, 2, 3 Emergency Action Level: Note: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded. SG2.l a. Loss of ALL offsite and ALL onsite AC power to 1A3 and 1A4 buses for 15 minutes or longer. AND

b. Indicated voltage is less than 105 VDC on BOTH Div 1 and Div 2 125 VDC buses for 15 minutes or longer.

Definitions: SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related. Basis: This IC addresses a concurrent and*ptolonged loss of both AC and Vital DC power. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A loss of Vital DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both AC and DC power will lead to multiple challenges to fission product barriers. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. The 15-minute emergency declaration clock begins at the point when both EAL thresholds are met. 117

APPENDIX A - ACRONYMS AND ABBREVIATIONS AC ...................................................................................................................... Alternating Current AOP ................................................................................................. Abnormal Operating Procedure ATWS ................................................................................... Anticipated Transient Without Scram BWR ............................................................................................................. Boiling Water Reactor CDE ...................................................................................................... Committed Dose Equivalent CFR ...................................................................................................... Code of Federal Regulations CNMT ........................................................................................................................... Containment DC .............................................................................................................................. Direct Current 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 FEMA ............................................................................. Federal Emergency Management Agency GE ...................................................................................................................... General Emergency HCL................................................................................................................... Heat Capacity Limit HPCI .............................................................................................. High Pressure Coolant Injection IC ........................................................................................................................ Initiating Condition ID ............................................................................................................................. Inside Diameter ISFSI ........................................................................... Independent Spent Fuel Storage Installation Keff .................................................................................... Effective Neutron Multiplication Factor LCO ................................. ;............................................................. Limiting Condition of Operation LOCA ........................................................................................................ Loss of Coolant Accident mR, mRem, mrem, mREM ........................ ~................................... milli-Roentgen Equivalent Man MW .................................................................................................................................... Megawatt NEI ............................................................................................................. Nuclear Energy Institute NRC .............................................................................................. Nuclear Regulatory Commission NORAD ................................................................. North American Aerospace Defense Command NOUE .............................................................................................. Notification Of Unusual Event NUMARC 1 .....*......*.*.........*.......*............*........*....*... Nuclear Management and Resources Council OBE. .................................................................................... :................. Operating Basis Earthquake OCA ............................................................................................................. Owner Controlled Area ODAM ......................................................................................... Offsite Dose Assessment Manual PA .............................................................................................................................. Protected Area PAG....................................................................................................... Protective Action Guideline PRA/PSA .................................... Probabilistic Risk Assessment/ Probabilistic Safety Assessment PWR ........................................................................................................ Pressurized Water Reactor PSIG ................................................................................................. Pounds per Square Inch Gauge R ......................................................................................................................................... Roentgen RCIC ............................................................................................... Reactor Core Isolation Cooling RCS ............................................................................................................. Reactor Coolant System Rem, rem, REM ...................................................................................... Roentgen Equivalent Man 1 NUMARC was a predecessor organization of the Nuclear Energy Institute (NEI). A-1

RPS ......................................................................................................... Reactor Protection System RPV ............................................................................................................. Reactor Pressure Vessel RWCU .......................................................................................................... Reactor Water Cleanup SCBA ... ...... .. ... ............. ... ... .................... .. ..... ..... ............. ....... Self-Contained Breathing Apparatus SPDS ............................................................................................ Safety Parameter Display System TEDE ............................................................................................. Total Effective Dose Equivalent TAF ..................................................................................................................... Top of Active Fuel TSC .......................................................................................................... Technical Support Center UFSAR. ................................................................................ Updated Final Safety Analysis Report A-2

APPENDIX B- DEFINITIONS The following definitions are taken from Title 10, Code of Federal Regulations, and related regulatory guidance documents. Alert: Events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA PAG exposure levels. General Emergency: Events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area. Notification of Unusual Event (NOUE): Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of SAFETY SYSTEMS occurs. Site Area Emergency: Events are in progress or have occurred which involve actual or likely major failures of plant functions needed forprotection of the public or HOSTILE ACTION that

.results in intentional damage or malicious acts; 1) toward site personnel or equipment that could lead to the likely failure of or; 2) that prevent effective access to, equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA PAG exposure levels beyond the site boundary.

The following are key terms necessary for overall understanding the DAEC emergency . classification scheme. Emergency Action Level (EAL): A pre-determined, site-specific, observable threshold for an Initiating Condition that, when met or exceeded, places the plant in a given emergency classification level. Emergency Classification Level (ECL): One of a set of names or titles established by the US Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions according to (1) potential or actual effects or consequences, and (2) resulting onsite and offsite response actions. The emergency classification levels, in ascending order of severity, are: Notification of Unusual Event (NOUE) Alert Site Area Emergency (SAE) General Emergency (GE) B-1

Fission Product Barrier Threshold: A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier. Initiating Condition (IC): An event or condition that aligns with the definition of one of the four emergency classification levels by virtue of the potential or actual effects or consequences. Selected terms used in Initiating Condition and Emergency Action Level statements are set in all capital letters ( e.g., ALL CAPS). These words are defined terms that have specific meanings as used in this document. The definitions of these terms are provided below. CONFINEMENT BOUNDARY: The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. This corresponds to the pressure boundary for the Dry Shielded Canister (DSC) shell (including the inner bottom cover plate) base metal and associated confinement boundary welds. CONTAINMENT CLOSURE: Procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions. For DAEC, this is considered to be Secondary Containment as required by Technical Specifications. DESIGN BASIS EARTHQUAKE (DBE): A DBE is vibratory ground motion for which certain (generally, safety-related) structures, systems, and components must be designed to remain functional. EXPLOSION: A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events may require a post-event inspection to determine if the attributes of an explosion are present. n FIRE: Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed. HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station. HOSTILE ACTION: An act toward a nuclear power plant or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the nuclear power plant. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). 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. B-2

IMMINENT: The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI): A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. OPERATING BASIS EARTHQUAKE (OBE): An OBE is vibratory ground motion for which those features of a nuclear power plant necessary for continued operation without undue risk to the health and safety of the public will remain functional. OWNER CONTROLLED AREA: This term is typically taken to mean the site property owned by or otherwise under the control of the licensee. PROJECTILE: An object directed toward a nuclear power plant that could cause concern for its continued operability, reliability, or personnel safety. PROTECTED AREA: The area under continuous access monitoring and control, and armed protection as described in the site Security Plan. REFUELING PATHWAY: Includes all the cavities, tubes, canals and pools through which irradiated fuel may be moved, but not including the reactor vessel. SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related. SECURITY CONDITION: Any Security Event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A SECURITY CONDITION does not involve a HOSTILE ACTION. SITE BOUNDARY: That line beyond which the land is neither owned, nor leased, nor otherwise controlled by the Company. UFSAR Figure 1.2-1 identifies the DAEC SITE BOUNDARY. UNISOLABLE: An open or breached system line that cannot be isolated, remotely or locally. UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. ' VISIBLE DAMAGE: Damage to a component or structure that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure. Damage resulting from an equipment failure and limited to the failed component (i.e., the failure did not cause damage to a structure or any other equipment) is not VISIBLE DAMAGE. B-3

ATTACHMENT 3 NEXTERA ENERGY DUANE ARNOLD, LLC DUANE ARNOLD ENERGY CENTER RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION RELATING TO LICENSE AMENDMENT REQUEST TSCR-166 UPDATED DEVIATIONS AND DIFFERENCES MATRIX 100 pages follow

UPDATED DAEC DEVIATIONS AND DIFFERENCES MATRIX TABLE OF CONTENTS

GENERAL COMMENT

S ................................................................................................................................ 1 ABNORMAL RAD LEVELS/ RADIOACTIVE EFFLUENT ICS/EALS ................................................................... 5 COLD SHUTDOWN/ REFUELING SYSTEM MALFUNCTION ICS/EALS ........................................................ 20 INDEPENDENT SPENT FUEL STORAGE INSTALLATION {ISFSI) ICS/EALS .................................................... 36 FISSION PRODUCT BARRIER ICS/EALS ....................................................................................................... 38 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS .............................................. 47 SYSTEM MALFUNCTION ICS/EALS ............................................................................................................. 63 APPENDIX A-ACRONYMS AND ABBREVIATIONS .................................................................................... 84 APPENDIX B - DEFINITIONS ....................................................................................................................... 89 APPENDIX C - PERMANENTLY DE FUELED ICS/EALS ............................................:..................................... 98

UPDATED DAEC DEVIATIONS AND DIFFERENCES MAliRIX

GENERAL COMMENT

S Page 1

DAEC DEVIATIONS AND DIFFERENCES MATRIX GLOBAL#l References to NEI 99-01 Replaced with DAEC Difference Convert generic guidance to DAEC specific. None GL0BAL#2 Effective date Replaced with TBD, 2018 Difference Convert generic guidance to DAEC specific. None GL0BAL#3 Defined terms in Appendix B; Defined terms in Appendix B; Difference All defined terms in Appendix B used in the Title Case Upper Case document are in upper case (CAPs) to None indicate that the terms are defined. GL0BAL#4 PWR specific references PWR references removed Difference DAEC is a BWR None GLOBAL#S Recognition Category A- Recognition Category R- Difference DAEC implemented the optional Abnormal Radiation Abnormal Radiation designation of "R" for radiological related Levels/Radiological Effluent Levels/Radiological Effluent items to maintain continuity with previous None category and Emergency Action category and Emergency Action practice at DAEC. Levels; AU, AA, AS, and AG Levels; RU, RA, RS, and RG GL0BAL#6 Permanently Defueled Section Deleted references to Difference Not Applicable to DAEC None Permanently Defueled Station GL0BAL#7 Acknowledgments, Notice and Deleted Difference Not Applicable to DAEC None Executive Summary GL0BAL#8 Parameters or indications listed Some parameters or indications Difference Tables or bullets were created to present in EALs listed in EALs were placed in DAEC-specific information in a manner None tables or bulletized lists. familiar to and desired by scheme users. GL0BAL#9 Site specific information or "Site specific information or Difference Compliance with intent of the guidance. indication statements indications" were replaced with None DAEC-specific information or indications where applicable. GL0BAL#10 Operating Mode Applicability lists Operating Mode Applicability lists Difference Mode numbers used for consistency with mode names (i.e., Power mode numbers (i.e., 1, 2, etc.) DAEC procedures and training. None Operation, Startup) GL0BAL#11 Developer's Notes Developer's Notes deleted Difference Developer's notes are not reflected in the None implementation of the EALs. Gl0BAL#12 Example EAL statement "Example" deleted from Difference In adopting the EAL, the "example" status None statement is no longer applicable. GL0BAL#13 The following terms: "all, any, or, Consistently capitalized and Difference Capitalized and balded conditional terms in either are sometimes capitalized balded the following terms: "ALL, ICs and EALs for consistency based on user None and/or balded in ICs and EALs ANY, OR, EITHER" in ICs and EALs. feedback. GL0BAL#14 Defined terms are only listed in Defined terms are also listed as in Difference Aid to the user to present all needed APPENDIX B - DEFINITIONS separate section of each IC/EAL information within the same section of the None where the terms are used. Basis document. GL0BAL#15 Term "emergency buses" Replaced with "essential buses" Difference Changed to reflect DAEC nomenclature None 2

DAEC DEVIATIONS AND DIFFERENCES MATRIX COVER PAGE Development of Emergency Duane Arnold Emergency Action Difference Changes made to adapt the generic NEI None Action Levels for Non-Passive Level Technical Bases Document guidance to a DAEC-specific document Reactors Introduction Acknowledgments, Notice and Deleted Difference Not Applicable to DAEC None Executive Summary TOC 1. Regulatory Background 1. Basis for Emergency Action Difference Title change None levels TOC 1.1 Operating Reactors 1.1 Regulatory Background Difference Title change None TOC 1.2 Permanently Defueled Station Deleted section Difference Not Applicable to DAEC None TOC 1.3 Independent Spent Fuel 1.2 Independent Spent Fuel Difference Re-numbered None Storage Installation (ISFSI) Storage Installation (ISFSI) TOC 1.4 NRC Order EA-12-051 1.3 NRC Order EA-12-051 Difference Re-numbered None TOC 1.5 Applicability of Advance and Deleted section Difference Not Applicable to DAEC None Small Modular Reactor Designs TOC 3.Design of the NEI 99-01 3. Design of the DAEC Emergency Difference Title Change None Emergency Classification Scheme Classification Scheme

• TOC 3.3 NSSS Design Differences Deleted section Difference Changes made to adapt the generic NEI None guidance to a DAEC-specific document TOC 3.4 Organization and Changed to 3.3 DAEC 3.4 Difference Changes made to adapt the generic NEI None Presentation of Generic Organization and Presentation of guidance to a DAEC-specific document Information Generic Information TOC 4.0 Site-Specific Scheme 4.0 DAEC Scheme Development Difference Title chan_ge None Development TOC 4.4; 4.5; 4.6; 4.8 Deleted sections Difference Changes made to adapt the generic NEI None guidance to a DAEC-specific document TOC 4.7 Developer and User Feedback None TOC Appendix C-Permanently Deleted section Difference Changes made to adapt the generic NEI None Defueled Station ICs/EALs guidance to a DAEC-specific document 1.1 Regulatory Background Regulatory Background Difference Changes made to adapt the generic NEI None guidance to a DAEC-specific document and removed developer information 1.2 Permanently Defueled Station Section deleted Difference Not Applicable to DAEC None 1.3 1.3 Independent Spent Fuel 1.2 Independent Spent Fuel Difference Re-numbered section. None Storage Installation (ISFSI) Storage Installation (ISFSI) 3

DAEC DEVIATIONS AND DIFFERENCES MATRIX

.* ' ' '*:~ ,.  ::*  ;' ' ' ;,, >>,

                                                                                       ,,    *~>  ':;;        ,,,,",         0,e
                                                                                                            ,,                                           Justificath:in                         Validation
   'Section                                      Rev.                  ,'
                     ""  NEI, 99-01                  6                                                     -"c'

• • ~**' ,'<},.  : ,,: . ' '.'i ' ',... ;,~,:.'  ;;;.' '\ .*. ,:? )' ,, ';;;, ',1.JDAE~ik:: , ::,.;:;r }:[Ih~n:~~  : ,{ .~J?, '

                                                                                                                                                           j~ ' *)i,{
                                                                                                                                                            '-, ,t      ,*:iX*,:* * ', '.,,    *. ,.,,# ,::;,

1.4 1.4 NRC Order EA-12-051 1.3 NRC Order EA-12-051 . Difference Re-numbered and removed wording to add None these readings (DAEC installation completed). 1.5 Applicability to Advanced and Section deleted Difference Not Applicable to DAEC None Small Modular Reactor Designs i 2 KEY TERMINOLOGY USED IN NEI KEY TERMINOLOGY USED IN Difference Minor changes to reflect DAEC-specific None 99-01 DAEC EAL SCHEME implementation. 3 DESIGN OF THE NEI 99-01 DESIGN OF THE DAEC Difference Changes made to adapt the generic NEI None EMERGENCY CLASSIFICATION EMERGENCY CLASSIFICATION guidance to a DAEC-specific document SCHEME SCHEME 3.1 Assignment of Emergency Assignment of Emergency Difference Changes made to adapt the generic NEI None Classification Levels (ECLs) Classification Levels (ECLs) guidance to a DAEC-specific document, removed references to PWRs, and removed developer information. 3.2 Types of Initiating Conditions and Types of Initiating Conditions and Verbatim None Emergency Action Levels Emergency Action Levels 3.3 Text referring to NSSS design Deleted Difference Guidance is now DAEC specific None differences for various types or plants; Developer guidance 3.4 Organization and Presentation of DAEC-Specific Organization and Difference Renumbered to 3.3, made DAEC-specific, None Generic Information Presentation of Generic and deleted developer information Information 3.5 Mode of Applicability Matrix; Deleted '1Permanently Defueled" Difference Renumbered to 3.4, removed PWR Vl Typical BWR Operating Modes section of matrix; replaced information, removed permanently Typical BWR Operating Modes defueled, and inserted DAEC Operating with DAEC-specific Operating Modes to comply with the document intent. Modes 4 Site Specific Scheme Development of the DAEC Difference Updated to reflect DAEC specific scheme None Development Guidance Emergency Classification Scheme development process. 5 GUIDANCE ON MAKING GUIDANCE ON USING THE DAEC Difference Guidance is now DAEC specific None EMERGENCY CLASSIFICATIONS EALS 6-11 Recognition Category IC/EAL removed Difference Matrixes were intended for use by EAL None Matrixes developers. Inclusion in licensee scheme is not desired. 4

DAEC DEVIATIONS AND DIFFERENCES MATRIX ABNORMAL RAD LEVELS/ RADIOACTIVE EFFLUENT ICS/EALS 5

DAEC DEVIATIONS AND DIFFERENCES MATRIX Val'i'dationJ::

                                                                                                      ./       . \

Recognition Category: AUl RUl Difference Global Comment #5 None Initiating Condition: Release of Release of gaseous or liquid Difference Global Comment #9 None gaseous or liquid radioactivity radioactivity greater than 2 times .-1 greater than 2 times the (site- the ODAM limits for 60 minutes

, specific effluent release or longer. -;

<( controlling document) limits for 60 minutes or longer. Operating Mode of Applicability: Operating Mode of Applicability: Verbatim None All All 6

                                                                                                                                 -- - - - - - - - - ~

DAEC DEVIATIONS AND DIFFERENCES MATRIX

                                                                                              '~;*. Jusfificaiio~'.

(1) Reading on ANY effluent (1) Reading on ANY Table R-1 Difference See Global Comments #8, 9, 12, & 13. V2 radiation monitor greater effluent radiation monitor than 2 times the (site- greater than column "NOUE" Reworded EAL statement to remove specific effluent release for 60 minutes or longer: operator confusion as to whether they controlling document) needed to multiply the values of the limits for 60 minutes or [inserted Table R=! of DAEC- following table by 2 or if the value provided longer: (site-specific specific radiation monitors already was 2X. Wording now matches monitor list and threshold and threshold values] wording of RSl and RGl allowing for easier values corresponding to 2 operator progression through the EALs.

          ,times the controlling document limits)

(2) Reading on ANY effluent (2) Reading on ANY effluent Difference Global Comment #13 None radiation monitor greater radiation monitor greater than 2 times the alarm than 2 times the alarm

....       setpoint established by a      setpoint established by a
 =

-~

~

0 CJ current radioactivity discharge permit for 60 minutes or longer. current radioactivity discharge permit for 60 minutes or longer. (3) Sample analysis for a (3) Sample analysis for a gaseous Difference Global Comment #9 None gaseous or liquid release or liquid release indicates a indicates a concentration concentration or release rate or release rate greater greater than 2 times the than 2 times the (site- ODAM limits for 60 minutes specific effluent release or longer. , controlling document) limits for 60 minutes or longer. Intent and meaning of the EALs are not altered. 7

DAEC DEVIATIONS AND DIFFERENCES MATRIX Sectibh Justification I Validation.# J Recognition Category: AU2 RU2 Difference Global Comment #5 & 14 None Initiating Condition: UNPLANNED UNPLANNED loss ohyater level Verbatim None loss of water level above above irradiated fuel. 0 irradiated fuel. Operating Mode of Applicability: Operating Mode of Applicability: Verbatim None All All (1) a. UNPLANNED water level (1) a UNPLANNED water level Difference Global Comment #9, 12 & 13 V3 drop in the REFUELING drop in the REFUELING PATHWAY as indicated by PATHWAY as indicated by ANY of the following: ANY of the following: N

, (site-specific level

• Report to control ct indications). room (visual observation)
• Fuel pool level indication (Ll-3413) less than 36 feet and lowering
• WR GEMAC Floodup indication (Ll-4541) coming on scale AND AND 8

DAEC DEVIATIONS AND DIFFERENCES MATRIX

..:~:,section

b. UNPLANNED increase in b. UNPLANNED rise in area Difference Global Comments #9 & 13 V4 area radiation levels as radiation levels as indicated by ANY of the indicated by ANY of the following radiation following radiation monitors. monitors.

(site-specific list of area

• Spent Fuel Pool Area, radiation monitors) Rl-9178

                                                    *                ' Floor, RI-North Refuel 9163

• New Fuel Vault Area,

-+l C Rl-9153 South Refuel Floor, RI-N 0 u

)

9164 NW Drywell Area Hi Range Rad Monitor, RIM-9184A

• South Drywell Area Hi Range Rad Monitor, RIM-9184B Intent and meaning of the EALs are not altered.

9

DAEC DEVIATIONS AND DIFFERENCES MATRIX

 . Section I         NEI 99-01 Re~.'      Et
                                       ,;o,O,<          *"

DAEC.*,

                                                                                      .,,,,//,,..Change Justification   ',<'

Validation #' Recognition Category: AAl RAl Difference Global Comment #5 & 14 None Initiating condition: Release of Release of gaseous or liquid Verbatim None gaseous or liquid radioactivity radioactivity resulting in offsite resulting in offsite dose greater dose great.er than 10 mrem TEDE than 10 mrem TEDE or 50 mrem or SO mrem thyroid CDE. thyroid CDE. Operating Mode of Applicability: Operating Mode of Applicability: Verbatim None All All (1) Reading on ANY of the (1) Reading on ANY Table R-1 Difference Global Comment #8, 9, 12 & 13 vs following radiation radiation monitor greater monitors greater than the than column "Alert" for 15 reading shown for 15 minutes or longer: minutes or longer: (site-specific monitor list [inserted Table R-1 of DAEC-and threshold values) specific radiation monitors and threshold vaJues] .-1 (2) Dose assessment using (2) Dose assessment using actual Difference Global Comment #9 None ~ actual meteorology meteorology indicates doses Added bracketed 'Preferred' to reinforce indicates doses greater greater than 10 mrem TEDE the 4th Note of the IC than 10 mrem TEDE or 50 or 50 mrem thyroid CDE at or mrem thyroid CDE at or beyond the SITE BOUNDARY. beyond (site-specific dose [Preferred] receptor point). (3) Analysis of a liquid (3) Analysis of a liquid effluent Difference Global Comment #9 effluent sample indicates sample indicates a a concentration or release concentration or release rate rate that would result in that would result in doses doses greater than 10 greater than 10 mrem TEDE mrem TEDE or 50 mrem or 50 mrem thyroid CDE at or thyroid CDE at or beyond beyond the SITE BOUNDARY (site-specific dose for one hour of exposure. receptor point) for one hour of exposure. 10

DAEC DEVIATIONS AND DIFFERENCES MATRIX

    * ;Secti(?lf
             .--~. I*     N El 99-0H~ev ,' :6*

Justification*.

                                                                                                                    ' ,-.   '/*

(4) Field survey results (4) Field survey results indicate Difference Global Comment #9 None indicate EITHER of the EITHER of the following at or following at or beyond beyond the SITE BOUNDARY: (site-specific dose

• Closed window dose receptor point): rates greater than 10

  ,.J C:

• Closed window dose rates greater than 10 m R/h r expected to continue for 60 minutes

  - 0 u
  . -1
  ~

mR/hr expected to continue for 60

• or longer.

Analyses of field survey minutes or longer. samples indicate thyroid

• Analyses of field survey CDE greater than SO samples indicate mrem for one hour of thyroid CDE greater inhalation.

than SO mrem for one hour of inhalation. Intent and meaning of the EALs are not altered. 11

DAEC DEVIATIONS AND DIFFERENCES MATRIX !; ,Se~ti~i, ;* I': .. *.. }. ~El ~~-,01.,8e~. ~'*"~ ,r,* (iii

                                                                               ' * ... DAEC<
                                                                           .**:>: !c< "" ' ,, *; '.   ;;. >,; .). \~Chang~.: . 
                                                                                                                                              * . 1Ju$ti(!ca~i.o.~ . :       ~alidation # 1, Recognition Category: AA2                     RA2                                                Difference   Global Comment #5 & 14                          None Initiating Condition: Significant             Significant lowering of water                      Verbatim                                                     None lowering of water level above, or             level above, or damage to, damage to, irradiated fuel.                   irradiated fuel.

Operating Mode of Applicability: Operating Mode of Applicability: Verbatim None All All (1) Uncovery of irradiated fuel (1) Uncovery of irradiated fuel in Verbatim None in the REFUELING the REFUELING PATHWAY.

                       )

PATHWAY. (2) Damage to irradiated fuel (2) Damage to irradiated fuel Difference Global Comment #8, 9, 12 & 13 VG resulting in a release of resulting in a release of radioactivity from the fuel radioactivity from the fuel as as indicated by ANY of the indicated by Hi Rad alarm for following radiation ANY of the following ARMs: monitors:

• Spent Fuel Pool Area, RI-9178 (site-specific listing of radiation
• North Refuel Floor, Rl-9163 monitors, and the associated
• New Fuel Vault Area, RI-N

  <t               readings, setpoints and/or                             9153
  <t alarms)

• South Refuel Floor, Rl-9164 OR Threshold values for the Drywell monitors Reading greater than 5 R/hr are only applicable in Mode 5 since the on ANY of the following calculated radiation levels from damage to radiation monitors (in Mode irradiated fuel would be masked by the 5 only): typical background levels on these
• NW Drywell Area Hi Range monitors during plant operation, and Rad Monitor, RIM-9184A mechanical damage to a fuel assembly in
• South Drywell Area Hi the vessel can only happen with the reactor Range ~ad Mo~itor, RIM- head removed (Mode 5).

9184B . j (3) Lowering of spent fuel pool (3) Lowering of spent Difference Global Comment #9 V7 level to (site-specific Level fuel pool level to 2 value). [See Developer 25.17 feet Intent and meaning of the EALs are not Notes altered. 12

DAEC DEVIATIONS AND DIFFERENCES MATRIX Recognition Category: AA3 RA3 Difference Global Comment #5 & 14 None Initiating Condition: Radiation Radiation levels that impede Difference Reworded IC to reflect non-applicability of None levels that impede access to access to areas necessary for EAL#2. equipment necessary for normal normal plant operation. plant operations, cooldown or shutdown. Operating Mode Applicability: All Operating Mode Applicability: All Verbatim None (1) Dose rate greater than 15 (1) Dose rate greater than 15 Difference Global Comment #9, 12 & 13 None mR/hr in ANY of the mR/hr in ANY of the following areas: following areas:

• Control Room
• Control Room ARM (RM-
• Central Alarm Station 9162)

         * (other site-specific

• Central Alarm Station (by areas/rooms) survey)

M (2) An UNPLANNED event Not used at DAEC - Difference EALs RA3 and HAS are not applicable to V8 ~ results in radiation levels DAEC because an evaluation has shown that prohibit or impede that there are no rooms or areas that access to any of the contain equipment which require a following plant rooms or manual/local action as specified in areas: operating procedures used for normal plant operation, cooldown and shutdown. All (site-specific list of plant rooms areas outside the Control Room that or areas with entry-related mode contain equipment necessary for normal applicability identified) plant operation, cooldown and shutdown do not require physical access to operate. Intent and meaning of the EALs are not altered. 13

DAEC DEVIATIONS AND DIFFERENCES MATRIX JusHtication ' . Recognition Category: ASl RSl Difference Global Comment #5 & 14 None Initiating Condition: Release of Release of gaseous radioactivity Verbatim None gaseous radioactivity resulting in resulting in offsite dose greater offsite dose greater than 100 than 100 mrem TEDE or 500 mrem TEDE or 500 mrem thyroid mrem thyroid CDE. CDE. Operating Mode Applicability: All Operating Mode Applicability: All Verbatim None (1) Reading on ANY of the (1) Reading on ANY Table R-1 Difference Global Comment #8, 9, 12 & 13 V9 following radiation effluent radiation monitor monitors greater than the greater than column "SAE" reading shown for 15 for 15 minutes or longer. minutes or longer: ..-1 (site-specific monitor list and [inserted Table R-1 of DAEC-Ill < threshold values) specific radiation monitors and threshold values] (2) Dose assessment using (2) Dose assessment using actual Difference Global Comment #3 & 9 None actual meteorology meteorology indicates doses Added bracketed 'Preferred' to reinforce indicates doses greater greater than 100 mrem TEDE the 4th Note of the IC than 100 mrem TEDE or or 500 mrem thyroid CDE at 500 mrem thyroid CDE at or beyond the SITE or beyond (site-specific BOUNDARY. dose receptor point). [Preferred] 14

DAEC DEVIATIONS AND DIFFERENCES MATRIX

Sectior(.* .. **'bAEC Validation J:

-~:"[;; .., * -<, ,;,~' ~-, (3) Field survey results (3) Field survey results indicate Difference Global Comment #3, 9, & 13 None indicate EITHER of the EITHER of the following at or following at or beyond beyond the SITE BOUNDARY: (site-specific dose receptor point):

• Closed window dose rates greater than 100
• Closed window dose rates greater than 100 mR/hr expected to mR/hr expected to continue for 60 minutes continue for 60 or longer. minutes or longer.
• Analyses of field survey samples indicate thyroid
• Analyses of fi~ld survey samples indicate C

CDE greater than 500 mrem for one hour of thyroid CDE greater than 500 mrem for one

           -  0 u
            . -1 II)
            <C inhalation.                       hour of inhalation.

Intent and meaning of the EALs are not altered. 15

DAEC DEVIATIONS AND DIFFERENCES MATRIX Recognition Category: AS2 RS2 Difference Global Comment #5 None Initiating Condition: Spent fuel Spent fuel pool level at 16.36 feet Difference Global Comment #9 VlO pool level at (site-specific Level 3 description). Operating Mode Applicability: All Operating Mode Applicability: All Verbatim None N II) (1) Lowering of spent fuel pool (1) Lowering of spent fuel pool Difference Global Comment #9 & 12 VlO <C level to (site-specific Level level to 16.36 feet 3 value). ' Intent and meaning of the EALs are not altered. 16

DAEC DEVIATIONS AND DIFFERENCES MATRIX

,z~:Section'
    '     ~,.,_:,_.J Justification '           .. Validation tt;'.

Recognition Category: AG1 RG1 Difference Global Comment #5 & 14 None Initiating Condition: Release of Release of gaseous radioactivity Verbatim None gaseous radioactivity resulting in resulting in offsite dose greater offsite dose greater than 1,000 than 1,000 mrem TEDE or 5,000 mrem TEDE or 5,000 mrem mrem thyroid CDE. thyroid CDE. Operating Mode Applicability: All Operating Mode Applicability: All Verbatim None (1) Reading on ANY of the (1) Reading on ANY Table R-1 Difference Global Comment #8, 9, 12 & 13 V9 following radiation effluent radiation monitor monitors greater than the greater than column "GE" for ,-f reading shown for 15 15 minutes or longer. 1.1' minutes or longer: <C (site-specific monitor list and [inserted Table R-1 of DAEC-threshold values) specific radiation *IJ1onitors and threshold values] (2) Dose assessment using (2) Dose assessment using actual Difference Global Comment #3 & 9 None actual meteorology meteorology indicates doses Added bracketed 'Preferred' to reinforce indicates doses greater greater than 1,000 mrem the 4th Note of the IC than 1,000 mrem TEDE or TEDE or 5,000 mrem thyroid 5,000 mrem thyroid CDE at CDE at or beyond the SITE or beyond (site-specific BOUNDARY. [Preferred] dose receptor point). 17

DAEC DEVIATIONS AND DIFFERENCES MATRIX

• \.,* *S~ci:idn . 1;'.;'Justificatiori, v*-r",,

(3) Field survey results (3) Field survey results indicate Difference Global Comment #3 & 9 None indicate EITHER of the EITHER of the following at or following at or beyond beyond the SITE BOUNDARY: (site-specific dose receptor

• Closed window dose point): rates greater than 1,000
• Closed window dose mR/hr expected to
• rates greater than 1,000 continue for 60 minutes mR/hr expected to or longer.

continue for 60 minutes

• Analyses of field survey or longer. samples indicate thyroid

-,iJ C:

• Analyses of field survey samples indicate thyroid COE greater than 5,000 mrem for one hour of

- 0 I.I CDE greater than 5,000 mrem for one hour of inhalation.

inhalation. Intent and meaning of the EALs are not altered. 18

DAEC DEVIATIONS AND DIFFERENCES MATRIX 1 :Sectiofr, , * ~* "'~NEl99-01d~ev;,.6*7::,:*'{*,~t;. ,' ' , ' '", ,,,, , ' ,, . ' , ', .

• j~ ,

                                                          , '       , -~  .;(,' Pf\EC ':r:* *~: 3 , ::_<:c;I;; ~~hange " :S; Ii.
                                                                             ,   .-. .      ,            .,, .,.  ,.1,, k.: .*, 8 , ,i:'

k .Ji :i;, Josi:ification' , '

                                                                                                                                                  ,   ,    , { , , , , , , . **;. ,,,;0 :,,;:,
                                                                                                                                                                                                       .. * . *. ' * .:validatibn'.'n, Recognition Category: AG2            RG2                                                     Difference              Global Comment #5                                                                         None Initiating Condition: Spent fuel     Spent fuel pool level cannot be                         Difference              Global Comment #9                                                                          VlO pool level cannot be restored to     restored to at least 16.36 feet for at least (site-specific Level 3      60 minutes or longer.

description) for 60 minutes or ,* longer. Operating Mode Applicability: All Operating Mode Applicability: All Verbatim None N (.?, (1) Spent fuel pool level cannot (1) Spent fuel pool level cannot Difference Global Comment #9 & 12 VlO

<t                       be restored to at least (site-        be restored to at least 16.36 specific Level 3 value) for 60        feet for 60 minutes or longer.

minutes or longer. Intent and meaning of the EALs are not altered. 19

DAEC DEVIATIONS AND DIFFERENCES MATRIX COLD SHUTDOWN/ REFUELING SYSTEM MALFUNCTION ICS/EALS 20

DAEC DEVIATIONS AND DIFFERENCES MATRIX Ii,. sectioif* :J,,

• NEl'99*:C>1 Re~'. 6(.' f *J ..

                                                               'iii ' :7ol(' DAEC "~~.
                                                                                       "::'[' .', .';.
                                                                                                   .*\L
                                                                                                        . :1c11an'~e '"~I: ,*'s'

• t? )llJustification'.;'.'"1 3 J' '

0 ' '

                                                                                                                                                                             * ~Valiclatiotf#
                                                                                                                                                                               'i- ~ O> - , ' ',_. ,

Recognition Category: CUl CUl Verbatim Global Comment #11, 14 None Initiating Condition: UNPLANNED UNPLANNED loss of RPV Difference Global Comment #4 None loss of (reactor vessel/RCS [PWR] inventory for 15 minutes or or RPV [BWR]) inventory for 15 longer minutes or longer. Operating Mode Applicability: Operating Mode Applicability: 4, Difference Global Comment #10 None Cold Shutdown, Refueling 5 (1) UNPLANNED loss of (1) UNPLANNED loss of reactor Difference Global Comment #4 & 12 None reactor coolant results in coolant results in RPV level (reactor vessel/RCS [PWR] less than a required lower or RPV [BWR]) level less limit for 15 minutes or than a required lower limit longer.

    .-1

, for 15 minutes or longer.

u (2) a. (Reactor vessel/RCS [PWR] (2) a. RPV level cannot be Difference Global Comment #4 None or RPV [BWR]) level cannot monitored. be monitored. AND AND

b. UNPLANNED increase in b. UNPLANNED level rise in Difference Global Comment #9 None (site-specific sump and/or Drywell/Reactor Building tank) levels. Equipment or Floor Drain sump, or Suppression Pool.

Intent and meaning of the EALs are not altered. Recognition Category: CU2 CU2 Verbatim Global Comment #11, 14 None Initiating Condition: Loss of all Loss of all but one AC power Difference Global comment #15 None but one AC power source to source to essential buses for 15 emergency buses for 15 minutes minutes or longer. or longer. N

, Operating Mode Applicability: Operating Mode Applicability: 4, Difference Global Comment #10 None u Cold Shutdown, Refueling, 5, Defueled Defueled (1) a. AC power capability to (1) a. AC power capability to Difference Global Comment #9, 12, & 13 Vll (site-specific emergency 1A3 and 1A4 buses is buses) is reduced to a reduced to a single power 21

DAEC DEVIATIONS AND DIFFERENCES MATRIX S~ction N.EI 9~7p1 ~ev. 6.. **Justification.,

                                                                                       '"_,,~4'}~ ** ,-_~ *' . * ,,
                                                                                                                    **:Validation#
                                                                                                                     '<  <";,.,,,., 4*

single power source for 15 source for 15 minutes or minutes or longer. longer. AND AND

b. Any additional single b. Any additional single power source failure will power source failure will result in loss of all AC result in loss of ALL AC power to SAFElY SYSTEMS. power to SAFElY SYSTEMS.

Intent and meaning of the EALs are not altered. 22

DAEC DEVIATIONS AND DIFFERENCES MATRIX

s~ction '

NEfgg:.01 R:ev. 51; '.~, (, ',fi, DAEt;:\', "C~0. , Y' . 'fha~je ]: .t: \'i ' :iJustificalion '1,,'~, * ':,,,": *,. , ,,:: *,* ' ~:validation '" ', if ,,

                       '     c,.:' ,         ' ;    ',        ,,         :    *,  ... <;     ','                                         .                                            y     ,:,  ,,, '

Recognition Category: CU3 CU3 Verbatim Global Comment #11, 14 None Initiating Condition: UNPLANNED UNPLANNED increase in RCS Verbatim None increase in RCS temperature. temperature. Operating Mode Applicability: Operating Mode Applicability: 4, Difference Global Comment #10 None Cold Shutdown, Refueling 5 (1) UNPLANNED increase in (1) UNPLANNED increase in RCS Difference Global Comment #9 & 12 Vl RCS temperature to temperature to greater than greater than (site-specific 212°F Technical Specification cold shutdown temperature limit). (2) Loss of ALL RCS (2) Loss of ALL RCS temperature Difference Global Comment #4 & 13 None temperature and (reactor and RPV level indication for rt) vessel/RCS [PWR] or RPV 15 minutes or longer

, [BWR]) level indication for u

15 minutes or longer. Intent and meaning of the EALs are not altered. 23

DAEC DEVIATIONS AND DIFFERENCES MATRIX N~(99-01 Rev .. 't(\* . Justific:ation Recognition Category: CU4 CU4- Verbatim Global Comment #11, 14 None Initiating Condition: Loss of Vital Loss of Vital DC power for 15 Verbatim None DC power for 15 minutes or minutes or longer.

                                                              ~

longer. Operating Mode Applicability: Operating Mode Applicability: 4, Difference Global Comment #10 None Cold Shutdown, Refueling 5 (1) Indicated voltage is less (1) Indicated voltage is less than Difference Global Comment #9, 12, 13 V12 o::t' u than (site-specific bus 105 voe on BOTH Div 1 and voltage value) on required Div 2 125 voe buses for 15 Vital DC buses for 15 minutes or longer minutes or longer. Intent and meaning of the EALs are not altered. 24

DAEC DEVIATIONS AND DIFFERENCES MATRIX Recognition Category: CU5 CU5 Verbatim None Initiating Condition: Loss of all Loss of all onsite or offsite Verbatim None onsite or offsite communications communications capabilities. capabilities. Operating Mode Applicability: Operating Mode Applicability: 4, Difference Global Comment #10 None Cold Shutdown, Refueling, 5, Defueled Defueled (1) Loss of ALL of the following (1) Loss of ALL of the following Difference Global Comment #9, 12 & 13 V13 onsite communication onsite communication methods: methods: (site-specific list of

• Plant Operations Radio communications methods) System
• In-Plant Phone System Ln
• Plant Paging System

, (Gaitronics)

~ (2) Loss of ALL of the following (2) Loss of ALL of the following Difference Global Comment #9 & 13 V13 ORO communications offsite response organization V14 methods: communications methods: (site-specific list of

• DAE CAIi-Ca II,phone communications methods)
• All telephone fines (PBX and commercial)
• Cell Phones (including fixed cell phone system)
• Control Room fixed satellite phone system
• FTS Phone system 25

DAEC DEVIATIONS AND DIFFERENCES MATRIX {lchang1:i:v;

                                                                                          *,                                        Vafitlation]#'

(3) Loss of ALL of the following (3) Loss of ALL of the following Difference Global Comment #9, 12 & 13 V13 NRC communications NRC communications methods: methods: C 0 (site-specific list of communications methods) FTS Phone system All telephone lines (PBX and commercial) ~ u,

• Cell Phones (including u fixed cell phone system)
• Control Room fixed satellite phone system Intent and meaning of the EALs are not altered .

26

DAEC DEVIATIONS AND DIFFERENCES MATRIX .. '* ~ N~l

                              ' *-~~-

99-0f.Rev.* ~;:,.

                                                   ,,            .,,,,_ ' ~
                                                                            " ~ '.           ~./,          . *~ ..    . '    -.-:_ '

Jus(ificatioi,,

                                                                                                                                                          . *.i, Vali~ation

. ,~* .. ;-

                                                                                                                                                                                  #:.I
                                                                                                                                                                                  ~.:'

Recognition Category: CA1 CA1 Verbatim Global Comment #11, 14 None Initiating Condition: Loss of Loss of RPV inventory. 'Difference Global Comment #4 None (reactor vessel/RCS [PWR] or RPV [BWR]) inventory. Operating Mode Applicability: Operating Mode Applicability: 4, Difference Global Comment #10 None Cold Shutdown, Refueling 5 (1) Loss of (reactor vessel/RCS (1) Loss of RPV inventpry as Difference Global Comment #4, 9 & 12 V15 [PWR] or RPV [BWR]) indicated by level less than inventory as indicated by 119.5 inches level less than (site-specific level). (2) a. (Reactor vessel/RCS [PWR] (2) a. RPV level cannot be Difference Global Comment #4 None or RPV [BWR]) level cannot monitored for 15 minutes be monitored for 15 or longer

  ""5 minutes or longer AND                           AND                                      Difference      Global Comment #4, 9 & 13                         None

b. UNPLANNED increase in b. UNPLANNED level rise in (site-specific sump and/or Drywell/Reactor Building tank) levels due to a loss of Equipment or Floor Drain (reactor vessel/RCS [PWR] sump, or Suppression Pool or RPV [BWR]) inventory. due to a loss of RPV inventory.

Intent and meaning of the EALs are not altered. 27

DAEC DEVIATIONS AND DIFFERENCES MATRIX Recognition Category: CA2 CA2 Verbatim Global Comment #11, 14 None Initiating Condition: Loss of all Loss of all offsite and all onsite Difference Global Comment #15 None offsite and all onsite AC power to AC power to essential buses for emergency buses for 15 minutes 15 minutes or longer. or longer. Operating Mode Applicability: Operating Mode Applicability: 4, Difference Global Comment #10 None N <t Cold Shutdown, Refueling, 5, Defueled u Defueled "- (1) Loss of ALL offsite and ALL (1) Loss of ALL offsite and ALL Difference Global Comment #9, 12 & 13 V12 onsite AC Power to (site- onsite AC Power to 1A3 and specific emergency buses) 1A4 for 15 minutes or longer. for 15 minutes or longer. Intent and meaning of the EALs are not altered. 28

DAEC DEVIATIONS AND DIFFERENCES MATRIX

                                                                                     ,v,r, Recognition Category: CA3                          CA3                                                      Verbatim   Global Comment #11, 14                     None Initiating Condition: Inability to                 Inability to maintain the plant in                       Verbatim                                              None maintain the plant in cold                         cold shutdown.

shutdown. ., Operating Mode Applicability: Operating Mode Applicability: 4; Difference Global Comment #10 None Cold Shutdown, Refueling 5 (1) UNPLANNED increase in (1) UNPLANNED increase in RCS Difference Global Comment #9 & 12 Vl RCS temperature to temperature to greater than greater than (site-specific 212°F for greater than the Technical Specification duration specified in Table C-cold shutdown 2. temperature limit) for greater than the duration specified in the following table. Table: RCS Heat-up Duration Thre h*"'* Containment ,~w.-u ...

                                                                           -       C-2 RCS Heat-up Duration Th Difference Global Comment #4                          None rt)

<[ u RCS Status Closure Status __ Duration

                                                            .,         ,. __,,___.      Containment Closure               Changed "RCS Status" to "RCS Integrity" to intact (but not at                                                                    Status                    match current site nomenclature Not applicable       t:n  .....,!-

• reduced inventory

[PWRJ) Intact Not Applicable Established N~Mll/1,CJtes* Established Not intact (or at reduced inventory [PWR]) Not Established 0 minutes Not Established

                                                       .* If ari,~cs heat removal system is in operatior

• If an RCS heat removal system is in operatio 1 wwim t dfiYWcs 1emperature fs being reduced, frame and RCS temperature is being reduce , tnelfl't 'Mlot app 1ca e.

applicable. (2) UNPLANNED RCS pressure (2) UNPLANNED RCS pressure Difference Global Comment #4 & 9 V16 increase greater than (site- increase greater than 10 psig Added "due to a loss of RCS cooling" to specific pressure, reading). due to a loss of RCS cooling. clarify the intent of the EAL (This EAL does not apply during water-solid plant conditions. [PWR]) Intent and meaning of the EALs are not altered. 29

DAEC DEVIATIONS AND DIFFERENCES MATRIX Validation 1# Recognition Category: CA6 CA6 Verbatim Global Comment #11, 14 None Initiating Condition: Hazardous Hazardous event affecting a Verbatim None event affecting a SAFETY SYSTEM SAFETY SYSTEM needed for the needed for the current operating current operating mode. mode. Operating Mode Applicability: Operating Mode Applicability: 4, Difference Global Comment #10 None Cold Shutdown, Refueling 5 (1) a. The occurrence of ANY of (1) a. The occurrence of ANY of Difference Global Comment #9, 12 & 13 None the following hazardous the Table C-3 hazardous events: events:

• Seismic event (earthquake)
• Seismic event (earthquake)
• Internal or external flooding event
• Internal or external
• High winds or tornado strike
• FIRE flooding event EXPLOSION
• Other events with similar hazard characteristics as determined by the

       *. High winds or tornado         Shift Manager or Emergency Director strike

• FIRE
• EXPLOSION

       * (site specific hazards)

• Other events with similar hazard characteristics as determined by the Shift Manager 30

                                                                                                                                      -1 DAEC DEVIATIONS AND DIFFERENCES MATRIX AND                               AND

b. EITHER of the following: b. 1. Event damage has Deviation Adopted the revised EAL wording provided V17

1. Event damage has caused indications of in approved EAL FAQ 2016-02.

caused indications of degraded degraded performance performance in one in at least one train of a train of a SAFElY SAFElY SYSTEM needed SYSTEM needed for for the current the current operating operating mode. mode. AND OR 2. EITHER of the Deviation Adopted the revised EAL wording provided V17

1. The event has caused following: in approved EAL FAQ 2016-02.

VISIBLE DAMAGE to a

• Event damage has SAFElY SYSTEM
• caused, indications Difference Added the following clarification to the V18 component or structure of degraded Basis from EALFAQ 2018-04:

-...:C needed for the current operating mode. performance to a second train of the An event affecting a single-train SAFElY SYSTEM (i.e., there are indications of -50 u U) SAFElY SYSTEM needed for the degraded performance and/or VISIBLE DAMAGE affecting the one train) would not current operating be classified under SAS because the two-mode, or train impact criteria that underlie the EALs

• The event has and Bases would not be met. If an event resulted in VISIBLE affects a single-train SAFElY SYSTEM, then DAMAGE to the the emergency classification should be second train of a made based on plant SAFElY SYSTEM parameters/symptoms meeting the EALs needed for the for another IC. Depending upon the current operating circumstances, classification may also occur
• mode. based on Shift Manager/Emergency Director judgement.

Intent and meaning of the EALs are not altered.

                                                              '1 31

DAEC DEVIATIONS AND DIFFERENCES MATRIX Recognition Category: CSl CSl Verbatim Global Comment #11, 14 None Initiating Condition: Loss of Loss of reactor vessel/RCS Difference Global Comment #4 None (reactor vessel/RCS [PWR] or RPV inventory affecting core decay [BWR]) inventory affecting core heat removal capability. decay heat removal capability. Operating Mode Applicability: Operating Mode Applicability: 4, Difference Global Comment #10 None Cold Shutdown, Refueling 5 (1) a. CONTAINMENT CLOSURE (1) a. CONTAINMENT CLOSURE Difference Global Comment #9 & 12 V19 not established. not established. ,-f AND AND en b. (Reactor vessel/RCS [PWR] b. RPV level less than +64 u or RPV [BWR]) level less inches than (site-specific level). (2) a. CONTAINMENT CLOSURE (2) a. CONTAINMENT CLOSURE Difference Global Comment #4 & 9 V19 established. established. AND AND -

b. (Reactor vessel/RCS [PWR] b. RPV level less than +15 or RPV [BWR]) level less inches than (site-specific level).

32

DAEC DEVIATIONS AND DIFFERENCES MATRIX (3) a. (Reactor vessel/RCS [PWR] (3) a. RPV level cannot be Difference Global Comment #4 None or RPV [BWR]) level cannot monitored for 30 minutes be monitored for 30 or longer. minutes or longer. AND AND

b. Core uncovery is indicated b. Core uncovery is indicated Difference Global Comment #9 &13 VG by ANY of the following: by ANY of the following:

~ C 0

          * (Site-specific radiation monitor) reading greater than (site-specific value)

• Drywell Monitor (9184A/B) reading greater than 5.0 R/hr

~ .-1

• Erratic source range
• UNPLANNED level rise in II) u monitor indication Drywell/Reactor Building

[PWR] Equipment or Floor Drain

• UNPLANNED increase in sump, or Suppression (site-specific sump Pool of sufficient and/or tank) levels of magnitude to indicate sufficient magnitude to core uncovery Intent and meaning of the EALs are not indicate core uncovery altered.

          * (Other site-specific indications) 33

DAEC DEVIATIONS AND DIFFERENCES MATRIX Recognition Category: CGl CGl Verbatim Global Comment #11, 14 None Initiating Condition: Loss of Loss of reactor vessel/RCS Difference Global Comment #4 None (reactor vessel/RCS [PWR] or RPV inventory affecting fuel clad [BWR]) inventory affecting fuel integrity with containment clad integrity with containment challenged. challenged. Operating Mode Applicability: Operating Mode Applicability: 4, Difference Global Comment #10 None Cold Shutdown, Refueling 5 (1) a. (Reactor vessel/RCS [PWR] (1) a. RPV level less than +15 Difference Global Comment #4, 9, 12 & 13 V19 .-1 or RPV [BWR]) level less inches for 30 minutes or ~ than (site-specific level) for longer. u 30 minutes or longer. AND AND b. ANY indication from the

b. ANY indication from the Containment Challenge Containment Challenge Table (see below).

Table (see below). (2) a. (Reactor vessel/RCS [PWR] (2) a. RPV level cannot be Difference Global Comment #4 None or RPV [BWR]) level cannot monitored for 30 minutes be monitored for 30 or longer. minutes or longer. 34

DAEC DEVIATIONS AND DIFFERENCES MATRIX AND AND Difference Global Comment #8, 9 & 13 V6

b. Core uncovery is indicated b. Core uncovery is indicated by ANY of the following: by ANY of the following:

          * (Site-specific radiation

• Drywell Monitor monitor) reading greater (9184A/B) reading than (site-specific value) greater than 5.0 R/hr
• Erratic source range
• Erratic source range monitor indication monitor indication

[PWR]

• UNPLANNED level rise
• UNPLANNED increase in in Drywell/Reactor (site-specific sump Building Equipment or and/or tank) levels of Floor Drain sump, or sufficient magnitude to Suppression Pool of indicate core uncovery sufficient magnitude to AND indicate core uncovery AND C. ANY indication from the c. ANY indication from the Difference Global Comment #9 None Containment Challenge Secondary Containment Table (see below). Challenge Table C-1.

Containment Challenge Table Table C-1 Containment Challenge T. Difference Global Comment #9 V20 ONTAINMENT CLOSURE not established*

• CONTAINMENT CLOSURE not established V21 xplosive mixture) exists inside containment
• Drywell Hydrogen or Torus Hydrogen gre AND Drywell Oxygen or Torus Oxygen grE NPLANNED increase in containment pressure econdary containment radiation monitor reading
• UNPLANNED increase in containment pre ite specific value) [BWR]
• Secondary containment radiation monito safe operating limits (MSOL) of EOP 3, Ta

                                              *If CONTAINMENT CLOSURE is

• If CONTAINMENT CLOSURE is re-established prior to exceeding Verbatim re-established prior to exceeding the 30-minute time limit, then the 30-minute time limit, then declaration of a General Intent and meaning of the EALs are not declaration of a General Emergency is not reqtJired. altered.

Emergency is not required. , 35

DAEC DEVIATIONS AND DIFFERENCES MATRIX INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS 36

DAEC DEVIATIONS AND DIFFERENCES MATRIX AE(!' x:;1\;-e.Y<, Recognition Category: E-HUl E-HUl Verbatim None Initiating Condition: Damage to a Damage to a loaded cask Verbatim None loaded cask CONFINEMENT CONFINEMENT BOUNDARY. BOUNDARY. Operating Mode Applicability: All Operating Mode Applicability: All Verbatim None (1) Damage to a loaded cask (1) Damage to a loaded cask Difference Global Comment #8, 9, 12 & 14 V22 CONFINEMENT BOUNDARY CONFINEMENT BOUNDARY as indicated by an on- as indicated by a radiation contact radiation reading reading greater than the greater than (2 times the values shown on Table E-1 on site-specific cask specific the spent fuel cask. technical specification .-t allowable radiation level)

, Table E~ 1 Cask tDose

c I on the surface of the spent Rates LIJ fuel cask. 800 61BTDSC mrem/hr 3 feet from 200 HSM mrem/hr Surface Outside HSMDoor- 40 Centerline mrem/hr ofDSC Intent and meaning of the EALs are not altered.

37

DAEC DEVIATIONS AND DIFFERENCES MATRIX FISSION PRODUCT BARRIER ICS/EALS The following section is configured in a manner that is different from the Fission Product Barrier Tables in the DAEC EAL Technical Bases Document. Where the Technical Bases Document evaluates all three fission product barriers simultaneously for a specific sub-category, this matrix presents each fission product barrier individually for all sub-categories. The significance of this presentation is that where the fission product barrier table in the Technical Bases Document moves vertically through the sub-categories, this matrix moves horizontally. 38

DAEC DEVIATIONS AND DIFFERENCES MATRIX Fission Product Barrier Emergency Classifications '.111'" ** *.::~f::r .. .::!\I.El 99~.01£~ev:< 6:1.:;,'. * .,!f ;::*,;1 :f }{~:r *. ,:i1c~n**.;DAE~,;IJ;.: ',,;()~,;. c; 'P, < ,' ._. t~t,zi I;, .c;;.~~iJg!=! :,~j f;, . , :}~;~)' :\ J~.s~ifica\t~n .:::~::i~i c

                                                                                                                                                                                                           <'. i

~i

• Validation,'~*;,J;>l,:,,F-

                                                                                                                                                                                                                 /. ' "l,;:)}.>:' ', '
                                                                                                                                                                                                                                       #~*'

Table 9-F-1: Recognition Category "F" Initiating Condition Matrix Site Area General Alert Emergency Emergency Any Loss or Loss or Loss of any two ,, any Potential Potential barriers and , Loss or Deleted per developer note. Mode Loss of either Loss of any applicability carried over onto Table 9-F the Fuel Clad two barriers. Potential Loss Deleted Difference None EAL listing. or RCS barrier. of the third barrier. Global Comment #11 Op Modes: Op Modes: Op Modes: Power Power Power Operation, Hot Operation, Operation, Hot Standby, Hot Standby, Standby, Startup, Hot Startup, Hot Startup, Hot Shutdown Shutdown Shutdown Table 9-F-2: BWR EAL Fission Product Barrier Table F-1: DAEC EAL Fission Product Renumbered and re-labeled due to Table Thresholds for LOSS or POTENTIAL LOSS of Barrier Table Thresholds for LOSS or Difference deletion ofTables 9-F-1 & 3. None Barriers POTENTIAL LOSS of Barriers Added Global Comment #9 Table 9-F-3: PWR EAL Fission Product Barrier Table Thresholds for LOSS or POTENTIAL LOSS of Deleted Difference Global Comment #4 None Barriers Basis Information For BWR EAL Fission Product Deleted Devel~per Notes Difference Transform generic NEI 99-01 guidance Barrier Table 9-F Developer Notes.

                                                                                                         ..                                                    into DAEC-specific application .

None Figure 9-F-4: PWR Containment Integrity or Deleted Difference Global Comment #4 None Bypass Example 39

DAEC DEVIATIONS AND DIFFERENCES MATRIX Thresholds for LOSS or POTENTIAL LOSS of Fuel Clad Barrier

-'   Table g::f    *                  . , NEl,99~0t Rev. 6 *,     :~     ,. . .*           .           DAEC.;'    .    ..   '       Change          Ju~tification Loss                  Potential Loss                    Loss                Potential Loss

1. RCS Activity A. (Site-specific Not Applicable A. Coolant activity Not Applicable Difference General Comment #9 indications that reactor greater than 300 coolant activity is µCi/gm dose greater than 300 equivalent 1-131

                      . µCi/gm dose equivalent 1-131).

2*.. RPV Wat~r '*' A. Primary containment A. RPV water level A. SAG entry is A. RPV water level cannot Difference EPFAQ 2015-004 Level ,, flooding required. cannot be restored reqyired. be restored and maintained Vl5 and maintained above above +15 inches OR cannot General Comment #9, 13 (site-specific RPV be determined. water level corresponding to the top of active fuel) or cannot be

              '                                     determined.

3. Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Verbatim None 4., Primary . A. Primary Not Applicable A. Drywell Monitor Not Applicable Difference V23 Containment containment (9184A/B) Global Comment #9 f!adia~ion radiation monitor reading greater reading greater than 1250 R/hr.

than (site-specific OR value). B. Torus Monitor (9185A/B) reading greater than 125 R/hr s: Other ' * , ** A. (site-specific as A. (site-specific as A. Fuel damage Not Applicable Difference Global Comment #9 l~~i~atJ~ns ,. applicable) applicable) assessment Core damage assessment indicates at least procedure. 5% fuel clad damage. 40

DAEC DEVIATIONS AND DIFFERENCES MATRIX Thresholds for LOSS or POTENTIAL LOSS of Fuel Clad Barrier

                                      . : =1# NEl;gg.io(*Re~/6
                                                                                                          ' DAEC .:~k:'
                                                                                                                    -1"L
                                                                                                                                                      . ,i;, ]Justificati'on
                                                                                                                                                          ""'  * : ~-'"i * - -:,,;1* ' ;

Loss Potential Loss Loss Potential Loss 6."';:Eme:rgency. A. ANY condition in the A. ANY condition in A. ANY condition in A. ANY condition in the Verbatim None '. Dif'ectc:lr * :.t,fit opinion of the the opinion of the the opinion of the opinion of the Emergency Jtidgn,~nt * *** Emergency Director Emergency Director Emergency Director Director that indicates

c.

;~l:,..:.t "'"/f'i .J-that indicates Loss of that indicates that indicates Loss Potential Loss of the Fuel the Fuel Clad Barrier. Potential Loss of of the Fuel Clad Clad Barrier.

the Fuel Clad Barrier. Barrier. 41

DAEC DEVIATIONS AND DIFFERENCES MATRIX Thresholds for LOSS or POTENTIAL LOSS of RCS Barrier

    .Table9~F                                                                                                                  C~ange Loss               Potential Loss                 Loss                Potential Loss 1.' Primary:,;.          A. Primary containment    Not Applicable           A. Primary             Not Applicable             Difference   V24 Containment

-*-:. * ,;r/'_;'.,

pressure greater than containment pressure . Global Comment #9 Pressure {site-specific value) due greater than 2 psig Rel)amed to to RCS leakage. due to RCS leakage.

1. P.rirnari Conl:ainmerit ..

Con.ditions \.: *

2. RPV Water** . A. RPV water level Not Applicable A. RPV water level Not Applicable Difference V19

• Level- cannot be restored and cannot be restored Global Comment #9, 13 maintained above {site- and maintained above specific RPV water level +15 inches OR cannot corresponding to the be determined.

top of active fuel) or cannot be determined. 3: *,RCS Leak. A. UNISOLABLE break in A. UNISOLABLE A. UNISOLABLE break A. UNISOLABLE primary Difference V25

  */}ate** *~'1'

• ANY of the following: primary system system leakage that Global Comment #9 in Main Steam,

{site-specific systems leakage that HPCI, Feedwater, results in exceeding with potential for high- results in RWCU, or RCIC as the Max Normal Added site-specific energy line breaks) exceeding EITHER indicated by the Operating Limit indication of an OR of the following: failure of both (MNOL) of EOP 3, unisolable steam line B. Emergency RPV 1. Max Normal isolation valves in Table 6 for EITHER of break which includes Depressurization. Operating ANY one line to the following: failure of both isolation Temperature close AND EITHER: valves to LOSS 3.A.

• Temperature OR
• High MSL flow OR

2. Max Normal or steam tunnel Operating Area

• Radiation Level temperature Radiation Level. annunciators OR
• Direct report of steam release OR B. Emergency RPV De pressurization

                                                                                ,required.

42

DAEC DEVIATIONS AND DIFFERENCES MATRIX Thresholds for LOSS or POTENTIAL LOSS of RCS Barrier A. Primary Not Applicable *A. Drywell Monitor Not Applicable Difference Global Comment #9 containment radiation (9184A/B) reading V23 monitor reading greater than 5 R/hr greater than (site- after reactor specific value). shutdown A. (site-specific as A. (site-specific as Not Applicable Not Applicable Difference Global Comment #9 applicable) applicable) A. ANY condition in A. ANY condition in A. ANY condition in A. ANY condition in the Verbatim None the opinion of the the opinion of the the opinion of the opinion of the Emergency Director Emergency Emergency Emergency Director that indicates Loss Director that Director that that indicates Potential of the RCS Barrier. indicates Potential indicates Loss of Loss of the RCS Barrier. Loss of the RCS the RCS Barrier. Barrier. 43

DAEC DEVIATIONS AND DIFFERENCES MATRIX Thresholds for LOSS or POTENTIAL LOSS of Containment Barrier Ji.i'stificatior( Loss Potential Loss Loss Potential Loss lF~rirT)~i;v :-:I( . A. UNPLANNED rapid drop A. Primary containment A. UNPLANNED rapid A. Torus pressure Difference Global Comment #9 Coiltainmentf .:*. in primary containment pressure greater than drop in Drywell greater than 53 psig V20

c.~n~i~ig;}~. :f"

• pressure following (site-specific value) pressure following OR V26 primary containment OR Drywell pressure rise V27 B. Drywell or Torus H2 pressure rise B. (site-specific OR cannot be OR explosive mixture) exists B. Drywell pressure Primary Containment determined to be B. Primary containment inside primary response not Isolation Failure Loss 3.A less than 6% and pressure response not containment *'consistent with LOCA and 3.B moved to sub-Drywell OR Torus consistent with LOCA OR ~onditions. category 1 "Primary 02 cannot be conditions. C. HCTL exceeded. OR Containment Conditions" determined to be C. UNISOLABLE direct as Losses 1.C and 1.D to less than 5%

downstream pathway consolidate concepts into to the environment OR single sub-category exists after primary C. HCL (Graph 4 of containment isolation EOP2) signal exceeded. OR D. Intentional primary containment venting per EOPs 2.-R~V;'-'1,,,ter:;}/. Not Applicable A. Primary Not Applicable A. SAG entry is Difference EPFAQ 2015-004 Level .  :*,,

• containment required.

flooding required. 44

DAEC DEVIATIONS AND DIFFERENCES MATRIX Thresholds for LOSS or POTENTIAL LOSS of Containment Barrier Loss Potential Loss Loss Potential Loss A. UNISOLABLE direct Not Applicable A. 'UNISOLABLE primary Not Applicable Difference Global Comment #9 downstream pathway system leakage that V28 to the environment results in exceeding exists after primary the Max Safe Primary Containment containment isolation Operating Limit Isolation Failure Loss 3.A signal (MSOL) of EOP 3, and 3.B moved to sub-OR Table 6 for EITHER of category 1 "Primary B. Intentional primary the following: Containment Conditions" containment venting

• Temperature as Losses 1.C and 1.D to per EOPs consolidate concepts into OR OR single sub-category C. UNISOLABLE primary
• Radiation Level system leakage that results in exceeding EITHER of the following:

1. Max Safe Operating Temperature.

OR

2. Max Safe Operating Area Radiation Level.

Not Applicable A. Primary containment Ndt Applicable A. Drywell Monitor Difference Global Comment #9 radiation monitor (9184A/B) reading V23 reading greater than greater than 5000 (site-specific value). R/hr. OR B. Torus Monitor (9185A/B) reading greater than 500 R/hr 45

DAEC DEVIATIONS AND DIFFERENCES MATRIX Thresholds for LOSS or POTENTIAL LOSS of Containment Barrier Table 9-F-2 *~:: N El 99-.01 _Rev'..* 6

                                                              *.                              'z.,.      ,,-,;

DAEC

                                                                                                                         ..                           CJ,ange ' z Justification .*

Sub:categ6iy *.

• Loss Potential Loss Loss Potential Loss

5. Other A. (site-specific as A. (site-specific as Not Applicable A. Fuel damage Difference Global Comment #9 Indications

   ' ' '"     ~       ,
                        -- ..         applicable)                      applicable)                                             assessment                         Core damage assessment

.;:~_

indicates at least procedure. 20% fuel clad damage .

           ."t

. 6. Emergencyc- : A. ANY condition in the B. ANY condition in the C. ANY condition in the D. ANY condition in the Verbatim None Direttor '. opinion of the opinion of the opinion of the opinion of the Judgment *

     ~ V    ~;

Emergency Director Emergency Director Emergency Director Emergency Director

  '.     ;'{s

__ ,,_\ that indicates Loss of that indicates that indicates Loss that indicates the Containment Potential Loss of the of the Containment Potential Loss of the Barrier. Containment Barrier. Containment

                         ..                                             Barrier.                                               Barrier.

46

DAEC DEVIATIONS AND DIFFERENCES MATRIX HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS 47

DAEC DEVIATIONS AND DIFFERENCES MATRIX Section ,:., NEI 99-0lRev/.6 I 'j. Change !,**. .Justification . l Val.idation # l Recognition Category: HUl HUl Verbatim Global Comment #11, 14 None Initiating Condition: Confirmed Confirmed SECURITY CONDITION Verbatim None SECURITY CONDITION or threat. or threat. Operating Mode Applicability: All Operating Mode Applicability: All Verbatim None (1) A SECURITY CONDITION (1) A SECURITY CONDITION that Difference Global Comment #9 & 12 None that does not involve a does not involve a HOSTILE HOSTILE ACTION as ACTION as reported by DAEC reported by the (site- Security Shift Supervision. speeific security shift supervision). .-1 (2) Notification of a credible (2) Notification of a credible Difference Global Comment #9 None

, security threat directed at security threat directed at

c -

the site. DAEC. (3) A validated notification (3) A validated notification from Verbatim None None from the NRC providing the NRC providing information of an aircraft information of an aircraft threat. threat. Intent and meaning of the EALs are not altered. 48

                                                  . DAEC DEVIATIONS AND DIFFERENCES MATRIX
                      ,:NEl,99-01 Rev. ,6 -                 . ::DAEC                                         Justificat16n            VaficJation)

Recognition Category: HU2 HU2 Verbatim Global Comment #11, 14 None Initiating Condition: Seismic Seismic event greater than OBE Verbatim None event greater than OBE levels. levels. Operating Mode Applicability: All Operating Mode Applicability: All Verbatim None (1) Seismic event greater than (1) Seismic event gr~ater than Difference Global Comment #9 & 12 V29 Operating Basis Operating Basis Earthquake Earthquake (QBE) as (OBE) as indicated by receipt N indicated by: of the Amber Operating

, (site-specific indication that a Basis Earthquake Light and

c seismic event met or exceeded the wailing seismic alarm on QBE limits) 1C35.

Intent and meaning of the EALs are not altered. 49 L - -- -

DAEC DEVIATIONS AND DIFFERENCES MATRIX Sect.ion * .NEI 99-01 Rev; 6 DAE.C *** j . Change *Justification*  ! Vctlidation # j Recognition Category: HU3 HU3 Verbatim Global Comment #11, 14 None Initiating Condition: Hazardous Hazardous event. Verbatim None event. Operating Mode Applicability: All Operating Mode Applicability: All Verbatim None (1) A tornado strike within the (1) A tornado strike within the Verbatim Global Comment #12 None PROTECTED AREA. PROTECTED AREA. (2) Internal room or area (2) Internal room or area Verbatim None flooding of a magnitude flooding of a magnitude sufficient to require sufficient to require manual manual or automatic or automatic electrical electrical isolation of a isolation of a SAFETY SYSTEM SAFETY SYSTEM component needed for the component needed for the current operating mode. current operating mode. (3) Movement of personnel (3) Movement of personnel Verbatim None within the PROTECTED within the PROTECTED AREA AREA is impeded due to an is impeded due to an offsite M offsite event involving event involving hazardous

, hazardous materials (e.g., materials (e.g., an offsite

c an offsite chemical spill or chemical spill or toxic gas toxic gas release). release).

(4) A hazardous event that (4) A hazardous event that Verbatim None results in on-site results in on-site conditions conditions sufficient to sufficient to prohibit the plant prohibit the plant staff staff from accessing the site from accessing the site via via personal vehicles. personal vehicles. (5) (Site-specific list of natural Difference Global Comment #9 or technological hazard events) .. Intent and meaning of the EALs are not altered. 50

DAEC DEVIATIONS AND DIFFERENCES MATRIX

• se'd:fon Validation

                                                                                                                                       ' ,_. 'i-. ,.,.. .*
                                                                                                                                                 ~

Recognition Category: HU4 HU4 Verbatim Global Comment #11, 14 None Initiating Condition: FIRE FIRE potentially degrading the Verbatim None potentially degrading the level of level of safety of the plant. safety of the plant. Operating Mode Applicability: All Operating Mode Applicability: All Verbatim None (1) a. A FIRE is NOT extinguished (1) a. A FIRE is NOT extinguished Difference Global Comment #12 & 13 None within 15-minutes of ANY within 15-minutes of ANY of the following FIRE of the following FIRE detection indications: detection indications:

• Report from the field
• Report from the field (i.e., visual observation) (i.e., visual
• Receipt of multiple observation)

(more than 1) fire

• Receipt of multiple alarms or indications (more than 1) fire
• Field verification of a alarms or indications single fire alarm
• Field verification ofa AND single fire alarm o:t' AND

)

c b. The FIRE is located within b. The FIRE is located within Difference Global Comment #8, 9, & 13 None ANY of the following plant ANY Table H-1 plant rooms or areas: rooms or areas.

(site-specific list of plant rooms Table H-1 Fire Areas

• 1631 D6 and Day Tank Rooms or areas)
• 1621 D6 and Day Tank Rooms
• Battery Rooms
• Essential Switchgear Rooms
• Cable Spreading Room
• Torus Room
• Intake Structure
• Pumphouse
• Drywell
• Torus
• NE, NW, SE Corner Rooms
• HPCI Room
• RCICRoom
• RHR Valve Room
• North CRD Area
• South CRD Area
• CSTs
• Control Building

                                                             * -Remote Shutdown Panel 1C388 Area

• Panel lCSS/56 Area
• SB6T Room 51

DAEC'DEVIATIONS AND DIFFERENCES MATRIX Sec~ion DAEC:. . . , ( Change Justification j Validation # j (2) a. Receipt of a single fire (2) a. Receipt of a single fire Difference Global Comment #8, 9 & 13 None alarm (i.e., no other alarm with no other indications of a FIRE). indications of a FIRE. AND AND

b. The FIRE is located within b. The FIRE is located within ANY of the following plant ANY Table H-1 plant rooms rooms or areas: or areas.

(site-specific list of plant rooms or areas) AND AND

c. The existence of a FIRE is c. The existence of a FIRE is Verbatim N/A None not verified within 30- not verified within 30-minutes of alarm receipt. minutes of alarm receipt.

C 0 (3) A FIRE within the plant or ISFSI [for plants with an ISFSI outside the plant (3) A FIRE within the plant or ISFSI PROTECTED AREA not extinguished within 60 Difference Global Comment #9 None ~

,::!'                Protected Area]                      minutes of the initial

I: PROTECTED AREA not report, alarm ~r indication.

extinguished within 60-minutes of the initial report, alarm or indication. (4) A FIRE within the plant or (4) A FIRE within the plant or Difference Global Comment #9 None ISFSI [for plants with an ISFSI PROTECTED AREA ISFSI outside the plant that requires firefighting Protected Area] support by an offsite fire PROTECTED AREA that response agency to requires firefighting extinguish. support by an offsite fire response agency to extinguish. Basis revised to include NFPA-805 in the discussion of Appendix R basis for the EAL thresholds. Intent and meaning of the EALs are not altered. 52

DAEC DEVIATIONS AND DIFFERENCES MATRIX

    **Section
 '.,, > *'~ ." *'e,.{?-~>"'"-> -~,, ,*"-,

• J,4stification
• Recognition Category: HU7 HU7 Verbatim Global Comment #11, 14 None Initiating Condition: Other Other conditions exist which in Difference NOUE versus (NO)UE, DAEC uses the full None conditions exist which in the the judgment of the Emergency NOUEterm judgment of the Emergency Director warrant declaration of a Director warrant declaration of a NOUE.

(NO) UE. Operating Mode Applicability: All Operating Mode Applicability: All Verbatim None (1) Other conditions exist (1) Other conditions exist Verbatim Global Comment #3, 12, 14 None which in the judgment of which in the judgment of the Emergency Director the Emergency Director indicate that events are in indicate that events are in

 ,...                                             progress or have occurred          progress or have occurred

, which indicate a potential which indicate a potential

c:

degradation of the level of degradation of the level of safety of the plant or safety of the plant or indicate a security threat indicate a security threat to facility protection has to facility protection has been initiated. No releases been initiated. No releases of radioactive material of radioactive material requiring offsite response requiring offsite response or monitoring are or monitoring are expected unless further expected unless further degradation of safety degradation of SAFETY systems occurs. SYSTEMS occurs. 53 L

DAEC'oEVIATIONS AND DIFFERENCES MATRIX

                                                                       '*.. ~:?chan'ge*
                                                                              "*f"*'" , -
                                                                                                        Justification           Va(idaticin]f Recognition Category: HAl         HAl                                      Verbatim   Global Comment #11, 14                    None Initiating Condition: HOSTILE     HOSTILE ACTION within the                Verbatim                                             None ACTION within the OWNER           OWNER CONTROLLED AREA or CONTROLLED AREA or airborne       airborne attack threat within 30 attack threat within 30 minutes. minutes.

Operating Mode Applicability: All Operating Mode Applicability: All Verbatim None (1) A HOSTILE ACTION is (1) A HOSTILE ACTION is Difference Global Comment #9, 12, 14 None occurring or has occurred occurring or has occurred .... within the OWNER within the OWNER <I: CONTROLLED AREA as CONTROLLED AREA as

c reported by the (site- reported by the DAEC specific security shift Security Shift Supervision.

supervision). (2) A validated notification (2) A validated notification Verbatim from NRC of an aircraft from NRC of an aircraft attack threat within 30 attack threat within 30 minutes of the site. mihutes of the site. Intent and meaning of the EALs are not

                                                            -                              altered.

54

DAEC DEVIATIONS AND DIFFERENCES MATRIX NEI 99-01 Rev.* 6 *:'f Justification Recognition Category: HAS Not used at DAEC Difference EALs RA3 and HAS are not applicable to vs

                                                        ,!                     DAEC because an evaluation has shown that there are no rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. All areas outside the Control Room that contain equipment necessary for normal plant operation, cooldown and shutdown do not require physical access to operate.

Initiating Condition: Gaseous Not used at DAEC Difference None release impeding access to equipment necessary for normal plant operations, cooldown or shutdown. Operating Mode Applicability: All Not used at DAEC Difference None (1) a. Release of a toxic, Not used at DAEC

• Difference None corrosive, asphyxiant or flammable gas into any of the following plant rooms i.n or areas:

c:i:

c (site-specific list of plant rooms  !

or areas with entry-related mode applicability identified) AND

b. Entry into the room or area is prohibited or impeded.

55

DAEC DEVIATIONS AND DIFFERENCES MATRIX Recognition Category: HA6 HAS Difference Renumbered to align with other similar ICs None Initiating Condition: Control Control Room evacuation Verbatim None Room evacuation resulting in resulting in transfer of plant transfer of plant control to control to alternate locations. alternate locations. Operating Mode' Applicability: All Operating Mode Applicability: All Verbatim None \Cl <t (1) An event has resulted in (1) An event has resulted in plant Difference Global Comment #9 & 12 V30

c plant control being control being transferred transferred from the from the Control Room to the Control Room to (site- Remote Shutdown Panel specific remote shutdown (1C388).

panels and local control stations). Intent and meaning of the EALs are not altered. 56

DAEC DEVIATIONS AND DIFFERENCES MATRIX ili~n,ijn~~~A Recognition Category: HA7 HAG Difference Global Comment #11, 14 None Renumbered to align with other similar ICs Initiating Condition: Other Other conditions exist which in Verbatim None conditions exist which in the the judgment of the Emergency judgment of the Emergency Director warrant declaration of Director warrant declaration of an Alert. an Alert. Operating Mode Applicability: All Operating Mode Applicability: All Verbatim None (i) Other conditions exist (1) Other conditions exist which,- Verbatim Global Comment #12 None which, in the judgment of in the judgment of the the Emergency Director, Emergency Director, indicate indicate that events are in that events are in progress or progress or have occurred have occurred which involve which involve an actual or an actual or potential potential substantial substantial degradation of

I:

degradation ofthe level of the level of safety of the safety of the plant or a plant or a security event that security event that involves probable life involves probable life threatening risk to site threatening risk to site persqnnel or damage to site personnel or damage to equipment because of site equipment because of HOSTILE ACTION. Any HOSTILE ACTION. Any releases are expe~ted to be. releases are expected to limited to small fractions of be limited to small the EPA Protective Action fractions of the EPA Guideline exposure levels. Protective Action Guideline exposure levels. Intent and meaning of the EALs are not altered 57

DAEC DEVIATIONS AND DIFFERENCES MATRIX SectiO:n NEI 99-01 Rev.:::6, DAEC .. * . ; -'.Change Justification Validation # Recognition Category: HS1 HS1 Verbatim Global Comment #11, 14 None Initiating Condition: HOSTILE HOSTILE ACTION within the Verbatim None ACTION within the PROTECTED PROTECTED AREA. AREA. Operating Mode Applicability: All Operating Mode Applicability: All Verbatim None (1) A HOSTILE ACTION is (1) A HOSTILE ACTION is Difference Global Comment #9 & 12 None occurring or has occurred occurring or has occurred II)

c within the PROTECTED within the PROTECTED AREA AREA as reported by the as reported by the DAEC (site-specific security shift Security Shift Supervision.

supervision). Intent and meaning of the EALs are not altered. 58

DAEC DEVIATIONS AND DIFFERENCES MATRIX

   .Sectidh                                                                                                                             Validati.on #

Recognition Category: HSG HSS Difference Renumbered to align with other similar ICs None Initiating Condition: Inability to Inability to control a key safety Verbatim None control a key safety function function from outside the Control from outside the Control Room. Room. Operating Mode Applicability: All Operating Mode Applicability: All Verbatim None Note: The Emergency Director Note: The Emergency Director Global Comment #9 V30 should declare the Site Area should declare the Site Area Emergency promptly upon Emergency promptly upon determining that (site specific determining that 20 minutes has number of) minutes has been been exceeded, or will likely be exceeded, or will likely be exceeded. exceeded. (1) a. An event has resulted in (1) a. An eveht has resulted in Difference Global Comment #9, 12 None plant control being plant control being transferred from the transferred from the ID Ill Control Room to (site- Control Room to the

c specific remote shutdown Remote Shutdown Panel panels and local control (1C388}.

stations). AND AND Difference Global Comment #4, 9 V30

b. Control of ANY of the b. Control of ANY of the following key safety following key safety functions is not functions is not reestablished within (site- reestablished within 20 specific number of minutes.

minutes).

• Reactivity control
• Reactivity control
• RPV water level
• Core cooling [PWR] /
• RCS heat removal RPV water level [BWR]
• RCS heat removal Intent and meaning of the EALs are not altered.

59

• J)istific.1tio'n *

                                                                                                         ,,, Y"    * ~-*

Recognition Category: HS7 Recognition Category.: HS6 Difference Global Comment #11, 14 None Renumbered to align with other similar ICs Initiating Condition: Other Initiating Condition: Other Verbatim None conditions exist which in the conditions exist which in the judgment of the Emergency judgment of the Emergency Director warrant declaration of a Director warrant declaration of a Site Area Emergency. Site Area Emergency. Operating Mode Applicability: All Operating Mode Applicability: Verbatim None ALL (1) Other conditions exist (1) Other conditions exist which Verbatim Global Comment #12 None which in the judgment of in the judgment of the the Emergency Director Emergency Director indicate indicate that events are in that events are in progress or progress or have occurred have occurred which involve which involve actual or actual or likely major failures likely major failures of of plant functions needed for plant functions needed for protection of the public or protection of the public or HOSTILE ACTION that results In

c HOSTILE ACTION that in intentional damage or results in intentional malicious acts, (1) toward site damage or malicious acts, personnel or equipment that (1) toward site personnel could lead to the likely failure or equipment that could of or, (2) that prevent lead to the likely failure of effective access to equipment or, (2) that prevent needed for the protection of e

effective access to the public. Any releases are equipment needed for the not expected to result in protection of the public. exposure levels which exceed Any releases are not EPA Protective Action ' expected to result in Guideline exposure levels exposure levels which beyond the site boundary. exceed EPA Protective Action Guideline exposure Intent and meaning of the EALs are not levels beyond the site altered boundary. 60

DAEC DEVIATIONS AND DIFFERENCES MATRIX

             ~~I 99~.ot Rev.:j,                                                                        fostification Recognition Category: HGl             HGl                                  Verbatim   Global Comment #11, 14                 None Initiating Condition: HOSTILE         HOSTILE ACTION resulting in loss     Verbatim                                          None ACTION resulting in loss of           of physical control of the facility.

physical control ofthe facility. Operating Mode Applicability: All Operating Mode Applicability: All Verbatim None (1) a. A HOSTILE ACTION is (1) a. A HOSTILE ACTION is Difference Global Comment #9, 12 None occurring or has occurred occurring or has occurred within the PROTECTED within the PROTECTED AREA as reported by the AREA as reported by the (site-specific security shift DA.EC Security Shift supervision). Supervision. AND AND Difference Global Comment #4, 9 None

b. EITHER of the following b. EITHER of the following

.-4 has occurred: has occurred: I!>

c

1. ANY of the following 1. ANY of the following safety functions cannot safety functions cannot be controlled or be controlled or maintained. maintained.

• Reactivity control
• Reactivity control
• Core cooling [PWR] /
• RPV water level RPV water level
• RCS heat removal

[BWR]

• RCS heat removal OR OR Verbatim None

2. Damage to spent fuel 2. Damage to spent fuel has occurred or is has occurred or is IMMINENT. IMMINENT.

Intent and meaning of the EALs are not

                                                    .                                    altered .

61

DAEC DEVIATIONS AND DIFFERENCES MATRIX Section . I NEI 99.-01 Rev. :.6

                           ';,1,, " ,o,o' " ., *, ,,,_\

DAEC .. "'. 'Justificatio'n'* .'I Vali'daticm #(I Recognition Category: HG7 HG6 Difference Global Comment #11, 14 None Renumbered to align with other similar ICs Initiating Condition: Other Other conditions exist which in Verbatim None conditions exist which in the the judgment of the Emergency judgment of the Emergency Director warrant declaration of a Director warrant declaration of a General Emergency. General Emergency. Operating Mode Applicability: All Operating Mode Applicability: All Verbatim None (1) Other conditions exist (1) Other conditions exist which Verbatim Global Comment #12 None which in the judgment of in the judgment of the the Emergency Director Emergency Director indicate indicate that events are in that events are in progress or progress or have occurred have occurred which involve which involve actual or actual or IMMINENT IMMINENT substantial substantial core degradation c.i,

c core degradation or or meltin_g with P,Otential for melting with potential for loss of containment integrity loss of containment or HOSTILE ACTION that integrity or HOSTILE results in an actual loss of ACTION that results in an physical control of the actual loss of physical facility. Releases can be control of the facility. reasonably expected to Releases can be reasonably exceed EPA Protective Action expected to exceed EPA Guideline exposure levels Protective Action Guideline offsite for more than the exposure levels offsite for immediate site area.

more than the immediate site area. Intent and meaning of the EALs are not altered 62

DAEC DEVIATIONS AND DIFFERENCES MATRIX SYSTEM MALFUNCTION ICS/EALS 63

DAEC DEVIATIONS AND DIFFERENCES MATRIX Sedipn .** .. .. ~e1 99-01 Rev.~6.

                      '     '              --*~                                                              . Justification

• I Val~dation # I Recognition Category: SUl SUl Verbatim None Initiating Condition: Loss of all Loss of ALL offsite AC power Difference Global Comment #15 None offsite AC power capability to capability to essential buses for emergency buses for 15 minutes 15 minutes or longer.

or longer. .... Operating Mode Applicability: Operating Mode Applicability: 1, Difference Global Comment #10 None

, Power Operation, Startup, Hot 2, 3 Ill '*'

Standby, Hot Shutdown (1) Loss of ALL offsite AC (1) Loss of ALL offsite AC power Difference Global Comment #9 & 12 None power capability to (site- capability to 1A3 AND 1A4 specific emergency buses) buses for 15 minutes or for 15 minutes or longer. longer. Intent and meaning of the EALs are not altered. 64

DAEC DEVIATIONS AND DIFFERENCES MATRIX

     . SecticM
        " ,p ,,:<:,,.-.-.??/,',,
                                                                                                                                    . Juitification* ..           Vafi"Hation #

Recognition Category: SU2 SU3 Difference Global Comment #11, 14 None Renumbered to align with other similar ICs Initiating Condition: UNPLANNED UNPLANNED loss of Control Verbatim None loss of Control Room indications Room indications for 15 minutes for 15 minutes or longer. or longer. Operating Mode Applicability: Operating Mode Applicability: 1, Difference Global Comment #10 None Power Operation, Startup, Hot 2, 3 Standby, Hot Shutdown (1) a. An UNPLANNED event (1) a. An UNPLANNED event Difference Global Comment #12 None results in the inability to results in the inability to monitor one or more of monitor one or more of the following parameters the Table S-1 parameters from within the Control from within the Control Room for 15 minutes or Room for 15 minutes or longer. longer. N Table S-1 Safety System Difference Global Comment #4, 9 None VI [BWR parameter [PWR Parameters list] parameter list] Reactor Power Reactor Power

• Reactor Power RPV Water Level RCS Level
• RPV Water Level RPV Pressure RCS Pressure
• RPV Pressure Primary Containment In-Core/Core Exit
• Primary Containment Pressure Pressure Temperature Suppression Pool Levels in at least
• Suppression Pool Level Level (site-specific
• Suppression Pool number) two Temperature steam generators Suppression Pool Steam Temperature Generator Auxiliary or Emergency Feed Water Flow Intent and meaning of the EALs are not altered.

65

Recognition Category: SU3 SU4 Verbatim Global Comment #11, 14R None Renumbered IC to align with other similar ICs Initiating Condition: Reactor Reactor coolant activity greater Verbatim None coolant activity greater than than Technical Specification Technical Specification allowable allowable limits. limits. Operating Mode Applicability: Operating Mode Applicability: 1, Difference Global Comment #10 None Power Operation, Startup, Hot 2, 3 Standby, Hot Shutdown (1) (Site-specific radiation (1) Pretreatment Offgas System Difference Global Comment #9 & 12 None m

, monitor) reading greater (RM-4104) Hi-Hi Radiation II) than (site-specific value). Alarm (2) Sample analysis indicates (2) Sample analysis Difference Global Comment #9 V31 that a reactor coolant indicates that reactor activity value is greater coolant_ specific than an allowable limit activity is greater specified in Technical than 2.0 µCi/gm dose Specifications. equivalent 1-131 for 12 hours or longer.

Intent and meaning of the EALs are not altered. J 66

DAEC DEVIATIONS AND DIFFERENCES MATRIX

     .Section.          ;NEI 99-01.Rev> 5.*                                                                                                 v.ditJation:#

Recognition Category: SU4 SUS Verbatim Global Comment #11, 14 None Renumbered to align with other similar ICs Initiating Condition: RCS leakage RCS leakage for 15 minutes or Verbatim None for 15 minutes or longer. longer. Operating Mode Applicability: Operating Mode Applicability: 1, Difference Global Comment #10 None Power Operation, Startup, Hot 2, 3 Standby, Hot Shutdown (1) RCS unidentified or (1) RCS unidentified or pressure Difference Global Comment #9 & 12 V32 pressure boundary leakage boundary leakage. greater , greater than (site-specific than 10 gpm for 15 minutes value) for 15 minutes or or longer. longer. o::t' (2) RCS identified leakage (2) RCS identified leakage greater Difference Global Comment #9 V32 II) greater than (site-specific than 25 gpm for 15 minutes value) for 15 minutes or or longer. longer. (3) Leakage from the RCS to a (3) Leakage from the RCS to a Verbatim None location outside location outside containment containment greater than greater than 25 gpm for 15 25 gpm for 15 minutes or minutes or longer. longer. Intent and meaning of the EALs are not altered. 67

DAEC DEVIATIONS AND DIFFERENCES MATRIX

                                                                                                  ,;, ;~ ,*

J'<UStifiCatiOJ1 Recognition Category: SUS SU6 Verbatim Global Comment #11, 14 None Renumbered to align with other similar ICs Initiating Condition: Automatic or Automatic or manual scram fails Difference Global Comment #4 None manual (trip [PWR] / scram to shutdown the reactor. [BWR]) fails to shutdown the reactor. Operating Mode Applicability: Operating Mode Applicability: 1, Difference Global Comment #10 V33 Power Operation 2 DAEC can be up to 12% power in STARTUP Mode, so Mode 2 applicability added (1) a. An automatic (trip [PWR] / (1) a. An automatic scram did Difference Global Comment #4 & 12 None scram [BWR]) did not no't shutdown the reactor. u,

, shutdown the reactor.

V, AND AND Difference Global Comment #9 None

b. A subsequent manual b. ANY of the following manual action taken at the reactor actions taken at 1COS are control consoles is successful in lowering reactor successful in shutting down power below 5% power the reactor.

• Manual Scram Pushbuttons
• Mode Switch to Shutdown
• Alternate Rod Insertion (ARI) 68

DAEC DEVIATIONS AND DIFFERENCES MATRIX

• . *:sectiqn (2) a. A manual trip ([PWR] / (2) a. A manual scram did not Difference Global Comment #4 None scram [BWR]) did not shutdown the reactor.

shutdown the reactor. AND AND- - None

b. EITHER of the following: lb. 1. EITHER of the following Difference Global Comment #9

1. A subsequent manual subsequent manual actions action taken at the taken at lCOS are successful reactor control consoles in lowering reactor power is successful in shutting below 5% power

-,.J C

down the reactor.

• Manual Scram Pushbuttons Mode Switch to Shutdown

-u, 0 u VI

• Alternate Rod Insertion (ARI)

OR OR Difference Global Comment #4 None

2. A subsequent automatic 2. A subsequent automatic (trip [PWR] / scram scram is successful in

[BWR]) is successful in shutting down the reactor. shutting down the reactor. Intent and meaning of the EALs are not altered. 69

DAEC DEVIATIONS AND DIFFERENCES MATRIX I*: Section . I . Justification'.')' l Validation # I Recognition Category: SU6 SU7 Verbatim Global Comment #14 None Renumbered to align with other similar ICs Initiating Condition: Loss of all Loss of ALL onsite or offsite Difference Global Comment #13 None onsite or offsite communications communications cap'c1bilities. capabilities. Operating Mode Applicability: Operating Mode Applicability: 1, Difference Global Comment #10 None Power Operation, Startup, Hot 2, 3 Standby, Hot Shutdown (1) Loss of ALL of the following (1) Loss of ALL of the following Difference Global Comment #9, 12 & 13 V16 Onsite communication Onsite communication methods: methods: (site-specific list of

• Plant Operations Radio communications methods) System
• In-Plant Phone System

  \Cl

• Plant Paging System V, (Gaitronics)

(2) Loss of ALL of the following (2) Loss of ALL of the following Difference Global Comment #9 & 13 V13, V'.1:4 ORO communications offsite response organization methods: communications methods: (site-specific list of

• DAEC All-Call phone communications methods)
• All telephone lines (PBX and commercial)
• Cell Phones (including fixed cell phone system)
• Control Room fixed satellite phone system
• FTS Phone system 70

DAEC DEVIATIOI\JS AND DIFFERENCES MATRIX (3) Loss of ALL of the following (4) Loss of ALL of the following Difference Global Comment #9 & 13 V13, V14 NRC communications NRC communications 0 C: methods: (site-specific list of communications methods) methods: FTS Phone system All telephone lines (PBX ~ U) and commercial) Ill

• Cell Phones (including fixed cell phone system)
• Control Room fixed satellite phone system Intent and meaning of the EALs are not altered.

71

VaJidatio,;t# .. Recognition Category: SU7 Not Applicable Difference Global Comment #4 None

                                                       ':                This IC and EALs are only applicable to PWR plants.

Initiating Condition: Failure to isolate containment or loss of containment pressure control. [PWR] Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown (1) a. Failure of containment to isolate when required by an actuation signal. AND ,... b. ALL required penetrations

, are not closed within 15 11'1 minutes of the actuation signal.

(1) a. Containment pressure greater than (site-specific pressure).

                                                       ~

AND

b. Less than one full train of (site-specific system or equipment) is operating per design for 15 minutes or longer.

72

DAEC DEVIATIONS AND DIFFERENCES MATRIX Justificat(qn j Va'l~atiorf~ j Recognition Category: SAl SAl

• Verbatim Global Comment #11, 14 None Initiating Condition: Loss of all Loss of ALL but one AC power Difference Global Comment #15 None but one AC power source to source to essential buses for 15 emergency buses for 15 minutes minutes or longer.

or longer. Operating Mode Applicability: Operating Mode Appljcability: 1, Difference Global Comment #10 None Power Operation, Startup, Hot 2, 3 Standby, Hot Shutdown (1) a. AC power capability to (1) a. AC power capability to Difference Global Comment #9, 12 None ....<I: (site-specific emergency 1A3 and 1A4 buses is Ill buses) is reduced to a reduced to a single single power source for 15 power source for 15 minutes or longer. minutes or longer. AND AND Difference Global Comment #13 None

b. Any additional single a. ANY additional single power source failure will power source failure will result in a loss of all AC result in a loss of ALL AC power to SAFETY SYSTEMS. power to SAFETY SYSTEMS.

Intent and meaning of the EALs are not altered. 73

DAEC DEVIATIONS AND DIFFERENCES MATRIX Section .. : I\IEI 99-01 Rev: 6*. Justification Recognition Category: SA2 SA3 Difference Global Comment #11, 14 None Renumbered to align with other similar ICs Initiating Condition: UNPLANNED UNPLANNED loss of Control Verbatim None loss of Control Room indications Room indications for 15 minutes for 15 minutes or longer with a or longer with a significant significant transient in progress. transient in progress. Operating Mode Applicability: Operating Mode Applicability: 1, Difference Global Comment #10 None Power Operation, Startup, Hot 2, 3 Standby, Hot Shutdown (1) a. An UNPLANNED event (1) a. An UNPLANNED event Verbatim Global Comment #12 None results in the inability to results in the inability to monitor one or more of monitor one or more Table the following parameters S-1 parameters from from within the Control within the Control Room Room for 15 minutes or for 15 minutes or longer longer. N <t [BWR [PWR parameter Table S-1 Safety System Difference Global Comment #4, 8 None VI parameter list] list] Parameters Reactor Power Reactor Power RPVWater RCS Level

• Reactor Power Level
• RPV Water Level RPV Pressure RCS Pressure
• RPV Pressure Primary Containment In-Core/Core Exit Temperature
• Primary Containment Pressure Pressure Suppression Levels in at least
• Suppression Pool Level Pool Level (site-specific
• Suppression Pool number) steam Temperature generators Suppression Steam Generator Pool Auxiliary or Temperature Emergency Feed Water Flow AND AND 74

DAEC DEVIATIONS AND DIFFERENCES MATRIX

• J.ustification * .Validation

                                                                                                                    '* ,~_,_;" '.,.

b. ANY of the following b. Any of the Table S-2 Difference Global Comment #4, 9 None transient events in transient events are in progress. progress

• Automatic or manual runback greater than Table S-2 Significant

~ C 25% thermal reactor power

• Automatic Transients or manual 0

u ~ VI

• Electrical load rejection greater than 25% full electrical load
• run back greater than 25%

thermal reactor power Electrical load rejection

• Reactor scram [BWR] / greater than 25% full trip [PWR] electrical load
• ECCS (SI) actuation
• Reactor scram
• Thermal power
• ECCS actuation oscillations greater than
• Thermal power oscillations 10% [BWR] greater than 10%

Intent and meaning of the EALs are not altered. 75

DAEC DEVIATIONS AND DIFFERENCES MATRIX

                                                      ,,,~"~ec
                                                           ;;-<<,,t,J;;

Recognition Category: SAS SAG Difference Global Comment #11, 14 None Renumbered to align with other similar ICs Initiating Condition: Automatic or Automatic or manual scram fails Difference Global Comment #4 & 9 None manual (trip [PWR] / scram to shutdown the reactor, and [BWR]) fails to shutdown the subsequent manual actions taken reactor, and subsequent manual at the reactor control consoles actions taken at the reactor are not successful in shutting control consoles are not down the reactor. successful in shutting down the reactor. Operating Mode Applicability: Operating Mode Applicability: 1, Difference Global Comment #10 V33 Power Operation 2 DAEC can be up to 12% power in STARTUP Mode, so Mode 2 applicability added (1) a. An automatic or manual (1) a. An automatic or manual Difference Global Comment #4, 9 & 12 None (trip [PWR] / scram [BWR]) scram did not shutdown LI) <C II) did not shutdown the the reactor. reactor. AND AND Difference Global Comment #9 None

b. Manual actions taken at b. ALL of the following the reactor control manual actions taken at consoles are not successful lCOS are not successful in shutting down the in lowering reactor reactor. power below 5% power Manual Scram Pushbuttons

• Mode Switch to Shutdown
• Alternate Rod Insertion (ARI)

Intent and meaning of the EALs are not altered. 76

DAEC DEVIATIO~S AND DIFFERENCES MATRIX

                                     ~~~              :~e'                             ~~~
                                                     ,,gf,;;,J/y,-,

Recognition Category: SA9 SA8 Difference Global Comment #11, 14 None Renumbered to align with other similar ICs Initiating Condition: Hazardous Hazardous event affecting a Verbatim None event affecting a SAFETY SYSTEM SAFETY SYSTEM needed for the needed for the current operating current operating mode. mode. Operating Mode Applicability: Operating Mode Applicability: 1, Difference Global Comment #10 None Power Operation, Startup, Hot 2, 3 Standby, Hot Shutdown (1) a. The occurrence of ANY of (1) a. The occurrence of ANY of Difference Global Comment #12 & 13 None the following hazardous the Table S-3 hazardous events: events:

• Seismic event Table S-3 Hazardous Events Difference Global Comment #8 & 9 None C"I (earthquake}
• Seismic event (earthquake}

<C II)

• Internal or external
• Internal or external flooding flooding event event
• High winds or tornado
• High winds or tornado strike strike
• FIRE
• FIRE
• EXPLOSION
• EXPLOSION
• Other events with similar

           * (site-specific hazards}        hazard characteristics as

• Other events with determined by the Shift similar hazard Manager or Emergency characteristics as Director determined by the Shift Manager 77

      }N.El 99~01,*Rev~;;:6 *
         ,' ~ "' ' ' ' '/<~ > ' >

• Chaiige*
• J.ustificatidn AND
• AND

b. EITHER of the following: b. 1. Event da-~age has Deviation Adopted the revised EAL structure and V17

1. Event damage has caused ini:lications of wording provided in approved EAL FAQ caused indications of degraded 2016-02.

degraded performance performance in one in at least one train of a train of a SAFETY SAFETY SYSTEM needed SYSTEM needed for for the current the current operating operating mode. mode. AND OR 2. EITHER of the following: Deviation Adopted the revised EAL wording provided V17

2. The event has caused

• Event damage has in approved EAL FAQ 2016-02 VISIBLE DAMAGE to a caused indications SAFETY SYSTEM of degraded component or structure performance to a Difference Added the following clarification to the V18 C

0 needed for the current operating mode. second train of the SAFETY SYSTEM Basis from EALFAQ 2018-04: An event affecting a single-train SAFETY ~ needed for the SYSTEM (i.e., there are indications of en <C current operating degraded performance and/or VISIBLE U'I mode, or DAMAGE affecting the one train) would not

• The event has be classified under SA8 because the two-resulted in VISIBLE train impact criteria that underlie the EALs
• DAMApE to the
• and Bases would not be met. If an event second train of a affects a single-train SAFETY SYSTEM, then SAFETY SYSTEM the emergency classification should be needed for the made based on plant current operating parameters/symptoms meeting the EALs mode. for another IC. Depending upon the circumstances, classification may also occur based on Shift Manager/Emergency Director judgement.

Intent and meaning of the EALs are not altered. 78

DAEC DEVIATIONS AND DIFFERENCES MATRIX Recognition Category: SSl SSl Verbatim Global Comment #11, 14 None Initiating Condition: Loss of all Loss of ALL offsite and ALL onsite Difference Global Comment #13, 15 None offsite and all onsite AC power to AC power to essential buses for emergency buses for 15 minutes 15 minutes or longer. or longer. Operating Mode Applicability: Operating Mode Applicability: 1, Difference Global Comment #10 None .-4 11'1 11'1 Power Operation, Startup, Hot 2, 3 Standby, Hot Shutdown (1) Loss of ALL offsite and ALL (1) Loss of ALL offsite and ALL Difference Global Comment #9, 12 & 13 None onsite AC power to (site- onsite AC power to 1A3 and specific emergency buses) 1A4 buses for 15 minutes or for 15 minutes or longer. longer. Intent and meaning of the EALs are not altered. 79

DAEC ,DEVIATIONS AND DIFFERENCES MATRIX

    *:$ection        " NEI .!i>9-0l*~ev..6                        DAEC Recognition Category: SSS           SS6                                Difference Global Comment #11, 14                     None Renumbered to align with other similar ICs Initiating Condition: Inability to  Inability to shutdown the reactor  Difference Global Comment #4                          None shutdown the reactor causing a      causing a challenge to RPV water challenge to (core cooling [PWR]    level or RCS heat removal.
              / RPV water level [BWR]) or RCS heat removal.

Operating Mode Applicability: Operating Mode Applicability: 1, Difference Global Comment #10 V33 Power Operation 2 DAEC can be up to 12% power in STARTUP Mode, so Mode 2 applicability added (1) a. An automatic or manual (1) a. An automatic or manual Difference Global Comment #4, 9 & 12 None (trip [PWR] / scram [BWR]) scram did not shutdown did not shutdown the the reactor. reactor. u, AND AND Verbatim None 11'1 11'1

b. All manual actions to b. All manual actions to shutdown the reactor have shutdown the reactor been unsuccessful. have been unsuccessful.

AND AND Difference Global Comment #9 V34 C. EITHER of the following C. EITHER of the following V27 conditions exist: conditions exi~t:

                      * (Site-specific indication

• RPV level cannot be of an inability to restored and maintained adequately remove heat above -25 inches.

from the core) OR

                      * (Site-specific indication

• HCL (Graph 4 of EOP 2) of an inability to exceeded.

adequately remove heat from the RCS) Intent and meaning of the EALs are not altered. 80

DAEC DEVIATIONS AND DIFFERENCES MATRIX

 , sectipn           NEI 99~01 Rev. 6..

• Recognition Category: SS8 SS2 Difference Global Comment #11, 14 None Renumbered to align with other similar ICs Initiating Condition: Loss of all Loss of ALL Vital DC power for 15 Difference Global Comment #13 None Vital DC power for 15 minutes or minutes or longer.

longer. Operating Mode Applicability: Oper~ting Mode Applicability: 1, Difference Global Comment #10 None Power Operation, Startup, Hot 2, 3 co VI VI Standby, Hot Shutdown (1) Indicated voltage is less (1) Indicated voltage is less Difference Global Comment #9 & 12 V12 than (site-specific bus than 105 VDC on BOTH voltage value) on ALL (site- Div 1 and Div 2 125 voe specific Vital DC busses) for buses for 15 minutes or 15 minutes or longer. longer. Intent and meaning of the EALs are not altered. 81

DAEC DEVIATIONS AND DIFFERENCES MATRIX

 .'.Section*          *NEI 99-01:R.ev. :G                           DAEC.     .~ *

• Ch~nge.. ~ustification Recognition Category: SGl SGl Verbatim Global Comment #11, 14 None Initiating Condition: Prolonged Prolonged loss of ALL offsite and Difference Global Comment #13, 15 None loss of all offsite and all onsite AC ALL onsite AC power to essential power to emergency buses. buses.

Operating Mode Applicability: Operating Mode Applicability: 1, Difference Global Comment #10 None Power Operation, Startup, Hot 2, 3 Standby, Hot Shutdown (1) a. Loss of ALL offsite and ALL (1) a. Loss of ALL offsite and ALL Difference Global Comment #9 & 13 None onsite AC power to (site- onsite AC power to 1A3 specific emergency buses). and 1A4 buses ,-1 I.!:! V, AND AND Difference Global Comment #9 & 13

b. EITHER of the following: b. EITHER of the following:

• Restoration of at least
• Restoration of at least one AC emergency bus one AC essential bus in in less than (site-specific less than 4 hours is not hours) is not likely. likely.

OR

                     *   (Site-specific indication

• RPV level cannot be Difference Global Comment #9 V34 of an inability to restored and maintained adequately remove heat above from the core) ,-25 inches.

Intent and meaning of the EALs are not altered. s 82

DAEC DEVIATIONS AND DIFFERENCES MATRIX NEI 99-0fRev: :6,

,;J;,' , ..... ; ... ~** * ',~ ,, '

                                                                ,' ,DAEC/(                 :Change c*,. *;

Justifica~iol'I Validation,#' Recognition Category: SG8 SG2 Difference Global Comment #11, 14 None Renumbered to align with other similar ICs Initiating Condition: Loss of all AC Loss of ALL AC and Vital DC Verbatim Global Comment #13 None and Vital DC power sources for power sources for 15 minutes or 15 minutes or longer. longer. Operating Mode Applicability: Operating Mode Applicability: 1, Difference Global Comment #10 None Power Operation, Startup, Hot 2, 3 Standby, Hot Shutdown (1) a. Loss of ALL offsite and ALL (1) a. Loss of ALL offsite and ALL Difference Global Comment #9, 12, 13 None co onsite AC power to (site- onsite AC power to 1A3 ~ 11'1 specific emergency buses) and 1A4 buses for 15 for 15 minutes or longer. minutes or longer. AND AND Difference Global Comment #9 & 13 V12

b. Indicated voltage is less b. Indicated voltage is less than (site-specific bus than 105 VDC on BOTH Div voltage value) on ALL (site- 1 a*nd Div 2 125 voe buses specific Vital DC busses) for for 15 minutes or longer.

15 minutes or longer. ~ Intent and meaning of the EALs are not altered. 83

DAEC DEVIATIONS AND DIFFERENCES MATRIX APPENDIX A-ACRONYMS AND ABBREVIATIONS 84

DAEC DEVIATIONS AND DIFFERENCES MATRIX li~ff~Jf,s,¢#1~nr*,\lfitl{.? ,':;t(~'.E!:~9~PJ;~Iie'6JJ~f,rJ?::}t~jJJbF~~*tlJ~tf{Ji:}Jl~ilJ'?AE.~ ;,::ii& *C::}J1i~:~,h~ng)j,1.;, ~-,,.;;xn:J,,.:3iilf~HI. ~@J!:~c1ti~"~!lfL

            -""<<i;Y, , ,             *,;r.,"-,*'" ,. ,)y'\.4:-., ,,.,-, _                     ,,e. .    ;i;/"YL*;,~- t,"H*"' ,,.,,.>>,~,,*,4-<-'*-,r,,          ,,~ *, * , .r ,,-,,,/!.*:,,,-,, ,,d_,.,f;,_,,*,-,

AC.......Alternating Current AC ....... Alternating Current Verbatim N/A AOP ...... Abnormal Operating AOP...... Abnormal Operating Verbatim N/A Procedure Procedure APRM ... Average Power Range Difference Not used N/A Meter ATWS ... Anticipated Transient ATWS ... Anticipated Transient Verbatim N/A Without Scram Without Scram B&W.... Babcock and Wilcox Difference Not used N/A BIIT...... Boron Injection Initiating Difference Not used N/A Temperature II) z BWR.... Boiling Water Reactor BWR .... Boiling Water Reactor Verbatim N/A 0 CDE ...... Committed Dose CDE ...... Committed Dose Verbatim N/A

          ~               Equivalent                                      Equivalent w              CFR ...... Code of Federal                       CFR ...... Code of Federal                   Verbatim                                                       N/A Cl:

ca Regulations Regulations ca

          <C Q              CTMT/CNMT... Containment                                                                      Difference                       Not used                      N/A z              CSF...... Critical Safety Function                                                            Difference                       Not used                      N/A
          <C II)

CSFST... Critical Safety Function Difference Not used N/A 2 z Status Tree 0 DBA...... Design Basis Accident Difference Not used N/A Cl: u DC........ Direct Current DC ........ Direct Current Verbatim N/A

          <C I             EAL. ...... Emergency Action Level               EAL. ...... Emergency Action Level           Verbatim                                                       N/A
          <C
         ><              ECCS .... Emergency Core Cooling                 ECCS.... Emergency Core Cooling              Verbatim                                                       N/A Q               System                                           System zw C.

ECL.. ..... Emergency Classification ECL. ...... Emergency Classification Verbatim N/A C.

          <C             Level                                            Level EOF ...... Emergency Operations                  EOF ...... Emergency Operations              Verbatim                                                       N/A Facility                                         Facility EOP ...... Emergency Operating                   EOP ...... Emergency Operating               Verbatim                                                       N/A Procedure                                        Procedure EPA. ..... Environmental Protection              EPA...... Environmental Protection           Verbatim                                                       N/A Agency                                           Agency EPG ..... Emergency Procedure                    EPG ..... Emergency Procedure Verbatim                                                       N/A Guideline                                        Guideline EPIP ..... Emergency Planning                                                                 Difference                       Not used                      N/A Implementing Procedure 85

DAEC DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. '6

                                                                    . DAEC::          I . Change 1. '.            Justification I Validation#: j EPR ...... Evolutionary Power                                               Difference Not used                            N/A Reactor EPRI. .... Electric Power Research                                          Difference Not used                            N/A Institute ERG ..... Emergency Response                                                Difference Not used                            N/A Guideline FEMA ... Federal Emergency           FEMA... Federal Emergency              Verbatim                                       N/A Management Agency                    Management Agency N/A

,iJ C 0 FSAR .... Final Safety Analysis Report GE ........ General Emergency GE ........ General Emergency Difference Verbatim Not used N/A ~ V, HCTL.. .. Heat Capacity HCL....Heat C~pacity __Limit Difference Updated to reflect DAEC EOPs N/A z Temperature Limit 0 ~ HPCI. .... High Pressure Coolant HPCI. .... High Pressure Coolant Verbatim N/A $ Injection Injection LI.I a:: HSI ........ Human System Interface Difference Not used N/A cc cc IC.......... lnitiating Condition IC.......... lnitiating Condition Verbatim N/A <[ C ID ......... lnside Diameter ID ......... lnside Diameter Verbatim N/A z <[ IPEEE ... lndividual Plant Difference Not used N/A II) 2 Examination of External Events z (Generic Letter 88-20) 0 ISFSl .... lndependent Spent Fuel ISFSl .... lndependent Spent Fuel Verbatim N/A a:: u Storage Installation Storage Installation <( I <[ Keff..... Effective Neutron Keff..... Effective Neutron Verbatim N/A X Multiplication Factor Multiplication Factor 25 LCO ..... Limited Condition of LCO ..... Limited Condition of Verbatim N/A z LI.I c.. Operation Operation c.. <( LOCA... Loss of Coolant Accident LOCA... Loss of Coolant Accident Verbatim N/A MCR.... Main Control Room Difference Not used N/A MSIV... Main Steam Isolation Difference Not used N/A Valve MSL..... Main Stem Line Difference Not used

• N/A mR, mRem, mrem, mREM .... milli- mR, mRem, mrem, mREM .... milli- Verbatim N/A Roentgen Equivalent Man Roentgen Equivalent Man MW..... Megawatt MW..... Megawatt Verbatim N/A NEI.. ..... Nuclear Energy Institute NEI.. ..... Nuclear Energy Institute Verbatim N/A NPP...... Nuclear Power Plant Difference Not used N/A 86

DAEC DEVIATIONS AND DIFFERENCES MATRIX tit. Sef!J~n.

        '    *I**,*
                                                                                             "i~Hf. ;~;;tf}PAa.~~~~.t . ]~~,. ,. ;,; * :t~h~Q~~:

• J,1j{is.**, >;:;!1: .,:. ~;~i~ifi~a,fi9n *'jf.:::

Rev.:,6~. Validation'!#: 1."8 /';'NEI 99~01

                         , . *~,:.~....~. , ,,,;'\:'$_:-;..    ;- ~c,,;,. , ,:. , ::.:;}j,                                                                                                  * ),H;: 1          ' , ".;;. ~  - ,:.*r,;;/7,;,fi' NRC. .... Nuclear Regulatory                                              NRC..... Nuclear Regulatory                   Verbatim                                                                          N/A Agency                                                                    Agency NSSS.... Nuclear Steam Supply                                                                                           Difference  Not used                                                              N/A System NORAD... North American                                                   NORAD ... North American                                                                                                        N/A Aerospace Defense Command                                                 Aerospace Defense Command (NO)UE ... (Notification of) Unusual                                      NOUE ... Notification of Unusual              Difference  DAEC uses full NOUE terminology                                       N/A
 -,.J C:

Event NUMARC.. .. Nuclear Management Event NUMARC. ... Nuclear Management Verbatim N/A

 -z 0

0 CJ I ll and Resources Council OBE ..... Operating Basis Earthquake and Resources Council OBE ..... Operating Basis Earthquake Verbatim N/A

  ~              OCA ..... Owner Controlled Area                                           OCA..... Owner Controlled Area                Verbatim                                                                          N/A UJ             ODCM/ODAM .... Offsite Dose                                               ODAM ... Offsite Dose Assessment              Difference  DAEC uses ODAM                                                        N/A c:::

ca Calculation (Assessment) Manual Manual ca

  <              ORO ..... Offsite Response                                                                                              Difference  Not used                                                              N/A C                                                                                                                '

z Organization s Ill PA ......... Protected Area PA......... Protected Area Verbatim N/A

  ~              PACS .... Priority Information and                                                                                      Difference  Not used                                                              N/A z

0 Control System c::: PAG ...... Protective Action PAG ...... Protective Action Verbatim N/A u

  <C I            Guideline                                                                 Guideline
  <              PICS..... Process Information and                                                                                       Difference  Not used                                                              N/A X

c Control System z UJ PRA/PSA... Probabilistic Risk PRA/PSA. .. Probabilistic Risk Verbatim N/A Cl. Cl. Assessment/Probabilistic Safety Assessment/Probabilistic Safety

  <C Assessment                                                                Assessment PWR .... Pressurized Water Reactor                                        PWR .... Pressurized Water Reactor            Verbatim                                                                          N/A PS ......... Protection System                                                                                          Difference  Not used                                                              N/A PSIG .... Pounds per Square Inch                                          PSIG .... Pounds per Square Inch              Verbatim                                                                          N/A R.......... Roentgen                                                      R.......... Roentgen                          Verbatim                                                                          N/A RCC. ... Reactor Control Console                                                                                      , Difference  Not used                                                              N/A RCIC ... Reactor Core Isolation                                           RCIC. .. Reactor Core Isolation               Verbatim                                                                          N/A Cooling                                                                    Cooling 87

DAEC DEVIATIONS AND DIFFERENCES MATRIX Section ::I

                                    ,.,  ,,,   *..                    < ,~ y
                                                                                       ',,             ..             Justification I Va!.~dation n.J RCS ..... Reactor Coolant System     RCS ..... Reactor Coolant System       Verbatim                                          N/A Rem, rem, REM ... Roentgen           Rem, rem, REM ... Roentgen             Verbatim                                          N/A Equivalent Man                       Equivalent Man RETS .... Radiological Effluent                                             Difference    Not used                            N/A Technical Specifications RPS ...... Reactor Protection System RPS ...... Reactor Protection System   Verbatim                                          N/A

+J C 0 RPV ...... Reactor Pressure Vessel RVLIS ... Reactor Vessel Level RPV ...... Reactor Pressure Vessel Verbatim Difference Not used N/A N/A ~ Instrumentation System VI z RWCU ... Reactor Water Cleanup RWCU ... Reactor Water Cleanup Verbatim N/A 0 SAR ....... Safety Analysis Report Difference Not used N/A ~ SAS ........ Safety Automation Difference Not used N/A LI.I System a: cc SBO .......Station Blackout Difference Not used N/A cc <C SCBA .....Self-Contained Breathing SCBA ..... Self-Contained Breathing Verbatim N/A C z Apparatus Apparatus <C VI SG ..........Steam Generator Difference Not used N/A ~ N/A > SI. ..........Safety Injection Difference Not used z SICS ...... Safety Information Difference Not used N/A 0 a: u Control System <C I SPDS ..... Safety Parameter Display SPDS..... Safety Parameter Display Verbatim N/A <C System System X c SRO ....... Senior Reactor Operator Difference Not used N/A z LI.I TEDE .....Total Effective Dose TEDE.. ... Total Effective Dose Verbatim N/A Q. Q. Equivalent Equivalent <C TOAF .....Top of Active Fuel TAF .....Top of Active Fuel Difference Updated to reflect DAEC EOPs N/A TSC ........Technical Support TSC. .......Technical Support Verbatim N/A System System

                -                                    UFSAR .... Final SafetyAnalysis Difference    Used in Section 3.1                 N/A Report WOG ..... Westinghouse Owners                                               Difference    Not used                            N/A Group 88

DAEC DEVIATIONS AND DIFFERENCES MATRIX APPENDIX B - DEFINITIONS 89

DAEC DEVIATIONS AND DIFFERENCES MATRIX IC ,. NEI 99:..01Rev. 6* *'~Justification Alert: Events are in progress or have occurred Alert: Events are in progress or have occurred Verbatim None which involve an actual or potential which involve an actual or potential substantial degradation of the level of safety substantial degradation of the level of safety of the plant or a security event that involves of the plant or a security event that involves probable life threatening risk to site probable life threatening risk to site personnel or damage to site equipment personnel or damage to site equipment because of HOSTILE ACTION. Any releases are because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of expected to be limited to small fractions of the EPA PAG exposure levels. the EPA PAG exposure levels. V'I General Emergency: Events are in progress or General Emergency: Events are in progress or Verbatim None 2 0 have occurred which involve actual or have occurred which involve actual or j:: IMMINENT substantial core degradation or IMMINENT substantial core degradation or z u::: melting with potential for loss of melting with potential for loss of w C containment integrity or HOSTILE ACTION containment integrity or HOSTILE ACTION I co that results in an actual loss of physical th<!t results in an actual loss of physical X control of the facility. Releases can be control of the facility. Releases can be 2i 2 w reasonably expected to exceed EPA PAG reasonably expected to exceed EPA PAG Q.. Q.. exposure levels offsite for more than the expos1Jre levels offsite for more than the ct immediate site area. immediate site*"area. Notification of Unusual Event: Events are in Unusual Event: Events are in progress or have Difference See Global Comment #3 None progress or have occurred which indicate a occurred which indicate a potential potential degradation of the level of safety of degradation of the level of safety of the plant the plant or indicate a security threat to or indicate a security threat to facility facility protection has been initiated. No protection has been initiated. No releases of releases of radioactive material requiring radioactive material requiring offsite offsite response or monitoring are expected response or monitoring are expected unless unless further degradation of safety systems further degradation of SAFETY SYSTEMS occurs. occurs. 90

DAEC DEVIATIONS AND DIFFERENCES MATRIX

                                                                           , uDAE'e" *,: :              Gij*ang~'"
                                                                                                                    ,fJustificatlor) .*.

Site Area Emergency: Events are in progress Site Area Emergency: Events are in progress Verbatim None or have occurred which involve actual or or have occurred which involve actual or likely major failures of plant functions needed likely major failures of plant functions needed for protection of the public or HOSTILE for protection of the public or HOSTILE ACTION that results in intentional damage or ACTION that results in intentional damage or malicious acts; 1) toward site personnel or malicious acts; 1) toward site personnel or equipment that could lead to the likely failure equipment that could lead to the likely failure of or; 2) that prevent effective access to, of or; 2) that prevent effective access to, equipment needed for the protection of the equipment needed for th~ protection of the public. Any releases are not expected to public. Any releases are not expected to result in exposure levels which exceed EPA result in exposure levels which exceed EPA Vl PAG exposure levels beyond the site PAG exposure levels beyond the site 2 0 boundary. boundary. E Emergency Action Level (EAL): A pre- Emergency Action Level (EAL): A pre- Verbatim None 2 u: LLI determined, site-specific, observable determined, site-specific, observable Q threshold for an Initiating Condition that, threshold for an Initiating Condition that, I 1:1:1 when met or exceeded, places the plant in a when met or exceeded, places the plant in a >< given emergency classification level. given emergency classification level. 2i 2 LLI Emergency Classification Level (ECL): One of a Emergency Classification Level (ECL): One of a Verbatim None

a. set of names or titles established by the US set of names or titles established by the US a.

<( Nuclear Regulatory Commission (NRC) for Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions grouping off-normal events or conditions according to (1) potential or actual effects or according to (1) potential or actual effects or consequences, and (2) resulting onsite and consequences, and (2) resulting onsite and offsite response actions. The emergency offsite response actions. The emergency classification levels, in ascending order of classification levels, in ascending order of severity, are: sev~rity, are:

• Notification of Unusual Event (NOLIE)
• Notification of Unusual Event (NOLIE)
• Alert
• Alert
• Site Area Emergency (SAE)
• Site Are_a Emergency (SAE)
• General Emergency (GE)
• General Emergency (GE) 91

DAEC DEVIATIONS AND DIFFERENCES MATRIX IC l* <N E199-01 Rev .J(,. * . *: :.j DAEC

• 1 Change , .. *~ ... Justification I Validation # I Fission Product Barrier Threshold: A pre- Fission Product Barrier Threshold: A pre- Verbatim None determined, site-specific, observable determined, site-specific, 9bservable threshold indicating the loss or potential loss threshold indicating the loss or potential loss of a fission product barrier. of a fission product barrier.

Initiating Condition (IC): An event or Initiating Condition (IC): An event or Verbatim None condition that aligns with the definition of condition that aligns with the definition of one of the four emergency classification one of the four emergency classification levels by virtue of the potential or actual levels by virtue of the potential or actual effects or consequences. effects or consequences. CONFINEMENT BOUNDARY: (Insert a site- CONFINEMENT BOUNDARY: The barrier(s) Difference Removed developer notes None v, specific definition for this term.) Developer between spent fuel and the E:nvironment and added site-specific 2 0 Note - The barrier(s) between spent fuel and once the spent fuel is processed for dry language. E the environment once the spent fuel is storage. This corresponds to the pressure 2 i:i: processed for dry storage. boundary for the Dry Shielded Canister (DSC) w Q shell (including the inner bottom cover plate) I a:I base metal ang associated confinement X *. boundary welds. c 2 CONTAINMENT CLOSURE: (Insert a site- CONTAINMENT CLOSURE: Site specific Difference Removed developer notes None w Cl. Cl. specific definition for this term.) Developer procedurally defined actions taken to secure and added existing

<C    Note - The procedurally defined conditions            containment and its associated structures,                        definition from present or actions taken to secure containment                systems, and components as a functional                           EALs.

(primary or secondary for BWR) and its barrier to fission product release under associated structures, systems, and existing plant conditions. For DAEC, this is components as a functional barrier to fission considered to be Secondary Containment as product release under shutdown conditions. required by Technical Specifications. DESIGN BASIS EARTHQUAKE (DBE): A DBE is Difference Added term used in HU2 None vibratory ground motion for which certain versus use of footnotes (generally, safety-related) structures, systems, and components must be designed to remain functional.

                                                                            ,,    92

Justification ,V VaUdat~on #'

                                                                                                                                        ,,... ~ ;fp OS'-M'"

EXPLOSION: A rapid, violent and catastrophic EXPLOSION: A rapid, violent, and catastrophic Verbatim None failure of a piece of equipment due to failure of a piece of equipment due to combustion, chemical reaction or combustion, chemical reaction, or overpressurization. A release of steam (from overpressurization. A release of steam (from high energy lines or components) or an high energy lines or components) or an electrical component failure (caused by short electrical component failure (caused by short circuits, grounding, arcing, etc.) should not circuits, grounding, arcing, etc.) should not automatically be considered an explosion. automatically be considered an explosion. Such events may require a post-event Such events may require a post-event V) inspection to determine if the attributes of an inspection to determine if the attributes of an z explosion are present. explosion are present. 0 j:: z u:: FAULTED: The term applied to a steam Difference Term not used for BWRs None LI.I C generator that has a steam leak on the I r:c secondary side of sufficient size to cause an X uncontrolled drop in steam generator i5 z pressure or the steam generator to become LI.I Q. completely depressurized. Developer Note - Q. <C This term is applicable to PWRs only. FIRE: Combustion characterized by heat and FIRE: Combustion characterized by heat and Verbatim None light. Sources of smoke such as slipping drive light. Sources of smoke such as slipping drive belts or overheated electrical equipment do belts or overheated electrical equipment do not constitute FIRES. Observation offlame is not constitute FIRES. Observation of flame is preferred but is NOT required if large preferred but is NOT required if large quantities of smoke and heat are observed. quantities of smoke and heat are observed. HOSTAGE: A person(s) held as leverage HOSTAGE: A person(s) held as leverage Verbatim None against the station to ensure that demands against the station to ensure that demands will be met by the station. will be met by the station. 93

HOSTILE ACTION: An act toward a NPP or its HOSTILE ACTION: An act toward a nuclear Difference Spelled out 'NPP' in 2 None personnel that includes the use of violent power plant or its personnel that includes the places force to destroy equipment, take HOSTAGES, use of violent force to destroy equipment, and/or intimidate the licensee to achieve an take HOSTAGES, and/or intimidate the end. This includes attack by air, land, or water licensee to achieve an end. This includes using guns, explosives, PROJECTILEs, vehicles, attack by air, land, or water using guns, or other devices used to deliver destructive explosives, PROJECTILEs, vehicles, or other force. Other acts that satisfy the overall devices used to deliver destructive force. intent may be included. HOSTILE ACTION Other acts that satisfy the overall intent may should not be construed to include acts of be included. HOSTILE ACTION should not be civil disobedience or felonious acts that are construed to include acts of civil not part of a concerted attack on the NPP. disobedience or felonious acts that are not Non-terrorism-based EALs should be used to part of a concerted attack on the nuclear address such activities (i.e., this may include power plant. Non-terrorism-based EALs violent acts between individuals in the owner should be used to address such activities (i.e., controlled area). this may include violent acts between individuals in the owner controlled area). HOSTILE FORCE: One or more individuals who HOSTILE FORCE: One or more individuals who Verbatim None are engaged in a determined assault, overtly are engaged in a determined assault, overtly or by stealth and deception, equipped with or by stealth and deception, equipped with suitable weapons capable of killing, maiming, suitable weapons capable of killing, maiming, or causing destruction. or causing destruction. IMMINENT: The trajectory of events or IMMINENT: The trajectory of events or Verbatim None conditions is such that an EAL will be met conditions is such that an EAL will be met within a relatively short period of time within a relatively short period of time regardless of mitigation or corrective actions. regardless of mitigation or corrective actions. INDEPENDENT SPENT FUEL STORAGE INDEPENDENT SPENT FUEL STORAGE Verbatim None INSTALLATION (ISFSI): A complex that is INSTALLATION (ISFSI): A complex that is designed and constructed for the interim designed and constructed for the interim storage of spent nuclear fuel and other storage of spent nuclear fuel and other radioactive materials associated with spent radioactive materials associated with spent fuel storage. fuel storage. 94

DAEC DEVIATIONS AND DIFFERENCES MATRIX NORMAL LEVELS: As applied to radiological Difference Term not used in this EAL None IC/EALs, the highest reading in the past scheme twenty-four hours excluding the current peak value. OPERATING BASIS EARTHQUAKE (OBE): An Difference Added term used in HU2 None OBE is vibratory ground motion for which versus use of footnotes those features of a nuclear power plant necessary for continued operation without undue risk to the health and safety of the public will remain functional. OWNER CONTROLLED AREA: (Insert a site- OWNER CONTROLLED AREA: The site Difference Definition from developer None specific definition for this term.) Developer property owned by or otherwise under the notes used. Developer Ill z Note-This term is typically taken to mean contror of the licensee. Notes deleted. 0 the site property owned by, or otherwise " j:: z i:i: under the control of, the licensee. In some LL.I cases, it may be appropriate for a licensee to Q I define a smaller area with a perimeter closer cc to the plant Protected Area perimeter (e.g., a X i5 site with a large OCA where some portions of z LL.I the boundary may be a significant distance Cl. Cl. from the Protected Area). In these cases, <( developers should consider using the boundary defined by the Restricted or Secured Owner Controlled Area (ROCA/SOCA). The area and boundary selected for scheme use must be consistent with the description of the same area and boundary contained in the Security Plan. PROJECTILE: An object directed toward a NPP PROJECTILE: An object directed toward a Difference Spelled out 'NPP' None that could cause concern for its continued nuclear power plant that could cause concern operability, reliability, or personnel safety. for ,its continued operability, reliability, or personnel safety. 95

DAEC .DEVIATIONS AND DIFFERENCES MATRIX NEI 99-01 Rev~;6 .. Justification I VaUdation # I PROTECTED AREA: (Insert a site-specific PROTECTED AREA: The area under Difference Definition from developer None definition for this term.) Developer Note - continuous access monitoring and control, notes used. Developer This term is typically taken to mean the area and armed protection as described in the site Notes deleted. under continuous access monitoring and Security Plan. control, and armed protection as described in the site Security Plan. REFUELING PATHWAY: (Insert a site-specific REFUELING PATHWAY: The reactor refueling Difference DAEC-specific definition None definition for this term.) Developer Note - cavity, spent fuel pool, and fuel transfer supplied. Developer This description should include all the canal. Notes deleted. cavities, tubes, canals and pools through which irradiated fuel may be moved, but not including the reactor vessel. RUPTURE(D): The condition of a steam Difference Not used None generator in which primary-to-secondary V, leakage is of sufficient magnitude to require a z safety injection. Developer Note - This term 0 j:: is applicable to PWRs only. z u::: LI.I C SAFETY SYSTEM: A system required for safe SAFETY SYSTEM: A system required for safe Difference Removed developer notes None I ca plant operation, cooling down the plant plant operation, cooling down the plant and clarified last sentence. X and/or placing it in the cold shutdown and/or placing it in the cold shutdown c z LI.I condition, including the ECCS. These are condition, including the ECCS. These systems Cl. typically systems classified as safety-related. are classified as safety-related. Cl. <C Developer Note - This term may be modified to include the attributes of "safety-related" in accordance with 10 CFR 50.2 or other site-specific terminology, if desired. SECURITY CONDITION: Any Security Event as SE~URITY CONDITION: Any Security Event as Verbatim None listed in the approved security contingency listed in the approved security contingency plan that constitutes a threat/compromise to plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or site security, threat/risk to site personnel, or a potential degradation to the level of safety a potential degradation to the level of safety 96

DAEC DEVIATIONS AND DIFFERENCES MATRIX Justificatior( *

• vaUclation

                                                                                                                                               - c.,{"'": :

of the plant. A SECURITY CONDITION does of the plant. A SECURITY CONDITION does not involve a HOSTILE ACTION. not involve a HOSTILE ACTION. SITE BOUNDARY: That line beyond which the

• Difference Defined term from ODCM None land is neither owned, nor leased, nor needed for several EALs otherwise controlled by the Company. UFSAR Figure 1.2-1 identifies the DA~C SITE BOUNDARY.

UNISOLABLE: An open or breached system UNISOLABLE: An open or breached system Verbatim None line that cannot be isolated, remotely or line* that cannot be isolated, remotely or locally. locally. UNPLANNED: A parameter change or an UNPLANNED: A-parameter change or an Verbatim N/A event that is not 1) the result of an intended event that is not 1) the result of an intended evolution or 2} an expected plant response to evolution or 2) an expected plant response to a transient. The cause of the parameter a transient. The cause of the parameter 11'1 change or event may be known or unknown. change or event may be known or unknown. 2 0 E VISIBLE DAMAGE: Damage to a component or VISIBLE DAMAGE: Damage to a component or Deviation Updated to reflect V17 2 i:! structure that is readily observable without structure that is readily observable without wording and guidance of w Q measurements, testing, or analysis. The visual measurements, testing, or analysis. The visual approved EAL FAQ 2016-I a:i impact of the damage is sufficient to cause impact of the damage is sufficient to cause 02. The updated wording X concern regarding the operability or concern regarding the operability or clarifies damage 2i 2 reliability of the affected component or reliability of the affected component or assessment meriting an w Q. structure. structure. Damage resulting from an ALERT declaration as used Q. <( equipment failure and limited to the failed in ICs using this definition component (i.e., the failure did not cause (CA6 and SA9). damage to a structure or any .other equipment} is not VISIBLE DAMAGE. 97

DAEC DEVIATIONS AND DIFFERENCES MATRIX APPENDIX C - Permanently_ Defueled ICs/EALs

DAEC DEVIATIONS AND DIFFERENCES MATRIX

                                                                                          *Justification *
                                                                                                           . Validation #
                                                                                                           '   ?,~ ,. '

Appendix C - Permanently Not used at DAEC Difference Not applicable to DAEC None

>          Defueled ICs/EALs C

QJ !l

;     <t E w QJ Q. -

a I "C J u.!! '

)(

"C .... C QJ QJ GJ C CL CL <t 99

                                                          -~l ATTACHMENT 4 NEXTERA ENERGY DUANE ARNOLD, LLC DUANE ARNOLD ENERGY CENTER RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION RELATING TO LICENSE AMENDMENT REQUEST TSCR-166 UPDATED SUPPORTING TECHNICAL INFORMATION 248 pages follow

Definitions 1.1 Table 1.1-1 (page 1 of 1) MODES REACTOR MODE AVERAGE SWITCH REACTOR MODE TITLE POSITION COOLANT TEMPERATURE (°F) 1 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(bl 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. DAEC 1.1-8 Amendment 223

Development of EAL Threshold values from NEE-323-CALC-003 Calculated values are provided in Calc-003 as shown below. Values ror the RU1 Gaseous EALs were detennined and are shown below. Table 1 - Gaseous Effluent Se/points Detector RU1 Threshold Loc.ition (}.!Ci/cc) Offg3S stack Kaman 10 1.97E-01 Turbine Building Vent Kaman2 7.74E-04 Reactor Building Vent Kaman4 6.00E-04 Reactor Building Vent Kaman 6 9.60E-04 Reactor Building Vent Kaman8 9.60E-04 LLRPSF Building Vent Kaman 12 1.1 9E-03 Values for the Liqu*d Effluent RU1 EALs were determined and are shown beloW. TDble 2 - liquid Effluent Se/points Loc.ition Equipment RU1 Unusual ID Event Level C S GSW RE-4767 1.53E+03 RHRSW/ESW RE-1997 8.42E+02 RHRSW Dilution Line* RE-4268 1.06E+03 The values are rounded for ease of operator use and to provide a step-wise progression through the emergency classification levels. The resulting values used in the DAEC RU1 .1 EAL are shown in the NOUE column below: I Table R Effluent Monitor Classification Thresholds Monitor GE SAE Alert NOUE Reactor Build ing ventilatio n rad m onitor 1.l E+OO uci/cc 1.lE-01 uci/cc 1.lE-02 uci/cc 8.0 E-04 uci/cc (Kaman 3/4, 5/6, 7/8) Ill Turbine Building ventilation rad monitor

J 1.4E+OO uci/cc 1.4E-01 uci/ cc 1.4E-02 uci/cc 8 .0E-04 uci/cc 0 (Kaman 1/2)

QI Ill ro Offgas Stack rad monitor l'.) 4.5 E+03 uci/cc 4.5E+02 uci/cc 4.SE+Ol uci/cc 2.0 E-01 uci/cc (Kaman 9/10) LLRPSF rad monitor (Kaman 12)

                                                                  -          l .4E-01 uci/cc 1.4E-02 uci/cc 1.2E-03 uci/cc GSW rad monitor (RIS-4767)
                                                                  --                 -        1.7E+04 cps    1.5E+03 cps
  -0
  *-:J RHRSW & ESW rad monitor 0-

J (RM-1997)

                                                                  -                   -       1.2E+04 cps    8.4E+02 cps RHRSW & ESW Rupture Disc rad monitor (RM-4268)                                                  -                   -       1.8E+04 cps    1.0E+03 cps

CALC NO. NEE-323-CALC-003 (>> ENERCON CALCULATION COVER

     .      .                                   .                                          REV. 00
          ~cell"etice~E~e_ry proje~t. fllery fi.oy.             SHEET PAGE NO. 1 of 9 Client:        Duane Arnold Energy Center Documentation of the RU 1 Emergency Action

Title:

Levels Project Identifier: NEE-323 Item Cover Sheet Items Yes No 1 Does this calculation contain any open assumptions, including preliminary

                                                                                                              ~       D information, that require confirmation? (If YES, identify the assumptions.)

2 Does this calculation serve as an "Alternate Calculation"? (If YES, identify the design D ~ verified calculation.) Design Verified Calculation No. -- 3 Does this calculation supersede an existing Calculation? (If YES, identify the design D ~ verified calculation.) Superseded Calculation No. -- Scope of Revision: Initial Issue ' Revision Impact on Results: Initial Issue Study Calculation D Final Calculation ~ Safety-Related D Non-Safety-Related ~ (Print Name and Sign) Originator: Jay Bhatt Date: 12/12/17 Design Verifier1 (Reviewer if NSR): Ryan Skaggs Date: 12/12/17 Approver: Aaron Holloway Date: 12/12/17 Note 1: For non-safety-related calculation, design verification can be substituted by review.

CALC NO. NEE-323-CALC-003 Q ENERCON Ex~elle~nce-Eve_ry pr9ject. ~ve,y dqy. CALCULATION REVISION STATUS SHEET REV. 00 CALCULATION REVISION STATUS REVISION DATE DESCRIPTION 00 12/12/17 Initial Issue PAGE REVISION STATUS PAGE NO. REVISION PAGE NO. REVISION All 00 APPENDIX/ATTACHMENT REVISION STATUS APPENDIX NO. NO~OF REVISION ATTACHMENT NO.OF REVISION

                        ..       PAGES                   NO.          NO.            PAGES             NO.

1 4 00 2 18 00 3 9 00 Page 2 of9

a ENER*CON Ex.c~fl,;nce.~Eyery pr~,j~(. ¥V,;ry r;Joy, TABLE OF CONTENTS CALCNO. REV. NEE-323-CALC-003 00 Section Page No~

1.0 Purpose and Scope

4 2.0 Summary of Results and Conclusions 4 3.0 References 5 4.0 Assumptions 5 5.0 Design Inputs 6 6.0 Methodology 6 7.0 Calculations 8 8.0 Computer Software 9 9.0 Impact Assessment 9

                                                                                      #of List of Attachments                                                                   Pages Attachment 1 - Calculation Preparation Checklist                                 4 Attachment 2 - Gas Effluent Setpoints                                           18 Attachment 3 - Liquid Effluent Setpoints                                         9 Page 3 of9

CALC a ENERCON

      ~xo;f!~t:ce_:Ev~.'Y projec~. Every dCfY.

Documentation of RU1 Emergenc, NO. Action Levels REV. NEE-323-CALC-003 00

1.0 Purpose and Scope

The Duane Arnold Energy Center site is implementing the guidance of Revision 6 to the Document NEI 99-01, "Development of Emergency Action Levels for Non-Passive Reactors," which is the industry-developed methodology for emergency classification for the current operating fleet. Changes to the definitions of the condition for entry into the Emergency Action Level (EAL) RU1 result in the development of a new entry threshold value for this EAL. This calculation provides calculated threshold values for the following EALs (from NEI 99-01, Rev. 6). Note that NEI 99-01 designates abnormal radiological conditions as "AU," NEE has adopted the "RU" designation permitted under the guidance. (1) Reading on ANY effluent radiation monitor greater than 2 times the (site-specific effluent release controlling document) limits for 60 minutes or longer. (2) Reading on ANY effluent radiation monitor greater than 2 times the alarm setpoint established by a current radioactivity discharge permit for 60 minutes or longer. This calculation uses the latest radiation monitor setpoints to determine the resultant EAL thresholds. Page 4 of9

CALC a ENl(RCON r, ..

          ~ic,~l[e[l~e_~E~_ry projet;t. Eve:Y <J<?Y-Documentation of RU1 Emergenc)

Action Levels NO. REV. NEE-323-CALC-003 00 2.0 Summary of Results and Conclusions Values for the RU1 Gaseous EALs were determined and are shown below. Table 1 - Gaseous Effluent Setpoints Detector RU1 Threshold Location (1,1Ci/cc) L~ .* ' \Hffg~~/;St~~~:; :.*. :. . Kaman 10*** **

                                                                                  ,,  " ... ~ \ ~ ,' ""::;  '

Turbine Building Vent Kaman 2 7.74E-04

      !i: Re~~for ~~11d1r19V~nft< **

Reactor Building Vent Kaman 6 9.60E-04 1 . .

         .~e,~l~~..~.~.,i~.~'-'-,_i'-'*~-..;:..~*-0...V-'-e._11=\=---'-'--~--'-'~-'--"-*             ~**.-'-"<=-:(;;_:-:---"-'*_,_\_*. _*.*_,.:.*:"'-::;:_:.§)~.6~9'-'--~=-;0_,1.o.....;*-'-'-.:,=\.,z~*,<:--"-'*

,;11 LLRPSF Building Vent Kaman 12 1.19E-03 Values for the Liquid Effluent RU1 EALs were determined and are shown below.

• Table 2 - Liquid Effluent Setpoints Location Equipment RU1 Unusual ID Event Level ,,

(cps)

             !~$W . . .::*> ...                      . ____ . . >* ...... * .. *<---* ,<.: ...~9-4761,*... . ::::J'._qg~tP~.-~;J RHRSW/ESW                                                                            RE-1997                                              8.42E+02 l R}IRSW t>_ilutiqn Line~; . :                                          ..              8!=-42_68:                            . .            1.061;+03                         .*< .!
*RE-4268 was previously known as the RHRSW Rupture Disk 3.0 References 3.1 NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors." November 2012.

3.2 DAEC Offsite Dose Assessment Manual (ODAM), Rev. 37. 3.3 Plant Chemistry Procedure PCP 8.3, Alarm Setpoints and Background Determination for KAMAN Normal Range Monitors. 3.4 Plant Chemistry Procedure PCP 8.7, Alarm Setpoints For Liquid Rad Monitors. 3.5 Technical Specifications, Section 5.5.4, Radioactive Effluent Controls Program. 3.6 DAEC Emergency Plan, Section 'I', Rev. 27 Page 5 of9

CALC a ENERCON

               ~xct;ll~!1~~-Ev~ry project. Eve:1 qqy.

Documentation of RU1 Emergenc, Action Levels NO. REV. NEE-323-CALC-003 00 4.0 Assumptions It is assumed that the current setpoint for the Kaman 4 monitor is 3.00E-04 µCi/cc. The latest setpoint determination received is from 3/4/2016 which exceeds the 18 month frequency specified by PCP 8.3. 5.0 Design Inputs 5.1 The setpoint determinations from Attachment 2 and Attachment 3, represent the latest responses at the associated gaseous and liquid effluent monitors. While the three most recent surveillances for each monitor are included for information, only the latest setpoint is used to determine the EAL threshold. It should be noted that the "RM" equipment designations are equivalent to the "RE" equipment IDs. 5.2 The gaseous effluent equipment ID number, monitor common name and range are taken from DAEC Emergency Plan Section "I" and ODAM Figure 3-1, and are presented in Table 3. Table 3- Gaseous Effluent Design Inputs Monitor Monitor Location Equipment ID Range f .Tu~~~:.ic::,~< *-*** L ,* * . * - ".: _.,, * ~. ' r:~g:; Common Name ft . T . !s:;':~:~ .<:

                                                    'KAMAN 3/4 ~* ** * ~ . RE-7645 RE-7644 . .                                      .-~ ..**.. ' *>: '.

(1,1Ci/cc)

                                                                                                                                               *~;:~; ~~;:: ** :

[ Reijctor Building *' .KAMAN.5/6 . ., RE-7647, RE~-7646: * . 1E~07- :1E+OS , 1~:.j>r: . Veri,t;.*_*,,,,,,~'..**.~*,,*;,i~K,A.ry!AN71a.* ,;i\*j.*:.:;;'.,.,'*,:8(76;4~:~BJ:~7p48i,,,,,,,,),,,.*.:),i.(,;;,;'_..*.. ',*L.:;.* . ,.::,.*:l LLRPSF Building KAMAN 12 RE-8801 1E 3E-01 Vent 5.3 The liquid effluent equipment ID number, and rarige are taken from ODAM Table 1-2, and are presented in Table 4. Table 4 - Uquid Effluent Design Inputs Location Equipment Monitor Range ID (cps)

                                                .. **GsW I    < - -*.-
                                                                                                'RE-4767                     1E-01 -. 1E+06
                    '~-..--' _.~ ~ ~ - - = - - ~ - - ~ - ~ - - - ' - - - ' , , - - ' - - - " - " ' _ , , _ _ _ _ _ - " - " - ' - ' - - - ' - ~ - ' - ' - -

RHRSW/ESW RE-1997 1E 1E+06 Page 6 of9

CALC NEE-323-CALC-003 Documentation of RU1 Emergenc1 NO. Action Levels ------ - - - - - - - - - - 1 REV. 00 6.0 Methodology The alarm setpoint of a radioactive noble gas effluent monitor is calculated on the basis of whole body dose equivalent rate offsite of 500 mrem/yr per the ODAM. The alarm setpoint for liquid radwaste effluent line provides automatic isolation when 1O times the water effluent concentration listed in 10 CFR 20 Appendix B, Table 2, is being exceeded in the unrestricted area per the ODAM. These setpoints are in accordance with Technical Specifications limits specified in 5.5.4b and 5.5.4g. This calculation considers historical setpoint determination for gaseous release (PCP 8.3) and liquid effluent (PCP 8.7). The latest three setpoints for each monitor were reviewed. Due to the high variance for some of the monitors, the latest alarm setpoint is used to determine the EAL thresholds. Page 7 of9

CALC NEE-323-CALC-003 0 - ENERCON

       ¢~c:~!(~t;~~¥"!~..'Y P..rojet;~. _Eve!Y cf~~

Documentation of RU1 Emergenc~ Action Levels NO. REV. 00 7 .0 Calculation 7 .1 Gaseous Setpoints Plant Chemistry Procedure PCP 8.3 is used by Chemistry Technicians to calculate setpoints for building vent KAMAN monitors at least once every 18 months. The three latest setpoint determinations for each location are shown in Attachment 2 for information. It should be noted that where the original PCP 8.3 setpoint calculation sheet is unavailable, the value is taken from the associated monitor calibration procedure. Thresholds corresponding to the latest setpoints are calculated and presented here. For example the latest PCP 8.3 setpoint for Offgas stack is 9.84E-02

            µCi/cc. This value is d_oubled to 1.97E-01 µCi/cc to correspond to the RU1 threshold. The remaining threshold values are shown in Table 5.

Table 5 - Gaseous Effluent Setpoints and Thresholds Detector Latest PCP 8.3 RU1 Threshold Location Setpoint {µCi/cc) {µCi/cc) Turbine Building Kaman 2 3.87E-04 7.74E-04 Vent Reactor Building Kaman 6 4.80E-04 9.60E-04 Vent LLRPSF Building Kaman 12 5.95E-04 1.19E-03 Vent 7.2 Liquid Setpoints As a result of variability in the isotopic mix of reactor water, backgrmmd radiation levels and detector efficiencies, the calculated liquid effluent setpoints will fluctuate over time. Chemistry Technicians perform effluent liquid radiation monitor setpoint calculations at least once per 18 months with guidance provided by Plant Page 8 of9

CALC NEE-323-CALC-003 QENERCON Documentation of RU1 Emergenc~ NO.

     §xcell~_nce~Ev~ry proje~t. f.vety c!'!Y.                       Action Levels REV.        00 Chemistry Procedure PCP 8.7. The three latest setpoint determinations for each location are shown in Attachment 3. It should be noted that where the original PCP 8.7 setpoint calculation sheet is unavailable, the value is taken from the associated monitor calibration procedure.

Thresholds corresponding to the latest setpoints are calculated and presented here. For example the latest PCP 8. 7 setpoint for the RHRSW Dilution Line is 421 cps. This value is doubled to 842 cps to correspond to the RU1 threshold. The remaining threshold values are shown in Table 6. Table 6 - Liquid Effluent Setpoints and Thresholds Location Latest PCP 8. 7 RU1 Threshold Setpoint (cps) (cps) i .:

               ', GSW ,                                                 7;65E+02 .             1.53E+O~

i~,-** - * - * ~ - - - ~ - - - - ~ ~ ~ - ~ ~ - - - - ~ - - - RHRSW/ESW 4.21E+02 8.42E+02 l .. : ......;:. **,:'.' . -; *,*.'*** t RHRSW . Dilution . i Line , *

• 5:301=+02 \ '1~06E+03 . :*

(, .. _. 8.0 Computer Software None. 9.0 Impact Assessment This calculation is based on "realistic" conditions for the purpose of declaring EALs, rather than typical conservative "bounding" type design basis analyses. The calculation documents the order of magnitude setpoints to assist Operations and Emergency Response personnel in determining an unusual event in accordance with NEI 99-01 Rev. 6. Page 9 of9

CALC a ENERCON

               ~XCf:l{~nc~:-E~.'Y proj"e.t;t. E~e!Y d(!Y.

Attachment 1 CALCULATION PREPARATION CHECKLIST NO. REV. NEE-323-CALC-003 00 CHECKLIST ITEMS1 YES NO N/A GENERAL REQUIREMENTS

1. If the calculation is being performed to a client procedure, is the procedure being used the latest revision? D D 181 The calculation is being prepared to ENERCON's procedures.

2. Are the proper forms being used and are they the latest revision? 181 D D

3. Have the appropriate client review forms/checklists been completed? D D 181 The calculation is being prepared to ENERCON's procedures.

4. Are all pages properly identified with a calculation number, calculation revision and page number con~istent with the requirements of the client's procedure? 181 D D

5. Is all information legible and reproducible? 181 D D

6. Is the calculation presented in a logical and orderly manner? 181 D D

7. Is there an existing calculation that should be revised or voided? D 181 D This is a new calculation to support implementing NEI 99-01 Rev. 6

8. Is it possible to alter an existing calculation instead of preparing a new calculation for this situation? D 181 D

9. If an existing calculation is being used for design inputs, are the key design inputs, assumptions and engineering judgments used in that calculation valid and do they 181 D D applyJo the calculation revision being performed.

10. Is the format of the calculation consistent with applicable procedures and expectations? 181 D D 11 . .. Were design inpuUoutput documents properly updated to reference this calculation? D D 181

12. Can the calcul/:Jtion logic, methodology and presentation be properly understood without referring back to the originator for clarification? 181 D D OBJECTIVE AND SCOPE

13. Does the calculation provide a clear concise statement of the problem and objective of the calculation? 181 D D

14. Does the calculation provide a clear statement of quality classification? 181 D D

15. Is the reason for performing and the end use of the calculation understood? 181 D D

16. Does the calculation provide the basis for information found in the plant's license basis? 181 D D

17. If so, is this documented in the calculation?

D D 181

18. Does the calculation provide the basis for information found in the plant's design basis documentation? D 181 D Page 1 of4

CALC Attachment 1 NEE-323-CALC-003 F.:ll ENERCON

             ~XCf!ll~[Jce-Eve_ry proje.ct. Every day.

CALCULATION PREPARATION NO. CHECKLIST REV. 00 CHECKLIST ITEMS1 YES NO N/A

19. If so, is this documented in the calculation? D D f8l

20. Does the calculation otherwise support information found in the plant's design basis documentation? D f8l D

21. If so, is this documented in the calculation? D D f8l

22. Has the appropriate design or license basis documentation been revised, or has the change notice or change request documents being prepared for submittal? D D f8l DESIGN INPUTS

23. Are design inputs clearly identified? f8l D D

24. Are design inputs retrievable or have they been added as attachments? f8l D D

25. If Attachments are used as design inputs or assumptions are the Attachments traceable and verifiable? f8l D D

26. Are design inputs clearly distinguished from assumptions? f8l D D

27. Does the calculation rely on Attachments for design inputs or assumptions? If yes, are the attachments properly referenced in the calculation? f8l D D

28. Are input sources (including industry codes and standards) appropriately selected and are they consistent with the quality classification and objective of the calculation? f8l D D

                                                                                                                  *:l

29. Are input sources (including industry codes and standards) consistent with the plant's design and license basis? f8l D D

30. If applicable, do design inputs adequately address actual plant conditions? f8l D D

31. Are input values reasonable and correctly applied? f8l D D

32. Are design input sources approved? f8l D D

33. Does the calculation reference the latest revision of the design input source? f8l D D

34. Were all applicable plant operating modes considered? f8l D D ASSUMPTIONS

35. Are assumptions reasonable/appropriate to the objective? D D f8l

36. Is adequate justification/basis for all assumptions provided? D D f8l

37. Are any engineering judgments used? D f8l D

38. Are engineering judgments clearly identified as such? D D f8l

39. If engineering judgments are utilized as design inputs, are they reasonable and can they be quantified or substantiated by reference to site or industry standards, D D f8l engineering principles, physical laws or other appropriate criteria?

Page 2 of4

CALC Attachment 1 NEE-323-CALC-003 C ENERCON Exc_el/ence,-Eve_ry r.roject. Every 1ar. CALCULATION PREPARATION CHECKLIST NO. REV. 00 CHECKLIST ITEMS1 YES NO N/A METHODOLOGY

40. Is the methodology used in the calculation described or implied in the plant's licensing basis? D ~ D

41. If the methodology used differs from that described in the plant's licensing basis, has the appropriate license document change notice been initiated? D D ~

42. Is the methodology used consistent with the stated objective? ~ D D

43. Is the methodology used appropriate when considering the quality classification of the calculation and intended use of the results? ~ D D BODY OF CALCULATION

44. Are equations used in the calculation consistent with recognized engineering practice and the plant's design and license basis? ~ D D

45. Is there reasonable justification provided for the use of equations not in common use?

D D ~

46. Are the mathematical operations performed properly and documented in a logical fashion? ~ D D

47. Is the math performed correctly? ~ D D

48. Have adjustment factors, uncertainties and empirical correlations used in the analysis been correctly applied? ~ D D

49. Has proper consideration been given to results that may be overly sensitive to very small changes in input? ~ D D SOFTWARE/COMPUTER CODES

50. Are computer codes or software languages used in the preparation of the calculation?

D ~ D

51. Have the requirements of CSP 3.09 for use of computer codes or software languages, including verification of accuracy and applicability been met? D D ~

52. Are the codes properly identified along with source vendor, organization, and revision level? D D ~

53. Is the computer code applicable for the analysis being performed? D .D ~

54. If applicable, does the computer model adequately consider actual plant conditions? D D ~

55. Are the inputs to the computer code clearly identified and consistent with the inputs and assumptions documented in the calculation? D D ~

56. Is the computer output clearly identified? D D ~

57. Does the computer output clearly identify the appropriate units? D D ~

Page 3 of4

CALC Attachment 1 NEE-323-CALC-003 0 ENERCC>N

            ~xc~f[~[!qe-Eve_ry project. Evefy cJ.ay.

CALCULATION PREPARATION CHECKLIST NO. REV. 00 CHECKLIST ITEMS1 I YES I NO I N/A

58. Are the computer outputs reasonable when compared to the inputs and what was expected? I D I D I 181

59. Was the computer output reviewed for ERROR or WARNING messages that could invalidate the results? I D I D I 181 RES ULTS AND CONCLUSIONS I I I

60. Is adequate acceptance criteria specified? I 181 I D I D

61. Are the stated acceptance criteria consistent with the purpose of the calculation, and intended use? I 181 I D I D

62. Are the stated acceptance criteria consistent with the plant's design basis, applicable licensing commitments and industry codes, and standards? I 181 I D I D

63. Do the calculation results and conclusions meet the stated acceptance criteria? I 181 I D I D

64. Are the results represented in the proper units with an appropriate tolerance, if applicable? I 181 I D I D

65. Are the calculation results and conclusions reasonable when considered against the stated inputs and objectives? I 181 I D I D

66. Is sufficient conservatism applied to the outputs and conclusions? I 181 I D I D

67. Do the calculation results and conclusions affect any other calculations? I D I 181 I D

68. If so, have the affected calculations been revised? I D I D I 181

69. Does the calculation contain any conceptual, unconfirmed or open assumptions requiring later confirmation? I D I 181 I D

70. If so, are they properly identified? I D I D I 181 DESIGN REVIEW I I I

71. Have alternate calculation methods been used to verify calculation results? I D I D I 181 No, a Design Review was performed.

Note:

1. Where required, provide clarification/justification for answers to the questions in the space provided below each question. An explanation is required for any questions answered as "No' or "N/A".

Originator: Jay Bhatt 12/12/17 Print Name and Sign Date Page 4 of4

                                                                                                                              ** . Attactiment.2 -

L,L,i o*

• YCJ-lf') 3Y7 fj NEE-323-CALC-003 Page tot 18
• PLANT CHl::NJl~rt~Y-P~Q.CEDURES 32QO M,\NLiA~** PCP 8.3 ALARIUI SETPOINTS AND .BACKGFiOUND Rev. 33 DETERMINATIO.N FOR KAMAN NORMAL Page 14of 14 RANGE MONITOR$ . .

ATTACHMENT 2 . KAMAN OFFGAS STACK GASEOUS DETECTOR HI HI.SETPOINT

1. $ample W. Kio .* CAAR.A1A-£ .. 2. Sample No. .* /'?* :S:S<c,o.

3. Sample Date .

• 8-">6.;n 4. Sa~nme
• D8s<f . . _*. - -.-.--....;.c.;..c::5,.;;.MWT-'--,..,,..,,.,...0.,...t,,

6. CounlDate ... e;,3c-r, _ -l q _ .... 7. CquntT1me O"'llT . . ...

• e?t>~n

8. Monitor Reading (µCi/cc~ . . .!?>> e ;c, ., . ?, 7o e*i 9. Pr'pcess FIO\Y Rate (CFM) S517 A~ ;?;o, oco

10. Sample Volume (riiL) ~ sve ~ .c~ .

Flow meter 10 #..;,_*:...Le.:..-._7..;;.?-...;.4:.:....*_ _..;._..;,____ Cal Due Date . .f o:-C..-:1 ') . 11 12 13 Isotope . ** µCj(nil. Dose FactorStack ki . rnrem sec

                                                                                               *

• _yr~µCi_ 1<1 xbFSi Xe 133 4;09E-5
• Kf85in .. : 1.81E:4 1-*!.!!Kr:...::8~8_ _.J--.;..:...;.._ _-1-1:.:.;:9::..:1~E~~3......-'--'-----,'---'--'---1----,-...,.....,.i 14. Bkg = i11strumerit background Xe .135 2.84E-4 8kg = 2, 1$' ,~., *µCi/cc Kr87 -* .* .. - .. 6:97E-4 .: .

  *.._*-~Xe:::..*.!!1.3~8-*..........* ~----'--'---l-'1.:.:..0:::.,Be.:E:..:~3'---------,--:.--'-i----,"-,--'---'l *

• Tliese dose 'faciors are from ODAM:

Xe 135m. .3.39E-4. stack release ata distance of 1260 - Xe 133m 3.61E-5 metersNNWofDA~C . , Ar41-* .. - ; 1.32E-3 . v£_ , * - 1-!-!N:..;;1~3.;.;.. .;_..:.:..~..;.,-:4.....

l.;..t>_.___ ..,.

1 ....;.;;i..,.'-*.:.;1.0.:;;:8::=E:...:-3=-.-

                                                                                                       .* ...;.__.;.;_....;..._        __,.+-.I'-.sz=.'""e--..,..,0,,,..-, ..~ Nbitrtirily .set equal to xe~138 1*tc Lki                   = .J,.':J\e"'                                                           ,I                                   15a. Li          1k.

• DFSi) JSZ.-e- 10 E ki _.* = C ms ) = t.'i/ e*"

                                                                                                                                                                        * * 'l ;Z_f> _i.v *e.

15b.

• E (ki
• DFSi) ( #15a. )  ;.n.(:-10 *

    ** 16: _limit~ L = -*..-*.
                                              * {06 x                                     L .ki' *
                                                                                  * . * . .. . . = ( # * °) x 1:06                             .      . .*           - .* .

(Th/3 L_es$ of t#'i 51> OR 34:36)

                                        *.          F                         L (k. DFS.)                                                  9
                           . =L =

l,.imit 1.06 '(. "".1'2.. q. o ) = ~-8Ye~-z.: / ( ro,-CC{} ) .

       *n            Hi HiALA.RM                = A x ( ins                              > =*(1 :o)(                    't .~~e .::z"'          >=

_Til8 ~adioactive gas flow corresponding to the HI Hi setpoint:

• • . / r~rtormefby: .*-. :ef?#~- -.

bate: ;5,-.,-:3q..,,,,7_ .*.

. ** lrjdependent Verification by; *                                             ~-&k~~-0~)~)~::t\*                                                                                  ***. Date:8*3Q~)1 114.

Page 2 of 18 Attachment 2 NEE-323-CALC-003 PCP 8.3 ALARM SETPOINTS AND BACKGROUND Rev.33 DETERMINATION FOR KAMAN NORMAL Page 14 of 14 RANGE MONITORS ATTACHMENT 2 KAMAN OFFGAS STACK GASEOUS DETECTOR HI HI SETPOINT

1. Sample LO. t( / 0 ,' C~ar *M61.,f 2. Sample No. } lo - };;/U) . .

3. Sample Date ;J --d:{..;:... f (g

• 4. Sample Ti.me .. ) 5?. 0 . 5. ~WT / C//f

6. Count Date :2 - :J.[o - Ilo 7. Count Time } '? ? ~ *

8. Monitor Reading (µCi/cc) G,& / e - 7 9. Process Flow Rate (CFM) ) O OOO I

10. Sample Volume (mL) '15J ODD ill L Flow meter ID #_=l-___,_?-"';J.'-'-q__.___ _ _ __ Cal Due Date ID - G, -17 11 12 13 Isotope* µCi/ml Dose Factor Stack k; mrem sec yr- µCi

    '* Xe 133                                                   4.09E-5                                                   -....
    'i-:K:.;:.r-=8=5m~*-~-=.JL.:<..J;b.,_,__~~1~.8~1=E4--"----'-----l_,_""'-='"'"'-':-.Y~

i-,:K:.;:.r-=8.;:;.8_ _+-=...-::,,,..,..,.--::d---c1""".9;-:,1-=-E-'""'3_ _ _"'----,--t--,,..--=,.,.--,,--:c::-1 14. Bkg = instrument background Xe 135 2.84E4 i-:K:.::r:...8:..;;7~____::~~;,,,=-.=:=....,,,4--6-.9,-7.,;-E--4--'---,,-------hil:-'=~:.....:.,,r:,"-

                                                                                                                        ,         Bkg =    I.Otoe   -fo' µCi/cc Xe 138                                               1.08E-3

• These dose factors are frolll ODAM:

Xe 135m 3.39E4

• stack release at a distance of 1260

       ~~~~~~=----,~~~------+.~=-:--;-.:;~, meters Xe 133m '                                            3.61E-5
                                                                                                                                    ' NNW of DAEC Ar 41                                                1,32E-3                                                   ,

x (The Less of#15b OR 3436)

                                                                                                     )     ~     1).-'.l~ -If; Limit= L = - --*- ~ - ( / 000 1 06
                                                                                            ) =         ~()L        I efr /, ofo e --1 "

(/f){JOO )

17. HiHiALARM = Ax( #16) =(1.0)( Lo6e-/ )= __ J.o<oe-{ .µCi/cc"-,

The r'aqioactive gas flow corresponding to the HI Hi setpoint: Perto,medby µ~/~Da!ea-?.HEi' --

                         * * * * ' *.~/? ?~

lndepend.ent Verification b y . ~ ~ ~ * ~

                                                                                                             * * .*                     . .Date:Z.
                                                                                                                                             .      . 2G ..Uc..

75 ff¥

                                                                                                       ;;;.;m                   *       '*

iiii L 4#

NEE-323-CALC-003 Attachment 2 Page 3 of 18 SURVEILLANCE TEST PROCEDURE STP NS791013 DAEC DUANE ARNOLD ENERGY CENTER TITLE: K10 CALIBRATION Page Rev. 10 of 68 17 Prerequisites Performance Date: i fi.. Af(t.. Z.0 I t..\ INITIALS 6.0 PREREQUISITES 6.1 Make a copy of the EMS database display. 6.2 From the Chemistry Supervisor or designee, obtain and record the following alarm setpoints. (Values will be used to confirm AS FOUND data.) 6.2.1 HI 8. Go ~. lo µCi/cc ~ (CHEM) 6.2.2 HIHI ;J.,_ (;t..\ ~ - l µCi/cc 4':::\ (CHEM) 6.3 From the Chemistry Supervisor or designee, obtain and record the desired New HI alarm setpoint. (Value will be used for the AS LEFT setpoint.) 6.3.1 Desired HI ijJ60 [ . . . 6 µCi/cc ~ (CHEM) 6.4 Verify Sr-90 0.09 µCi source (UID #687) is available for use. .J~ (CHEM) 6.5 Verify the KAMAN/EMS 'IDT:time and the HPGe System Computer time are .;(~~ within +/- 30 seconds. *- * (CHEM) NOTE .,tc...., When Kaman point sources are decay corrected, decay is to be from the date marked on the source to the test date.. 6.6 pecay correct permissible range (8.5E4 - 9.0E4 cpm) for UID #687 and ¥c:-, record below. (CHEM) PERMISSIBLE DECAY CORRECTED RANGE:

                                                                                                      ~

1. 80 f~ cpm to lf .r ' " ~ cpm (IV) it*,

NEE-323-CALC-003 LJ 04 7 clO(~, Attachment 2

                                                                                      <£                                      Page4 of 18 PLANT CHEMISTRY PROCEDURES 3200 MANUAL                                                         PCP 8.3 ALARM SETPOINTS AND BACKGROUND                                                                  Rev.33 DETERMINATION FOR KAMAN NORMAL                                                                  Page 13 of 14 RANGE MONITORS ATTACHMENT 1 VENT MONITORS GASEOUS DETECTOR HI HI SETPOINT

1. Sample 1.D. J<..-.J_, C.V\o..rv-.-....ar 2. Sample No ) 7- ;t--14 (r.

3. Sample Date s*

• e;. *I-, 4. Sample Time i.>'12..- t'[l O 5: MWT _._iq_._,_\.,_\_ __

6. Count Date _ _,'5,__:"'-e;;_ n - ' - - - - - - - - 7. Count Time -c-: -::Y'"':\.,,_.[_ _ _ _ _ _ _ __

8. MonitorReading(µCi/cc) ~ ;-?o<;:;_**7 9. ProcessFlowRate(CFM) -r.;ioc,o

10. Sample Volume (ml) .:.../7 (,n O 11 12 13 <

Isotope k; Dose Factor Vent Product

                  *;Ci/ml                          DFV;                                  Flow Meter ID#...:L=--_?,,::..c=.....CX-1------

mrem m3 k; x DFV; yr µCi Cal Due Date: ( 0 - l.){1 -{ ] Xe 133 294 Kr85m 1.17E3 Kr88 1.47E4 14. 8kg = ln~trument_!?a..round Xe 135 1.81E3 8kg = I, 1fo ~ - µCi/cc Kr87 5.92E3 Xe 138 8.83E3 15. X/Q = 4.3 x 10-e sec/m~ Xe 135m 3.12E3 (atmospheric dispersion) Xe 133m 2.51E2 Ar41 8.84E3 N 13 LJ. "ll f - 1..1 8.83E3 ** Ll.l{i '.S ** Arbitrarily set equal to Xe-138 16.:: k; = Ll.1ll:::.-':'\' 16a. L (k; DFV,) = I..\.\~ E-., L k, y;ll f'Cl 13 E-i ( #16 ) I 16b. = = c,,= L(k,

• DFV,) ( #16a ) '-\.\.~ f* ~

17. Limit= L =

1.06 X L ki 1.06 x (The lesser of #16b OR 1.81 E-4) = (F)(X IQ) L (ki

• DFVi) ( #9 )( #15 Limit= l = l.0 6 ( I. iJ E-L\ ) = 3, 81 ( .,. L( '\.

( 'llO\Jl) )( '-!.~ i: -b ) 18.HiHiALARM=Ax(#17 )=(1.0)( 3.~1 E:-L\ )= 3.£] E°':\;.iCi/cc " The radioactive gas flow corresponding to the Hi Hi setpoint: Performed by: _ _~-~,;;;..;__:__ _ _ _ _ _ _ _ _ _ _Date: oS-o.(' - ( J

                                                   ~                                                          r.*      h     ,,

Independent Verification b y : * - - - ~ ~ " ' - - ' " ' - - - ' - - - - - - - - - - - D a t e :  :; * ::) -* J 76

NEE-323-CALC-003 Attachment 2 Page 5 of 18 PLANT CHEMISTRY PROCEDURES 3200 MANUAL PCP 8.3 ALARM SETPOINTS AND BACKGROUND Rev.33 DETERMINATION FOR KAMAN NORMAL Page 13 of 14 RANGE MONITORS ATTACHMENT 1 VENT MONITORS GASEOUS DETECTOR HI HI SETPOINT

1. Sample 1.D. KaoogY) 2 2. Sample No. _ __.l...:G.._-...,'3"""Lf:.......,_(_----=,-----

3_ Sample Date '2.- S-- IG 4. Sample Time JJ: I'} 5. MWT J4H

6. Count Date --=2.:a..-_5'---..,_l~ - - - - - - - - 7. Count Time -:--=~j\~**___ l'""'--:1=---------

8. Monitor Reading (µCi/cc) j.G,S e - :is 9. Process Flow Rate (CFM} -......,-7....:..=2~,c.=-G=-C=-----

10. Sample Volume (ml} 45: cco 11 12 13 '

Isotope k, Dose Factor Vent Product

                      µCi/ml                          DFV;                                        Flow Meter ID#                    L('2. 9 mrem yr m:

uCi k, X DFV; Cal Due Date: IC

• 11 Xe 133 /1'41-C, iA,...11:....i 294 '""l<" * .I 1./'.__ I Kr85m 1.17E3 Kr88 1.47E4 14. 8kg = Instrument background Xe 135 1.81E3 Bkg = 4,, 3S e- 1 µCi/cc Kr87 5.92E3 Xe 138 8.83E3 15. X/Q = 4.3 x 10-Q sec/ma Xe 135m 3.12E3 (atmospheric dispersion)

Xe 133m 2.51E2 Ar41 . I/ 8.84E3 'L.t N 13 ,.05',.,. --i 8.83E3 ** (i, .7.3 ~ -5 ** Arbitrarily set equal to Xe-138 '16. ~ k, = ,.05.~ -'i 16a. I: (k; DFV1) = ll,.L3~-s " Ik #16 ) 7 .c.S" c:t <j = -4

                                                                                           \.\~e 1Gb.             '

I:(k,

• DFV,) #16a )

                                                     =     G, . l.3 l"- S" 1.06             X       L'.- -
                                            - -(ki      ki
                                                            --                       1.06

17. Limit= L = x (The  !~r@R 1.81 E-4) =

(F)(X IQ) L

• DFVi ) ( #9 )( #15 Limit= L =

1.06 ( -1.\ 3.'6] e-\.\

                                              )( '-/ .3 ,q c, - If.)        I , \ '3 e.        )   =

18. HiHiALARM =Ax( #17 ) =(1.0){ 3-l51et# )= 3.'87e-'t µCi/cc The radioactive gas flow corresponding to the Hi H1 setpoint:

Date: _ _ 2_-_5'_--_I_,_ Independent Verification by:_ _ _ _ _ _ _ _-+(--::~~i;;;......;,,_.c..:....;:_L_,4-- _ _ _Date:_ _2._--_'r_l _<-_ 98

NEE-323-CALC-003 Attachment 2 Page 6 of 18 PLANT CHEMISTRY PROCEDURES 3200 MANUAL PCP8.3 ALARM SETPOINTS AND BACKGROUND Rev.33 DETERMINATION FOR KAMAN NORMAL Page 13 of 14 RANGE MONITORS ATTACHMENT 1 VENT MONITORS GASEOUS DETECTOR HI HI SETPOINT

1. Samplel.D. ....L,s=.;..'"'".......,~~;i.,...___ _ _ _ _ 2. Sample No. /4-J.fefr'~

3. SampleDate =s="&::*l~ " 4. SampleTime Jz.:11_ - 13;03 5.MWT 1

                                                                                                                                                  !1o7

6. Count Date S .. tr*,~ 7. CountTime /'.) 2 '1

8. Monitor Reading (µCi/cc) 2 .. J{( e, .. , 9. Process Flow Rate (CFM) ]d. O o o o.

1 Sample Volume (ml) '1<f yo D 11 12 13 Isotope ki Dose Factor Vent Product

                         µCi/ml                                     DFV1 m3 Flow Meter ID#          L 7I 5 mrmn yr                        µCi k1 x DFV1 Cal Due Date:           7 .. 2. 2 - / (:..

Xe 133 1110 .-1,l! 294 NONF.. Kr85m 1.17E3 I Kr88 1.47E4 I 14. Bkg = Instrument back~und Xe 135 1.81E3 I Bkg = /,. /0 l!...' µCi/cc Kr87 5.92E3 I Xe 138 8.83E3 15. X/Q = 4.3 x 10"" sec/ma Xe 135m 3.12E3 (atmospheric dispersion) Xe 133m 2.51E2 I Ar41 fVQ,.Jf. 8.84E3 .vo.,.,,e. N 13 7, &'I e. *'\ 8.83E3 ** c;,.~1 ** ,.. ** Arbitrarily set equal to Xe-138 ,. 1s. r kt = '1,\'t.fc... , 16a. !: (kt DFV1) = '.,i;t_.s 1sb. r ki ( #16 ) 1. ~,*" = J,l'Sc.*"

            'l(ki

• DFV,)

                                   =

( #16a )

                                                                   =         ,.*u.~*<

17. Limit= L =

1.06 X r ki 1.06 x (The lesser of#16b OR 1.81E-4) - (F)(XIQ) r (ki

• DFVi) ( #9 )( #15 ) ,

Limit=L =

• I.0 6

4./ ~ "' ( l,l'~t*'i ) = *,, ~1c.-"'I ( ,~ooC )(

• e- ) *

18. HiHiALARM =Ax( #17 ) =(1.0)( J,i'1t*c.f )= 3,V"/£*"' µCi/cc The radioactive gas flow corresponding to the Hi Hi setpoint:

Performed by:_ _ _ i_<_L_\\_V\-+-________Date: S'- &°"" I '1 ( Independent Verification by:_

                                                   ---.,     A.::

___..__,.~

                                                                         ~/\A - .
                                                               .........~~  r '~~

_ _ _ _ _ _ _ _ _Date:

                                                                                                                                 ,:;      ~ lti,
                                                                                                                                 -;:J .... 0' n

NEE-323-CALC-003 Attachment 2 Page 7 of 18 PCP 8.3 ALARM SETPOINTS A.ND BACKGROUND Rev.*33 PETERMIN.A.TION FOR *~MAN NORMAL Page 13 of 14 RAN~E MONITORS ATTACHMENT 1 VENT MONITORS GASEOUS .DETECTOR HI HI SETPOINT . .

1. $ample Lt:>. -=-~:...:..;::;,::...:._'-'----'--""""-'-_..:. 2. Sample No_. ..*. /&, ~ 155] .

3. Sample Date -........,..,._...,.-,--~ '4. SampleTirile * * / }k/ 7 * . * * *

• 5. M.WT 1 Jq I I .

6. Count Date 8 ..Monitor Reading (µCi/cc) - -.. . . . .~~~-***-* *_ *l._* ~-r~~- ~~r~6w

10. ~ample Volume (ml)

R_. aie cctkf{q ~* ~- f/00 000 '. 11 . ., 12

• 13 * . .* . * *. * * . **

Isotope k; Dose Factor Verit Product

                                 µCiimL                                   *oi=v1                                                  Flow Meier           m# L 17 -a °t.

mrem ~ k; X DFV1

                                                            . yr .                             µCi               :

Cal[)ue Date: /tJf

                                                                                                                                                                 ,,rr Gl/7       . *. .

Xe 133 . i\n,\, .~ u . , ~' '\ 294 troNe... *.1,~ .. -=~* rJ Kr85m U7E3 *. Kr88 1A7E4 14: B.k.g =ln~trument b ....,groi.Jn.d .. Xe 135 1.81E3 8kg = (oi(qc:.,- * µCi/cc Kr87 > 5.92E3 .

 .Xe 138

• 8,83E3 15. X/Q = 4.3 x 10"' sec/rn 3 Xe 135m 3..12E3.*.* (atmospherjc dispersion) 2:51E2

 .Xe 133m Ar41.*.**.

N 13 G,, Jl i e-:-'I: 8.84E3 8.83E3

                                                                       ...       QI I
                                                                                                              ,:;-:~VIP~ ** Ai'bitrariJYset equal to Xe-138 16.}:k; =         l11.1q e.,-q
                                               .        16a.

• E(k; DFV1) =
• 11 1./ 7~-~

16b. *. I ki

                      ..*.* .. *. * . =
                                        .             c #16 )             =            t,Jqe-r:r ;_                        l. J3e-~

. I(k1 ~ DFV1) ( #16a ) s.~'?er5'"'---:

17; LimiJ = L = I.0 6 .X L . ki' 1.0,6,

                                                            -'---'-'----'---'-- ~'---'----'-'--- x(Thelesserof#16bOR 1.81E-4) = *

(F)(XIQ) .I (ki *DFVij  :( #~ )( #15 )

       *. Li~M             =(J_,/:)Di)O                     1.;t'1;?e.-(,, pc /,I ~e""/) ~                                                    '!,,oOe -~

18. Hi Hi ALARM :;: A X ( #.17 ) := (1.0) ( *31eo. e -"1 ) = ::,.0De ~~ pCi/cc

    . The !lldioactive gai; flow corresponding to the H.i Hi setpoiht
          . *. . ... [m1 .* .~ ~VI ~ ..* (,i,,')                                                                                       .                         /

PerlQnned by pwt'tvfL . . ** . ... *. . .*. *. *. . ' *. Date'  ?, *J{-'J:J . lndep~ndent Vei'ific~tion'by: . .

                                                              ~l3
                                                                                             -~"2  .*     ,*     -   .,  .
                                                                                                                                                '*. . Date:        3- Y- ~ .

14

NEE-323-CALC-003 Attachment 2 Page 8 of 18 PLANT CHEMISTRY PROCEDURES 3200 MANUAL PCP8.3

          ~ALARM SETPOINTS AND BACKGROUND                                                                    Rev.33 DETERMINATION FOR KAMAN NORMAL                                                                    Page 13 of 14 RANGE MONITORS ATTACHMENT 1 Isotope                     ki               Dose Factor Vent            Product
                     µCi/ml                             DFV1
                                             !!l!fil!!           m3        k1 X DFV1 yr               µCi Xe 133 Kr85m 294 1.17E3                               ~-

Kr8S ~ 1.47E4 ~ 14. Bkg = lnstruw~~ack~und

                     \'<:>               1.81E3                               ~                   8kg = :J"- '/ 5 v -         µCi/cc Xe 135 Kr87 Xe 138 5.92E3 8.83E3                               R.            15. XJQ = 4.3 x 10-IS sec/m:$
                     ,H                  3.'12E3                              ~                   (atmospheric dispersion) * **

Xe 135m Xe 133m It,...... 2.51E2 ~ Ar41 ,:4 ~ 8.84E3 .,,..:,_ N 13 16.Z:ki =

                    .'~\
                       ... IJl/1.

8.83E3 **

                                         'I &ii. L (k, DFV;)  =       .vn     ~-
                                                                              ~-            ** Arbitrarily set equal to Xe-138 16 b.      :E ki                  = (   #16       )   =       JJ/t-     ===

L(ki

• DFV I ) ( #l 6a ) ---,SW

17. Limit=L=

1.06 x L ki 1.06 x(Thelesserof#16bOR 1.81E-4) = (F)(X IQ} L (ki

• DFVi) ( #9 )( #15 )

Limit=L = ( 'fq,d(){J 1 06

                                                .)(    f.'3f-~) ( /.fl(G-f)                 =

1~ f;,()t3-f

18. Hi Hi ALARM =.Ax { #17 ) = (1.0)( f. ~0 £-lf )= 'fr 8()tf-'/µci/cc The radioactive gas flow corresponding to the Hi Hi setpoint:

Performed by: -?ltt6"7ff' Date: / 2 ~ tf- / f Independent Verification by: ~9'.YY\. ~ Date: (J.-'1 . . 1{

NEE-323-CALC-003 Attachment 2 Page 9 of 18 PLANT CHEMISTRY PROCEDURES 3200 MANUAL

• PCP 8.3 ALARM SETPOINTS AND BACKGROUND Rev.31 DETERMINATION FOR KAMAN NORMAL Page 13 of 14 RANGE MONITORS ATTACHMENT 1 VENT MONITORS GASEOUS DETECTOR HI HI SETPOINT

1. Sample I.D. /<-'I --1 's  !.'5'1 ~tJ<f- I 2. Sample No. I 3 - I 1'5'7

3. Sample Date 3-/-I J 4. Sample Time /~Jo 5. MWT' J"J/u

6. Count Date -:J-1-/3 7-.-C-o-un-t~Tr~m~e-------- 1338

8. Monitor Reading (µCi/cc) /. ?G 1; - 8 9, Process Flow Rate (CFM) 1O. Sample Volume (ml) 11 12 13 Isotope kt Dose Factor Vent Product

                         µCi/ml                        DFV1                                         FlowMeterlD# L 7t;..o m3           kt x DFV; ID!fil!l yr                    µCi                             CaI Due Date:        S- 7-/  Y Xe 133              /U"ON'v!.           294                                     /.)CJ,,.; fr Kr85m              ~Oe,,,>i'1Fi e.o     1.17E3                                .J.:~"il:F-rc*.f Kr88                    I              1.47E4                                                     14. 8kg = Instrument background Xe 135                                  1.81E3                                                           8kg = 5,"'k :z. C-,        µCi/cc Kr87                                   5.92E3
 -Xe 138                                 8.83E3                                                     15. XIQ = 4.3 x 10.e sec/m3 Xe 135m                                 3.12E3                                                           (atmospheric dispersion)*

Xe 133m 2.51E2 Ar41 8.84E3 N 13 V 8.83E3 ** * ** Arbitrarily set equal to Xe-138

16. :t 1<; ::: AJ/l'r 16a. :t Cki DFV1) = 11)/4

( #16 ) /;//;9-- A/6:- 16b. 'L k1 = = =

         °L(kr

• DFV1) ( #16a ) IV/r 17, Limit= L =

1.06 (F)(X IQ) X

                                            'L
                                                   'L ki' (ki

• DFVi) (

1.06

                                                                                  #9 )( #15             )

x (The lesser of#16b 08=

                                                                                                                        -y Limit= L =

1.06 ( / ~, E.

                                                                                     -'-4
                                                                                                 ) =        '-I. lo ii               \

( ti 3£>00 )( tj, ?,rt,::i'- ) ,(>

                                                                                                               - <.f 0 _~
                                                                                              ) = __'-1_._~_

18. Hi Hi ALARM = A x ( #17 ) = (1.0) ( 4. '!PE '-i The radioactive gas flow corresponding to the Hi Hi setpoint:

_ _ __,uCi/cc

                                                                                                                                   '\

19. Q = 472 (A * #17)#9 **

Q : 472 (1.0) ( '-/, U, E, If ) ( q !, DOC> ) \. a= ..!l, I I ~ t' W µCi/sec z..,11 E:"' v..'-~/'Se..c... P.erformed by: '7d;r-- ~ Date:~l-13 Independent Verification by:--_ _..... &_'--'---k-t....__A-__--------------*Date: s-,-r~

NEE-323-CALC-003 Attachment 2 Page 10 of 18 PLANT CHEMISTRY PROCEDURES 3200 MANUAL PCP 8.3 ALARM SETPOINTS AND BACKGROUND Rev.33 DETERMINATION FOR KAMAN NORMAL Page 13 of 14 RANGE MONITORS ATTACHMENT 1 VENT MONITORS GASEOUS DETECTOR HI HI SETPOINT

1. Sample I.D. K-*4' CIA<<~* vv'°',,.. 2. Sarrtple No. _l_lc_~_~_ 0_~_L_l_l___..,..__ _

3. Sample Date 11-'30 *14' 4. Sample Time I? I~* 5. MWT .........1q...,_,_1.._(_ __

6. Count Date ___;l:...:.t_-_.*:3c...:o:...;,.-.!..:lb:::..___ _ _ _ _ _ 7. Count Time _._.i'5,_p""---7'--,,_ _ _ _ _ _ __

8. Monitor Reading (µCi/cc) ;2 ~ E.--, 9. Process Flow Rate (CFM) --'~'"""3,,,...0_C_'_(._,_ _ __

10. Sample Volume (ml) 4~000 11 12 13 Isotope k, Dose Factor Vent Product

                           ;..Ci/ml                                      DFV;                                     Flow Meter ID# L * -!~           'l LU!sl.!!l                 m3          k, x DFV, yr                    µC1                           Cal Due Date: ID- (., - I t Xe 133          ,:,/1.,.-.,,_ ';J:~ ":, J294                                               r(r,'"1.~,i-,f; IJ Kr85m                                            1.17E3 Kr88                                             1.47E4                                                         14. Bkg = Instrument background Xe 135                                           1.81E3                                                               Bkg = ,.(e'"7~ ,                       µCi/cc Kr87                                             5.92E3 Xe 138                                           8.83E3                                                         1 5. X/Q = 4.3 x 10,t; sec/ma Xe 135m                                          3.12E3                                                 ...           (atmospheric dispersion)

Xe 133m 2.51E2 Ar41 N 13 ~., 8.84E3 8.83E3 ** . .* Arbitrarily set equal to Xe-138 .. 16. ~ k; = !VI.A- 16a. 'I: (k1 DFVi\ = A.I?'A.,. 1Gb. :r k.  :: ( #16 )

                                                                       =

1:(k,. OFV;} ( #16a )

17. Limit= L =

1.06 X :E ki 1.06 x (The lesser of #16b OR ~ = (F)(X /~) I: (ki

• DFVi) #9 )( #] 5 )

1.06 LI ) Limit = L = ( ernooc,

                                                               )( ,._, ,"2 ,;;
                                                                   "'l   .)=

_ 1 4'

                                                                                     ) (  I . CC t E.

• I =

18. Hi Hi ALARM = A x ( #17 ) = (1.0) ( 4. ~o {~...X--_;,,._______ Date: / f I 4, 1ndependent Verification by:._..:;:0---:;._

                                                        -*'°/ -   _ _ _ _ _ _ _ _ _ _ _ _ _ Date:._ _ _ _ __

II- >D*-/l-77

NEE-323-CALC-003 Attachment 2 Page 11 of 18 PLANT CHEMISTRY PROCEDURES 3200 MANUAL PCP 8.3 ALARM SETPOINTS AND BACKGROUND Rev.33 DETERMINATION FOR KAMAN NORMAL Page 13 of 14 RANGE MONITORS ATTACHMENT 1 VENT MONITORS GASEOUS DETECTOR HI HI SETPOINT

1. Sample I.D. N~10r \ t,t:F\ .\.s-lo 2. Sample No. ll:, - ti) °d.S

3. Sample Date \..\ - \ \p- \$ 4. Sample Time ,....--:,llo=C()"'-"";,,,,......----,..,.....,~ s.MVvr _\"-'9'"""l.,_I_ __

6. Count Date 4- l\o- \S' 7. Count Time __,_Ib.=..,...14--'------=----=----

8. Monitor Reading (µCi/cc) l.10 E- f 9. Process Flow Rate (CFM) _.,..,.._,Gr,_,,~~*=O=D,,..0"'-----

1o. Sample Volume (ml) ~.rj,.S c 4 11 12 13 Isotope ki Dose Factor Vent Product

                      µCi/ml                          DFV1                                      Flow Meter ID#_               __..;...L_,_a_°t__,___

mrrun m3 k, x DFV1 yr µCi Cal Due Date: ___l'-D_-_to_-_\f_ Xe 133 t-1~ Jlt>,Jt~ {294 roou~~ltl Kr85m 1.17E3 Kr88 1.47E4 14. Bkg = Instrument background Xe 135 1.81E3 Bkg = Ci/cc Kr87 5.92E3 Xe 138 8.83E3 15. X/Q = 4.3 x 10~ sec/m~ Xe 135m 3.12E3 (atmospheric dispersion) Xe 133m 2.51E2 Ar41 8.84E3 N 13 ,,/ 8.83E3 "* \ / ... Arbitrarily set equal to Xe-138 16.!:k; = t,..:)IA 16a. !: (krDFV1) = ~\.l; 16 b. I k1 = ( #16 ) = = I(k.

• OF\/,)

17. Umit=L=

1.06 (F)(X IQ) ( #16a X L

                                                  )

I; (ki

• DF\/i) ki

( 1.06

                                                                            ~9 )( #_IS               )

x(Thelesserof#16b0

                                                                                                                                    . e  1.81E-4 =

6 Limit=l= I.0 . (UHE::.-t..\) = l..\:~OE:.-Y

                       <<\3;oo0              )< '-\.'3i;;. - L:>  )

18. HiHiALARM =Ax( #17 ) =(1.0)( \.\,tgOt:-L\ )= l..\S~Dt.-4 µCi/cc The radioactive gas flow corresponding to the Hi Hi seteoint:

                                                                                                                                                        *I
                         ~=1f                                                                          L\- lCo- \~
                                                  ~r Performed by:                                                                           Date:

Independent Verifical;on bye Date: f-/p-/'§

NEE-323-CALC-003 Attachment 2 Page 12 of 18 SURVEILLANCE TEST PROCEDURE STP NS791009 DAEC DUAlfE ARNOLD ENERGY CENTER TITLE: K6 CALIBRATION Page Rev. 64 of 66 14 Performance Date: q_c.,,,~ INITIALS 7 .15.6 Record the following AS LEFT values:

a. AS LEFT HI-HI ALARM SETPOINT (from Step 7.15.4):

                                         ~ ~M,*l3 L( /l0,-t(_ '1, ~u            e ~;:   µCi/cc

b. AS LEFT HI ALARM SETPOINT (from Prerequisite 6.3.1):

                                                            µCi/cc

c. AS LEFT BACKGROUND (from Step 7.13.50 or7.14.49): &

7 ~ 'is'"~ E.- 7 µCi/cc 7.15.7 At the Kaman EMS IDT, verify the following has been correctly entered into the EMS database:

a. HI-HI alarm setpoint (from Step 7.15.6.a) Y<.r/ ~

b. HI alarm setpoint (from Step 7.15.6.b) &~

c. Background concentration {from Step 7.15.6.c)

                                                                                                         >fp{ .,

7.15.8 Update database values on the status boar/and in Labstats.v' WI... 7.15.9 Attach completed setpoint calculation documentation (Step 7.15.4) to this STP. (PRINT / SIGN)

                  /.......o.,.. . .'t :CsA.A.c..S     I ~ "'-A~J~,.              '\.-t.-l3 I :,DC\..

R,"1,.,a_,J P~ I ~ 4-£-G 1)12-Performed by:

                                                        ------                       Date:     Time:       lnit.

L{ 0 '-l :J Lt (p l (p Page 13 of 18 NEE-323-CALC-003 Attachment 2

                                                                                                                                                                                                                             .      PCP 8.3 ALARM SETPOINTS AND BACKGROUND                                                                                                                                                                           Rev.33 DETERMINATION FOR KAMAN NORMAL RANGE MONITORS '                                                                                                                              ' ..                                                       Page 13 of 14 ATTACHMENT 1
                          . VENT MONITORS GASEOUS DETE:CTOR HI HI SET:POINT

1. SamplHD: )(;& Chw; Q'.lqy 2. San:iple No. I },-}~4~

3. Sample Date * ** 4-1 2 :.cff

• 4. SarripleTime )2~9 5. MWT __,_lc;J-'-'o=a:...--_ __;_

6, Count Date __ t.;...-)u:;2"""'.:.....,17'----~--'--- 7. Count Time *.. --,--'-'12.:::::S=<..i.!f:-=*_*-'----...------'--

8. Monitor Rea9ing ~µ~i/cc) 'ij 10 e-:':l.. 9. Process Flow Rate (CFM) *, 93000 , .

10. Sample VohJrrie (ml) Lj7soo *.

11 12 13 . Isotope( ki Dose Factor Vent Product pCilriiL . DFVi. Flow Meter ID# l...,...:7z9 mrem m3 k, X DFVi yr

                                                                                                                             µCi                                                                   Cal Due Date:                     j() .- eo-1  :J Xe 133          *11'1&/\~

• I - ~ 294* .... tri<..,1 ... * ~L.'Lf-,;;I

                                                  ', c Kf85m                  '
                                                             .1'17E3                                                                                                       I Kr88                                            1A7E4.                                                                                                                                14. Bkg =Instrument background Xe 135                                          1.81E3.                                                                                                                                             8kg =

• Y.N7 e. - 7 µCi/cc Kr87 5,92E3 *,*,

Xe138 8.83E3 15. X/Q ::;4.3 x 10'1, sec/m~ Xe 135ni. 3.12E3 .(~tme>spheric dispersion) xe 133m : * .2.51E2' Ar41 8'84E3

            *N 13                        .. 1.

8.83E3 *** ,,

                                                                                                                                                                         --.,,                     ** Arbitrarily set equal to Xe-138 16; :E k1 .=t ' .*NJ&.                          16a.

• 1: (k1 DFV1) = r-Jf1L 1: ki . . . ( #16 ) =

16b. . ,.

             .

• L(k,
• DFV, )

( *. #16a . ) . = 17 L*1mft-L-1.06 x*

• t ki I .06

                                                                                                                                                                                                       .). x(Thelesser.of#16b0R 1.81E-4) =
                 .       "" - (F)(X IQ)                                       L           (l<i

• pFVL) ( #9 )( #15 1.06 -"( L{ *. -ti Umit=L = .

( 93;uCO

                                                                                     )(~:

3~*io-~) ( 1/:Zle * ) = <*JSOe .*.

18. Hi Hi ALARM =Ax ( #17 ) =(1,0) ( l.f .fa e~L\ >=

• Lf, 156e.;.q 11ci1~ ~

The radioactive gas flow correspcinaing to the Hi Hi setpoir:it Perforin<:d by:._::.&:.£<' ~~Ll*~~* ~B:::.'*~*::7-,5,'

                                                         * . J . ,, , . . . . * .                   ::..:::** *~:L::::*.*
                                                                                              *~wki..a::*    ..c....*.            ~:...:...*<,....:~'----'--: Date:

lndep~~dent v~rificati?n by:.;;_**_..:....,...;.......:...**. ..;..* .....:...**...:...*.~~"""

                                                                                                                . ..';;;;..*.::.*.*c;;..**.*-"-.***-:...:...***.!+-"'-.*"-~...:...**.*:,,,..,.,..         * ;.;;.***..,.._._o;;ite:
                                                                                                                                                                                . ..,.,.,..;...***.;,-:,....*                        4. *"' \ ~ '.': l 1 80

NEE-323-CALC-003 Attachment 2 Page 14 of 18 ALARM SETPOINTS AND BACKGROUND Rev.33 DETERMINATION FOR KAMAN NORMAL Page 13 of 14 RANGE MONITORS ATTACHMENT 1 VENT MONITORS GASEOUS DETECTOR HI HI SETPOINT

1. Sampla l.D. i<'amttl\ ~ 2. Sample No. _/_S""_... _5'?__3_'1'_______

3. Sample Date -,.D-.,_--,,f,....¥/$,,,__ _4-.-Sa_m_p-le-T-im_e_ 1/30 5. MWT

6. Count Date 10 --.z-,.,- 7. Count Time // ~9

8. Monitor Reading (µCi/cc) 4S°9e-* 31 9. Process Flow Ra-te_(_C~FM-)"'-'-----.ZJ....-='3,-0:"X>--,,,__----

1O. Sample Volume (ml) '<£ e 'f 11 12 13 Isotope k; Dose Factor Vent Product

                     µCi/ml                            DFV1                                     Flow Meter ID#                   I... 7 24 mrem              ma            ki x DFV1 yr              µCi                            Cal Due Date: /0-6 -/ 7 Xe 133         {I/ 1'7,! 'Zl.tt,rt, 294                                    /l'l'Ae.

Kr85m 1.17E3 ~-n-rih-Kr88 1.47E4 I 14. 8kg = Instrument background Xe 135 1.81E3 8kg = :.~3e - 7 µCi/cc Kr87 5.92E3 Xe 138 8.83E3 15. X/Q = 4.3 x 10-ll sec/m 3 Xe135m 3.12E3 " (atmospheric dispersion) Xe 133m , 2.51E2 Ar41 r:' 8.84E3 \' N 13 'f, (,,2. e.-r

• 8.83E3 ** £/,6ie. ...... ** Arbitrarily set equal to Xe-138

16. 1: k1 = 1/.1., *,)(! ~ Co) 16a. :E (k1 DFV1) = ' ( ~ ()-S. u,

_c, 'f 16 b. L 1<; = c #16 ) = 11.IR'le = / ,(3e-L(k,

• DFV,) ( #16a ) f.r;g~-5 1.06 x L ki 1.06

17. Limit= L= x(Thelesserof#16bOR 1.B1E-4) =

(F)(X IQ) L (ki

• DFVi) ( #9 )( #15 ) ,o~cti..\~

Limit= L = ('r,ocO 1.06

                                                  )(¥,3,e.-C,      )

( ,.(~e.-f ) = -.if °j,ub e -'f

18. Hi Hi ALARM = A x ( #17 ) = (1.0)( -~.bO-e. -'1 )= .,,.40(!.. µCi/cc The radioactive gas flow corresponding to the Hi Hi setpoint:

Performed by: ~ - - Date: Independent Verification by:_ _ _ _ _ _~ s-* ,~ A--" __ _ _ _ _ _ _ _ _Date: tv - c>- I,') Q._ ,c~ 72

NEE-323-CALC-003 Attachment 2 Page 15 of 18 PLANT CHEMISTRY PROCEDURES 3200 MANUAL PCP 8.3 ALARM SETPOINTS AND BACKGROUND Rev.33 DETERMINATION FOR KAMAN NORMAL Page 13 of 14 RANGE MONITORS ATTACHMENT 1 VENT MONITORS GASEOUS DETECTOR HI HI SETPOINT

1. Samplel.D.

3. Sample Date K"'""""' 1

                      -j----z.-~---,'1---4-.-Sa_m_p-le-T-im_e_

2. SampleNo.

I IO3 I'.\- '2.LfC,C

                                                                                              --'----5.-MWf----,-,,..0-.-/-

6. Count Date 4 . . 7....'1 ~ 14 7. Count Time 1/....

_ _.._l..... 0_..,....., _ _ _ _ __

8. Monitor Reading (µCi/cc) /. 2 2. 1= - , 9. Process Flow Rate (CFM) 4 3 coo v:-

10. Sample Volume (ml) 48 C:,oO 11 12 13 Isotope ki Dose Factor Vent Product

                     µCi/ml .                        DFV1                                 Flow Meter ID#          l. 7 6 0 mrem                 ~          k1 X DFV1 yr                µCi                        Cal Due Date:          S-7-14 Xe 133         nJl>IV E..          294                              ',t.1(>µ£_

Kr65m £0eA7jF,"e.. 1.17E3 IT,p.,,7,'7"JeJ Kr68

• 1.47E4 I 14. 8kg = Instrument background Xe 135 . 1.81E3 8kg = µ{A µCi/cc Kr67 5.92E3 Xe 138 B.83E3 16. X/Q = 4.3 x 10-1> sec/m;,

Xe 135m 3.12E3 (atmospheric dispersion) Xe 133m 2.51E2 Ar41 8.64E3 I N 13 8.83E3 ** ** Arbitrarily set equal to Xe-} 38

16. L ki = ,v/+ ,. ' 16a. E (ki DFV1) = ,,v/4-16b. L ki- = ( #16 ) = /VIA-. ';v/4-I(ki

• DFV1) { #16a )

17. Limit=L=

1.06 x

                                           ~

L ki 1.06 x(Thelesserof#16b0R 1.81E-4) = (F)(XIQ) Lo, (ki

• DFVi) ( #9 )( #15 )

1.06 _I( . -'( limit= L = ( 'I 30cJG> )( ¥.3 ft'~ ) ( I " \ e. IQ,

                                                                                       ) - -

LI -a O

l'J:/

e v'

                                                                                        '              -'(

18. HiHiALARM =Ax( #17 ) =*(1.0)( ¥.&e-'1 )= &/.Boe µCi/cc/

The radioactive gas flow corresponding to the Hi Hi setpoint: Perfonned by: .6?6?~~r::i:l*"f:--.-- Date: Independent Verification by:_ _~ . c . . : ; ;...a='-"""~--------D,ate:

                                                          ,_..;;.._*                                              'f* 'Z-f-lE- ;

NEE-323-CALC-003

                         ~~o~ LtOL\5 rJ...c')9{-,o 1 Attachment 2                                                               Page 16 of 18 PLANT Cijj:MIST~Y PRPCEDl,l~E~ ~2~P MANUA._                                                                          PCP 8.3 ALARM SETPOINTS AND BACKGROUND                                                                                       Rev.~3 DETERMINATION FOR KAMAN NORMAL                                                                                       Page 13 of 14 RANGE MONITORS                 *.

ATTACHMENT 1 VENT MONITORS .GASEOUS DETECTOR HI i-U SETPOINT

1. Sample 1.0. )<..,. I :;;,_ C.hi~"'"""":c"".. 2. Sample No. _.1_7_-_J.._.-~--*_°l_(p"----'----,--..,.;-.....

3. Sample Date 4 -~ * \ 7 4. Sample Time l 3;2..:l.. 5. MWT 19 \ D

6. Count Date y ..,.::i..o -17 7: CountTimei ,--___.?....,3_.</:_.6""--,----'--'-----

8. MonitorReading (µCi/cc) g, 4, 1 ** ?;" 9. Process Flow Rate (CFM) .....,...:I~'5_o_D_D_ _ _ __

10. Sample Volume (ml) 4<1r-, 600 11 12 13 '

Isotope k1 Dose F~ctqr Vent Product

                         µCi/ml                                DFV1                                           Flow Meter ID#           L---1 ~c\

mrem ma k1 x DFV1 yr

                                                                           µCi                                Cal Due Date:.           I (:)

• le -* I!

Xe 133 *.No>> .. ~+- 294 N o,vi ::c.t....~-P. Kr85m 1.17E3 .. Kr88 1.47E4 14. 8kg = Instrument background Xe 135 1.81E3 8kg = 1-\,C~G.--r µCi/cc . "Kr87 5.92E3 Xe 138 8.83E3 15. X/Q = 4.3 x 10-ti seclm" Xe135m 3.12E3 (atmospheric dispersion) Xe .133m 2.51E2 Ar41 8.84E3

 .N 13                      \:,               8.83E3 ..                                          ?            ** Arbltr.:irily set equal to Xe-138 16;2'.ki     = ... N1A:                 *1sa. L (k; DFV;) =

• NY.Ai 16b.

L l<i

                                   =* (

( #.16 )

                                                              -        N/1+/-                -        N{A:

L(k, ~DFV,) #16a )

17. Limit= L=

1.06 (F)(XIQ) X L L (ki

• DFVi)

                                                                   .ki

( 1.06

                                                                                           #9 )( #JS             )

x(The lessei of #16b o@ = Limit =L =

                                                  . 1.06
                                . * . *..* .*. *. * **          .. ..

• 1 _ .

( r- G I .~ T -c u ) = ( -,5Do<::;, )( Lj,',f.- \.f' ) . 1

18. HiHi_ALARM=Ax( #17 )=(1.0)( &$.c106.-L( )= 5','}';,<3-L{µCi/cc The radioactive gas flow COffesponding to the Hi Hi setpoint:

Performed by:._ _--'----'-----'--'~;_--'*=-**.'-. *-=-*---'--'----'.Date: L[ - d-..0 ~ 17 Independent Verification by:--,-7!~-- A_//..-_.*----.-~,...,...~----.----Date: . /' *

                                                                                                                              *.. w~2..o--1~          . *.. *.*

94

Attachment 2 Page 17of 18

                           , :et.Ar-i:tCHEivllST~Y-,P~dceriURE~ .3200 tviANuAi:J ,;, *. /} :/ PCP 8;3
                          ' .:*, :*. . . *.: :, : ;_ ,,. :** **.<... *-:* <'\: :\:: :.:* .:.,/.: *},::.f.::,,t, .::-"\.~_);::_i:,: \: *,-, ;. *:~*'.;t: ~ ~ ::,..:-:t)~;; :~ ~,,' '. :*. *\~':)::,* :....fy> *(, ..

1 1 ** ALAR.IVI SETPOINTS AND BACKGROUND Rev.33 DETERMINA:rioN FOR KAMAN NORMAL Page 13 of_ 14 RANGE nilqNiTORS ATTACHMENT 1 VENT MONITORS GASEOUS DETECTOR HI HI SETPOINT

1. SamphelD.

                                        .                        ~-.         d
                                                                . . :_ ,,,_,~,/I .
                                                                                        ;z.;.   * ..

• 2. Sample No.

3. $ample Date IJ--/1*/5':' >'4.*samplEi!Time

• 13~'1}? 5.MWT. * /"l/0 * *

6. Gaunt OatE:i 11~19-/~ 7. Count Timl:l __ * * * /:&; t!- 7 *_. ._ .. *.

8. Monitor Reading (µCi/cc)* * 'S",-:; e'.":'7 9. Process Flow Rate (GfM)  ?,~>

10. SampleVolurne(r'nL). -.-=3,...,q=_*..,.2/r-e..,.,4,.,..,_, *..- - - -

--'----'-----'....:.1~1~----'-'-'-~-~-~--1~2-'--'---~---'1=3-*~ * ' Isotope

                                                         µCi/ml k;                       Dose Factor Vent inrerri DFV; . . .

m 3 Product k1 x DFV1 Fiow MeterlD# l ? I? . ___..__--.-'-~ yr µCi Cal Due Date: . '1. ;...{ Z--f7 l--'.X-e~1~3~3---'-:-..4: ~~.. -:..-... - -..1-2....:.9_4..,_.- ..- *.------:...,-___..1-~----i Kr85m * \"4~ >* *1.17E3 * \~ l--'-:..;;Kr:..a8a:::8_ _4--\->,.~C\"----'----l--'1,;,.;..4..:..7..;;;E;.;_4.._.*------'-1--..,,.\~..:.-..;.._-'-I 14, 8kg ='.lnstriJmentbackground Xe ..135 -\:~, 1,81E3 . \ .>. Elkg = 5, -J1e -'1 _. µCi/cc Kr at . . .'\17'..c - .5.92E3 .* \ ~

• _x....

1-X:..;;e;:;..**..:.,13c.::8_.- - - * ~,,,,A;...-'1-'8;:;..:8.;..;3'-E'-"-3_._.- - - - - - - *\_,,~--......,* 15. X/Q =4._3 x 10.i; sec/m~ 1-X:..:.e;:;..*.,;...;13=5=m'"""*,......,+----**.-*"'"':_.*'\:...,~..a-1r--f-,3""";"""12"-"E_3_.--~---i--....,_,..,_.\.,...,."1;'"h--1* .... (atr'nospheiic dispersion) Xei:133rri '<'\~ 2.51E2 .<. . . : '\.3, Ar41 . °'I 8.84E3 > _.** '- . !-!.:.N.:.:..1;::.3---'---'--*.<.:.'J...:..'11-",e,-*..--,--+-s=.:**~83c.::E:.::.3_*_*.._* .* .._.---,--_,.;....___;,,_*-'-:'lf.,._i.:..:ee=*...:*S'-. ---1. *' Arbitraiily set equal to Xe-1.38 1s::Ek;=* 5,.i.//,/,i_,/1 16a. L(k;DFVi)= i,/."1,1>-l".l

          .                      I:            k1                          ( #1.6 ) .                                ').-'11?" 7                                              I,13a -'1 16b.                       .                      *. *.         =     .                           =            4,ic-~-,                               =

Z:(k1. DFV1) ( #16a . ) 1.06 x _L . ki 1.06

17. limit= L = ) x (The less~r of #16b OR 1.81E,4) =

(F)(XJQ) '. .*}:* (ki. -~ DFVi) * ( #9 )( #15 4 Limit= L =*. * . > J.06 ._ . _.. -*(:. *_* ( /J"3e f )  ;,_ - }~ ](e.,;.. ( ,~ )( t/..,3x1..-;, .-*. )

'                                                         . ' ... ' ' ..* .. _ ' ' .                                         ' .. . .;,'1                                                     ' *,         ,* - ..,

18. Hi Hi.ALARM= Ax( #17 ) =(1.0)( '7~1/e *_._ ).= '3,'1le *,.** µCiicc

          . The. radioactive 'ga:s flow ~~responding to the! Hi Hi setpoint:

I Perfoi:medby: .. - ~ Date: //.d?--/S-lndep(;in9erit \/erificatie>ll by:.,..,:~__,.~:_\_._--*_-_ill-=>-',,,..---..,..*--.**- 7

                                                                                                                                    .. _**,....kt~i---...-*_i*..,..,___._-_-.--_*-~---,-Date: ... _:                             //-19-J ';::,

162

NEE-323-CALC-003 Attachment 2 Page 18 of 18 PLANT CHEMISTRY PROCEDURES 3200 MANUAL PCP 8.3 ALARM SETPOINTS AND BACKGROUND Rev.33 DETERMINATION FOR KAMAN NORMAL Page 13of 14 RANGE MONITORS ATTACHMENT 1 VENT MONITORS GASEOUS DETECTOR HI HI SETPOINT

1. Sample 1.0. - ~ : . n . . ~ ~ ~ ~ : _ __ _ _ 2. Sample No. \ "f- 4 3&'"

3. Sample Date - - - - - - - - 5. MWT I ~t.J t

6. Count Date -~---..-,_......__,.-,-----:---- 7. CountTime __,~l l.~)'--'1._-=-"_ _ _ _ _ __

8. Monitor Reading (µCifcc) -1 9. Process Flow Rate (CFM) _ _7_,Sia..,o-"_,,o _ _ __

10. Sample Volume (ml) '1i:l:Oo 11 12 13 Isotope k1 Dose Factor Vent Product

                        µCifml                      DFV1 m3 Flow Meter    ID#_L-=-7-~_C,_-,-

k1 x DFV1

                                        !!l!:M1 yr                   µCi                        Cal Due Date:. _ _        J_0_-_1_-~I-~._

Xe 133 Kr85m

                        ~ '.d-
                                . 294 1.17E3
                                                       ,J JI -             /I./  IA-,

Kr88 1.47E4 14. Bkg = lnstrpment backm.ound Xe 135 1.81E3 Bkg = 4, to\ l:..-7 µCi/cc Kr87 5.92E3 Xe 138 . 8.83E3 \ 15. XfQ = 4.3 x 10-11 secJm~ Xe 135m 3.12E3 \ (atmospheric dispersion) Xe 133m 2.51E2 \ 1' Ar41 N 13 16.tki = I" ,Iv 16a. B.84E3 8.83E3 ** r (ki OFV1l == J.. \~ JV JJ,,, ** Arbitrarily set equal to Xe-138 16b. :E ki ( #16 )

E(ki

• DFV I ) ( #I 6a )

                                                   =

17. Limit=L= - - -

1.06 X L ki

                                        -~-(-ki_ _ _ _ - - - - - -

J .06 x(Thelesserof#16bOR 1.81E-4):: (F)(X IQ) f..

• DFVi ) ( #9 )( #I S )

                                                                                                                              /          s. 'iS" e,"'1 Umit=L    =                    I.OG                       (  /,~/4-"( ) =                   5, '15 ~ .$                l~'t
                        <,f'oo9              )(   '{,3.,,,0-"1               ,,.                 a- ~'~                s-.,s-c.. c..-f

18. HiHiALARM = Ax( #17 ) =(1.0)( 5.q Sct.'*S )= -5 .. 1.)Z:.: **1a'l>cc The radioactive gas flow corresponding to the Hi Hi setpoint:

Performed by: l~ . Date: Independent Verification by:, _ _ _ _ _*_ _(_ _*_c._l~_A--

                                                                   ) _* 37'=,            _ _ _Date:_ _I_-_z._z_-_,_4-_*

5\&..r~ l\:5l S T'o P 17'AS'

                                                                                                                           \,.&-1)     i.~~

NEE-323-CALC-003 Attachment 3 Page 1 of9 PCP 8.7 ALARM SETPOINTS FOR LIQUID RAD Rev. 17 MONITORS Page 9 of 11 ATTACHMENT 1 Page 1 of 3 LIQUID EFFLUENT RAD10ACTIV1TY MONITOR SETPOINT

1. Sample No. /7 .... 5qqc; . 2. Sample Date & Time f-z.,$-17 / 00?~

3. Stream/Monitor Description ~ Rn1 -L/767

4. Effluent Monitor Reading {cps),--:::.-;--:~/=0-----------------

5. Effluent Flow (gpm) 9~ Ot> *

6. Average effluent :flow during time represented by sample, F1 {gpm)_----"=.,v__;,/~,1-;;.,-----,----,,.---

7. Average dilution (discharge canal) flow during time re2resented by sample, F2 (gpm) LJ/A--

8. Monitor calibration factor, g, (cps/µCi/ml) _ _z=-*::o-'-:;:;Cf.-e_

......S : - - - - - - - - - - - - - - - -

10. Fraction to apply as a safety margin, A = 0.5 Setpoint  :::lOx[! (KI.,K,+ WEC,)

xgx F Fi 2 xA]+Bkg = Setpoint=l ox[(l5)(S)(?) x(lO)J+(4) (16)(6) 1 1 J Setpoint = sx[(l 5)(S)(7) + (4) (16)(6)

                                     ..:z,...          C,

( J. o/pe,, }( '1,t'IG )( ,t/M, Set-point = Sx[ (  ; ] +( /0 )

l I Cf )( ..v/r

                         . - .      =2--#~-z:s-n

11. Setpoint = ~ - 5' 4.C> v"'

Fractional Ch~nge = New value - Previous Value= ( 11 ) - ( 9} = ( '5c./l> ) - ( 7 ~ ) Previous Value ( 9 ) ( ",G,.S- )

12. Fractional Change= .... 0,1,..1f v' If fractional change is greater than +/-0.3, adopt a new monitor alarm setting.

Continuous Monitor Hi Alarm = Setpoint

13. Monitor Hi Alarm = '7/, < ./

14. Radwaste Monitor Hi Alarm= .16 (11} = .16 ( ) = uM" ops . /

65

NEE-323-CALC-003 Attachment 3 Page 2of9 ALARM SETPOINTS FOR LIQUID RAD Rev. 17 MONITORS Page 9 of 11 ATTACHMENT 1 Page 1 of 3 LIQUID EFFLUENT RADIOACTIVITY MONITOR SETPOINT

 '- 1.         Sample No.        \ 0 - ']\9'6           /          . 2. Sample Date & Time l7.-l-l'5            (11  ~o'J "3.         Stream/Monitor Description GS'uJ f9i l'V\.Ol\'rt'l~                  sm-   4J(p1 v
   "-4.
  ."-. 5.

Effluent Monitor Reading (cps)_ _ 1o_~l--~'-~"- - - - - - - - - - - - - Effluent Flow (gpm) CC ,_ * * *

   ', 6.       Average effluent flow during time* represented by sample, F1 (gpm)

• N ,/

   '7.         Average dilution (discharge canal) flow during ~!Dyepresented by sample, F2 (gpm) , Nls,
  " 8.         Monitor calibration factor, g, (cps/µCi/mL) ~ - * "2.,l°t               e~     v"" *
 "'--.9.       Previous alarm value setpoint (cps)__1. . .w""""""$__,,,c...f-S _.V'--------------
"'- 10.        Fraction to apply as a safety margin,                   A = 0.5 Setpoint =  10x['r.,(K"i,K, 1 + WEC;)

xg xF Fi 2 xA]+Bkg = Setpoint=l ox[(l S)(S)(?) x(l o)]+(4) (16)(6) 7 J Setpolnt = sx[(l S)(S)( ) + (4) (16)(6)

                        = sx[( cpf'3 \:.~:i, Setpoint C
                                             )(

111. 5'-\

Z.,l<t e((>

                                                          )(
                                                                )(    l\~
                                                                    "1vt.
                                                                                 ; J+ (       )a -       }

-~ 11. Setpoint = ___ 9_9.._2____/ Fractional Change = New value - Previous Value = ( 11 ) - ( 9 ) = ( si 2. )-{ rfo5 ) Previous Value ( 9 ) ( ,G~ )

 *"* 12.       Fractional Change    = *- .. '2..2~ *V""'

U!)ractional change is greater than +/-0.3, adopt a new monitor alarm setting. Continuous Monitor Hi Alarm = Setpoint

13. Monitor Hi Alarm = '7C, "5 cf s /

14. Radwaste Monitor Hi Alarm= .16 (11) = .16 ( ) = N/A cps,/

24

s-,\-tt.\ .... NEE-323-CALC-003 PCP 8.7 ALARM SETPOINTS FOR ,LIQUID RAD Rev. 17 MONITORS Page 9 of 11

                                                  ' ATTACHMENT 1                                                  Page 1 of 3 LIQUID EFFLUENT RADIOACTIVITY MONITOR SETPOINT

1. Sample No. 14 - t,'6? v 2. Sample Date & Time 3.-;;>..'-?'-(Ll /oo"30""'

3. Stream/Monitor Description__G""'*":=S"""-v-/"~---4....,;...;leL..;1.;.__v v

    ~

Effluent Monitor Reading (cps} Effluent Flow (gpm) 0\ <DO D v

6. Average effluent:flow during time represented by sample, F1 (gpm)__._rv/......._A....__ _ _ _ _~v

7. Average dilution (discharge canal) flow during time represented by sample, F2 (gpm)....,ri(A-.....___V"

    @       Monitor calibration factor, g, (cps/µCi/mL)              ;:i , I q . 4> / . . . _________.;. , . .

9. Previous alarm value setpoint (cps) ~ ::l? ':I V'

10. Fraction to apply as a safety margin, A= 0.5 v Setpoint = JOx [ I ' K' **>:,_g-~ J

                                                   .* ,: -f:2 xA +Bkg          =      Setpoint=l ox[ (I(16)(6)

S)(S)(?) x(l O)] +(4)

• I.,(K, + WEC1) *
• Fi , * .

l

• I (16)(6)

                                      >]

Setpoint = sx[(lS)(S)(7 + (4) ** *

11. Setpoint = _ _ 7_<..e_S ___v' Fractional Change= New value - Previous Value= ( 11 ) - ( 9} = C7f.o? ) - ( ~.). ~~ }

Previous Value { 9 ) . ( ~ ~'5lf }

12. Fractional Change= - 0 , (oL:,"'

If fractional change is greater than +/-0.3, adopt a new monitor alarm setting. Continuous Monitor Hi Alarm =.Setpoint

13. Monitor Hi Alarm = 7 I.ii 15 .. *v7 1 *

14. . Radwaste Monitor Hi Alarm = .16* (11) = .16 ( ,-¢A..) = cy/tt cps /

• NEE-323 -CALC -003 Att ac hment 3 P aae 4 of 9 SURVEILLANCE TEST PROCEDURE STP NS790305 DAEC DUANE ARNOLD ENERGY CENTER TITLE: RHRSW RADIATION MONITOR CALIBRATION IRM-1997 I

Page Rev. 6 of 18 14 Prerequisites Performance Date: c:::).. - l ~- l 7 INITIALS 6.0 PREREQUISITES 6.1 From the Chemistry Supervisor, obtain the current UPSCALE HI alarm setpoint. Record below and in the trip column of the step indicated . Step 7.1 .10 lQ ~ ~ cps 6.2 From the Chemistry Supervisor, obtain the current high voltage setting. Record below and in the step indicated. _ __st-ep_1_.1_ .2_5_ ,_ _ _D__vD_c_ _ _ _ _ _ _ __ ~ W 1 S NOTE ~ Original Transfer Cal ibration Count Rate is the count rate of the 8 µCi source taken from the last time that the mockup was used to determine the detector efficiency. This can be found in the Effluent Monitor Alarm Setpoint book. It is then decay corrected to the date that this STP is being performed . 6.3 From the Chemistry Supervisor, obtain the following source information and record below: L\- 6.3.1 Original Trans Cal Count Rate ~ * ~q t: cps 6.3.2 Source Number _ LI\_\ _ [ )_~_ __,(p_~_ I_ _C_ s -l37 6.3.3 Original Date of Cal Count Rate ~ * ~4 - l 5 6.3.4 Geometry _ _ _ p__O_: _'""\_+_ ______ 6.3.5 Old Efficiency [Q , 4? ~ e -/ µCi/cc/cps Ch~ 6.4 Decay correct the Original Transfer Calibration Count Rate. Record and transfer the value to the step indicated below: Y ~ Decay Corrected Transfer Count Rate ~*~( E cps (Transfer to Step 7.1.37.) 6.5 As directed by PCP 8.7, analyze a sample of unfiltered reactor water and calculate the UPSCALE HI setpoint. Record below and in the trip column of the table listed . Step 7.1.28 qd\ l cps 10

NEE-323-CALC-003 Attachment 3 Page 5 of 9 wlo l(os2 g7~ </ PCP 8.7 ALARM SETPOINTS FOR LIQUID RAD Rev. 17 MONITORS Page 9 of 11 ATTACHMENT 1 Page 1 of 3 LIQUID EFFLUENT RADIOACTIVITY MONITOR SETPOINT

1. Sample No. t!"- If~& 2. Sample Date & Tir.pe g-zq-~-/ c:,,o.:.7

3. Stream/Monitor Description lfm-/91) (/2-1HU"w/ esw J

4. Effluent Monitor Reading ( c p s ) - . - , . ~ ~ - - - o = - - - - - - - - - - - - - - - - - -

5. Effluent Flow (gpm)

• qg@ *

6. Average effluent flow during time represented by sample, F1 (gpm)_./-'1/'.~'/_A_ _ _ ___,..,.--

7. Average dilution (discharge canal) flow during time represented by sample, F2 (gpm) ,;t//A

8. Monitor calibration factor, g, (cps/µCi/mL)-r--.----/,..... S""'/...,.e'-~------------

9. Previous alarm value setpoint (cps)_ _ _ t;_,_q_______________

10. Fraction to apply as a safety margin, A= 0.5 Setpo1nt . = }Ox [ "i. '..K, i..1(K1 -. WEC1)

F2 xg x -xA F;

                                                                   ]
                                                                     +Bkg    =     Setpoint=l ox[(l(16)(6)

S)(S)(?) x(l O)] +( 4) Setpoint = sx[0 5X8)(7 (16)(6)

                                    )]+     (4)
                                        -'2..-

1.10~ Set1)omt = 5x [< ~ 1s*1elP )< v/4

                       ..:._.:*"-'----=-----'-'--~-----'--: - -- - ' -
                                                                               ))] + C .30             )

( ~ )C ,vlfo

                                         /5°'3

• j '/

11 . St

• t e pain =

                       ~ 5"7?,3~                                                                a 4
                                                                                           ,; 6,, 1 ;rtP---b
                                                                                                             ,/'

Fractional Change = New value - Previous Value = ( 11 ) - ( 9 ) = ( -~ ) - ( IP I'-/ ) Previous Value ( 9 ) ( hf l/ )

12. Fractional Change= ~ - 0

                                                                 .o, If fractional change is greater than +/-0.3, adopt a new monitor alarm setting.

Continuous Monitor Hi Alarm = Setpoint

13. Monitor Hi Alarm = (.p I ~

14. Radwaste Monitor Hi Alarm= .16 (11) = .16 ( ) = _M_~__cps 33

NEE-323-CALC-003 Attachment 3 Page 6 of9 ALARM SETPOINTS FOR LIQUID RAD Rev. 17 MONITORS Page 9of 11 ATTACHMENT 1 Page 1 of3 LIQUID EFFLUENT RADIOACTIVITY MONITOR SETPOINT

1. SampleNo. /?- f;,~4- .2. SampleDate&Time 2-14-\G.. (O'V"Z...(

3. Stream/Monitor Description t2.. \-\-£-. Sc..J I ~- <;-.,J"""""~ 12.<I'\.. (°19 2 4.

5. Effluent Monitor Reading (cps)....,.....,._,..._~~- Effluent Flow {gpm) 12.i.-\-,ts..J 'A-' =- t;~..C r--=------+-=---....-------- (-t-t2_S:;,J ...~, ::.. 4~

6. Average effluent flow during time represented by sample, . 1 (gpm} ,.v 4-

7. Average dilution (discharge canal) flow during time represented by sample, F 2 (gpm) ,u 14--

8. Monitor calibration factor, g, (cps/µCl/mL)_...,..l.:...:;,,,_..S-f'-.;;...;.'-~-~-~------------

9. Previous alarm value setpoint (cps}_ _ _ _ <o_l4- _ _ _ _ _ _ _ _ _ _ _ _ _ __

10. Fraction to apply as a safety margin, A= 0.5 S etpomt = lOx

[ I., K, . ..: F,. . ] xg x-xA +Bkg = Setpoint=l ox[(lS)(S)(?) x(l o)J+(4) IiK, + fYEC,).. Fj (16)(6) . Setpoint::: sx[{lS)(S)(7 + (4) (16)(6)

                                     >]

Setpoint = sx[-'-(_t_,<o_*7_£_*-_>..:....)(____l,_S__l__~~...;..-_.)...:..(_ _ ,..,....:.t_o.__,_)J + ( *2. ~ ) ( l l\ .. ei~ )( 1\Jlk) ,.

11. Setpoint = S d... !? v""

Fractional Change= New value - Previous Value= ( 11 l-C 9) = { S4-3, ) - C G:>>C.:1--) Previous Value ( 9 ) ( b IA- )

12. Fractional Change= *- D ... l'Z..-v If fractional change is greater than +/-0.3, adopt a new monitor alarm setting.

Continuous Monitor Hi Alarm=- Setpoint '::=b OL-t:> <; ... , f&,. . .,, v"""

13. Monitor Hi Alarm=

14. Radwaste Monitor Hi Alarm= .16 (11) = .16 ( .... jA---) = tv /A;- cps

NEE-323-CALC-003 Attachment 3 Page 7 of 9 PCP 8.7 ALARM SETPOlNTS FOR LIQUID RAD Rev. 17 MONITORS Page 9 of 11 ATTACHMENT 1 Page 1 of 3 LIQUID EFFLUENT RADIOACTIVITY MONITOR SETPOINT

1. Sample No. J 7..- 31C::, . 2. Sample Date & Time /-/Jr/7 / ,t;;o/C,

3. Stream/Monitor Description /{/JJ - ~I? /<th&"~l£sv..1 Pt ?,,h'.9>' L.,,;.e (. llbru-r~)

4. Effluent Monitor Reading (cps)_ _ _~'O_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __

Effluent Flow (gpm) 'J(.,.&10

0) Average effluent :flow during time represented by sample, F1 (gpm)_~&-~

Average dilution (discharge canal) flow during time repres~nted by sample, F2 (gpm) __l.:~1?::...____ _ ___,....._ e/4

8. Monitor calibration factor, g. (cps/µCi/ml). l,.</"l e *

9. Previous alarm value setpo1nt (cps)_ _S___3_0 _ _ _ _ _ _ _ _ _ _ _ _ _ _ __

10. Fraction to apply as a safety margin, . A = 0.5 Setpoint =10x[ 'r.,K, xg x F2 xA]+Bkg = Setpoint=l ox[(lS)(S)(?) x(l o)]+(4)

              , *            'i:.1(K1 + WEC,)         Fj                                                 (16)(6)                   .*

Setpoint = sx[(l5)(S)(7

          .     .           (16)(6)
                                       ))+  (4)

Y'?"--J Setpoint == 5x[(_____ ._-'C________________ 41r

                                            )C //1ze-°">c-=-------)              )] +*(          /b               )

( 13-r. (J )( ,/4

11. Setpoint = _---c-5'_1_5-'.,__~--

Fractional Change = New value - Previous Value = ( 11 ) - ( 9) = ( 5~~S- ) - { S'~ ) Previous Value ( 9 ) ( G '%> )

12. Fractional Change = ti, IO S"'

If fractional change is greater than +/-0.3, adopt a new monitor alarm setting. Continuous Monitor Hi Alarm = Setpoint

13. Monitor Hi Alarm = 53P

14. Radwaste Monitor Hi Alarm =.16 (11) = .16 ( ) = Vll.,_'A"_cps

                                                                                   ---J.<&.....

24

NEE-323-CALC-003 Attachment 3 Page 8 of9

• .*..... ::. . :. ,.PLANT CHEMISTRY

                                                   *PROCEDURES                   ..3200 MANUAL PCP 8.7 ALARM SETPOINTS FOR LIQUID RAD                                                                Rev. 17 MONITORS                                                                                     Page 9 of 11 ATTACHMENT 1                                              Page 1 of 3 LIQUID EFFLUENT RADIOACTIVITY MONITOR SETPOINT

1. Sample No. 1'5-&t,5. . 2. SampleDate&Time 10-1q_.,c:;/ool9

3. Stream/Monitor Description R,H:g~w/E:s-~ R.v.p+..,.,,...e... R.ev1 -4:;J..~ci'

4. Effluent Monitor Reading (cps)_~.;.1..-'c,=*-------,,~..,,...----,.------,-,..-,;;--------

5. EffluentFlow(gpm) ~H.t~S../" A - '-1'60t>Wro,RH.R~"",. 6 4~0ft::p~

6. Average effluent flow during time represented by sample, F1 (gpm) N'ffl: ,

7. Average dilution (discharge canal) flow during time represented by sample, F2 (gpm) rvf.fl

8. Monitor calibration factor. g, (cps/µCi/ml) ~ * ;).. q ~ '-" * .

9. Previous alarm value setpoint (cps)__'ii:"'-=I.D._3::a.-:---:::-::--...:--------------

10. Fraction to apply as a safety margin, A= 0.5

• Setpoint=l ox[(l S)(S)(72x(l O)]+(4)

(] 6)(6) Setpoint = sx[(lS)(S)(7)] + (4) (16)(6) Setpomt C\ 'l:\{)E, .. ~ )( ~ . ?,.C\f.4, )(

                               * "" Sx[(--'---'-----'-------'-'---'--~                     rv/A.         )] *I- (

ao

                                         < l 74 .C\. 7                       )(          rv/A          .)

11. Setpoint = ___Ll.....:_ {;,?5 I _ _ __ V Fractional Change= New value - Previous Value= C 11 ) " ( 9) = ( ~ f.R~ ) - ( ~to~ )

• Previous Value ( 9 ) ( 'it~~ )

12. Fractional Change = - 0. ;l..~(Q ./

Continuous Monitor Hi Alarm = Setpoint If fractional change is greater than +/-0.3, adopt a new monitor alarm setting.

13. Monitor Hi Alarm = '6~ ~ . /

14. Radwaste Monitor Hi Alarm= .16 (11) == .16 (

28

NE:E-323-CALC-003 Attachment 3 Page 9 of9 ALARM SETPOINTS FOR LIQUID RAD PCP8.7 Rev. 17

                                                                                                                          \)

MONITORS Page 9 of 11 ATTACHMENT 1 Page 1 of 3 LIQUID EFFLUENT RADIOACTIVITY MONITOR SETPOINT

1. Sample No. 14- -en:b 4- 2. Sample Date & Time *2.-\4-\4--* / oo c.. I

3. Stream/Monitor Description 12-~<2..SyJ /f..5tN 'R.u.PTl.4.CZ-L~ Q.~ 4'2.'=>'e> *

4. Effluent Monitor Reading {cps) *2.; o -

5. Effluent Flow (gpm) 12~JZ.-S.J 'A', *4~':\C=: 1 ~~11J '~- 4e>otx-~ *

6. Average effluent flow during time represented by sample, F1 (gpm) #J J.o.- r..

7. Average dilution (discharge canal) flow during time represented by sample, F2 (gpm) ,.) ;w

8. Monitor calibration factor, g, (cps/µCi/ml) __'--=,-..";2_,:.~--'---<-P__ v_,__________

9. Previous alarm value setpoint (cps)____ e-=(p;;..~=----v"-------------

10. Fraction to apply as a safety margin, A= 0.5 Setpoint~1 ox[(lS)(S)(?) x(l O)]+(4)

(16)(6) Setpoint = sx[(l S)(S)(?)] + ( 4) . (16)(6)

                                                                                                         )

11. Setpoint= _ _ _~0? _ _ __ v Fractional Change= New value - Previous Value= ( 11 )- { 9) = ( 'b 0 7l- { ~\o*7 }

Previous Value ( 9 ) ( e,to 7)

12. Fractional Change= -o, 0*7 ./"

If fractional change is greater than +/-0.3, adopt a new monitor alarm setting., Continuous Monitor Hi Alarm = Setpoint *=i:>OLb S~-r f o'.,.-..""T v"

13. Monitor Hi Alarm = Bfo :1? /

14. Radwaste Monitor Hi Alarm= .16 (11) = .16 ( ""f A-) = ,J /p,,- cps

Spent Fuel Storage Pool Water Level 3.7.8 3.7 PLANT SYSTEMS 3.7.8 Spent Fuel Storage Pool Water Level LCO 3.7.8 The spent fue l storage pool water level shall be~ 36 ft. APPLICABILITY: During movement of irradiated fuel assemblies in the spent fuel storage pool. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Spent fuel storage A .1 ----------NOTE-*---- pool water level not LCO 3.0.3 is not within limit. applicable. Suspend movement of Immediately irradiated fuel assemblies in the spent fuel storage pool. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.8 .1 Verify the spent fuel storage pool water In accordance with level is ~ 36 ft. the Surveillance Frequency Control Program DAEC 3.7-18 Amendment 280

IAOP 981 FUEL HANDLING EVENT PROBABLE ANNUNCIATORS 1C03A A1 FUEL POOL EXHAUST HIGH-HIGH RADIATION 81 FUEL POOL EXHAUST HIGH RADIATION 1C048 86 NEW FUEL STORAGE AREA ARM HI RAD 1C04B C6 SPENT FUEL STORAGE AREA ARM HI RAD 1C05B ca PCIS GROUP "3" ISOLATION INITIATED 1C09A A2 NW DRYWELL RADIATION LEVEL HI-HI B2 NW DRYWELL RADIATION LEVEL HI 1C09B A2 SOUTH DRYWELL RADIATION LEVEL HI-HI B2 SOUTH DRYWELL RADIATION LEVEL HI 1C35A A1 REFUELING FLOOR NORTH END HI RADIATION A2 REFUELING FLOOR SOUTH END HI RADIATION PROBABLE INDICATIONS

1. Lowering cavity and/or Spent Fuel Pool level on the 5th floor. Visual

2. Lowering cavity level Floodup Range on level indicator, Ll-4541 (at 1C04).

3. Lowering Skimmer Surge Tank level on level indicator, Ll-3412 (at 1C04). not used in EAL

4. Lowering Fuel Pool level on level indicator, Ll-3413 (at 1C04).

5. Rising radiation levels on any of the following ARMs:

Spent Fuel Pool Area, Rl-9178 North Refuel Floor, Rl-9163 New Fuel Vault Area, Rl-9153 South Refuel Floor, Rl-9164

6. Rising Drywell radiation levels on either of the following (at 1C09):

NW Drywell Area Hi Range Rad Monitor, RIM-9184A South Drywell Area Hi Range Rad Monitor, RIM-91848 IAOP 981 Page 6 of 8 Rev. 6

Development of EAL Threshold values from NEE-323-CALC-004 Calculated values are provided in Calc-004 as shown below. Table 2- Reccmmended RA1 Uquld EALs Rad Monitor Equip. Modesl,2,3 Modes4,5 cps cps GSW RE-4767 2.32Et4 1.04Et4

                                                        . ,"\. " ,,~*

i

                }   iHRSW/ESW             'i!E\M'     1;60Et4          }.20E+3 RHRSW Dilution Line      RE-4268    2.42E+4          1.09E+4 The following table of threshold values was developed for use in the DAEC EAL scheme by averaging the separate Mode 1-3 and Mode 4-5 thresholds from Calc-004, and then rounding the average values for ease of EAL evaluator use, as well as to provide a step-wise progression through the emergency classification.

GSW rad monitor (RIS-4767) 2.0E-!04cps i5- RHRSW & ESW rad monitor (RM-1997) 1.0E-!04cps

 '.::i RHRSW & ESW Rupture Disc rad monitor (RM-4268)                          2.0E+o4cps

CALC NO. NEE-323-CALC-004 CALCULATION COVER arll 1* ENERCON

          ~xa;lfef_'l.ce...:.fvery proje~t. ~Ve,y qoy.             SHEET                       REV. 00 PAGE NO. 1 of 23 Client:         Duane Arnold Energy Center Revised Liquid Radiological EALs per NEI 99-

Title:

01 Project Identifier: NEE-323 Item Cover Sheet Items Yes No 1 Does this calculation contain any open assumptions, including preliminary D ~ information, that require confirmation? (If YES, identify the assumptions.) 2 Does this calculation serve as an "Alternate Calculation"? (If YES, identify the design D ~ verified calculation.) Design Verified Calculation No. _ _ 3 Does this calculation supersede an existing Calculation? (If YES, identify the design D ~ verified calculation.) Superseded Calculation No. -- Scope of Revision: Initial Issue Revision Impact on Results: Initial Issue Study Calculation D Final Calculation [g] Safety-Related D Non-Safety-Related [g] (Print Name and Sign) Originator: Jay Bhatt Date: 12/12/17 Design Verifier1 (Reviewer if NSR): Ryan Skaggs Date: 12/12/17 Approver: Zachary Rose Date: 12/12/17 Note 1: For non-safety-related calculation, design verification can be substituted by review.

CALC NO. NEE-323-CALC-004

~ ENERCON CALCULATION Ex~~lle_n~e-Eve.ry project Every dt;1y. REVISION STATUS SHEET             REV.         00 CALCULATION REVISION STATUS REVISION                                         DATE                           DESCRIPTION 00                                         12/12/17                          Initial Issue PAGE REVISION STATUS PAGE NO.                                       REVISION                PAGE NO.                 REVISION All                                           00 APPENDIX/ATTACHMENT REVISION STATUS APPENDIX NO.                         NO.OF                 REVISION       ATTACHMENT          NO.OF         REVISION PAGES                     NO.             NO.            PAGES            NO.

A 12 00 1 4 0 2 3 0 Page 2 of23

CALC NO. NEE-323-CALC-004 g) ENERCON TABLE OF CONTENTS Exc~llencf.>-E\:ery pr9Jecr. ~very c!ay, REV. 00 Section Page No.

1.0 Purpose and Scope

4 2.0 Summary of Results and Conclusions 4 3.0 References 5 4.0 Assumptions 5 5.0 Design Inputs 6 6.0 Methodology 8 7.0 Calculations 15 8.0 Computer Software 23 9.0 Impact Assessment 23

                                                                                     #of List of Attachments                                                                  Pages Attachment 1 - Calculation Preparation Checklist                                4 Attachment 2 - Monitor Efficiency                                               3 Page 3 of23

CALC a ENERCON

          *~xctrl!f.'Jce-E1:.e_ry proje_,;~.*Ev_e(y r/py.

Revised Liquid Radiological EALs per NEI 99-01 NO. REV. NEE-323-CALC-004 00

1.0 Purpose and Scope

The Duane Arnold Energy Center is implementing the guidance of Revision 6 to NEI 99-01, "Development of Emergency Action Levels for Non-Passive Reactors," which is the industry-developed methodology for emergency classification for the current operating fleet. Changes to the definitions of the condition for entry into the Emergency Action Level (EAL) RA1 result in the development of a new entry threshold value for this EAL. This calculation determines the liquid radiation monitor readings that correspond to the new EAL thresholds for the release of liquid radioactivity resulting in offsite dose greater than 1O mrem Total Effective Dose Equivalent (TEDE) or 50 mrem thyroid Committed Dose Equivalent (CDE) for one hour of exposure. 2.0 Summary of Results and Conclusions A spreadsheet was used to calculate the monitor counts per second (cps) reading necessary to reach offsite dose of 1O mrem TEDE or 50 mrem child thyroid organ dose as described in Section 7.0. The output from that spreadsheet is seen below. Table 1- Monitor Response for Liquid Radiological EAL Thresholds Rad Monitor Equip. Modes 1,2,3 Modes 4, 5 ID 2 H~ur.Decay 36 Hour Decay cps cps 10 mrem '. . . , 50 mrem Jo ~rem . SO mrem

                                                                           .. /L* TEDE

'--~"-'-'-----'-.:...:..c.;.;.....;...;..:c.;.;....__....c.=c.;.c.....c..c.c--~-'-l~ .. *: .

                                                                                                  * :, *c::Jhvroici.
                                                                                                                                \*JE(!I::: ,. *-/lhvr.oid General Service Water (GSW)                                  RE-4767              23,200                          49,100                       14,000 RHRSW Dilution Line                                  RE-4268              24,200                          51,300      10,900           14,650 For a given scenario, the threshold is always met for the TEDE dose before it is met for the organ dose.

The recommended RA1 Liquid EALs are: Page4 of23

CALC NEE-323-CALC-004 g!JI ENERCON Revised Liquid Radiological EALs NO. per NEI 99-01

       ~!C'rl(~n~e-Ev~ry wafe.tt! .E~e.ry ct_ay.

REV. 00 Table 2 - Recommended RA 1 Liquid EALs Rad Monitor Equip. Modes 1,2,3 Modes 4, 5 GSW RE-4767 2.32E+4 1.04E+4

                                                                                                     '       i
                                                                                                   '-', "->~ti RHRSW Dilution Line                     RE-4268        2.42E+4        l.09E+4 3.0 References 3.1 DAEC Offsite Dose Assessment Manual (ODAM), Rev. 37.

3.2 Federal Guidance Report No. 12, External Exposure to Radionuclides in Air, Water, and Soil, 1993. 3.3 Code of Federal Regulations, 10CFR20, January 2013. 3.4 NUREG-1940, RASCA~ 4: Description of Models and Methods, United St.~tes Nuclear Regulatory Commission, Office of Nuclear Security and Incident Response, 2012. 3.5 American National Stlndard Institute (ANSI/ANS). 1999. "Radioactive So.urea Term for Normal Operation of Light-Water Reactors," ANSI/ANS-18.1-1999, American Nuclear Society, La Grange Park, IL. 3.6 Plant Chemistry Procedure PCP 8.7, Alarm Setpoints for Liquid Rad. Monitors.

3. 7 NEI 99-01, "Development of Emergency Action Levels for Non-Passive Reactors", Rev. 6.

4.0 Assumptions 4.1 For the calculation determining the RA 1 EAL for Reactor Modes 1, 2 and 3, the source is assumed to have decayed for 2 hours before reaching the receptor. This decay time is appropriate to produce best estimate results for liquid effluent thresholds for the corresponding Reactor Modes. 4.2 For the calculation determining the RA1 EAL for Reactor Modes 4 and 5, the source is assumed to have decayed for 36 hours before reaching the receptor. This decay time is appropriate to produce best estimate results for liquid effluent thresholds for the corresponding Reactor Modes. 4.3 Per the ODAM, a mixing ratio of 5 is assumed when the effluent mixes with the water in the river. This correlates to a dilution ratio of 1/5 = 0.2. While the ODAM Page 5 of23

CALC g ENERCON fx.c.~ll~n.ce-;Ev~ry project. E~e_ry dqy. Revised Liquid Radiological EALs NO. per NEI 99-01 NEE-323-CALC-004 REV. 00 Section 2.3 states that a dilution factor of 10 can be used for drinking water, using a mixing ratio of 5 is appropriate for determining the EAL thresholds. 5.0 Design Inputs 5.1 Data for each of the Service Water Radiation monitors, taken from ODAM Table 1-2 and Attachment 2, is presented here: Table 3- Service Water Radiation Monitor Design Inputs Rad Monitor Equip. Range Efficiency Efficiency ID cps/µci/ml Source Document cps

                         ..*1.*
                            *, ,GSVit
                                . . /*,
                                                   *RE~4767    :0.1~106   * ..*.

RHRSW/ESW RE-1997 0.1-10 6 1.51E+06 Attachment 2 RHRSW Dilution Line* RE~4268 0.1:-106 2.29~+06. Attachmemt 2

                        *RE-4268 was previously known as the RHRSW Rupture Disk Page 6 of23

CALC NEE-323-CALC-004 0 ENERCON

    ~JfC5;f(~np:~~~.'Y proje._~t E~ery ~'!t Revised Liquid Radiological EALs NO.

per NEI 99-01 REV. 00 5.2 The isotopic mixture and half-lives used in this calculation are taken or developed from NUREG-1940 Table 1-2 and Table A-4. Table 4 - Isotopic Mixture and Half-Jives Isotope BWR Coolant Half-life Isotope BWR Coolant Half-life Concentration (hr) Concentration (hr) (Ci/g) (Ci/g)

           ~LA~g~-=:11~,o_rn~c"_.~~1:_oo~,E=.~-1~ .. '... 6.09E.-t:9i.                                  *. Na:-24          ', ; .;>f'.OOE:~09       \.;:J :50E+01 !

Ba-140 4.00E-10 3.05E+02 Np-239 8.00E-09 5.66E+01 Ce-144* 3.00E-12 6.82E+03 Rb-89 5.00E-09 2.54E-01

           !.. co~~~..                                  1.oofa1 o                        1_. 70E:t0,~ :.. ~u-103 , ..,

• 2.00E-11 **~A3E+02 ;

Co-60 2.00E-10 4.61E+04 Ru-106 3.00E-12 8.83E+03 Cs-134 3.00E-11 1.81E+04 Sr-90 7.00E-12 2.54E+05

                .. Cs~136' ,. ,'"        *'

2.QOE::l1

• 3.14E+02 Sr *. :. )4:0QE.:og . 9:5dE+OO  !

Cs-137* 8.00E-11 2.64E+05 Sr-92 1.00E-08 2.71E+OO t>.::cs;;f3s*.*::.

t. ( /c' ..,, .. ::,:,,' <:**

                                                                    .. -.o. Jt:.,r .

. :./.1..;. o...o,E *. i5:3aE::ot\ Te~129m

                                                                                                                      * <":\J4;00E~n.'\ s,,::ife,:...
                                                                                                                    .,_,. ,..,. i:,~*,:,;

v.r* 0

                                                                                                                                                   ,A *.8'* .*._:"0**6*.E*.*""*0**2* .*.:.:.

7(. ' Cu-64 . 3.00E-09 1.27E+01 Te-131m 1.00E-10 3.00E+01

           ' Fe~~9, . : . 3.0oE.:11                                                      1.07E+03 .. Te-132 . >: 1.ooE.:11 !:::7:82Et01 :

1-131 2.20E-09 1.93E+02 W-187 3.00E-10 2.39E+01 L :J:132 ',. . f'f9'=\Ht:: ... *2.29E:f99/,::. ,y;;91 ; ~.001::-11, **..\\{4ciE+o3 i

         -~            1-133                            1.SOE-08                         2.08E+01              Y-92                 6.00E-09            3.55E+OO i: *i*",1~~** *:_' .                      4,~oE::.oa                       *a.?GE"-Ol':'** :Y-$3 *....*: t4.doE~og*.                  :>:          .a 1 1E+ci 1 '1 1-135                            2.20E-08                         6.60E+OO           Zn-65                   1.00E-10            5.86E+03

( .. Mn-5,4 * .

• 3:50E-1 t Mn-56 2.50E-08 2.57E+OO Fe-55 1.00E-09 2.37E+04 Ni-63 1.00E-12 8.40E+05 5.3 Dose Coefficient from Water Immersion/ Annual Limit for Intake (ALI)

The dose coefficient for water immersion for each isotope are from Table 111.2 of FGR12. The Annual Limit for Intake (ALI), which represents the number of microcuries that would have to be ingested to cause a dose of 5 rem to an occupationally exposed worker is taken from 10CFR20, Appendix B, Table 1 Column One. Page 7 of23

CALC NEE-323-CALC-004 0 ENERCON

                ~xc<:11~.nce-E~~ry projecr.-.f'!ery, q~y.

Revised Liquid Radiological EALs per NEI 99-01 NO. REV. 00 Table 5 - FGR 12 and ALI Isotope FGR12 ALI Isotope FGR12 ALI tt*. µCi

                                                                 \.,

r I Sv rn 3

                                                                                                                                           / Bq*s Ag-llOm                         2.94E-16                       5.00E+02                          Na-24                       4.73E-16                  4.00E+03

ea-140 1.87E-17 6;00E+02 Np-239

• 1.70E-17 .. 2.00E+03 *.

Ce-141 7.62E-18 2.00E+03 P-32 1.90E-19 6.00E+02 3):>0E+02 . *~b~ij~ ';J:7/:'.': :i:3C>E~tGs '.'.{(°)< . 6:00~:t94? . : Co-58 1.03E-16 2.00E+03 Ru-103 4.89E-17 2.00E+03 Co-60 2.74E-J6 .. 2;QOE+02 Ru-106 *. *2.24E-17 2.00E+OZ

• Cr-51 3.30E-18 4.00E+04 Sr-89 1.49E-19 5.00E+02

                                                                                                      *sr~90
                                                                                                          * ,, A ' ,;\ ,~

l..46~:~?,Q.' ,. *. Li:H~~tl1Jr,J::: J Cs-136 2.31E-16 4.00E+02 Sr-91 7.48E-17 2.00E+03

• cs-.137 1.49E-:io . *1.00E+02 Sr-92 1.47E-'16. 3.00E+03 Cs-138 2;62E-16 3.00E+04 Te-129m 3.39E-18 5.00E+02

                                                                                   ,\*'"""'"**:
                                                                                   *   ' /[,_*,~ >-

Fe-59 1.29E-16 8.00E+02 Te-132 2.28E-17 7.00E+02 l* .,::131 3.98Ec.l7. *

• 9:00E+o1 .  ; w~1s1/' . 4.97E-11 . 2.00E+03 1-132 2.43E-16 9.00E+03 Y-91 5.44E-19 6.00E+02 2.81E-17 * '

3.00E+03,:

                                                                                                                                                           . . .. . ' : '      ~  ,*, -. t,*

i:

  ,.,    1-134                        2.82E-16                       3.00E+04                           y,.93                       1.03E-17                 1.00E+03

l. ,~135 1i73f.16 . *. 3.00E+03. , . Zn:65 - *6.29.E.:;t.7 . . 4:()0E+Q2i Mn-54 8.88E-17 2.00E+03 Zr-95 7.82E-17 1.00E+03

• M~-ss * *1;$6fJ6 .* 5,00E+03. Fe-55 . o:oo~+OO ~fOOl:+03;

                                                                                                                                                                  *, \; ' '

0 1."'\_*** ~-, ~. ,' i 1 Mo-99 1.58E-17 1.00E+03 H-3 O.OOE+OO 8.00E+04 r:.: , '~~~;~}<':. ' *, . I "* * / :** ' ,<

                                  ~-'---'"-'-'---~~~~-~-~-
                                                                                                     . iNj.:53*):;::;;,;. O:OOE+06'\ }';*:                   9:boE+CJ3:/

5.4 Dose transfer factors for radionuclides in effluent water for a child through the potable water pathway are taken from ODAM Appendix C, and shown in Table 9. 6.0 Methodology 6.1 General Approach With a given mixture of radionuclides, the dose received by an individual offsite is a function of the gross activity present in the liquid mixture. The resultant dose ~eceived by an offsite receptor is dependent not only on the gross radioactivity levels of the effluent but also upon the isotopic mixture present in the Page 8 of23

CALC NEE-323-CALC-004 0 . ENERCON

       ¥xCe!l~,:tc~~~'!~'Y proje.ct*.£~er}' ~<?Y.

Revised Liquid Radiological EALs per NEI 99-01 NO. REV. 00 liquid. This calculation predicts the relative contribution of each radionuclide to the gross radiation monitored by the liquid effluent monitor. For a liquid release, the only phenomenon affecting the mixture is radioactive decay. With the mixture known, a given gross output reading (cps) from a liquid effluent radiation monitor can be scaled to determine the concentration of each isotope present in the liquid effluent. The calculation then uses liquid dilution factors as described in the Offsite Dose Assessment Manual to determine the resultant concentration of radionuclides to which an individual offsite would be exposed. Dose conversion factors are used to determine the dose (mrem) to an individual offsite due to their exposure to the liquid mixture of radionuclides. With the given radionuclide mixture and dilution factors understood, an iterative process can be used to relate the liquid effluent monitor reading to a target offsite dose. Two types of radiation dose are calculated:

• Thyroid COE or Committed Dose Equivalent is the radiation dose to the thyroid due to an uptake of radioactive material. In this case, the uptake is limited to ingestion of radioactive material present in river water.
• TEDE or Total Effective Dose Equivalent is the summation of the Effective Dose Equivalent (EDE) and the Committed Effective Dose Equivalent (CEDE):

TEDE = EDE + CEDE. . ,.

• EDE is the dose due to an individual being directly exposed (by submersion) to the radiation present .in the liquid release. For this scenario, the individual is not actually immersed in the liquid, but boating above it so a correction factor is applied. * *
• CEDE is the sum of the COE for each organ of the body with weighting factors applied for each organ.

6.2 Scenario The ODAM described dose pathways focus on long term ingestion of radionuclides through various food pathways. This is in sharp contrast to the NEI thresholds which limit the exposure to one hour. To meet the prescribed one hour exposure scenario, the following scenario will be used:

• An adult and a child are fishing from a boat on the Cedar River downstream of the facility. While they are there, a radioactive liquid release from the facility oc-curs. The release lasts one hour.
• During that time frame each of the individuals ingests 500 milliliters of river water.
• The individuals leave the area one hour after the start of the exposure when they heed the announcement from the ERO siren system.

The pathways thus indicated are drinking water and boating. Page 9 of23

CALC NEE-323-CALC-004 0 ENERCON fic<rlf~!}.r;e.~El(e_ry projec~. E~e!Y ~ay. Revised Liquid Radiological EALs NO. per NEI 99-01 REV. 00 6.3 Radioactive Source The gross radioactive concentration is converted from cps to µCi/ml using the monitor efficiencies from Design Input 5.1. For example, the gross radioactive concentration in the GSW system with an indication of 10000 cps from the GSW Rad. Monitor is calculated as follows: 10000 f*r 1 µCi/ml 4.57E-03 µCi/ml 2.19E+06 f*r With the total concentration of the effluent known, given a mixture of isotopes where the relative amount of each isotope is known, the isotopic mixture in the effluent can be determined. The isotopic mixture and half-lives used in this calculation come from Design Input 5.2. In order to determine the radiation dose to the receptor, the concentrations affecting the receptor must be known. For Plant Modes 1, 2, and 3, a source decay time of two hours is assumed to account for transit time (Assumption 4.1 ). For Plant Modes 4 and 5, a source decay time of 36 hours is assumed (Assumption 4.2). The NUREG-1940 source is decayed using Design Input 5.2. The Ag-11 Om computation is displayed (indicated as row 7) as an example: Page 10 of23

CALC NEE-323-CALC-004 0 ENERCON Revised Liquid Radiological EALs NO. per NEI 99-01 ------------------l REV. 00 Table 6 - Source Decay b'ta~~i4ottr;~A:ooE::ioc.~ *.r::*:*t9JE+.Qz;;f::{?:2Jz71'.:l&3:)*.,:,';,1;;*'13J§s~'10:',;:J::;\;;:,;: ,,:r:;:0Ia:i:i§1:1~0,'o:'*,;<1 Ce-141 3.00E-11 7.80E+02 8.89E-04 2.99E-11 2.91E-11 [¢e;i44t;lJ,*~~f,§3'.(19~~;i,2'; *. }C'Ji:s:t1:tb3:if':i/tJ>iEtQ4".°:(::J.,P, 1;J{o91:::12'l.?~J;~;iTT~[fi-Qe-~1-i-;:_t_;;;-iiJ Co-58 1.00E-10 1.70E+03 4.0SE-04 9.99E-11 9.85E-11 Cr-51 3.00E-09 6.65E+02 1.04E-03 2.99E-09 2.89E-09 t,1,¢s~i~~<,'.;'f::§:QClE;;l.'tj;::*;\,iY~;I,EtQ4J< ,,;:{84~fo$**;\f }';3.QQtff:;~;  ;,~:()Q~1iij;cf't:;! Cs-136 2.00E-11 3.14E+02 2.20E-03 1.99E-11 1.85E-11 ttsfi~zf,*:;t<:slo:oEfti};':,'i>i:~41:fo$r.:c:;;ft.'i:i3E~o!>i;I;1*.t'},*,;1s:c:io~f:1J\;t;'c::i;'t :';1~:ooi:J-i'1'?'iil'iJJ Cs-138 1.00E-08 5.38E-01 1.29E+OO 7.59E-10 6.95E-29 f::i'(ul:&ii,l,'lt:{:>a:00(99?;i;;;:.':+/-:2.1Ho:r}:i,~;?t.$fifi:ie::ot,;:rit.:;,,:,.:,;t1H2;6§E~b9**,:.:,  :,4:2.oEftcfti~;t'.?i Fe-59 3.00E-11 1.07E+03 6.49E-04 3.00E-11 2.93E-11 v~*;:.41.~J.cf54;:,}~;(2'.2qe:o§;:I;;f* 1:;93E,i\ijg;t{;'?/:a;i;~E Q~(.:{;};;;;*~i:;::'i{1~~fo9 :. 0 f ***::'

• D 1:93EsCl91]:t:?¥i1 1-132 2.20E-08 2.29E+OO 3.02E-01 1.20E-08 4.16E-13 t;}:'i~;i~~ v;li;~:?zt({.;f:5QEfClij}?'.:::S2!Ci$'~+oi.~.1n)i3$3E{Cl2 :Y:';~;'i:~: *. :iA9E!QB ~f* tf:/1ii$2.f'.iQ§}i)f4:'1i 1-134 . 4.30E-08 8.76E-01 7.91E-01 8.83E-09. 1.83Ea20 i?*';ji;.l.3$;'.)*.;:.'.;'M;s2:io.E~8n*.'.i*z;.~.6()EtciQ:f:~,:::1:ros:i::p:i, }*;'f;;::';7;;"*1.7,SE;()&':i~ c;14:,::~:i.:;::>{,:s[Q2Efto:;:3;,;;;:~ri Mn-54 3.SOE-11 7.51E+03 9.23E-05 3.SOE-11 3.49E-11 i".'.',M~:$6 : :\;f':LSQE}O&"',.*:.** f(;'ij,{1:,/efo:CJ\:~. * * .~* i0fbf'.:+bi;f ':*. :.* ~;~!i;':,1:l46E;C)g'~l* ,rt*;i.)'~(l'.1.%.'.l.t;:i:z :*ttt:?i Mo-99 2.00E-09 6.60E+01 1.0SE-02 1.96E-09 1.37E-09

  ., ,,~~;z4l.**.*: ;12too:E~o9:.'f;; :;t,:'fSOE}Oli.*}~}*4:GZEfO:t.r)f 0:"%1{f:'{c:f;gzJ:4ij9<,f,cfJ* }':{;3}7~1:~10;*{                                           :,/;)

Np-239 8.00E-09 5.66E+01 1.22E-02 7.81E-09 5.15E-09

      ){'<\e;-32,;\; ;'.):c::4:PQE 1:;i,,,, t,5;83Etoi *.;:.)*'\};i§E*(){\;::,::;:J~?IfJ \~;§§'E~i:i'.i}.**

0 . 3:8.3.E7li trrrt; Rb-89 5.00E-09 2.54E-01 2.72E+OO 2.15E-11 1.26E-51 tr.Ru:;1.of,2JNt 'tqoe;11:;c: :i\:;9'.4=iEtOit,t: }fi3SEIQ*(0'/~;f~i1,P';io.Qt-'i1.** .,*. I:~;;./,; ',j:1:!:lsE:11 :Ji'}~?)J Ru-106 3.00E-12 8.83E+03 7.85E-05 3.00E-12 2.99E-12

           . $,::s~***** 2\:,t1:opE;io' t.*~~i1'.21E+03;~ h*s:~iE~4 *:c,_~\; .i~Q:Q9E~ii:E:~f;,);1i~;if;; .:;:,$.19;gQgt1+/-;; ,;1r,*~

1 Sr-90 7.00E-12 2.54E+05 2.72E-06 7.00E-12 7.00E-12 Sr-92 1.00E-08 2.71E+OO 2.56Es01 6.00E-09 1.0lE-12 Te-131m 1.00E-10 3.00E+Ol 2.31E-02 9.SSE-11 4.35E-11 W-187 3.00E-10 2.39E+01 2.90E-02 2.83E-10 1.06E-10 rt **:v::~1t:****;tt},t:001::11;,~:;-;J1Joe+o3*

• Ai94E~cl4\,'tls:j;zt#:oq~21.1.. :cs*:.4*l:;1i,si!:a!Q31:lii:J;,,c;tt(1 Y-92 6.00E-09 3.SSE+OO 1.95E-01 4.06E-09 5.34E-12 Fi.I;y,~~J't;§;i,~t***4:q9('Qil }:* '.*tQif'.!i:oi;; J: };'(GtsGe;.clieij ' ],;3.,:49E-69;fa;.,.f:S{\':' './3i:3sf".i6C
• t.sl Zn-65 1.00E-10 5.86E+03 1.18E-04 1.00E-10 9.96E-11 Jf'.,*:z,:95 ;;:rr:;::,1s:00E~i2>.*<t ' . :i'S54E{Q3,}1')4{~1Ei()4;1ffct*:,: .t7.J:l9J:if2:: *.. . ** *;:\ . * *7'.87Effi:**** ,i::l Fe-55 1.00E-09 2.37E+04 2.93E-05 1.00E-09 9.99E-10 f }s".tt:3\'!1",,{** l:1f;()OE:CJ8f/i'1:;;'.1;08:E°;!'QSt1,\{~;efoEfOq} '.\,1\t'op1:~o.s.*

Ni-63 1.00E-12 8.40E+05 8.25E-07 1.00E-12 1.00E-12 Page 11 of23

CALC NEE-323-CALC-004 0 ENERCON

         *~~C~ll~pc~~E'l(_e,ry_ pr.Oje_t:;. .E~<;!Y r:JC?Y.

Revised Liquid Radiological EALs NO. per NEI 99-01 REV. 00 6.4 TEDE Dose To determine the TEDE dose for a one hour exposure, two components are considered.

1. Direct exposure to the radioactive water present in the river is considered, commonly described as Effective Dose Equivalent (EDE).

2. Committed Effective Dose Equivalent (CEDE) is considered, which is the dose com-mitment due to the ingestion or inhalation of a mixture of radioactive material.

6.4.1 EDE Immersion dose is calculated with guidance provided in Federal Guidance Report 12 (FGR12). With the isotopic concentration at the receptor known (Table 6), the dose (mrem) at the receptor can be calculated: Where Xir= concentration of radionuclide i present in the water at the receptor (µCi/ mP i= . ' each isotope present in the liquid release hTi= factor from FGR12 for converting the liquid concentration to effective dose equivalent from Design Input 5.3 (mrem ml /'sec µCi) The dose coefficients in Design Input 5.3 from FGR12 relate the radioactive concentration of a liquid to the dose received by a person who is immersed in the liquid. FGR 12 also provides this statement about the relationship between immersion dose and dose received while boating: Exposure during boating activities The dose coefficients for immersion in contaminated water in Table 111.2 assume Immersion in an infinite pool and, thus, are appropriate for exposure while swimming. External exposure to contaminated water can also occur during boating activities. For photon exposure, a dose-reduction factor of 0.5 during boating activities is a reasonable value that Is unlikely to underestimate external dose equivalents. 6.4.2 CEDE To calculate the Committed Effective Dose Equivalent due to the consumption of contaminated water, the ALI values from Design Input 5.3 are used. The first column.of Page 12 of23

CALC NEE-323-CALC-004 0 E.NE.R;CQN f.~c_{f!(~!]~ff77f~.e/Y. prof~t. .E~ety r/_~y Revised Liquid Radiological EALs per NEI 99-01 NO. REV. 00 the table lists the ALI for each isotope. The ALI value is the Annual Limit for Intake and represents the number of microcuries that would have to be ingested to cause a dose of 5 rem to an occupationally exposed worker. Following the conversion of the ALI values to units of "mrem per µCi" the following equation can be used to determine radiation dose due to ingestion of a liquid mixture of radioisotopes.

                                           . Dose      = L/Xir

• v
• hED Where concentration of the radionuclide i present in the water at the receptor

(µCi/ml) each isotope present in the gaseous release factor converting the gas concentration to effective dose equivalent (mrem/µCi) v= volume of water ingested (ml) 10CFR20 includes a statement regarding the use of Table 1 Column One values for members of the public:

                    ... a factor of 2 to adjust the occupational values (derived for adults) so that they are applicable to other age groups.

Spreadsheets are used in section' 7 .1 to calculate EDE and CEDE from all of the isotopes in the mixture.

    , 6.5 Organ dose Methods to calculate Organ Dose are taken from ODAM section 2.6 titled "Accumulated Personal Maximum Dose". The guidance is provided here:

Page 13 of23

CALC NEE-323-CALC-004 (j EN Etttb N Revised Liquid Radiological EALs NO. per NEI 99-01 1-----------------t

         ~>fe!?t[e_(J~e~Every projec(, EVe!Y ~aY:

REV. 00 Dan =LA Dank k where the dose commitment (mrem) to organ n of age group a due to the isotopes identified in analysis k, where the analyses are those required by Table 7 _1-2. Thus the contribution to the dose from gamma emitters become available on a batch basis for batch releases and on a weekly basis for continuous releases. Similarly the contributions from H-3 is available on a monthly basis and the contributions from Fe-55, Sr-89, and Sr-90 become available on a quarterly basis. Dan = the dose commitment during the quarter-tO-date to organ n, induding whole body, of the maximally exposed person in age group a (mrem}

    &:a11i      =transfer factor relating a unit release of radionudide i (Ci) in a unit stream flow (gal/min) to dose commitment to organ n, or whole body, of an exposed person            tn mremgal]

[ age group a Ci min via environmental pathway e. Cu. = the concentration of radionuclide i in the undiluted liquid waste represented by sample k to be discharged (µCi/ml).

     !J. t1; =                    duration of radioactive release represented by sample k which occurs within time boundaries TB and TE and during which concentration Cik and flows F1k and F2k exist. (min.)                                *

3. 785

• 10-1 = conversion constant (3785 mUgal
• 10-6 Ci/µCi)

F 11; = flow in the radioactive waste release line (gal/min)" represented by sample k. F 21:.a = flow into which radioactive release represented by sample k is mixed in the river at the point of exposure or withdrawal of water for use (same units as F1k)" For this calculation,

                    *      ~tk is set to 60 minutes

• F1k!2ke is 0.2 per Assumption 4.3.

Page 14 of23

CALC a ENERCON

       ~.x{:~l[~!'~,:~,E~ry pr~j~t. ~~~efY. 9'.q);

Revised Liquid Radiological EALs per NEI 99-01 NO. REV. NEE-323-CALC-004 00

• Aeani values are taken from ODAM Appendix C (Design Input 5.4)
• Based on the scenario, only the child thyroid organ is considered, as this bounds the adult, and is used in Iowa as the basis for Protective Action Guidelines.
• sa*sed on the scenario, only the drinking water pathway is considered.

Spreadsheets are used in section 7.2 to calculate COE-Thyroid from all of the isotopes in the mixture. 7 .0 Calculation 7.1 TEDE Dose A Microsoft Excel spreadsheet uses an iterative process to determine the cps output readings from each of the three monitors that correspond to a TEDE Dose of 10 mrem. Sections of the spreadsheet are presented here. This is the spreadsheet for the RHRSW/ESW Monitor. Ingestion Dose: 8.24 mrem

                                                                      + Boating Dose:        1.79   mrem Resultant Total mRem:        10.0   mrem RM1997 FGR~ Units Conversion Factor:

I I 3.70E+15 FGRl2Boating/lmmersion ~

                                                                                          , , , Reduction Factor:L:::_J Monitorc[s:!X'    :,;.:'.16,00,0'.'.?; /J!cps
                                                                                            . 10CFR20 Ingestion    r--;-1 Monitor Efficiency:! 1.51f+6             I~                      AgeConsiderationfactor:L-.::....J uCr/ml Volume       Consumed:F'f{;;soo.: :Z'! ml Dilution Factor: !  0.20 Resultant River Gamma Concentration for the Given CPS Reading :

I lµCi/ml 1E+06Et'}S I 0.20

                                                                                              =

, 2J2£20~~

;1µCi/ml The mrem values seen above are calculated in the spreadsheet on the following pages.

As can be seen above, the dilution factor (0.20) is included in the efficiency equation to account for the fact that the concentration in the river will be only 20% of the concentration seen by the effluent radiation monitor per Assumption 4.3. Page 15 of23

CALC NEE-323-CALC-004 ~ ENERCON Revised Liquid Radiological EALs per NEI 99-01 NO.

       ~xc~ll~nce-Eve__ry project. E"'.ery day.

REV. 00 7.1.1 Boating Dose The concentrations for the individual isotopes are scaled to the gross concentration determined above. In this case, the value is 2.12E-03 µCi/ml. The value of 2.12E-03 is calculated based on the monitor cps reading entered by the spreadsheet user. Through an iterative process, the user enters the monitor cps necessary to determine the desired resultant total dose (in mrem). The dose coefficient for water immersion for each isotope in column B in Table 7 is taken from Design Input 5.3. FGR 12 displays dose factors in the SI units of SV m3/ bq sec. Traditional units of mrem ml/sec µCi sec are desired. FGR 12: 1 &lJ. ~ lE+OS mRem 1 ~ 1E+06 ml Q 3.70E+l5 mRem ml

                                                                                                                    =
                            ~            sec               &lJ. 2.7E-11    Q          ~       1E+06  µCi                     sec  µCi The conversion factor from SV m3/bq sec to mrem cm 3/µCi sec is 3. 70E+15.

The decayed mixture from column Eis taken from Table 6. Note that the concentration values present in the (starting) mixture do not affect the result. It is the ratios of the isotopes to the gross concentration (Section 7.1 - Column G) in the mixture that are needed. Per Section 6.4.1, a 0.5 dose-reduction factor is applied in Column I. For illustrative purposes, for the isotope AG-11 Om, the cell formulas are displayed. Table 7 - Boating Dose Decayed FGR 12: FGR12 Mix River Immersion Boating SVm3 Units mrem ml µCi  !!!r!ill!  !!!r!ill! b sec Conv. Ci sec ml Fraction Ci/ml Sec Hr

                                                                                    =E2/                              =H2*3600 2.94E-16            3.70E+l5     1.09E+O        Table 6             ,}it2.J}E0 3  =D2*G2 1.17E-7

• 0.5 Ag-llOm 2.94E-16 3.70E+l5 1.09E+o 1.00E-12 0.00% 1.81E-8 1.97E-8 3.S4E-5 Ba-140 1.87E-17 3.70E+l5 6.92E-2 3.98E-10 0.34% 7.20E-6 4.98E-7 8.97E-4 Ce-141 7.62E-18 3.70E+l5 2.82E-2 2.99E-11 0.03% 5.41E-7 1.53E-8 2.75E-5 Ce-144* 1.91E-18 3.70E+l5 7.07E-3 3.00E-12 0.00% 5.42E-8 3.83E-10 6.90E-7 Co-58 1.03E-16 3.70E+l5 3.81E-1 9.99E-11 0.09% 1.81E-6 6.88E-7 1.24E-3 Co-60 2.74E-16 3.70E+l5 1.0lE+o 2.00E-10 0.17% 3.62E-6 3.67E-6 6.60E-3 Cr-51 3.30E-18 3.70E+l5 1.22E-2 2.99E-09 2.55% 5.41E-5 6.61E-7 1.19E-3 Cs-134 1.64E-16 3.70E+l5 6.07E-1 3.00E-11 0.03% 5.42E-7 3.29E-7 S.92E-4 Cs-136 2.31E-16 3.70E+15 8.SSE-1 1.99E-11 0.02% 3.60E-7 3.08E-7 S.54E-4 Cs-137* 1.49E-20 3.70E+15 5.SlE-5 8.00E-11 0.07% 1.45E-6 7.97E-11 1.44E-7 Cs-138 2.62E-16 3.70E+l5 9.69E-1 7.59E-10 0.65% 1.37E-5 l.33E-5 2.39E-2 Cu-64 l.98E-17 3.70E+l5 7.33E-2 2.69E-09 2.29% 4.86E-5 3.S6E-6 6.41E-3 Fe-59 1.29E-16 3.70E+l5 4.77E-1 3.00E-11 0.03% 5.42E-7 2.59E-7 4.65E-4 1-131 3.98E-17 3.70E+l5 1.47E-1 2.18E-09 1.86% 3.9SE-5 5.82E-6 1.0SE-2 1-132 2.43E-16 3.70E+l5 8.99E-1 1.20E-08 10.26% 2.17E-4 1.95E-4 3.52E-1 1-133 6.39E-17 3.70E+15 2.36E-1 1.40E-08 11.97% 2.S4E-4 6.00E-5 1.08E-1 1-134 2.82E-16 3.70E+l5 l.04E+O 8.83E-09 7.54% 1.60E-4 1.67E-4 3.00E-1 1-135 1.73E-16 3.70E+15 6.40E-1 1.78E-08 15.21% 3.22E-4 2.06E-4 3.71E-1 Mn-54 8.88E-17 3.70E+l5 3.29E-1 3.SOE-11 0.03% 6.33E-7 2.08E-7 3.74E-4 Page 16 of 23

CALC NEE-323-CALC-004

0. ENERCON f~c_t;lie!J~e-Ev~ry pr.oje~t.. E~e!Y d_(JY.

Revised Liquid Radiological EALs NO. per NEI 99-01 REV. 00 Decayed FGR 12: FGR12 Mix River Immersion Boating SVm3 Units mrem ml µCi mrem mrem b sec Conv. Ci sec ml Fraction Ci/ml Sec Hr Mn-56 1.86E-16 3.70E+15 6.SSE-1 1.46E-08 12.43% 2.63E-4 1.SlE-4 3.26E-1 Mo-99 1.SSE-17 3.70E+l5 5.SSE-2 1.96E-09 1.67% 3.54E-5 2.07E-6 3.73E-3 Na-24 4.73E-16 3.70E+l5 1.75E+O l.S2E-09 1.56% 3.30E-5 5.77E-5 1.04E-1 Np-239 1.70E-17 3.70E+l5 6.29E-2 7.SlE-09 6.66% 1.41E-4 8.SSE-6 1.GOE-2 P-32 1.90E-19 3.70E+15 7.03E-4 3.99E-11 0.03% 7.22E-7 5.07E-10 9.13E-7 Rb-89 2.30E-16 3.70E+15 8.SlE-1 2.lSE-11 0.02% 3.S9E-7 3.31E-7 5.95E-4 Ru-103 4.89E-17 3.70E+l5 1.SlE-1 2.00E-11 0.02% 3.61E-7 6.53E-8 1.1SE-4 Ru-106 2.24E-17 3.70E+15 8.29E-2 3.00E-12 0.00% 5.42E-8 4.49E-9 8.09E-6 Sr-89 1.49E-19 3.70E+15 5.51E-4 9.99E-11 0.09% 1.SlE-6 9.96E-10 1.79E-6 Sr-90 1.46E-20 3.70E+15 5.40E-5 7.00E-12 0.01% 1.27E-7 6.84E-12 1.23E-8 Sr-91 7.48E-17 3.70E+15 2.77E-1 3.46E-09 2.95% 6.25E-5 1.73E-5 3.llE-2 Sr-92 1.47E-16 3.70E+l5 5.44E-1 6.00E-09 5.12% 1.0SE-4 5.90E-5 1.0GE-1 Te-129m 3.39E-18 3.70E+15 1.25E-2 3.99E-11 0.03% 7.22E-7 9.0GE-9 1.63E-5 Te-131m 1.52E-16 3.70E+l5 5.62E-1 9.SSE-11 0.08% 1.73E-6 9.71E-7 1.75E-3 Te-132 2.2SE-17 3.70E+15 8.44E-2 9.82E-12 0.01% 1.7SE-7 1.SOE-8 2.70E-5 W-187 4.97E-17 3.70E+15 1.84E-1 2.83E-10 0.24% 5.12E-6 9.41E-7 1.69E-3 Y-91 5.44E-19 3.70E+l5 2.0lE-3 4.00E-11 0.03% 7.23E-7 1.45E-9 2.62E-6 Y-92 2.SlE-17 3.70E+l5 1.04E-1 4.0GE-09 3.46% 7.34E-5 7.63E-6 1.37E-2 Y-93 1.03E-17 3.70E+15 3.SlE-2 3.49E-09 2.98% 6.31E-5 2.40E-6 4.33E-3 Zn-65 6.29E-17

• 3.70E+15 2.33E-1 1.00E-10 0.09% 1.81E-6 4.21E-7 7.57E-4 Zr-95 7.82E-17 3.70E+15 2.S9E-1 7.99E-12 0.01% 1.45E-7 4.18E-8 7.53E-5 Fe-55 O.OOE+OO 3.70E+15 O.OOE+O 1.00E-09 0.85% 1.SlE-5 O.OOE+O. O.OOE+O H-3 O.OOE+OO 3.70E+15 O.OOE+O 1.00E-08 8.53% 1.SlE-4 O.OOE+O O.OOE+O Ni-63 O.OOE+OO 3.70E+15 O.OOE+O 1.00E-12 0.00% 1.SlE-8 O.OOE+O O.OOE+O 1.17E-7 100.00% 2.12E-3 9.97E-4 1.79

                                                                                          - : :j;12e::a*               mrem 7.1.2               Ingestion Dose Ingestion dose is calculated in the section of the spreadsheet seen below. ALis (col-umn D) from Design Input 5.3 are converted to mrem/µCi factors (column E) as shown in the example below for Co-60 which has an ALI of 200 µCi:

5 -Fem 1000 mrem 25 mRem 2E+02 µCi 1 RefR µCi The resultant dose (in mrem) caused by ingesting 500 ml of the liquid is calculated per Section 6.4.2. The decayed mixture from column F is taken from Table 6. An example demonstrating dose caused by drinking 500 ml of water containing cobalt-60 with a concentration of 0.001 µCi/ml: Page 17 of23

CALC NEE-323-CALC-004

~   ENERCON                                 Revised Liquid Radiological EALs NO.

1-----------------l Excellenct-EW!ry prOJ<<l Every day. per NEI 99-01 REV. 00 25 mrem 0.001 ~ 500 ~ 12.5 mrem

       ---+----1----1--1----+--=-------

For an occupationally exposed worker, drinking those 500 milliliters of water contaminated with Co-60 would result in a dose of 12.5 mrem. Per Section 6.4.2, this is multiplied by 2, generating 25 mrem. Note that there are three hard-to-detect isotopes (HTDs) present in the mixture: Fe-55, H-3 and Ni-63. Because they do not emit gamma rays, they are not detected by the service water radiation monitor. Therefore, they are effectively removed from the gross gamma calculation calibrating the monitor response (Columns H, I, and J). The HTDs are then scaled back into the calculation of the applied dose. For illustrative purposes, for the isotope Ag-11 Om, the cell formulas are displayed . The decayed mixture from column F is taken from Table 6. Column M contains the multiplier for the 500 ml volume consumed and the 2x multiplier factor for members of the public. Table 8 - Ingestion Dose Gamma Gamma 10CFR20 mrem Ci Ci Gamma  !&i  !&i ALI Ci ml Fraction ml Fraction ml ml Fraction mrem

                                                 =1/                                        =H2/                          =K2/     =K2*E2 Ag-110m           5.00E+02                  Table 6  =F2*1 .17E-7     =F2             =12*2.12E-3    =J2 (D2/5000)                                      1.06E-7                       2.32E-3   *2*soo Ag-llOm            5.00E+02      1.00E+l   1.00E-12      0.0%      l.OOE-12     0.0%    1.99E-8    1.99E-8    0.0%    1.99E-04 Ba-140           6.00E+02      8.33E+O   3.98E-10      0.3%      3.98E-10     0.4%    7.95E-6    7.95E-6    0.3%    6.62E-02 Ce-141           2.00E+03      2.SOE+O   2.99E-ll      0.0%      2.99E-ll     0.0%    5.98E-7    5.98E-7    0.0%    1.49E-03 Ce-144*          3.00E+02       1.67E+l   3.00E-12      0.0%      3.00E-12     0.0%    5 .98E-8   5.98E-8    0.0%    9.97E-04 Co-58            1.00E+03      5 .00E+O  9.99E-ll      0.1%      9.99E-ll     0.1%    1.99E-6    1.99E-6    0.1%    9.97E-03 Co-60            2.00E+02      2.SOE+l   2.00E-10      0.2%      2.00E-10     0.2%    3.99E-6    3.99E-6    0.2%    9.98E-02 Cr-51           4.00E+04      1.25E-1   2.99E-09      2.6%       2.99E-9     2.8%    5.97E-5    5.97E-5    2.6%    7.47E-03 Cs-134           7.00E+Ol      7.14E+l   3.00E-11      0.0%      3.00E-11     0.0%    5.99E-7    5.99E-7    0.0%    4.28E-02 Cs-136           4.00E+02      1.25E+l   1.99E-ll      0.0%      1.99E-ll     0.0%    3.97E-7    3.97E-7    0.0%    4.97E-03 Cs-137*           1.00E+02      5 .00E+l  8.00E-11      0.1%      8 .00E-11    0.1%    1.GOE-6    1.GOE-6    0.1%    7.98E-02 Cs-138           3.00E+04      1.67E-l   7.59E-10      0.6%      7.59E-10     0.7%    l.SlE-5    1.SlE-5    0.7%    2.52E-03 Cu-64            1.00E+04      5.00E-1   2.69E-09      2.3%       2.69E-9     2.5%    5.37E-5    5.37E-5    2.3%    2.68E-02 Fe-59          8.00E+02       6.25E+O   3.00E-11      0.0%      3.00E-11     0.0%    5.98E-7    5.98E-7    0.0%    3.74E-03 1-131          9.00E+Ol       5.SGE+l   2.18E-09      1.9%       2.18E-9     2.1%    4.36E-5    4.36E-5    1.9%    2.42E+OO 1-132          9.00E+03       5.56E-1   1.20E-08     10.3%       1.20E-8    11.3%    2.40E-4    2.40E-4   10.3%    1.33E-Ol 1-133           5.00E+02      l .OOE+l  1.40E-08     12.0%       1.40E-8    13.2%    2.SOE-4    2.SOE-4   12.1%    2.80E+OO 1-134           3.00E+04      1.67E-1   8.83E-09      7.5%       8 .83E-9    8.3%    1.76E-4    1.76E-4    7.6%    2.94E-02 1-135           3.00E+03      1.67E+O   1.78E-08     15.2%       1.78E-8    16.8%    3.SGE-4    3.56E-4   15.3%    5.93E-Ol Mn-54            2.00E+03      2.SOE+O   3.SOE-11      0.0%      3.SOE-11     0.0%    6.98E-7    6.98E-7    0.0%    1.75E-03 Mn-56           5.00E+03       l .OOE+O  l.46E-08     12.4%       l.46E-8    13.7%    2.91E-4    2.91E-4   12.5%    2.91E-01 Mo-99            1.00E+03      5.00E+O   l .96E-09     1.7%       1.96E-9     1.8%    3.91E-5    3.91E-5    1.7%    1.95E-Ol Na-24           4.00E+03       l .25E+O  l .82E-09     1.6%       l .82E-9    1.7%    3.64E-5    3.64E-5    1.6%    4.SSE-02 Np-239           2.00E+03      2.SOE+O   7.SlE-09      6.7%       7.81E-9     7.4%    1.SGE-4    1.SGE-4    6.7%    3.89E-01 P-32          6.00E+02       8.33E+O   3.99E-ll      0.0%      3.99E-ll     0.0%    7.96E-7    7.96E-7    0.0",A; 6.64E-03 Rb-89           6.00E+04       8.33E-2   2.lSE-11      0.0%      2.lSE-11     0.0%    4.29E-7    4.29 E-7   0.0%    3.57E-05 Page 18 of 23

CALC NEE-323-CALC-004 ENERCON Revised Liquid Rad iological EALs NO. Excell~ncr:-E~ry pro;~,. Ev~ry day per NEI 99-01 t---- - - - - - - -- -----; REV. 00 Ga mma Gamma 10CFR20 mrem Ci Ci Gamma  !!9  !!9 ALI Ci ml Fractio n ml F raction ml ml Fraction mrem Ru- 103 2.00E+03 2.SOE+O 2.00E- 11 0.0"..<. 2.00E-11 0 .0"..<. 3.98E-7 3.98E-7 0.0"..<. 9.96E-04 Ru-106 2.00E+02 2.SOE+l 3.00E-12 0.0% 3.00E-12 0.0"..<. 5.99E-8 5.99E-8 0.0"..<. 1.SOE-03 Sr-89 5.00E+02 1.00E+l 9.99E-11 0.1% 9.99E- 11 0.1% 1.99E-6 l .99E-6 0.1% l.99E-02 Sr-90 4 .00E+Ol 1.25E+2 7.00E- 12 0.0% 7.00E-12 0.0% l .40E-7 l.40 E-7 0.0",6 l.7SE-02 Sr-91 2.00E+03 2.SOE+O 3.46E-09 2.9% 3.46E-9 3.3% 6.90E-5 6.90E-5 3.0",6 1.72E-01 Sr-92 3.00E+03 1.67E+O 6.00E-09 5.1% 6.00E-9 5.6% l .20E-4 1.20E-4 5.2% 1.99E-01 Te-129m 5.00E+02 1.00E+l 3.99E-11 0.0"..<. 3.99E- 11 0.0% 7.97E-7 7.97E-7 0.0"..<. 7.97E-03 Te-131m 6.00E+02 8.33E+O 9.SSE-11 0 .1% 9.SSE- 11 0 .1% 1.91E-6 1.91E-6 0 .1% 1.59E-02 Te-132 7.00E+02 7.14E+O 9 .82E-12 0 .0% 9.82E-12 0.0% 1.96E-7 1.96E-7 0 .0",6 1.40E-03 W-187 2.00E+03 2.SOE+O 2.83E-10 0.2% 2.83E-10 0.3% 5.GSE-6 5.GSE-6 0 .2% 1.41E-02 Y-91 6.00E+02 8.33E+O 4 .00E-11 0.0"..<. 4.00E- 11 0 .0",(, 7.97E-7 7.97E-7 0.0",{, 6.64E-03 Y-92 3.00E+03 1.67 E+O 4.0GE-09 3.5% 4.0GE-9 3.8% 8 .lOE-5 8.lOE-5 3 .5% 1.35E-01 Y-93 1.00E+03 5.00E+O 3.49E-09 3.0",(, 3.49E-9 3.3% 6.96E-5 6.96E-5 3.0"..<. 3.48E-Ol Zn-65 4.00E+02 1.25E+l 1.00E- 10 0.1% 1.00E-10 0.1% 1.99E-6 1.99E-6 0.1% 2.49E-02 Zr-95 1.00E+03 5.00E+O 7.99E-12 0.0"..<. 7.99E-12 0.0",{, 1.59E-7 1.59E-7 0 .0"..<. 7.97E-04 Fe-55 9.00E+03 5.SGE-1 1.00E-09 0 .9% l .81E-5 0.8% 1.00E-02 H-3 8.00E+04 6.25E-2 l.OOE-08 8.5% 1.81E-4 7.8% 1.13E-02 Ni-63 9.00E+03 5.SGE-1 1.00E- 12 0 .0",6 1.81E-8 0.0",{, 1.00E-05 1.17E-7 100",(, 1.0GE-07 100.0",(, 2.12E-3 2.32E-3 100.0% 8.24 2.12E-3 mrem The initial HTD value is scaled to the total value of the gammas present. Using those ratios , the new HTD concentrations are determined by multiplying the ratios by the revised sum of the gamma emitters. Hard-to-Detect Determination Concl Igammas Igammas Conc2 HTD Ci Ci Ratio HTD  !!9 .l&l ml ml ml ml Fe-55 l.OOE-9 1.17E-07 8.53E-3 Fe-55 2.12E-3 l.81E-5 H-3 1.00E-8 1.17E-07 8.53E-2 H-3 2.12E-3 l.81E-4 Ni-63 l.OOE-12 1.17E-07 8.53E-6 Ni-63 2.12E-3 l.81E-8 7.2 Organ Dose The organ dose calculation is similar to the TEDE calculation above in that it uses a spreadsheet to determine the monitor cps reading necessary to reach the EAL threshold . The calculation is also similar in that it uses liquid concentrations and dose conversion factors to determine dose. The first part of the spreadsheet is presented here showing the gross concentration developed from section 6.5. Page 19 of 23

CALC NEE-323-CALC-004 ENERCON Revised Liquid Radiolog ical EALs NO. 1--- - - - - - - - ------i Exctlltnce- Every pro1ect £11try day. per NEI 99-01 REV. 00 Child Thyroid: 50.0 mRem M onitor: RHRSW/ESW RM 1997

   -."'cua                               Decay Hours!_2__

I ODAM Conversion Factor: 13. 785E- 03 I ftS

     ~

ftS Monitor CPS:! Mon itor Efficiency: ! 1.51E+6 33,800 jcps

                                                                     !cps EKposure Time (Mins.): ~

L...:_J

    >                                                       uCi/ m l Dilution Facto:r: ...._ 0_.2_   0 ....,!                      *Combining Factor:! 2..271E-01 !

Resultant Gamm a Concentrati on for th e Given CPS Reading : 33,80011* I 1.51806 Eft5 I uCi/ ml I 0 .20 = 4.4,BE.!03 1uCi/ ml The dose value seen above is calculated in the spreadsheet on the following pages. As can be seen above, the dilution factor (0.20) is included in the efficiency equation to account for the fact that the concentration in the river will be only 20% of the concentration seen by the effluent rad monitor per Assumption 4.3. ,. The concentrations for the individual isotopes are scaled to the gross concentration determined above. In this case , the value is 4.48E-03 µCi/ml. This value is calculated based on the monitor cps reading entered by the spreadsheet user. Through an iterative process, the user enters the monitor cps necessary to determine the desired resultant total dose (in mrem). Resultant river concentrations are presented in the section of the spreadsheet seen on the next page. In this spreadsheet, all of the isotopes present in Append ix C of the ODAM (Design Input 5.4) are included. In many instances, there is no corresponding isotope available from the NUREG-1940 reference . The entire list was included to simplify the spreadsheet calculation . The decayed mixture is taken from Table 6. Note that the concentration values present in the (starting) mixture do not affect the resu lt. It is the ratios of the isotopes to the gross concentration in the mixture that are needed . These fractions are calculated in the same way as section 7 .1. As in section 7 .1.2, HTDs present in the mixture (Fe-55, H-3, and Ni-

63) are removed from the gross gamma calculation and then scaled back into the calculation of the applied dose (see below for scaling).

Page 20 of23

CALC NEE-323-CAlC-004 ENERCON Revised Liquid Radiological EALs NO. 1---------- ----------< Exceflenc~-Evtry prOJ<<t. Every day. per NEI 99-01 REV. 00 The ODAM Appendix C dose transfer factors for the thyroid (Design Input 5.4) are displayed on the second to last column from the right. Using ODAM calculation methods described in section 6.5, the H-3 dose component of the child thyroid pathway is calculated individually here as an example: 3.82E-04 µCi/ml

• 2.97E+01 mrem gal/ Ci min
• 3.785E-03 ml Ci/gal µCi
• 60 min = 2.58E-03 mrem Where 3.785E-03 is the conversion factor from the ODAM .

2.97E+01 is the H-3 dose transfer factor for the thyroid (Design Input 5.4). 3.82E-04 is the concentration of H-3 in the river corresponding to the monitor reading developed from Table 6 (see below for scaling). AND the dispersion term (0.20) is not included here because it has already been included above in the concentration calculation . Table 9 - Thyroid Dose Decayed River River River Dose Trans-Mix Mix Mix Mix Water fer Factor Gamma Gamma Thyroid Thyroid ODAM g g Gamma .l!9 Gamma .l!9 {mremgal) Dose Isotopes Qm Fraction Qm Fraction ml Fraction ml (Ci min) mrem H 3 1.00E-08 8.532% 0.000",{, 0.00% 3.82E-4 2.97E+l 2.SSE-3 C 14 0.000%. O.OOE+OO 0.000% O.OOE+OO 0.00% O.OOE+O , 6.19E+2 O.OOE+O NA24 1.82E-09 1.556% 1.82E-09 1.717% 7.69E-05 1.72% 7.69E-5 4.91E+2 8.57E-3 P 32 3.99E-11 0.034% 3.99E-11 0.038% 1.68E-06 0.04% 1.68E-6 O.OOE+O O.OOE+O CR 51 2.99E-09 2.554% 2.99E-09 2.819% 1.26E-04 2.82% 1.26E-4 1.23E+O 3.52E-5 MN54 3.50E-11 0.030",{, 3.SOE-11 0.033% 1.48E-06 0.03% 1.48E-6 O.OOE+O O.OOE+O MN56 1.46E-08 12.432% 1.46E-08 13.720% 6.14E-04 13.72% 6.14E-4 O.OOE+O O.OOE+O FE 55 1.00E-09 0.853% 0.000",{, 0.00% 3.82E-5 O.OOE+O O.OOE+O FE 59 3.00E-11 0.026% 3.00E-11 0.028% 1.26E-06 0.03% 1.26E-6 O.OOE+O 0.00E+O C058 9.99E-11 0.085% 9.99E-11 0.094% 4.21E-06 0.09% 4.21E-6 O.OOE+O O.OOE+O C060 2.00E-10 0.171% 2.00E-10 0.188% 8.43E-06 0.19% 8.43E-6 O.OOE+O O.OOE+O Nl63 1.00E-12 0.001% 0.000% 0.00% 3.82E-8 O.OOE+O O.OOE+O N165 0.000% O.OOE+OO 0.000% O.OOE+OO 0.00% O.OOE+O O.OOE+O O.OOE+O CU64 2.69E-09 2.295% 2.69E-09 2.532% 1.13E-04 2.53% 1.13E-4 O.OOE+O O.OOE+O ZN 65 1.00E-10 0.085% l .OOE-10 0.094% 4.21E-06 0.09% 4.21E-6 O.OOE+O O.OOE+O ZN 69 0.000",{, O.OOE+oO 0.000% O.OOE+OO 0.00% O.OOE+O O.OOE+O O.OOE+o BR83 0.000",{, O.OOE+OO 0.000% O.OOE+OO 0.00% O.OOE+O O.OOE+O O.OOE+O BR84 0.000",{, O.OOE+OO 0.000% O.OOE+OO 0.00% O.OOE+o O.OOE+O O.OOE+O BR 85 0.000",{, O.OOE+OO 0.000% O.OOE+OO 0.00% O.OOE+O 0.00E+O 0.00E+o RB 86 0.000% O.OOE+OO 0.000% O.OOE+OO 0.00% O.OOE+O O.OOE+O O.OOE+O RB 88 0.000% O.OOE+OO 0.000% O.OOE+OO 0.00% O.OOE+O O.OOE+O O.OOE+O RB 89 2.15E-11 0.018% 2.lSE-11 0.020",{, 9.0GE-07 0.02% 9.0GE-7 O.OOE+O O.OOE+O SR89 9.99E-11 0.085% 9.99E-11 0.094% 4.21E-06 0.09% 4.21E-6 O.OOE+O O.OOE+O SR90 7.00E-12 0.006% 7.00E-12 0.007% 2.95E-07 0.01% 2.95E-7 O.OOE+O O.OOE+O SR 91 3.46E-D9 2.950",{, 3.46E-09 3.255% 1.46E-04 3.26% 1.46E-4 O.OOE+O O.OOE+O SR92 6.00E-09 5.117% 6.00E-09 5.647% 2.53E-04 5.65% 2.53E-4 O.OOE+O O.OOE+O Y 90 0.000% O.OOE+OO 0.000% O.OOE+OO 0.00% O.OOE+O O.OOE+O O.OOE+O Y 91M 0.000% O.OOE+OO 0.000% O.OOE+OO 0.00% 0.00E+O O.OOE+O O.OOE+O Y 91 4.00E-11 0.034% 4.00E-11 0.038% 1.68E-06 0.04% 1.68E-6 O.OOE+O O.OOE+O Page 21 of 23

CALC NEE-323-CALC-004 ENERCON Revised Liquid Radiological EALs NO. Excellen~-Every pro;<<r. Ew:ry doy. per NEI 99-01 f--------------------1 REV. 00 Decayed River River River Dose Trans-Mix Mix Mix Mix Water fer Factor Gamma Gamma Thyroid Thyroid ODAM g g Gamma J.!Q Gamma J.!Q (mrem gall Dose Isotopes om Fraction am Fraction ml Fraction ml (Ci min) mrem Y 92 4.06E-09 3.465% 4.06E-09 3.824% l.71E-04 3.82% l.71E-4 O.OOE+O O.OOE+O Y 93 3.49E-09 2.975% 3.49E-09 3.283% l.47E-04 3.28% 1.47E-4 O.OOE+O O.OOE+O ZR95 7.99E-12 0.007% 7.99E-12 0.008% 3.37E-07 0.01% 3.37E-7 O.OOE+O O.OOE+O ZR 97 0.000% O.OOE+OO 0.000% O.OOE+OO 0.00% O.OOE+O O.OOE+O O.OOE+O NB 95 0.000% O.OOE+OO 0.000% O.OOE+OO 0.00% O.OOE+O O.OOE+O O.OOE+O M099 1.96E-09 1.671% l.96E-09 1.844% 8.26E-05 1.84% 8.26E-5 O.OOE+O O.OOE+O TC99M 0.000% O.OOE+OO O.OOO"A; O.OOE+OO 0.00% O.OOE+O O.OOE+O O.OOE+O TC101 0.000% O.OOE+OO 0.000% O.OOE+OO 0.00% O.OOE+O O.OOE+O O.OOE+O RU103 2.00E-11 0.017% 2.00E-11 0.019% 8.42E-07 0.02% 8.42E-7 O.OOE+O O.OOE+O RU105 0.000% O.OOE+OO O.OOO"A; O.OOE+OO 0.00% O.OOE+O O.OOE+O O.OOE+O RU106 3.00E-12 0.003% 3.00E-12 0.003% l.26E-07 0.00% 1.26E-7 O.OOE+O O.OOE+O AG110M 1.00E-12 0.001% l.OOE-12 0.001% 4.21E-08 0.00% 4.21E-8 O.OOE+O O.OOE+O TE125M 0.000% O.OOE+OO O.OOO"A; O.OOE+OO 0.00% O.OOE+O 8.09E+2 O.OOE+O TE127M 0.000% O.OOE+OO 0.000% O.OOE+OO 0.00% O.OOE+O l.76E+3 O.OOE+O TE127 0.000% O.OOE+OO 0.000% O.OOE+OO 0.00% O.OOE+O 1.41E+l O.OOE+O TE129M 3.99E-11 0.034% 3.99E-11 0.038% 1.68E-06 0.04% l.68E-6 3.94E+3 l .SlE-3 TE129 0.000% O.OOE+OO 0.000% O.OOE+OO 0.00% O.OOE+O 1.44E-5 O.OOE+O TE131M 9 .55E-11 0.081% 9.SSE-11 0.090% 4.02E-06 0.09% 4.02E-6 7.52E+2 6.87E-4 TE131 0.000% O.OOE+OO 0.000",4; O.OOE+OO 0.00% O.OOE+O O.OOE+O O.OOE+O TE132 9 .82E-12 0.008% 9.82E-12 0.009% 4.14E-07 0.01% 4.14E-7 l.35E+3 l.27E-4 1130 0.000% O.OOE+OO O.OOO"A; O.OOE+OO 0.00% O.OOE+O 4.32E+4 O.OOE+O I 131 2.18E-09 1.864% 2.lBE-09 2.057% 9.21 E-05 2.06% 9.21E-5 1.34E+6 2.80E+l 1132 1.20E-08 10.258% l.20E-08 11.321% 5.07E-04 11.32% S.07E-4 l.26E+l l.45E-3 1133 1.40E-08 11.973% 1.40E-08 13.213% 5.92E-04 13.21% 5.92E-4 l.56E+S 2.lOE+l 1134 8.83E-09 7.538% 8.83E-09 8.318% 3.72E-04 8.32% 3.72E-4 2.SSE-5 2.16E-9 I 135 1.78E-08 15.214% l.78E-08 1.6.790% 7.52E-04 16.79% 7.52E-4 5.73E+3 9.78E- l CS134 3.00E-11 0.026% 3.00E-11 0.028% l.26E-06 0.03% l.26E-6 O.OOE+O O.OOE+O CS136 1.99E-11 0.017% l.99E-11 0.019% 8.39E-07 0.02% 8.39E-7 O.OOE+O O.OOE+O CS137 8.00E-11 0.068% 8.00E-11 0.075% 3.37E-06 0.08% 3.37E-6 O.OOE+O O.OOE+O CS138 7.59E-10 0.647% 7.59E-10 0.714% 3.20E-05 0.71% 3.20E-5 O.OOE+O O.OOE+O BA139 0.000% O.OOE+OO 0.000% O.OOE+OO 0.00% O.OOE+O O.OOE+O O.OOE+O BA140 3.98E-10 0.340"A; 3.98E-10 0.375% 1.68E-05 0.37% 1.68E-5 O.OOE+O O.OOE+O BA141 O.OOO"A; O.OOE+OO 0.000",4; O.OOE+OO 0.00% O.OOE+O 0.00E+O O.OOE+O BA142 0.000% O.OOE+OO 0.000% O.OOE+OO 0.00% O.OOE+O O.OOE+O O.OOE+O LA140 0.000",4; O.OOE+OO 0.000% O.OOE+OO 0.00% 0.00E+O O.OOE+O O.OOE+O LA142 O.OOO"A; O.OOE+OO 0.000% O.OOE+OO 0.00% O.OOE+O O.OOE+O O.OOE+O CE141 2.99E-11 0.026% 2.99E-11 0.028% l.26E-06 0.03% l.26E-6 O.OOE+O O.OOE+O CE1 43 0.000% O.OOE+OO 0.000% O.OOE+OO 0.00% O.OOE+O O.OOE+O O.OOE+O CE144 3.00E-12 0.003% 3.00E-12 0.003% l.26E-07 0.00% l.26E-7 0.00E+O O.OOE+O PR1 43 0.000% O.OOE+OO 0.000% O.OOE+OO 0.00% O.OOE+O O.OOE+O O.OOE+O PR144 0.000% O.OOE+OO 0.000% O.OOE+OO 0.00% O.OOE+O O.OOE+O O.OOE+O ND147 0.000% O.OOE+OO 0.000% O.OOE+OO 0.00% O.OOE+O O.OOE+O O.OOE+O W 187 2.83E-10 0.242% 2.83E-10 0.267% l.19E-05 0.27% l.19E-5 0.00E+O O.OOE+O NP239 7.81 E-09 6.660% 7.81E-09 7.350"A; 3.29E-04 7.35% 3.29E-4 O.OOE+O O.OOE+O L 1.17E-07 100.0% 1.06E-07 100% 4.48E-03 100.00% 4.90E-3 l.SSE+6 5.00E+l 4.48E-3 Page 22 of 23

CALC NEE-323-CALC-004 ENERCON Revised Liquid Radiological EALs NO. 1-----------------l Exceffence-Every proJf!Ct. Every day. per NEI 99-01 REV. 00 The HTD isotopes displayed in this spreadsheet were scaled into the results as they were in the TEDE spreadsheet. Hard to Detect Determination HTD Canel 1gammas Ratio HTD 1gammas Conc2 Fe-55 l.OOE-9 l.17E-07 8.53E-3 Fe-55 4.48E-3 3.82E-5 H-3 l.OOE-8 1.17E-07 8.53E-2 H-3 4.48E-3 3.82E-4 Ni-63 l.OOE-12 1.17E-07 8.53E-6 Ni-63 4.48E-3 3.82E-8 Spreadsheet cases are run for all three monitors at decay times of 2 hours and 36 hours in consideration of EAL entry thresholds that are mode dependent. See Section 2.0 for results. 8.0 Computer Software No computer software was used in this calculation. 9.0 Impact Assessment This calculation is based on "realistic" assumptions for the purpose of declaring EALs, rather than typical conservative "bounding" type design basis analyses. The calculation results are intended to provide order of magnitude setpoints to assist Operations and Emergency Response personnel in determining the state of the three fission product barriers in accordance with NEI 99-01 Rev. 6. Page 23 of23

CALC NEE-323-CALC-004 g ENERCON

        ~x(.~l!~.ncfJ~Ev_~ry prcij~¢~ E~ery c!<?Y.

Appendix A NO. REV. 00 Ingestion Dose: 8.24 mrem

                                                                     +     Boating Dose:       1.79      mrem Resultant Total mRem:            10.0      mrem Monitor:                          GSW                                  RIS4767 Ill                        FGR12 Units          I    3.70E+15       I                   FGR12 Boating/Immersion      ~

L..:.::_J

 -..ara                 Conversion Factor: .__ _ _ __._                                              Reduction Factor:

GJ Monitor c[s: l ,,: *. 23,200*

                                                                                 .' l cps
  *-ra
   '-                  Monitor Efficiency:         l     2.19E+6       l~

10CFRZO Ingestion Age Considerat_ion Factor:

                                                                                                                        ~

L.:._j

  >                                                                       uCi/ml Decay Hrs:   I,   2 . * ... J Volume Consumed:            l '*    500 ; *    :1  ml Dilution Factor:  l   0.20       l Resultant River Gamma Concentration for the Given CPS Reading:

23200 I Ef,S 2.19E+06 I µCi/ml Ef,S I 0.20 -  :*.n"~ I µd/ml Page 1 of 12

CALC a ENERCON gxc':lt~nr:e~~~~.'Y. project E.lie[Y f/qy. Appendix A NO. REV. NEE-323-CALC-004 00 Child Thyroid: 50.0 mRem Monitor:** .. GSW RIS4767 U) I

  -.cra GJ                                             Decay Hours I
                                                               !         2 ODAM ConveB;on Fa<tDr. 3.785E-03
  *-ra Monitor CPS:                         49,100
   >            Monitor Efficiency:                 I 2.19E+6  I cps uCi/ml Dilution Factor:              !   0.20                               Combining Factor:   I 2.271E-01 Resultant Gamma Concentration for the Given CPS Reading :

49100 I~ I 2.19E+06 I uCi/ml

                                                                 ~
                                                                                . I 0.20 Page 2 of 12

CALC NEE-323.:.CALC-004 Cjl ENERCQN Appendix A NO. Sxcelf~!JCe-Eve_ry project, Ev.~ry ct.OY, REV. 00 Ingestion Dose: 9.78 mrem

                                                                                   +       Boating Dose:            0.19      mrem Resultant Total mRem.                      . 10.0      mrem Monitor::                                         GSW.                                        RIS4767 u,                       FGR12 Units           I       3 ,70E+15 FGR12 Boating/lm~ersion     ~

Reduction Factor: ~

  -.c GJ Conv~rsion Factor: ..__ _ _ __.

I . . . . ____. .;.-'-,*1_0_,_,4_-o_o_. -'--'--'--'---'I cps Monitor c[s: ... ra ra Monitor Efficiency: I 2.19E+6 I~ 10CFR20 Ingestion Age Consideration Factor:

                                                                                                                                            ~

l___:__J

   >                                                                                    uCi/ml Decay Hrs:   j   36 Volume Consumed:             I<::       \'so,r' - Iml Dilution Factor:  I  0.20 Resultant River Gamma Concentration for the Given CPS Reading :

10400 I 1* 2.19E+06 I µCi/ml f* I 0.20 Page 3 of 12

CALC NEE-323-CALC-004 Cl I E NEi~C:QN Appendix A NO.

           * 'g~qif!?.~~t:~~"!5!..'Y projec;~...E~eiy r/.t;fi REV.        00 Child Thyroid:  49.9  mRem Monitor:                                                                     RIS4767 u,
  .-c GJ                                                        Decay Hours !        36
  *-.RIRI..                        Monitor CPS:                            14,000 I          Icps
   >                  Monitor Efficiency:                        2.19E+6 uCi/ml Dilution Factor:                  I   0.20                              Combining Factor: I  2.271E-01 Resultant Gamma Concentration for the Given CPS Reading :

14000 I Ef*7 I 2.19E+06 I uCi/ml Ef*7 I 0.20 = \li~$l)ij/$ I ud/ml Page4 of 12 __J

CALC a E N E>ftCbN El " Excel/e11ce,'--;Ev~,x project. Ev1ry <jay. Appendix A NO. REV. NEE-323-CALC-004 00 Ingestion Dose: 8.24 mrem

                                                                     +    Boating Dose:     1.79         mrem Resultant Total mRem:       10.0         mrem RM1997 u,                            FGR12 Units         I 3 _70E+ 15 FGR12 Boating/Immersion                                         ~

L...:.::_J

    -.raa                     Conversion Factor: ..___ _ ___,                                     Reduction Factor:

CIJ

     *-.ra..                 Monitor Efficiency:        I  1.51E+6     !~

10CFR20 Ingestion Age Consideration Factor:

                                                                                                                                                        ~

L..:__J

     >                                                                   uCi/ml Dilution Factor:                                     I  0.20 Resultant River Gamma Concentration for the Given CPS Reading :

16000 I Ef** 1.51E+06 I µCi/ml Et* I 0.20 i.;,;;2.l2E~03'.;;;,, µCi/ml Page 5 of 12

CALC ~* NEE-323-CALC-004 ENERCON NO. 1w f"<eller,ce,...:Every pioi<<t. Every q9y. Appendix A REV. 00 Child Thyroid: 50.0 mRem RM1997 Ill I

 -.c CIJ                                            Decay Hours I ____

ODAM Conversion Factor: l.____2_ ___.__ _ _ _ _ _ _ ____.___ _ ___. 3.785E-03

  *-ta ta I ..

Monitor CPS: .... 33_,,,_80_0_ _ ___.! cps

  >             Monitor Efficiency:                j 1.51 E+6 Icps uCi/ml Dilution Factor:           I    0.20                                  Combining Factor:   I 2.271E-01 Resultant Gamma Concentration for the Given CPS Reading :

33800 I Ef*i I i.slE+OG I uCi/ml Ef*i Page 6 of 12 J

CALC NEE-323-CALC-004 0" ENERCON

           ~~e(!~!lr~-Eve_ry projet;t. Eve!Y <!_ay.

Appendix A NO. REV. 00 Ingestion Dose: 9.82 mrem

                                                                        +    Boating Dose:     0.19        mrem Resultant Total mRem:         10.0        mrem Monitor:                       RHRS\N/ESW                                 RM1997 u,                        FGR12 Units           I      3 _70E+ 15 FGR12 Boating/lmi_nersion Reduction Factor: ~
                                                                                                                                      ~
  -.rac Conversion Factor: ,__ _ _ __.

CIJ Monitor c[s: 1* * * *

   *-.ra..              Monitor Efficiency:         I       1.51 E+6      I W 10CFR20 Ingestion Age Consideration Factor:
                                                                                                                                      ~

L_:____j

   >                                                                        uCi/ml Decay Hrs:              1. .... 36 .: !

Volume Consumed:  !* soo ..... I ml Dilution Factor:  ! 0.20 Resultant River Gamma Concentration for the Given CPS Reading : 7200 I Ej35 1.51E+06 IµCi/ml E135 I 0.20 .;:*:*9.54E~04

                                                                                                               . *,**.** *.*****,, >I. µCi/ml Page 7 of 12

CALC a ENERCON~xc_t;l!~pre......:..Eve_ry projec;t Eveiy (j_qy. Appendix A NO. REV. NEE-323-CALC-004 00 Child Thyroid: 49.9 mRem Monitor: **RHRSW/ES\1\/ RM1997

                                                                                                                       "° I 3.785E-03 u,
 -.c cu                                                     Decay Hours I

I 36 ODAM Conve"'M RI I .. Monitor CPS: 9,650 8 RI I I Expm*.,Tlme lMln*J,

  >               Monitor Efficiency:                        1.51E+6      cps uCi/ml Dilution Factor:              I   0.20                                    Combining Factor:        I 2.271E-01 Resultant Gamma Concentration for the Given CPS Reading :

9650 I tf*j . I 1.51E+06

                                                                     ** 1 .a,m, tf*j I 0.20
                                                                                                        =

(ti8E~03*****1 uCi/ml

                                                                             - Page 8 of 12

CALC g' ENERCON

        ~xc':t{~f!fe-E':'~'Y proje.ct~ Evpy rJ.ay.

Appendix A NO. NEE-323-CALC-004 REV. 00 Ingestion Dose: 8.22 mrem

                                                                        +    Boating Dose:     1.79      mrem Resultant Total mRem:       10.0      mrem Monitor: * ~tlR~,W Qilu,~iqn, ~ine .                                   RM4268 u,                        FGR12 Units          I     3 .JOE+15 FGR12 Boating/Immersion     ~

Reduction Factor: ~

  -.rac Conversion Factor: .__ _ _ __,

CIJ Monitor c[s: L*

   *-ra
    ~

Monitor Efficiency: I 2.29E+6 I~ 10CFR20 Ingestion ~ Age Consideration Factor: ~

   >                                                                        uCi/ml Decay Hrs:!.: 2

• Volume Consumed: ... I ____,". .;._so_o_ __.l ml Dilution Factor: I 0.20 Resultant River Gamma Concentration for the Given CPS Reading :

24200 I~ 2.29E+06 I µCi/ml

                                                                            ~

I 0.20

                                                                      -Page 9 of 12

CALC NEE-323-CALC-004 QENERCON Appendix A NO.

        ~Cf!ll~!1ce~Eve_ry proje.tt. Eve_ry .d,qy.

REV. 00 Child Thyroid: 50.0 mRem Monitor: , 'RHRSW~Dik1tio11 Line* RM4268 I ___2_ ___.__ _ _ _ _ _ _ ___._I_ _ _ ____. u,

 -.c a,

n, Decay Hours .... I ____51_,_,_30_0_ _ ___.! cps ODAM Conversion Factor: 3.785E-03

  *-n, I ..

Monitor CPS: .... 8 Monitor Efficiency: I Icps Expo,umTime (Mins),

  >                                                   2.29E+6 uCi/ml Dilution Factor: I 0.20                                                    Combining Factor:   ! 2.271E-01 Resultant Gamma Concentration for the Given CPS Reading :

51300 IEfl5 I +oG I uCi/ml Efl5 I 0.20 Page 10 of 12

CALC NEE-323-CALC-004 g ENERCON Appendix A NO. REV. 00 Ingestion Dose: 9.80 mrem

                                                 +    Boating Dose:     0.19        mrem Resultant Total mRem:            10.0        mrem Monitor:

• RHRSW Qilu.ti<>n. ~ine RM4268 UI FGR12una, I 3.70E+15 FGRU Boating/I""."*"°*

Reduction Factor: G

 -.rac Conver,N>n Factor.

GJ D Monitor c[s: 1* . .10,900 *. cps

  *-...ra  Monitor Efficiency: !  2.29E+6 I~

10CFR20 .......... Age Consideration Factor:

  >                                                  uCi/ml Decay Hrs:             I ...36.* I Volume Consumed:    !   . 500. *      *I  ml Dilution Factor:                 ! 0.20 Resultant River Gamma Concentration for the Given CPS Reading :

10900 IEf*i 2.29E+06 I µCi/ml Ef*i I 0.20

                                                                          =
                                                                                .**.:*,*:* .. _:* .* ****,. :*I
                                                                                '. ':.9.52E-04:.'               µCi/ml Page 11 of 12

CALC g . ENERCON

       ~c_~l!~.nce-Eve_ry f?[Ojecr. E~ety rJ..ay.

Appendix A . NO. NEE-323-CALC-004 REV. 00 Child Thyroid: 49.9 mRem Monitor: : -RHRSW 1:>'ilutlor:a Line RM4268 Ill eo....,,** ""' I

  -ns GJ
  .D Decay Hours
                                                              !         36 ODAM                    3. 785E-03
  *-ns Monitor CPS:                           14,650 8
  >           Monitor Efficiency:                  ! 2.29E+6  !uCi/ml cps Expowmllmo(MIM.),

Dilution Factor:  ! 0.20 Combining Factor: ! 2.271E-01 Resultant Gamma Concentration for the Given CPS Reading : 14650 I Ef)5 I +o6 I uCi/ml Ef)5 I 0.20 Page 12 of 12

CALC Attachment 1 NEE-323-CALC-004

 ~ ENERCON   ~xc~lt~_nce-Eve_ry project. Every day.

CALCULATION PREPARATION NO. CHECKLIST REV. 00 CHECKLIST ITEMS 1 YES NO N/A GENERAL REQUIREMENTS

1. If the calculation is being performed to a client procedure, is the procedure being used the latest revision? D D 181 The calculation is being prepared to ENERCON's procedures.

2. Are the proper forms being used and are they the latest revision? 181 D D

3. Have the appropriate client review forms/checklists been completed? D D 181 The calculation is being prepared to ENERCON's procedures.

4. Are all pages properly identified with a calculation number, calculation revision and page number consistent with the requirements of the client's procedure? 181 D D

5. Is all information legible and reproducible? 181 D D

6. Is the calculation presented in a logical and orderly manner? 181 D D

7. Is there an existing calculation that should be revised or voided? D ~ D This is a new calculation to support implementing NEI 99-01 Rev. 6

8. Is it possible to alter an existing calculation instead of preparing a new calculation for this situation? D ~ D

9. If an existing calculation is being used for design inputs, are the key design inputs, assumptions and engineering judgments used in that calculation valid and do they D D 181 apply to the calculation revision being performed.

10. Is the format of the cal_culation consistent with applicable* procedures and expectations? 181 D D

11. Were design input/output documents properly updated to reference this calculation? D D 181

12. Can the calculation logic, methodology and presentation be properly understood without referring back to the originator for clarification? 181 D D OBJECTIVE AND SCOPE

13. Does the calculation provide a clear concise statement of the problem and objective of the calculation? 181 D D

14. Does the calculation provide a clear statement of quality classification? 181 D D

15. Is the reason for performing and the end use of the calculation understood? 181 D D

16. Does the calculation provide the basis for information found in the plant's license basis? D ~ D

17. If so, is this documented in the calculation?

D D 181

18. Does the calculation provide the basis for information found in the plant's design basis documentation? D ~ D Page 1 of4

CALC Attachment 1 NEE-323-CALC-004 g:(iENERCON NO. CALCULATION PREPARATION

          ~xc~l{e!lft?-E~_ry proje.¢.~. Every d_(!Y.

CHECKLIST REV. 00 CHECKLIST ITEMS1 YES NO N/A

19. If so, is this documented in the calculation? D D [8]

20. Does the calculation otherwise support information found in the plant's design basis documentation? D [8] D

21. If so, is this documented in the calculation? D D [8]

22. Has the appropriate design or license basis documentation been revised, or has the change notice or change request documents being prepared for submittal? D D [8]

DESIGN INPUTS

23. Are design inputs clearly identified? [8] D D

24. Are design inputs retrievable or have they been added as attachments? [8] D D

25. If Attachments are used as design inputs or assumptions are the Attachments traceable and verifiable?

[8] D D

26. Are design inputs clearly distinguished from assumptions? [8] D D

27. Does the calculation rely on Attachments for design inputs or assumptions? If yes, are the attachments properly referenced in the calculation?

[8] D D

28. Are input sources (including industry codes and standards) appropriately selected and are they consistent with the quality classification and objective of the calculation?

[8] D D

29. Are input sources (including industry codes and standards) consistent \Yith the plant's design and license basis?

[8] D D

30. If applicable, do design inputs adequately address actual plant conditions? [8] D D

31. Are input values reasonable and correctly applied? [8] D D

32. Are design input sources approved? [8] D D

33. Does the calculation reference the latest revision of the design input source? [8] D D

34. Were all applicable plant operating modes considered? [8] D D ASSUMPTIONS

35. Are assumptions reasonable/appropriate to the objective? [8] D D

36. Is adequate justification/basis for all assumptions provided? [8] D D

37. Are any engineering judgments used? D [8] D

38. Are engineering judgments clearly identified as such? D D [8]

39. If engineering judgments are utilized as design inputs, are they reasonable and can they be quantified or substantiated by reference to site or industry standards, D D [8]

engineering principles, physical laws or other appropriate criteria? Page 2 of4

CALC Attachment 1 NEE-323-CALC-004

~..L ENERCON                                      CALCULATION PREPARATION NO.
         ~XCf!ffence-Eve_ry projeci Every day.

CHECKLIST REV. 00 CHECKLIST ITEMS 1 YES NO N/A METHODOLOGY

40. Is the methodology used in the calculation described or implied in the plant's licensing basis? D ~ D

41. If the methodology used differs from that described in the plant's licensing basis, has the appropriate license document change notice been initiated? D D ~

42. Is the methodology used consistent with the stated objective? ~ D D

43. Is the methodology used appropriate when considering the quality classification of the calculation and intended use of the results? ~ D D BODY OF CALCULATION

44. Are equations used in the calculation consistent with recognized engineering practice and the plant's design and license basis? ~ D D

45. Is there reasonable justification provided for the use of equations not in common use? D D ~

46. Are the mathematical operations performed properly and documented in a logical fashion? ~ D D

47. Is the math performed correctly? ~ D D

48. Have adjustment factors, uncertainties and empirical correlations used in the analysis been correctly applied? ~ D D

49. Has proper consideration been given to results that may be overly sensitive to very small changes in input? ~ D D SOFTWARE/COMPUTER CODES

50. Are computer codes or software languages used in the preparation of the calculation? D ~ D

51. Have the requirements of CSP 3.09 for use of computer codes or software languages, including verification of accuracy and applicability been met? D D ~

52. Are the codes properly identified along with source vendor, organization, and revision level? D D ~

53. Is the computer code applicable for the analysis being performed? D D f8I

54. If applicable, does the computer model adequately consider actual plant conditions? D D f8I

55. Are the inputs to the computer code clearly identified and consistent with the inputs and assumptions documented in the calculation? D D f8I

56. Is the computer output clearly identified? D D f8I

57. Does the computer output clearly identify the appropriate units? D D f8I Page 3 of4

CALC Attachment 1 NEE-323-CALC-004 lf!I ENERCON

 ~
              ~XCf!l[~~~~~Eve_ry proje_i:t: E~ery d,qy.

CALCULATION PREPARATION NO. CHECKLIST REV. 00 CHECKLIST ITEMS 1 YES NO N/A

58. Are the computer outputs reasonable when compared to the inputs and what was expected? D D ~

59. Was the computer output reviewed for ERROR or WARNING messages that could invalidate the results? D D [gl RESULTS AND CONCLUSIONS

60. Is adequate acceptance criteria specified? [gl D D

61. Are the stated acceptance criteria consistent with the purpose of the calculation, and intended use?

[gl D D

62. Are the stated acceptance criteria consistent with the plant's design basis, applicable licensing commitments and industry codes, and standards?

[gl D D

63. Do the calculation results and conclusions meet the stated acceptance criteria? [gl D D

64. Are the results represented in the proper units with an appropriate tolerance, if applicable?

[gl D D

65. Are the calculation results and conclusions reasonable when considered against the stated inputs and objectives?

[gl D D

66. Is sufficient conservatism applied to the outputs and conclusions? [gl D D

67. Do the calculation results and conclusions affect any other calculations? D [gl D

          .*                                                                         ~

68. If so, have the affected calculations been revised? D D [gl

69. '. Does the calculation contain any conceptual, unconfirmed or open assumptions requiring later confirmation? D [gl D

70. If so, are they properly identified? D D [gl DESIGN REVIEW

71. Have alternate calculation methods been used to verify calculation results? D D [gl No, a Design Review was performed.

Note:

1. Where required, provide clarification/justification for answers to the questions in the space provided below each question. An explanation is required for any questions answered as "No' or "N/A".

Originator: Jay Bhatt 12/12/17 Print Name and Sign Date Page4 of4

NEE-323-CALC-004 Attachment 2 Page 1 of 3 PCP 8.7 ALARM SETPOINTS FOR LIQUID RAD Rev. 17 MONITORS Page 9 of 11 ATTACHMENT 1 Page 1 of 3 LIQUID EFFLUENT RADIOACTIVITY MONITOR SETPOINT

  "- 1.        Sample No.        \ I;, - 'J\99;          /         .       2. Sample Date & Time l?.-l-tS     (u   ,0'1
   '\. 3.      Stream/Monitor Description GS'uJ               (9i    ~~~ ~Y)')- 4:]4,1 v
   "'-4.       Effluent Monitor Reading (cps)___,....o___.._/__~.___.,- - - - - - - - - - - - -
  '"-5.        Effluent Flow (gpm)               ca , * *                    *                  *
   ', 6.       Average effluent flow during time* represented by sample, F1 (gpm)                  'N (:(       ,/
   "- 7.       Average dilution (discharge canal) flow durin~~~epresented by sample, F2 (gpm) N<<
  "' 8.        Monitor calibration factor, g, {cps/µCi/mL) ~ * "2,19                   e~ v *
 "--.9.        Previous alarm value setpoint (cps).__--,,..i.;~-$___,.c""'f=-$__,V:::"'--------------
"- 10.         Fraction to apply as a safety margin,                    A = 0.5 Setpoint =  lOx[!.,(K"i,K,WEC;) xg xFiFz xA]+Bkg
                                          +
                                                                               =    Setpoint=l ox[(lS)(S)(?) x(l O)]+(4)

(16)(6) 1 Setpolnt = sx[(lS)(S)(?)J + (4) (16)(6)

                        = sx[( C:,.Y,3 ~-~ )( :2,l<t e~

Setpoint C 117 , 5 L\ )(

                                                                )(     l\f..
                                                                     'Nvt.
                                                                                 ; J+ (      )0 -         )

-~ 11. Setpoint = --=9_9..._2____/ Fractional Change = New value - Previous Value = ( 11 ) - ( 9 ) = ( 5E\ 2. )-( r<eS: ) Previous Value ( 9 ) ( -, ~ 5' )

 -,, 12. Fractional Change     =.-       o  '2-2.~ *v Ufjractional change is greater than +/-0.3, adopt a new monitor alarm settln(;l.

Continuous Monitor Hi Alarm = Setpoint

13. Monitor Hi Alarm = --Zft> "5 cps /

14. Radwaste Monitor Hi Alarm::: .16 (11) = .16 ( ) = N/A cps,/

24

NEE-323-CALC-004 Attachment 2 Page 2of 3 wlo lfosz t7~ 'I PCP 8.7 ALARM SETPOINTS FOR LIQUID RAD Rev. 17 MONITORS Page 9 of 11 ATTACHMENT 1 Page 1 of 3 LIQUID EFFLUENT RADIOACTIVITY MONITOR SETPOINT

1. Sample No. IS-- 'f~h . 2. Sample Date & Tir.pe g'-zq-lS"'/ e:,c,-r:.7

3. Stream/Monitor Description ifm-/917 C~1MSw/ &SLJJ J

4. Effluent Monitor Reading {cps)__,....,.-=-g~c,=-------------------

5. Effluent Flow {gpm) 9x@

6. Average effluent flow during time represented by sample, F1 {gpm)_./...,i/.=-<%;-'-'A;...__ _ _ -=---

7. Average dilution {discharge canal) flow during time represented by sample, F2 (gpm) nl/A

8. Monitor calibration factor, g, (cps/µCi/ml),....---,-,-=l.=5...,_/...,,f._~_ _ _ * ----------

9. Previous alarm value setpoint (cps)_ _ _ b_t_l./_ _ _ _ _ _ _ _ _ _ _ _ _ __

10. Fraction to apply as a safety margin, A = 0.5 Setpoint=l ox[(l S)(S)(?) x(l O)] +( 4)

(16)(6) . Setpoint = sx[(lS)(S)(?)J+ (4) (16)(6)

                                     -'2-
                          /,/ ()l!,,
           *1.
                * [c . ~

C ~

                                            /,Ste" >< vlA Set,>oint = Sx _:__.::::..;:_.....:---~---~------=-
                                                    )(         ,v/A-
                                                                                ')] + (
                                                                                )
                                                                                                            )

15'3* "2 '/

11. Setpoint= ~ 5b'/,'35" . t;6'1, 9 ~v1---1f Fractional Change= New value - Previous Value = ( 11 ) - { 9} = { *Jd) -{ 6,I'-/ )

Previous Value ( 9 } ( h /q )

12. Fractional Change=-~--~-*--

                                   ~   . ~ - o .. o-,

If fractional change is greater than +/-0,3, adopt a new monitor alarm setting. Continuous Monitor Hi Alarm = Setpolnt

13. Monitor Hi Alarm = &2 I ~

14. Radwaste Monitor Hi Alarm = .16 (11} = .16 ( ) = ~M-~_ ___;cps 33

Nfi:E-323-CALC-004 Attachment 2 Page 3 of 3 ALARM SETPOINTS FOR LIQUID RAD PCP 8.7 Rev. 17

                                                                                                                                    \J MONITORS                                                                              Page 9 of 11 ATTACHMENT 1                                                    Page 1 of 3 LIQUID EFFLUENT RADIOACTIVITY MONITOR SETPOINT

1. Sample No. / 4- -enb 4- 2. Sample Date & Time *2 -\4-\A-- / oo "2.. l

3. Stream/Monitor Description 12-\+'2..S~ /f-51.N R,u.,~T'-<CZ...L ! R\N\ 4'2..b'o *

4. Effluent Monitor Reading (cps)

• o .

5. Effluent Flow (gpm) 12\,\JZ...S,.j 'A', *'itxx:;,'-iC:='= 1 ~~~J ' ~- 4-t!JOvc-~ *

6. Average effluent flow during time represented by sample, F1 (gpm) N JA- r.

7. Average dilution (discharge canal) flow during time represented by sample, F2 (gpm) ,J jw

8. Monitor calibration factor, g, (cps/µCi/mL) __'---=--'Z:.-~ ...-'-_~__ v____,__________

9. Previous alarm value setpoint (c.ps}_ _ _ _ ~....:<.:,:....:~:;;;.._----'v"-------------

10. Fraction to apply as a safety margin, A = 0.5 Setpoint=l ox[ (l(16)(6)

S)(S)(?) x(l O)] +( 4) J Setpoint = sx[<l S)(S)(?) + ( 4) (16)(6) Setpomt = 5x [/7,&7f-'">)( 2.2."1~ ~X /-J/ .4-

                                                                               )
                                                                                 >] + (                        )
                          < U I

• 8 <o >< I tv P.

11. Setpoint= _ _ _~OS _ _ __v' Fractional Change= New value - Previous Value= ( 11 ) - ( 9) = ( ~o7) - ( ~'°*~ )

Previous Value ( 9 } ( e, \D 7)

12. Fractional Change= *-o, 0*7 /

If fractional change is greater than +/-0.3, adopt a new monitor alarm setting .* Continuous Monitor Hi Alarm= Setpoint *:::t:>oLb, s~-c (>ct,...,-r V

13. Monitor Hi Alarm = B0 ~ ~

14. RadwasteMonitorHiAlarm=.16(11) = .16( ,..,jA-) = ,J /1-¥ cps

Development of EAL Threshold values from NEE-323-CALC-002 Due to elevated background radiation levels on these monitors during plant operation ( 10-12 R/hr), the calculated threshold va lue was rounded to 5 (minimum serviceable threshold value accounting for scale of monitor) for ease of use by the EAL evaluator, and the "in Mode 5 only" caveat is added to the EAL usage. The resultant EALs are: RA2.2 Reading greater than 5 R/hr on ANY of the following radiation monitors (in Mode 5 only):

• NW Drywell Area Hi Range Rad Monitor, RIM-9184A
• South Drywell Area Hi Range Rad Monitor, RIM-91848 CS1/CG1 Core uncovery is indicated by ANY of the following:
• Drywell Monitor (9184A/B) read ing greater than 5.0 R/hr I..
• CALC NO. NEE-323-CALC-002 g ENERCON

          ~Cf:llepCe_:._Every project. ~ve.ry day CALCULATION COVER SHEET                        REV. 00 PAGE NO.       1 of 28 Dose Rate Evaluation of Reactor Vessel                               Client:           Duane Arnold Energy Center

Title:

Water Levels During Refueling for EAL Thresholds Project Identifier: NEE-323 Item Cover Sheet Items Yes No 1 Does this calculation contain any open assumptions, including preliminary information, that require confirmation? (If YES, identify the assumptions.) D ~ 2 Does this calculation serve as an "Alternate Calculation"? (If YES, identify the design verified calculation.) D ~ Design Verified Calculation No. _ _ 3 Does this calculation supersede an existing Calculation? (If YES, identify the design verified calculation.) D ~ Superseded Calculation No. Scope of Revision: Initial Issue Revision Impact on Results: Initial Issue

                                                                                        '*a Study Calculation      D           Final Calculation    IZI Safety-Related  D        Non-Safety-Related      IZI (Print Name and Sign)

Originator: Jay Bhatt Date: 12/12/17 Design Verifier1 (Reviewer if NSR): Caleb Trainor Date: 12/12/17 Approver: Aaron Holloway Date: 12/12/17 Note 1: For non-safety-related calculation, design verification can be substituted by review.

CJ ENERCC>N

      ~c~ilerce-fve.ry projecr. fvery day.

CALCULATION REVISION STATUS SHEET CALC NO. NEE-323-CALC-002 REV. 00 CALCULATION REVISION STATUS REVISION DATE DESCRIPTION 00 12/12/17 Initial Issue PAGE REVISION STATUS PAGE NO. REVISION PAGE NO. REVISION All 00 APPENDIX/ATTACHMENT REVISION STATUS APPENDIX NO. NO.OF REVISION ATTACHMENT NO.OF REVISION PAGES NO.

• NO. PAGES NO.

A 1 00 1 5 00 8 2 00 C 1 00 Page 2 of28

a EN.E*RCON

         &ceiteni:e,-Every profecr. ¢very d_ay. TABLE OF CONTENTS CALC NO.

REV. NEE-323-CALC-002 00 Section Page No.

1.0 Purpose and Scope

4 2.0 Summary of Results and Conclusions 4 3.0 References 5 4.0 Assumptions 6 5.0 Design Inputs 8 6.0 Methodology 13 7.0 Calculations 14 8.0 Computer Software 27 9.0 Impact Assessment 28

                                                                                   #of List of Appendices                                                                 Pages Appendix A - Electronic File Listing                                          1 Appendix B - DAEAL.xlsx Sheets                                                2 Appendix C - SCALE Input                                                      1
                                                                                   #of List of Attachments                                                                Pages Attachment 1 - Calculation Preparation Checklist                              5 Page 3 of28

Dose Rate Evaluation of CALC NO. NEE-323-CALC-002 a ENERC:0.N

              &,cellem;e_~Eyery ~ro}¥ci, '!,veiy,day, Reactor Vessel Water Levels During Refueling for EAL REV.                           00 Thresholds

1.0 Purpose and Scope

The purpose of this calculation is to evaluate dose rates with water at the top of active fuel in the reactor vessel during cold shutdown or refueling operations in order to set Emergency Action Level (EAL) thresholds (RA2, CS1, CG1) per NEI 99-01 [Reference 3.5]. The dose rates are calculated at the locations of the drywall monitors 9184A/B so that dose rate measurements by these devices can be correlated to the water level in the core, upon failure of other water level detection systems. This calculation is nonsafety-related as the results of the calculation do not affect the design basis or safety-related systems structures or components. These results are best estimates based on as-built conditions and provide information to operators with respect to classifying an emergency, therefore no acceptance criteria is required. 2.0 Summary of Results and Conclusions The dose rates just prior to the core being uncovered (i.e. water at the top of the active fuel) are shown in the table below. Note that the results presented below are calculated dose rates and do not account for background radiation or any installed detector check sources. Table 1 - Dose Rate at Top of Active Fuel ,,, Model Description Drywell Monitor Drywell Monitor Drywell Monitor 9184A Reading 9184B Reading (9184A/B) Range (R/hr) , , , , , ,', @/hr) .. , ., , ., ',.,. (Rlhr)

 ,t;;;~:c1:- ~::;IJfau*'ort:;";;,:,;J~~;;,',?;0'.'*;: ' ;l::;::~:f!l;t~1;~J;:'.;;"'1~;:;;r~:l~'1;:~>:;~!;,f;~;~;:tJit,'§&'*;:;Y;':'.;;,~~'.}t) f8:;;,~f~;.;;;,r;~1-;fo;1E*1:~~t~J~ir:9,i 1

1 Head On 1.11 7.41E-01 I to IE+? 1 This value is off scale low. Page4 of28

Dose Rate Evaluation of CALC NO. NEE-323-CALC-002 Reactor Vessel Water Levels _ _ _ _ _ _ _ _ _ _ __ OENERCON

       . ~iellence.-Eve,y prp}ecr. §very day,  During Refueling for EAL   REV      oo Thresholds

• 3.0 References 3.1 "Standard Composition Library," ORNUNUREG/CSD-2N1/R6, Volume 3, Section M8, March 2000 3.2 CGDG-SCALE-6.1.2, Rev 00, Commercial Grade Dedication SCALE Version 6.1.2 3.3 CGDG-MCNP6-V1 .0, Rev 00, Commercial Grade Dedication MCNP6 Version 1.0 3.4 ANSI/ANS 6.1.1-1977, Neutron and Gamma Flux-To-Dose Conversion Factors 3.5 NEI 99-01, Rev. 6, "Development of Emergency Action Levels for Non-Passive Reactors" 3.6 I.RIM-V115-01, Rev. 10, "Victoreen Model 876A Containment Radiation Monitor Calibration" 3.7 NUREG 1940, "RASCAL 4: Descriptions of Models and Methods" 3.8 CAL-ROO-PUP-008, Rev. 03, "Non-LOCA Radiological Consequence Dose with Alternate Source Term" 3.9 RFP 110, Rev. 45, "Refueling Procedure- Reactor Pressure Vessel Disassembly" 3.1 OTechnical Specifications, Section 1.1 3.11 Technical Specifications, Section 4.2.1 3.12 NUREG 1754, "A New Comparative Analysis of LWR Fuel Design~"

3.13 BECH-M009, Rev. 14, "Equipment Locations Reactor Building Section-GG" 3.14 BECH-C405, Rev. 14, "Reactor Building Floor Plan@ El. 757'-6'."' 3.15 NG-17-0156, Proprietary Data Transmittal to EN ERCON . 3.16 BECH-M405, Sh 04, Rev. 24, "Instrument Points and Rack Locations Diagram Plans at Elevs 812' -0" & 833' -6"" 3.17 NG-88-0966, "G.E. Fuel Damage Documentation/Dose Rate Calculations" 3.18 C003-029, Rev. 0, '.'Drywell Cylindrical Shell & Cone" 3.19VS-01-06, Rev. 4, "Top Head Assembly" 3.20 BECH-C511, Rev. 5, "Reactor Building RPV Ped Dev. Elev. & Sect's" 3.21 BECH-C514, Rev. 1, "Drywell Interior Biological Shield Wall Reinforcing Sections" 3.22BECH-C-516, Rev. 6, "Drywell Interior Biological Shield Wall Plans El. 816'-3 %" to El 779' -1 %"" 3.23 BECH-M405, Sh 02, Rev. 71, "Instrument Points & Lines Diagrams Plan at Elev 757'-6"" 3.24 APED-B-31-2816-001, Rev. 5, "Outline Reactor Recirculating Pump" Page 5 of28

Dose Rate Evaluation of CALC NO NEE-323-CALC-002 lyl ENERCON Reactor Vessel Water Levelsi----*- - - * - - - - - - - - - 1

       ~iell~n~-F'(ery Aro}~Ci. ~V':,y day. During Refueling for EAL Thresholds           REV.         00 3.25 FSAR Section 4.3.2.1, and Section 9.1 3.26 CAL-M98-058, Rev. 1, "ADS Accumulator Size Verification" 4.0 Assumptions 4.1 The core is homogenized based on the typical 1Ox1 Ofuel assembly dimensions, taking into account the fuel rods and space between. Any small variations in fuel parameters will have a negligible effect on containment dose rates. The cladding is modeled as Zircaloy 4 in lieu of ZIRLO; this is acceptable due to the similarity of the materials.

4.2 Any non-fuel hardware, including rod end plugs, is ignored in the active fuel region. This is acceptable since the primary self-shielding occurs in the fuel itself, and there may be some unknown streaming effects through the non-fuel hardware. This homogenization takes into account the presence of water when calculating the isotopic weight fraction and homogenized density. For the case with the reactor vessel head in place, the region between the head and the active fuel region is homogenized based on the actual mass of the upper internals over the entire region. Homogenization of source regions and shields is acceptable due to the insignificant effects on the detector response given the model geometry. 4.3 The composition of the containment structure and components are based on the values in the SCALE standard composition library [Reference 3.1]. These material properties are commonly used in shielding applications, and are acceptable for modelling the structures and compon~ts used to determine the

        ., best estimate response at the detector loc*ations. 

4.4 The minimum period of decay after reactor shutdown before moving fuel is 60

• hours [Reference 3.8, Section 4.3.8]. This calculation assumes a decay time of 50 hours to allow EAL thresholds to be determined for reactor vessel conditions that exist prior to the commencement of fuel movement which is representative of the applicable operating modes (cold shutdown, refueling). This decay time is appropriate to produce best estimate results for both the head on and head off configurations.

4.5 The hardware in the upper internals region between the active fuel region, reactor recirculating pumps and reactor vessel head is assumed to be stainless steel type 304. While the actual composition of the hardware may vary slightly, small variations in the material will have a negligible effect on the dose rate response at the detectors. 4.6 It is assumed that the water below the active fuel region is liquid at a constant temperature. Using a density of 0.9982 g/cm 3 is common in shielding Page 6 of28

Dose Rate Evaluation of CALC NO. NEE-323-CALC-002 0 ENERCQN

    ~celle,:i~e--~~t!TY pro}e,cr. }v,;iy doy,_

Reactor Vessel Water Levels During Refueling for EAL REV. 00 Thresholds applications. Any water above this region would be steam with little shielding value.

4. 7 The source term is generated shortly after shutdown, therefore, the fuel gamma source term will predominate and the neutron-gamma and hardware activation can be neglected.

4.8 The high range detectors read out in roentgen per hour (R/h) which is a measurement of exposure rate, while the MCNP output is provided in mrem/h

     , which is a measurement of the equivalent dose rate that represents the biological effects of ionizing radiation. It is assumed that 1 R is approximately 1000 mrem.

This is acceptable as only the gamma source term is considered. 4.9 The roof of the Reactor Building is modeled as 0.5 inches of stainless steel. This will account for any scattering interactions that may contribute to the response at the detector. The magnitude of the detector response due to scattering off of the roof will be small due to the geometry and amount of shielding in the model, and is therefore acceptable. 4.1 OAutomatic Depressurization System Accumulators 1R003A/B/C located on the 775' -11 %" elevation are not included in the model. The size of the accumulators are 200 gallons [Reference 3.26]. This is relatively small compared to the geometry of the model, and the corresponding scatter interactions will not have a significant impact on the detector response'. Page 7 of28

Dose Rate Evaluation of CALC NO. NEE-323-CALC-002 C) EN.ERC.q>N Reactor Vessel Water Levels During Refueling for EAL Excelterce,-Every wo}~ct. §vefy ~oy. REV. 00 Thresholds 5.0 Design Inputs 5.1 Fuel Assembly Parameters The following fuel assembly parameters are used to homogenize the core in the MCNP model. They are based on typical fuel-assembly values for 1Ox1 O fuel. Table 2 - Design Input Fuel Assembly Parameters Parameters Value Unit Reference

 # of Assemblies in Core                                              368                                             3.11 Clad thickness                                                      0.026              [in]                          3.12 5.2 Model Dimensions The following elevations and dimensions are based on the associated drawings or other reference. Some parameters are estimated using drawing scales when exact dimensions are not provided.

Table 3- Design Input Dimensions Dimension ft in cm Reference

~111~~~~t;;J~!~:11~r ~~4ffi~:r:~

3 Pedestal outer radius 12 365.76 3.20 Reactro vessel thickness 5 12.70 3.15 Concrete around drywell spherical 36 9 1120.14 3.14

~sU~i~~;;!~1~~!$[iJ:ii~1,f~:s)::t,{ kf,¥:14':~r1~i,~,}J:;;~'0::i~ri'~'l,if":1~i('.:;J~~t~~t§)t~~~*~{);;,${;}2\';,:*~~:~0!'.~A<>:.':};;2~i~~l 4

Concrete around drywell cylindri-22 9 693.42 3.16 li~i~~iij;iji~,~~~~~~~~,:~r , . Reactor Roof Thickness Vessel Height 704.5 1789.43 ).15 Page 8 of28

Dose Rate Evaluation of CALC NO. NEE-323-CALC-002 It) ENERCON Excellence-Every projecr. Every day. Reactor Vessel Water Levels During Refueling for EAL REV. 00 Thresholds Dimension ft in cm Reference j Reacto1:*Y~~s~l_head thickn~ss . .. ,.. ,.. .. . }~9375. . .. !_0,:QQ....... . 3.19 Distance from vessel Oto bottom of 200.94 510.39 3.15, 3.12 active fuel i)iii ;hieiciinne~ racifus: -_ 9 3.21. Bio shield outer radius . 11 8.25

                                                                                                     -.--*-.,-,,~- - -

356.24 3.21 i_ Reactor .r,e~frclilating pulrip-_heigitt 17 2 523.24 .*. 3.24 S n*S', *' ' * ~N-, .*.-.n*w~--****a*.* ***** * -~- "" ~ Reactor recirculating pump radius 2 9 83.82 3.24 1 DetectorllE-9184.A,distance froJn .

                                                                               -4.                                             -121.92 *.
,_origiri_(x pla.ne)_ ___ .. ____ :. - : _ .

Detector RE-9184A distance from 13.33 406.29 3 .23 [Scaled] 0 fi.zj!!JiPl~~!l ... *.,., *.,

• 7_ * * -.- * **:

i Detector*RE-9184B distance from - .. 182.88 . * **3.23. [Scaled]:

!.~rigt.._ (:{~!~9-~L ______ . _____ ; _-__ /:~ ... _*                           6
                                                                                                                             . -      ......... - .. : - *-- :~--- --*                            '

Detector RE-9184B distance from - 12 _365 _76 3 .23 [S ca1ed] _origi.n(~plane)_ ___ *-*-------.--***--"" ..... *-- ............. ,. ***---**** ,......, .... .

; Re*actor R~circuJating Pump:i;p,. *-**                                                                                   .. 365.76'-

12 3 .2}. [~caled] .* [_79!A.4!~~!!c~fr..~!!! Qt!giJ!J~_-pJ~n~l ... Reactor Recirculating Pump IP-12 365.76 3.23 [Scaled] 201A disfancefrom origin (yplane) i Reactor l{ecirculating*Pump IP-:- . . .

                                                                              -_12                                             '-365.76.                                       3.23 [Scaled]
\ 201B. di~tithceJrom origin(x pjarie) .*

Reactor Recirculating Pump IP-

                                                                              -12                                              -365.76                                         3.23 [Scaled]

201B distance from origin (y plane) Table 4 - Design Input Elevations2 Dimension: ft. in cm Reference 0;00 _: . . . 3.13

\ D~eiIEqu~to! .... *.. **-*--                             ....... ____        . ?.~-~---******-':Q:t                                .,  *'-""-**-----~-- , __         --*---~*-***-*-,-.,.*--*--*-~--  -

Vessel 0 5.5 195.58 3.15

;Bottom-of pedestai eieva~on - ~- -~ -*~ -.- .- .*                                                                                 * -7095)3,

• 3.13 Top of cylindrical portion of drywell 855 2711.45 3.13 concrete F::i:~F:~rJi~~~f6f~iiiiafrig._* ::; T":<:::'fi:~ > . _**:s~1:: 0

                                                                  ,~;,                                                              . . 4,op$;:?:~ ;'.t:r: *c_;* ~Lrn:. : .*

Detector elevation 760 -184.15 3.17 _Top ofp~d~stalf bottom of j)fo ~1*ield. -* . __ .*_yi_Q_**_-

• 147.32 . - 3.20

.. Top_ of bio shield_ __.... __ . ____ -- ... _                                  816                             3.25                   1530.99                                        3.22
; React9rrecirculating pump botfom                                                748.

• _;_:;8:.5 . 3.13 2 All elevations listed in centimeters are relative to the equator of the drywell elevation of 766' 0.5" [Refer-ence 3.13].

Page 9 of28

Dose Rate Evaluation of CALC NO. NEE-323-CALC-002 0 ENERCON Exiell~,:ice~Ev~ry woJecr. fv'!iY do~ Reactor Vessel Water Levels During Refueling for EAL REV. 00 Thresholds 5.3 Core Isotopic Inventory Core isotopic activities in Ci/MWt are taken from Reference 3.7 Table 1-1. A table of the input values is shown in Table 5, below. The activities in Ci are determined by multiplying by the rated thermal power of 1912 MWt taken from Reference 3.10. Table 5 - Core Source Term Isotope Ci/MWt Ci Isotope Ci/MWt Ci

• • ***c. ,*:: .., * * ,:,* ';"v: c:**c**** ,* C*\l' .. : * ..* ;:'/" ***** * >.**. ***"!*''.*'* >"",, **:; . * ,-.;:*>*:** *cc.:* . .

                 . Bat13~ ... 4.74EfJ)4 .. 9;06E+07'":, . Rh.:.lOS* . *2L81E+04 p:J7E:t07.

Ba-140 4.76E+04 9.I0E+07 Ru-103 4.34E+o4 8.30E+07

                 -~Ce:141 .                 4.39£+04 _ B.39E+07 <.Ru-105 . *. ~:b6E+o4 ..5~85Et07.

Ce-143 4.00E+04 7.65E+07 Ru-106 1.55E+04 2.96E+07

                 ,e~5f:f~::.*.:~:,s1~fo1?:~:1.1:e+ow:                                     ~<~~~.;121:; 1*2fa9g+o3\::ra~sY~fg§.

• Cm-242 1.I2E+03 2.I4E+06 Sb-129 8.68E+o3 1.66E+07

                 .:Cs~134 *~4.70E+O:{ _8:99E+06; _Sr-89_ ; _2?41E+Q4 .._4:§fl~:+j}7.

Cs-136 l.49E+03 2.85E+06 Sr-90 2.39E+o3 4.57E+06

                'ei~1~t . . :               ~:2JE£Q:t,i'6:'.flt:+06:,                     X;\Sr~~,i '            5(Q{Ej:04_*~:s/t,§Ef07 .

1-131 2.67E+04 5.11E+07 Sr-92 3.24E+04 6.I9E+07 {Jfa2 : .. _3:s*~E+64 ,,:-7.42E+Q12 *,*Tc-:99fu:: :,4:37Ef04 . 8.36E{07. 1-133 5.42E+04 l.04E+08 Te-127 2.36E+03 4.5IE+06 _1-:!~4: . * *. 5.98~+~4 ..l.14~t08 ..t~12'1in *.3;97E+o2_i59E+OJ .. 1-135 5.I8E+04 9.90E+07 Te-129 8.26E+03 l.58E+07

                  *:t{fis3m' >              3.osf:{of.*              s:sJe+o~-~ :r~)'29Iri*                      'WnlfsE+<fr                \::3:2,fE*Q6.,

Kr-85 2.78E+02 5.32E+05 Te-131m 5.4IE+o3 l.03E+07 I(r.Zssm. 6:l7E*03_LI8E+/-Q7 _-cTe-:132 :_:)_)iE+04_i28E+07 _ Kr-87 l.23E+04 2.35E+07 Xe-131m 3.65E+02 6.98E+05

iq~s~.: i' *.*:l~7£):iifo4:~::,.3,;2,E+oii ~tx~-133 :$.~3Eio4 .,,f.!>~E+::os/

La-140 4.9IE+04 9.39E+07 Xe-133m 1.72E+03 3.29E+06

                    'l,~~141 __4:33E+Q4 : _8.28E+d'7* ', *Xe-135 l.A2E+o4 2:72E+07 La-142 4.2IE+o4 8.05E+07 Xe-135m l.I5E+04 2.20E+07 Mef.:.99_: __5,3QEfQ4/J,01E~rt8:*~>X:tt13S*: . . 4{s6E+o4.c-_S.*7ZEfo7__

Nb-95 4.50E+04 8.60E+07 Y-90 2.45E+03 4.68E+06

• Nd~147 : **1.75E+o~f-.3.35E+07, *Y.:.91.' :*:':!t11E+o4**_6'.b6Et07. >*

Np-239 5.69E+05 l.09E+09 Y-92 3.26E+04 6.23E+07

                 .Jlr;;:143 _:*_3.96Et04 7,57E+q7:. *::: ..Y-93 . ..22s*2E+04 **4~82E+07.

Pu-241 4.26E+03 8.I5E+06 Zr-95 4.44E+04 8.49E+07

                  .R1>is6:: \: :t29Ei . >J;6 fE4':os:~ :; it~'>"h* '\~:'.i3Et04 : :s:o9E/o'i:

Page 10 of 28

Dose Rate Evaluation of CALC NO. NEE-323-CALC-002 g ENERCON Excellence-Every projecr. ~veiy do)( Reactor Vessel Water Levels During Refueling for EAL REV. 00 Thresholds 5.4 Material Compositions The following compositions used in the MCNP model are taken or developed from the SCALE standard composition library [Reference 3.1] and are shown in Table 6. Table 6 - Scale Standard Compositions used in MCNP Model Material Isotope Weight Fraction

                                                                                                                                  * . :Zr: . . . *();9823
                                                                                                                                     .....:*: -;,,;. -' .** ,-, ' .. \ ,,..,   ~  '  * ,'.' *.,-.* -'L* ,_,,,. ~"'<:~: .. \ ,_* ,; . .: . ,,

J~~?~ g~~~31-. . .. ..... . . . -~~--- . .. . .. . ..2:~!1_~

                                                                                                                                           .Ct

__ ,.. ______ ---*--** ______ ___ 0.0010

                                                                                                                                                                                              ._      ..     *--~-'-***

Fe 0.0021

                                                                                                                              ... Hf.                                                *. 0:0001 U02                                                                                             U-235                                                   0.0348 (10.~!2-g(~~!Y . * " .. *:*.*{EI1r**                                                                                                ...                o:Sfl66 0                                              0.1186

_Air _____ ._ . __ .... ........f. _.... __ O.QQPJ 3 (1.21E-03 g/cm

                                   \,- '\,' "",'. __ ,,,_ ""'",* .,.., ~\'"\~ 't' "\,'* \';",""'.'.:**
                                                                                                                   )
                                                                                                                  >>'*W* "'

• N 0.7651 Water H

                                                                                                                              ~*-*****--*~""-- *-                      -          --**

0.1111

                                                                                                                                                                                             ._..~-~-------**.-*....- - ....

(0.99s2 gtcm3)

• 0. *o.s889 .

SS-304 Fe 0.6838 (7.94 g/cm~l .. *0:1900 ,. *.*

                                                                                                                                                                                        -* *,","' .* ' *.* '4-., *-*~.-... *** ***         *.,

Ni 0.0950

                                     '"                .   '       .,.,     ,       .*.                * ..            - . -*-Mn:* .....                                                      0.0200.
                                                  *   *   ,o*    ,, .... * * * * * * * * - . * * * ' -   ***-***M***,-*

Si

                                                                                                                            ,*****,******,-----*-**-****            *H       *******

0.0100

                                                                                                                                                                                      --H**--~--=*~*-,**-************                    -
                                                                                                                                         .C

• 0.0008 P 0.0004 Concrete.*

                                     -~-......---~'-'"****-~*****~- ..---~-* ..... ,

• b ..

                                                                                                                                   '"'" _,,~*-.,...,...,-, ~ ..*. ,. '"' ....        --  ~

0.5320:

                                                                                                                                                                                             --........+-,~~-*-'"-*' - *-*** *-*'

3 (2.30 g/cm ) Si 0.3370

                                      -[KE-N6j~egufar--. .-_-_ ~o~~------                                                                                                 . _ . -~-Q1'!2                                      __ .

Concrete Standard Al 0.0340 dVIixf * :~f. *....:.~~7Q~2Q~~:_-;~~-: Fe 0.0140

                                                                         * ---- *-                                                             Ii * *                               -        *o~oioff*                       * --

Carbon Steel C 0.0100

                                   .(7.82g/crii3-.f<-,---.---,-**:*~.Jie . . . -* 0;9~00."' .....

Page 11 of28

Dose Rate Evaluation of CALC NO. NEE-323-CALC-002 ID ENERCON Excelfen~~-fver; p~jeCr. *~v~iy_doy Reactor Vessel Water Levels During Refueling for EAL REV. 00

• Thresholds 5.5 Upper Internals The following weights are used in the MCNP model for the region between the active fuel and the reactor vessel head [Reference 3.9, Appendix 8.9]:
• The weight of stainless steel for the moisture separator is 83,000 lbs.
• The weight of stainless steel for the steam dryer is 50,000 lbs.

5.6 The drywell (9184 A/8) and torus (9185 A/8) radiation monitor ranges (1 to 107 R/hr) are taken from Reference 3.6. 5.7 ANSI/ANS-1977 Flux to Dose Factors Flux to dose conversion factors are taken from ANSI/ANS-6.1.1-1977 [Reference 3.4] and are shown in Table 7. Table 7-ANSl!ANS-6.1.1-1977 Flux to Dose Factors MeV mrem/hr/(y/cm2/s) MeV mrem/hr/(y/cm2/s)

                                       .* ,_ .. * ...*;:: 3.96E:-03 :;*~? *.. * . . .* O'.$, *. ,;*' .... ;* *._t,§s1*~Q3 * .

0.03 5.82E-04 1 1.98E-03

        *.0.95*-:.;..\>:                                                                                ... '.3247fi;03 .: . *...

0.07 2.58E-04 2.6 3.82E-03 Page 12 of28

Dose Rate Evaluation of CALC NO. NEE-323-CALC-002 ti ENE RCON a!l!l!

         ~iel(e,:i~~~E~~r;. pr9);.cr. ~V~iy day.

Reactor Vessel Water Levels During Refueling for EAL REV .. 00 Thresholds 6.0 Methodology The reactor source terms are decayed to 50 hours with ORIGEN-S of the SCALE 6.1 code package, Reference 3.2. The results are used to bin design input isotope specific activities into energy dependent photon bins. These energy specific photon emission bins are used as input for the energy distribution described by the MCNP source definitions. The MCNP6, Reference 3.3, Monte Carlo transport code is used to determine the dose rates via the flux to dose conversion factors in Table 7, while accounting for shielding and particle transport. The detailed engineering drawings are converted into MCNP surface and cell cards in the dimensions shown in Table 3 and Table 4. The radiation monitors of interest are modeled as point detectors to determine the expected dose rate for those detectors. The dose rates are calculated for two reactor refueling conditions:

1. With Head - the reactor is modeled with a 3.9375 inch carbon steel plate as indicated in Table 3, which is additional attenuation between the source and detector. The mass of the moisture separator and steam dryer is homogenized between the. active fuel region and the vessel head.

2. Without head - the reactor is modeled with air between the active fuel zone and containment.

3. A sensitivity case is run with a mirror surface at the top of the drywall to ensure the modeling of the drywall cap would not significantly affect the response at the. de-tector locations due to scattering.

Variance reduction is accomplished with a geometric importance map that is imposed on the homogenized core. In addition, cell based importance weighting and source biasing (see Section 7.5) are utilized to improve the variance reduction of the simple geometric scheme. A superimposed weight window mesh is utilized where necessary to improve variance. The weight windows are iteratively generated using the MCNP weight windows generator card. All final dose rates presented in this calculation include weight windows variance reduction. Page 13 of 28

Dose Rate Evaluation of CALC NO. NEE-323-CALC-002 0 E.N*., ER C: 0 l\l

       &!_cell~nce~fv~ry pro}ei:r. §ve;y d_oy Reactor Vessel Water Levels During Refueling for EAL                        REV.                00 Thresholds 7 .0 Calculation 7.1   Source Terms The ORIGEN-S input deck, DAECEAL.inp, is provided in Appendix C. This input produces a simple case where the isotopic composition from Table 5 is decayed. The isotope is specified in the 73$$ card using the special identifier described in Section F7.6.2 of the ORIGEN-S manual, and the activity in curies is specified in the 74** card. The time steps for the decay are given on the 60** card in hours. Although multiple time steps are calculated, the source term with 50 hours decay time is used in this calculation to model the core shortly after shutdown. The output of the decay is given in terms of photons/s/Energy-Group, which is automatically normalized in the MCNP input. The results of this calculation are summarized below in Table 8. These values are used in the MCNP input source definition.

Table 8 - Binned Total Core Source Term Energy Group Energy Boundaries (MeV) Photons/sec

                           !;l~;';}-;.,lf~i:'.zi:;1j,~6:~\ic::*:'~;1~j3~::s~i:,,fO~Ol6Q'.!Q$~%f: ,.:sW~~vi';J,; -~~Q~SEft;<X*~~

2 0.05-0.1 6.572E+18 totals 8.37E+19 Page 14 of28

Dose Rate Evaluation of CALC NO. NEE-323-CALC-002 m ENERCQN ~

      *. Excelfen!"_~Every proj~r. §v~iy d.ar.

Reactor Vessel Water Levels During Refueling for EAL REV. 00 Thresholds 7.2 MCNP Model Core Homogenization The source term is given for the entire core, therefore, the self-shielding from the assemblies is an important part of the dose rate response. For simplicity, the core is modeled as a three dimensional cylinder with a uniformly distributed spatial particle distribution. The calculations for determining the mass of fuel, cladding and water for the core and the resulting density are shown below. The inputs are based on the dimensions in Table 2. Assembly Width= (Array Size - 1) x pitch+ Rod OD = (10 -1)(0.51in) + 0.395in

                             = 4.985 in Active Fuel Region Area= (Assembly Width) 2 x Number of Assemblies in Core
                                = (4.985in) 2 x 368 = 9144.883 in2 Active Fuel Equivalent Radius= jActive Fuel Region Area/TC= j9144.883 in 2JTC
                               = 53.953 in Rod Volumeu 02                       = TC(Pellet Radius)2 x Active Length= TC(0.168 in) 2 (144 in)
                                     = 12.768 in 3 Rod Massu 02                  = p x V = (10.412~    g) (12.7682*-

• in3 ) (2.54 cm)3 in = 2178.54 g Number of Fuel Rods Assembly Massu 02 = Rod Mass x A bl = (2178.54 g)(92) ssem y

                                      = 200.43 kg
                                               =*TC (0D2    ID 2) x Active Length Clad Volume 4 -   4 2                2

_ ( ) [(0.395 in) (0.343 in) ] ( * ) _ *

                                               - TC                -                    144 m - 4.34 m 3 4               4 Rod Masszry- 4 = p                 X V = (6.56  !)  (4.34 in 3 ) (2.54   c;;;)

3

                                                                                                   = 466.5 g Page 15 of 28

Dose Rate Evaluation of CALC NO. NEE-323-CALC-002 C, ENERCON Reactor Vessel Water Levels

 . _Ei:cellence_~Every pro)~i:r.'gv~ry day.       During Refueling for EAL REV.              00 Thresholds Number of Fuel Rods Assembly Masszry- 4                            = Rod Mass x         Assem y bl          = (466.59)(92) = 42.92 kg Assembly H2 0 Volume
                                           = [(Assembly Width)2
                                           - n(Rod Radius) 2 x Number of Fuel Rods] x Active Length
                                           = [(4.985 in) 2 - (rr)(0.1975 in) 2 (92)](144 in)= 1955 in 3 Assembly MassH2 O                        = p x V = (0.9982- g) (1955)
                                                                               .     ( 2.54-.-

3 cm) = 31.98 kg cc m3 m Assembly Volume= Active Length x (Assembly Width)2 = (144 in)(4.985 in)2

                                   = 3578.4 in3 Total Mass              1000g/kg(200.43 + 42.92 + 31.98) kg Density=                                       =                                 3         = 4.70 g/cc Volume                     3578.4 in (2.54 ~)

3

                                                                .                   m The corresponding isotopic composition for the homogenized active fuel region is calculated based on the compositions in Table 6. An example calculation for the mass fraction of U-235 is included below.

Assembly Massu 0 2 Mass Fraction U235 = T lM ota ass x weight fraction U235 200.43 kg .

                                                  = (200.43 + 42.92 + 31.98) kg X 0 *0348 = 0*0253 The remaining calculations for the homogenization are done in the worksheet Compositions of the EXCEL workbook DAEAL.xlsx and are shown in Appendix B. The isotopic compositions are calculated with the water level above the top of the fuel. Note that the EXCEL workbook uses additional significant figures.

Page 16 of28

Dose Rate Evaluation of CALC NO. NEE-323-CALC-002 I ENERCON Reactor Vessel Water Levels,.__ _ _ _ _ _ _ _ _ _ __

      &ctlft:nce-Every proJ<<r. Every day.      During Refueling for EAL         REV.        00 Thresholds Table 9 - Homogenization of Active Fuel Region ZAIDNumber Atom                       Mass Fraction Active Fuel Region Homogenized 92235                        U-235                          0.0253 92238                        U-238                          0.6163

• 8016 0 0.1896 40000 Zr 0.1531 50000 Sn 0.0023 24000 Cr 0.0002 26000 Fe 0.0003 72000 Hf 0.0000 1001 H 0.0129 7.3 MCNP Model Upper Internals Homogenization For the case with the reactor vessel head in place, the steam dryer and moisture separator region are modeled as a discrete cylinder with a uniformly distributed homogenized material to account for the mass of stainless steel between the active fuel height and reactor vessel head. The homogenization accounts for the mass of metal from Section 5.5 (assumed stainless steel type 304 per Assumption 4.5) distributed evenly across the volume between the active fuel height (Z=1071.73 cm) and the head (Z=1985.01 cm).

Mass Upper Internals= (83000 lb+ 50000 lb) ( 453.59 ~) = 6.033 x 10 7 g The mass is divided by the volume of the region between the active fuel height and the reactor vessel head to determine the density. Density Upper Internals = Mass Upper Internals+ V

                                       = 6.033 x 10 7 g +  (913.28cm x (rr(235.43cm)2)       = 0.379.§!__

cc 7.4 MCNP Model Geometry The following MCNP model geometry is based on the containment dimensions summarized in Table 3 and Table 4. The model only focuses on the primary systems and components that provide shielding or reflection from the core to the radiation monitors. These components include the reactor vessel, recirculation pumps, pedestal, biological shield and drywell. VISED plots of the model geometry are provided in Figures 1-3. The MCNP surface cards with the model dimensions (cm) are shown in Figure 4, and the cell cards are shown in Figure 5 for the cases with no reactor vessel head. A VISED plot of the model with the reactor vessel head is shown in Figure 6. Areas that are not of interest Page 17 of 28

Dose Rate Evaluation of CALC NO. NEE-323-CALC-002 Reactor Vessel Water Levels,___ _ _ _ _ _ _ _ _ _ __ ENERCON Excellence-Every pro1e-cr. Every day. During Refueling for EAL REV. 00 Thresholds are given an importance of zero (white areas) so MCNP will not track particles in locations that will not contribute to the detector response. Figure 1 X-Z VISED Plot of Reactor Vessel (No Head) Air

                                                              + +-     12.70 cm 1789.43 cm Reactor
                                                                           +                             Vessel Biological Shield Homogenized Core           - -t-.....,..-1-----l-L Water Page 18 of 28

Dose Rate Evaluation of Reactor Vessel Water Levelst-C_A_LC _ N_O_._N_E_E_-3_2_3_-C _A 2-1 _ LC_-_o_o_ ENERCON

       &cellence-Every projec1. Every doy.       During Refueling for EAL          REV.          00 Thresholds Figure 2 Vised Plot of Drywe/1 and Reactor Building3 Reactor Building Drywell Radiation       +----~~+---                                                                         Pedestal Monitor Monitor Pump Pump 3

Radiation monitors are not on the same plane shown above. They are included for visualization purposes only. The VISED Plot was rotated around the Z axis until the Recirculating Pumps were visible. Page 19 of 28

Dose Rate Evaluation of CALC NO. NEE-323-CALC-002 ENERCON Reactor Vessel Water Levels1-----------------1 Excellence- Every projecr. Every doy. During Refueling for EAL REV. 00 Thresholds Figure 3 X-Y Vised Plot of Detectors and Reactor Recirculating Pumps at Elevation 760 '-0" 4 Radiation Monitor RE-9184A 0 ----- Radiation Monitor Pump IP- RE-9184B 201B 4 Detectors are included for visualization purposes only. Page 20 of 28

Dose Rate Evaluation of CALC NO. NEE-323-CALC-002 ENERCON Reactor Vessel Water Levelst------- -----------1 Excellenr:,:-Every projecr. Every do,. During Refueling for EAL REV. 00 Thresholds Figure 4 MCNP Model Surface Cards 5 c surfaces 1 rec 0 0 7 05.97 0 0 365 . 76 137 . 045 $ Active Fue l Region 2 rec 0 0 208 . 28 0 0 17 7 6 . 73 235 . 43 $ Reactor Press u re Vessel Inner Surface 3 rec 0 0 1 95 . 58 0 0 17 89 . 4 3 2 4 8 .1 3 $ Reactor Press u re Vesse l Outer Surface 4 rpp - 1120 .1 4 1 120.1 4 - 1120. 14 11 20 . 1 4 - 1120 . 14 82 1. 86 $ Concrete Sph er port drywell outer 5 so 960 . 12 $ Spher portion of drywe l l outer surface 6 so 958 . 21 $ Spher portion of drywe l l l i ner surface 7 pz - 709 . 93 $ Bottom of Pedestal Elevation 8 rec 0 0 - 709 . 93 0 0 857 . 25 243 . 84 $ Pedestal Inner Surface 9 rec O O - 709 . 93 0 0 85 7. 25 365 . 76 $ Pedestal Outer Surface 81 rec 0 0 14 7 . 32 0 0 1383 . 67 290 . 20 $ Bio Sh i eld Inner Surface 91 rec 0 0 1 47 . 32 0 0 1 383 . 6 7 356 . 2 4 $ Bio Shie l d Ou ter Surface 82 rec 365 . 76 365 . 76 - 528.32 0 0 523 . 24 83 . 82 $ Recirc Pump I P- 20 1A 92 rec - 365 .7 6 - 365 . 7 6 - 528 . 32 0 0 523 . 24 83 . 82 $ Recirc Pump IP - 201B 10 pz 1 95 . 58 $ Vesse l 0 11 pz 82 1. 86 $ Transition Spherica l to Cylindrical 12 rec 0 0 82 1. 86 0 0 1889.59 51 8 . 16 $ cy l i n po r t d r ywe ll concrete surface 1 3 rec O O 82 1. 86 0 0 1889 . 59 51 6 . 25 $ cy l i n port drywe ll l iner surface 1 4 r pp - 693.42 693.42 - 693 . 42 693 . 42 82 1. 86 271 1. 45 $ Concrete c y l in port drywell outer 15 pz 1 07 1 . 73 $ Water Elevatio n Surface 16 pz 1 985 . 01 $ Top of RPV (head l evel ) 17 rpp - 4267.2 4267 . 2 - 4267 . 2 426 7. 2 27 11 . 4 5 4006 . 85 $ Reactor b ui lding abo ve drywe l l 18 rpp - 426 7 .2 4267.2 - 426 7 . 2 4267 . 2 4 006 . 85 4008 . 1 2 $ Reactor b uil d ing r oof 19 pz 14 7 . 32 $ Top of Ped El evat i on / Bottom Bi o Shield 20 pz 1 530 . 99 $ Top of Ped Elevation/Bottom Bio Sh i e l d 28 re c 0 0 1 985 . 0 1 0 0 1 0 . 00 2 4 8 .1 3 $ Reactor He a d 101 pz 7 42 . 54 6 1 02 pz 7 79. 1 22 103 pz 8 1 5 . 698 1 0 4 pz 852.2 74 1 05 pz 888 . 85 1 06 pz 925 . 246 1 07 pz 962 . 002 1 08 pz 998 . 57 8 1 09 p z 1 035~ 1 54 110 p z 1 071. 73 5 The surface card for the MCNP model without the reactor vessel head does not have surface 28 . Page 21 of 28

Dose Rate Evaluation of Reactor Vessel Water Levelsr-C _A_L _ N_O_._ N_E_E-_3_2_ _C 3-_C_A_L_C_-o_o_2---1 ENERCON Excellence-Every proj~ r. Ev,:ry doy, During Refueling for EAL REV. 00 Thresholds Figure 5 MCNP Model Cell Cards (No Head) c cells 101 1 - 4 . 49 101 i mp:p=256 $ Active Fuel Regio n 102 1 - 4.4 9 - 1 101 - 102 imp:p=l28 $ Active Fuel Region 10 3 1 - 4.4 9 - 1 102 - 103 imp : p=64 $ Active Fuel Regio n 104 1 -4 . 49 - 1 103 - 104 imp : p=32 $ Active Fuel Region 105 1 - 4.4 9 -1 104 - 105 i mp:p=l6 $ Active Fuel Region 106 1 - 4 . 49 -1 105 - 106 imp : p=8 $ Active Fuel Region 107 1 -4.4 9 - 1 106 -107 imp : p=4 $ Active Fuel Regio n 10 8 1 - 4 . 49 - 1 107 -108 imp : p=3 $ Active Fuel Region 10 9 1 - 4 . 49 - 1 108 - 109 imp : p=2 $ Active Fuel Region 11 0 1 - 4.49 - 1 109 - 110 imp : p=l $ Active Fuel Region 2 2 - 0 . 9982 1 15 imp:p=256 $ Water Region 3 3 - 1.21E- 03 15 - 2 imp:p=256 $ Air Region inside vessel 4 4 - 7 . 94 2 16 imp:p=256 $ RPV Shell 7 5 - 2.3 5 - 4 imp:p=256 $ Concrete Surrou nding RPV spherical 8 5 - 2 . 3 -14 12 imp:p=256 $ Concrete Surrou nding RPV cylindrical 9 5 - 2 . 3 - 9 8 7 -1 9 imp : p=256 $ Pedestal 91 5 - 2 . 3 - 91 81 1 9 - 20 imp :p=256 $ Bio Shield 10 5 - 2.3 7 imp:p=256 $ Concrete at b ottom of pedestal 11 3 - 1. 21E 8 imp : p=256 $ Inside Pedestal Air 1 2 3 - 1. 21E 6 7 - 11 9 3 H8 H9 # 91 imp : p=256 $ Inside Spherical portion Air 13 3 - 1. 21E 1 3 3 #91 imp:p=256 $ Inside Cy l indrical portion Air 14 3 - 1. 21E 17 imp : p=256 $ Reactor Buil~ing above drywel l Air 15 4 - 7 . 94 2 - 18 imp:p=256 $ Reactor Build Roof Stainless S t eel 16 4 - 7. 94 6 11 imp : p=256 $ Con tainment Liner Spherical portion 17 4 - 7 . 94 13 - 12 imp:p=256 $ Containment Liner Cylin p ortion 18 4 -7. 94 - 82 i mp : p=256 $ Recirc Pump IP- 201A 1 9 4 - 7.94 - 92 imp : p=256 $ Recirc Pump I P- 201B 999 0 1 12 #3 #4 n # 8 i9 no Ul U2 H 3 U4 us U6 U7 #18 J/1 9 #91 imp:p=O $ Problem Boundary Page 22 of 28

Dose Rate Evaluation of CALC NO. NEE-323-CALC-002 I ENERCON Reactor Vessel Water L e v e l s f - - - - - - - - - - - - -~

                      &ct!llence-EVf!ry pro;<<r. Every~       During Refueling for EAL         REV.        00 Thresholds Figure 6 X-Z V SED Plot of Reactor Vessel (With Head)

Reactor Vessel Head Cell for the homogenization of the Upper Internals stainless steel 0.379 g/cm3 Homogenized Core 7.5 MCNP Source Definition The core source term is modeled as uniformly distributed throughout the homogenized core, and has an energy spectra based on the decayed core inventory (Section 7 .1 ). Only the gamma source term is taken into account for this evaluation. The source term is generated shortly after shutdown , therefore, the fuel gamma source term will predominate, and the neutron-gamma and hardware activation source terms can be neglected (Assumption 4.7). The source is defined on the MCNP sdef card using Page 23 of28

Dose Rate Evaluation of CALC NO. NEE-323-CALC-002 Reactor Vessel Water Levels _ __ _ _ _ _ _ _ _ _ _- - 1 ENERCON

         &ceflena-Evt!ry pro}t!C.r. Every day. During Refueling for EAL        REV.            00 Thresholds distributions to define the particle location and energy. The radius of the core is defined with the. rad parameter, which automatically creates a uniform distribution based on a cylindrical geometry. The ext and axs parameters define the direction and distance of the cylinder axis. These parameters combined define the core where the particles can be born . The erg parameter defines the energy spectrum of source particles, and is based on the results of the ORIGEN-S calculation discussed previously. This distribution is a histogram of energies represented by activities. These are automatically normalized by MCNP to create a probability distribution. The total activity is preserved in the tally multiplier. The MCNP source definition cards are shown below in Figure 7. The sb card is a source biasing card, which in this case biases the particle generation to the lower end of the core. Th is is a variance reduction technique to improve the statistical certainty in the results.

Figure 7 MCNP Source Definition Cards sdef rad=dl ext=d2 axs =O O 1 erg=d8 ~ Source Definition Card

                                                                                             - Radius= dl
                                                                                             -Extent= d2
                                                                                             - Axis = +Z
                                                                                             - Energy= dB sil 137.045                                                                           ~ Core Radius Distribution si2 h O 742.546 779.122 815 . 698 852 . 274 888 . 85 925 . 246 962.002                ~ Core Axial Distribution 998.578 1035.154 1071.73 sp2 0 1 1 1 1 1 1 1 1 1 1                                                             ~ Actual Uniform Distribution sb2 0 1 1 0 .1 0.1 0.1 0 . 01 0.01 0.01 0.001 0.001                                   ~ Biased to Bot Distribution c Fuel Gamma Spectra s i8 h l . OOOe-002 5.000e - 002 l.OOOe - 001 2 . 000e - 001 3 . 000e-001 4 . 000e - 001      ~ Source Energy Groups 6 . 000e-001 8.000e-001 l . OOOe+OOO l.330e+OOO l.660e+OOO 2 . 000e+OOO 2 . 500e+OOO 3 . 000e+OOO 4.000e+OOO 5.000e+OOO 6 . 500e+OOO 8 . 000e+OOO l . OOOe+OOl l .l OOe+OOl sp8       O.OOE+OO 2 . 028E+l9 6 . 572E+l8 l . 557E+l9 9 . 672E+l8 3 . 582E+l8 7.837E+l8 ~ Source Emi s sion on Energy Basis l . 373E+l9 2 . 132E+l8 4.942E+l7 3.579E+l8 6.576E+l6 7.518E+l6 l . 110E+l7 8.689E+l4 l.553E+l0 2.568E+08 3 . 792E+07 8 . 041E+06 4.352E+05 Page 24 of 28

Dose Rate Evaluation of CALC NO. NEE-323-CALC-002 lyl ENERCON Reactor Vessel Water Levels Excellence-Every project ~v~ry ~al During Refueling for EAL REV. 00 Thresholds 7.6 MCNP Tally Specification The tallies used in this evaluation are point detectors placed at approximate locations of radiation monitors RE-9184A, and RE-91848. Point detectors are chosen because they use quasi-deterministic dose calculations that will provide better results than surface or cell based tallies that require the particles to enter those regions. The inputs to this card are the coordinates of the dose points followed by an exclusion zone to reduce variance, as well as a multiplier card, which represents the total core activity in photons/sec. The tally cards are shown in Figure 8. Figure 8 MCNP Tally Cards f5c RE-9184A, and 9184B ~Tally Comment Card f5:p -121.92 406.29 -184.15 20 ~Tally 5 (point detector) 182.88 -365.76 -184.15 20 x y z exclusion fm5 8.370E+19 ~ Tally Multiplier (Total Activity) In addition, the flux is multiplied by ANSI/ANS flux-dose conversion factors [Reference 3.4]. This is specified in MCNP using the de/df cards. These are shown in Figure 9. Figure 9 ANSI/ANS-6.1.1-1977 Gamma Flux to Dose Conversion Factors C ------------------------------------------------------------------ C ANSI/ANS-6.1.1-1977 c Gamma Flux to Dose Conversion Factors c (mrem/hr)/(photons/cm2-s) C ------------------------------------------------------------------ deO .01 ;03 .05 .07 .10 .15 .20 .25 .30 .35 .40 ~Energy Bins for Flux

      .45 .50 .55 .60 .65 .70 .80 1. 1.4 1.8 2.2                                                 to Dose Conversion 2.6 2.8 3.25 3.75 4.25 4.75 5. 5.25 5.75 6.25 6.75 7.5 9. 11.

dfO 3.96E-03 5.82E-04 2.90E-04 2.58E-04 2.83E-04 3.79E-'04 ~Energy Dependent 5.0lE-04 6.31E-04 7.59E-04 8.78E-04 9.85E-04 1. 08E-03 Flux Multipliers 1.17E-03 1.27E-03 1. 36E-03 1.44E-03 1.52E-03 ' 1. 68E-03

1. 98E-03 2.51E-03 2.99E-03 3.42E-03 3.82E-03 4.0lE-03 4.41E-03 4.83E-03 5.23E-03 5.60E-03 5.80E-03 6.0lE-03 6.37E-03 6.74E-03 7. llE-03 7.66E-03 8.77E-03 1.03E-02 Page 25 of28

Dose Rate Evaluation of CALC NO. NEE-323-CALC-002 lyl ENERCON Reactor Vessel Water Levels

       &cellenFO~Every projei:r. {,very doy,  During Refueling for EAL REV.             00 Thresholds 7.7  MCNP Material Cards The MCNP material cards are provided in Figure 9. These are based on the compositions described in Table 6 or calculated in Section 7.2.

Figure 10 MCNP Material Cards ml 92235 -0.0253 $ Homogenized Active Fuel Region 92238 -0.6163 8016 -0.1896 40000 -0.1531 50000 -0.0023 24000 -0.0002 26000 -0.0003 1001 -0.0129 m2 1001 2 8016 1 $ Water m3 6012 -0.000126 $ Air 7014 -0.76508 8016 -0.234793 m4 6000 -0.0008 $ ss 304 14000 -0.01 15031 -0.00045 24000 -0.19 25055 -0.02 26000 -0.68375 28000 -0.095 m5 26000 -0.014 $ Reg-Concrete 1001 -0.01 13027 -0.034 20000 -0.044 8016 -0.532 14000 -0. 337 11023 -0. 029 m6 6012 -0.01 $ Carbon Steel 26Q56 -0.99 Page 26 of28

Dose Rate Evaluation of CALC NO. NEE-323-CALC-002 ly ENERCON £,:cellen0!._-fvery pr~J~r. §very d_oy. Reactor Vessel Water Levels During Refueling for EAL REV. 00 Thresholds 7.8 Results The dose rates are provided in Table 1O for the water level at the top of the fuel assemblies. The dose rate is slightly above the detectable response of 1 R/h (1 E+03 mrem/h) for the no head configuration, and below the detectable response for the configuration with the reactor vessel head in place for one of the detectors. The sensitivity case shows that there is no significant impact due to reflection from the drywell cap. Table 10 - Dose Rate Response (mrernlh) Configuration Dose Rate 1 fsd6 Dose Rate 2 fsd Tally File RE-9184A RE-9184B l.... *:~~g.II~~c1 .. :~;: .* h?J;g+o_~*::';; ',. ~9':$~r(\3 * .; \* h~8E+Q~.-P.- -:,;,:{,'.>7")*(%***-***** :.-~9-0ncln1{ /1 With Head 1.11E+03 10.16% 7.41E+02 8.24% .. ~Ohgm _ 1

!(. .Wi!h*** JI¢ag,(S~11,;::,-* :* i1**. o*HE*'+*03 - : .'.* .: --.*1**.*s'*.*...-2*

sitivity{'.:ase), I  : , 1*.*. 610*,. *.**.: .**.*_. , *_-*.**.*.**1*.:_1:6**1_ IC E+. . . o*2' *.*: * ,

                                                                                                                                                                   *'1***5* *5* *1{/.'.
                                                                                                                                                                  <": , *         'J,.~r:* * ,..*. \ a:o***fum:
                                                                                                                                                                                                            *..:1.:.'...; * **

8.0 Computer Software This calculation uses ORIGEN-S of the SCALE Version 6.1.2 code package [Reference 3.2] and MCNP Version 6.1.0 [Reference 3.3] in accordance with CSP 3.09. 6 Fraction standard deviation. Page 27 of28

Dose Rate Evaluation of CALC NO. NEE-323-CALC-002 (t{I ENERCON Reactor Vessel Water Levels

~     ... *
       ~ll(!,:ice-Er~ry proJ~ct: fv~fy d()y. During Refueling for EAL   REV. 00 Thresholds 9.0 Impact Assessment This calculation is based on "realistic" assumptions for the purpose of declaring EALs, rather than typical conservative "bounding" type design basis analyses. The calculation results are intended to provide order of magnitude dose rates to assist Operations and Emergency Response personnel in determination of core uncovery in accordance with NEI 99-01 Rev. 6.

Page 28 of28

CALC NEE-323-CALC-002 ti ENERCON ~

       * ~x_c~ll~[!Ce~~l(e_ry pro}~t.~Every ~C!Y.

Appendix A Electronic File Listing NO. REV. 00 Origen output: 07/26/2017 04:19 PM 82,114 DAECEAL.OUT MCNP output: Directory of \No head\ 08/16/2017 09:13 AM 327,680 dOnao Directory of \With Head\ 08/16/2017 10:01 AM 1,269,760 dOhgo Directory of \sensitivity\ 08/16/2017 03:54 AM 286,720 dOrdo Page 1 of 1

CALC NEE-323-CALC-002 O' ENERCON

                        '1!:xcellenc~~i""ry proj<!d tveiy day.

Appendix B DAEAL.xlsx Sheets NO. REV. 00 dli A B I C D E I Fl G i H /Ii J  ! K L I _11 ' i I I i 1

Mas I Weight Mass Fraction Active Fuel ZAIDNumber Atom I :Material Isotope Fraction Reference :Materials Region Homogenized

  .:tJ.___

(KG) Zry-4 Zr 0.9823 [1] U02 200.4 92235 U-235 0.0253 41J ________

  ......          (6.56 'l)crrf)                Sn             0.0145                   Zry-4       42.92         92238      U-238               0.6163

__ s Cr 0.001 Water 31.98 8016 0 0.1896 51 Fe 0.0021 40000 Zr 0.1531

• • 1*1---

 ""--i Hf             0.0001                                             50000       Sn                 0.0023 U02                        U-235             0.0348       [1]                                   24000       Cr                 0.0002
 ": _f_____                                  U-'238            0.8466                                             26000       Fe                 0.0003            ..

101 11 l_ _

*****-**-r------*

Air . (1.21E-03 0 N C 0.1186 0.0001 0.7651 [1] 72000 1001 Hf H 0.0000 0.0129 1.0000 121r . g/cm.3)

  ~1 141             Water 0

H 0.2348 0.1111 [1] -  !

  -1 gfcm.3)             0              0.8889 Jll=

17 I (0.9982 SS-304 (7.94 gfcrn3) Fe Cr 0.6838 0.19 [1] i

-----+-*-**--*                                  Ni             0.095

. 18 191 t---- l\.fn 0.02 "-.---+----* _20 ~----... Si 0.01 i C  ! _21 4--- 0.0008 221 p 0.0004

  --f----- Concrete                             0              0.532       [1]                                             :
 ?~-                            3 Si            0.337

_241---- (2.30 'l)cm. ) ' 251 Ca 0.044 _?~J**-- Al 0.034 I ~l Na 0.029 Fe 0.014 H 0.01 _?O Carbon Steel C 0.01 [1] i 311 (7.82 'l)cm3) Fe 0.99

 *--r-*-:

Page 1 of2

CALC a ENER,CON

                                                 * .ExCelfence-E1*e..ry projf'!cr. *£very flay.

Appendix B DAEAL.xlsx Sheets NO. REV. N EE-323-CALC-002 00

   ,:'!          A                   B               C         :           D                    E    i F i     G           H       I      J       K                               L                     i i

1  ! i

                                                                                           '         . '   i                       i            !       l                                                 i i

Weight ZAID l\faterial Isotope Reference Material l\llass(KG) Atom Mass Fraction Active Fuel Region Homogenized Fraction Number 2 3 Zry-4 Zr 0_9823 [1] U02 200.42 92235 U-235 =(H3/SUM(H3:H5))*D8 3 Sn Zry-4 U-238 4 (6.56 gjcrn ) 0.0145 42-92 92238 =(H3/SUM(H3:H5))*D9 5 Cr 0_001 Water 31.98 8016 0 =((H3/(SUM(H3:H5)))*D10)+((H5/(SUM(H3:H5))))*D15 6 Fe 0.0021 40000 Zr =($H$4/SUM($H$3:$H$S))"'D3 7 Hf 0_0001 50000 Sn =($H$4/SUM($H$3:$H$S))*D4 8 U02 U-235 0.0348 [1] 24000 Cr =($H$4/SUM($H$3:$H$5) )*D5 9 U-238 0_8466 26000 Fe =($H$4/SUM($H$3:$H$S))*D6 10 0 0.1186 72000 Hf =($H$4/SUM($H$3:$H$5)) *D7 .. C [1]

                     ----- Air 11                                                             0.0001                                                                 1001         H    =(H5/SUM(H3:H5))*D14

• • ~-- *----------- -

1 j (1..21E-03 gjcm3) N 12 0_7651 =SUM(L3:Ul) 13 0 0.2348 14 Water H 0_1111 [1] 15 (0.9982 g}cni) 0 0.8889 16

---

-- SS:.304 Fe 0.6838 m 17 (7.94 gjcm.3

                                         )          Cr          0.19 18                                                Ni           0.095 19
                                                  ?vfn          0.02 I

20 Si 0.01 -- I 21 C 0.0008 i 22 p 0.0004 23 Concrete 0 0.532 [1] 24 {2.30 3 gjcm ) Si 0.337 i 25 Ca 0.044 26 Al 0.034 27 Na 0.029 i 28 Fe 0.014 29 H 0.01 30 Carbon Steel C 0.01 [1] 31 (7.82 gjcm. 3

                                        )          Fe           0_99                                       !                                            i Page2 of2

CALC NEE-323-CALC-002 0 ENERCON

          ~c_trl(~r,J~e.....:..E~~fY proj~t. ~~ery dpy.

AppendixC SCALE Input NO. REV. 00 =origens 0$$ all 71 e t BWR Source Term DAEC EAL Analysis 3$$ 21 1 1 a4 27 a16 4 a33 19 e t 35$$ 0 t 54$$ a8 0 all 2 e 56$$ 0 6 a6 1 alO O a13 63 3 3 0 2 0 e 57** 0 a3 1-16 e 95$$ 0 t DAECEAL Ci Source Terms 60** 0 24 40 50 60 70 61** 5rl-8 1+6 1+4 65$$

'GRAM-ATOMS                     GRAMS              CURIES  WATTS-ALL WATTS-GAMMA 3Z                0 1 0                        1 0 0  1 0 0        3Z      6Z 3Z                1 1 1                        1 0 1  1 1 1        3Z      6Z 3Z                1 1 1                        1 1 1  1 1 1        3Z      6Z 81$$ 2 0 26 1 e 82$$ f2 83** 1.10E+07 1.00E+07 8.00E+06 6.50E+06 5.00E+06 4.00E+06 3.00E+06 2.50E+06 2.00E+06 1. 66E+06 1.33E+06 1. OOE+06 8.00E+05 6.00E+05 4.00E+05 3.00E+05 2.00E+05 1.00E+05 5.00E+04 1.00E+04 e 84** 2.00E+07 6.43E+06 3.00E+06 1.85E+06 1.40E+06 9.00E+05 4.00E+05 1.00E+05 1. 70E+04 3.00E+03 5.50E+02 1.00E+02 3.00E+Ol 1. OOE+Ol 3.05E+OO 1. 77E+OO 1.30E+OO 1.13E+OO 1. OOE+OO 8.00E-01 4.00E-01 3.25E-01 2.25E-01 1.00E-01 5.00E-02 3.00E-02 1.00E-02 1.00E-05 e 73$$ 561390 561400 581410 581430 581440 962429 551340 551360 551370 531310 531320 531330 531340 531350 360831 360850 360851 360870 360880 571400 571410 571420 420990 410950 601470 932390 591430 942410 370860 451050 441030 441050 441060 511270 511290 380890 380900 380910 380920 430991 521270 521271 521290 521291 521311 521320 541311 541330 541331 541350 541351 541380 390900 390910 390920 390930 400950 400970 74** 9.06E+07 9.lOE+07 8.39E+07 7.65E+07 6.77E+07 2.14E+06 8.99E+06 2.85E+06 6.21E+06 5.11E+07 7.42E+07 1.04E+08 1.14E+08 9.90E+07 5.83E+06 5.32E+05 1.18E+07 2.35E+07 3.25E+07 9.39E+07 8.28E+07 8.05E+07 1. 01E+08 8.60E+07 3.35E+07 1.09E+09 7.57E+07 8.15E+06 l.01E+05 5.37E+07 8.30E+07 5.85E+07 2.96E+07 4.57E+07 1.66E+07 4.61E+07 4.57E+06 5.76E+07 6.19E+07 8.36E+07 4.51E+06 7.59E+05 1.58E+07 3.21E+06 1. 03E+07 7.28E+07 6.98E+05 1. 04E+08 3.29E+06

2. 72E+07 2.20E+07 8.72E+07 4.68E+07 6.06E+07 6.23E+07 4.82E+07 8.49E+07 8.09E+07 75$$ 3 3 3 3 3 2 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 2 3 2 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 t

56$$ fO t end Page 1 of 1

CALC NEE-323-CALC-002 Attachment 1 NO. r~ E'NERCONExcell~nce-Fvery project.'.Every day. CALCULATION PREPARATION CHECKLIST REV. 0 CHECKLIST ITEMS1 YES NO NIA GENERAL REQUIREMENTS

1. If the calculation is being performed to a client procedure, is the procedure being used the latest revision? D D [gl The Calculation is performed in accordance with ENERCON procedures.

2. Are the proper forms being used and are they the latest revision? [gl D D The Calculation is performed in accordance with ENERCON procedures.

3. Have the appropriate client review forms/checklists been completed? D D [gl OAR will be performed after calculation submittal

4. Are all pages properly identified with a calculation number, calculation revision and page number consistent with the requirements of the client's procedure?

[gl D D

5. Is all information legible and reproducible? [gl D D

6. Is the calculation presented in a logical and orderly manner? [gl D D

7. Is there an existing calculation that should be revised or voided? D [gl D There is no existing calculation that should be revised or voided. **

8 .. Is it possible to alter an existing calculation instead of preparing a new calculation for this situation? D [gl D No existing calculation would be applicable.

9. If an existing calculation is being used for design inputs, are the key design inputs,

• r assumptions and engineering judgments used in that calculation valid and do they D D [gl apply to the calculation revision being performed.

No existing calculation is used for design inputs

10. Is the format of the calculation consistent with applicable procedures and expectations? - [gl D .D

11. Were design input/output documents properly updated to reference this calculation? D D [gl There are no design output documents.

12. Can the calculation logic, methodology and presentation be properly understood without referring back to the originator for clarification?

                                                                                                      - [gl    D    D OBJECTIVE AND SCOPE

13. Does the calculation provide a clear concise statement of the problem and objective of the calculation?

[gl D D

14. Does the calculation provide a clear statement of quality classification? [gl D D

15. Is the reason for performing and the end use of the calculation understood? [gl D D

16. Does the calculation provide the basis for information found in the plant's license basis? D ~ D This does not provide basis for license basis

17. If so, is this documented in the calculation? D D ~

Page 1 of5

CALC NEE-323-CALC-002 Attachment 1 NO. ltJ

 ~

ENERCON . . Ex~e!lence::-Ev~ry pro}~cr. £ve!Y _day. CALCULATION PREPARATION CHECKLIST REV. 0 CHECKLIST ITEMS1 I YES I NO I NIA See above

18. Does the calculation provide the basis for information found in the plant's design basis documentation? I D I D I t8l This does not provide basis for design basis

19. If so, is this documented in the calculation? I D I D I t8l See above

20. Does the calculation otherwise support information found in the plant's design basis documentation? I D I t8l I D This does not provide support for information found in design basis documentation

21. If so, is this documented in the calculation? I D I D I t8l See above

22. Has the appropriate design or license basis documentation been revised, or has the change notice or change request documents being prepared for submittal? I D I D I t8l See above DESIGN INPUTS I I I

23. Are design inputs clearly identified? I t8l I D I D

24. Are design inputs retrievable or have they been added as attachments? I t8l I D I D

25. If Attachments are used as design inputs or assumptions are the Attachments traceable and verifiable? I t8l I D I D

26. Are design inputs clearly distinguished from assumptions?

                                                                                              'I I t8l    I D I D

27. Does the calculation rely on Attachments for design inputs or assumptions? If yes, are the attachments properly referenced in the calculation? I t8l I D I D The Design !nformation Transmittal is included as an Attachment is properly referenc~d in the calculation

28. Are input sources (including industry codes and standards) appropriately selected I t8l D D and are they consistent with the quality classification and objective of the calculation? I I

29. Are input sources (including industry codes and standards) consistent with the plant's I [81' D D design and license basis? I I

30. If applicable, do design inputs adequately address actual plant conditions? I t8l I D I D

31. Are input values reasonable and correctly applied? I t8l I D I D

32. Are design input sources approved? I t8l I D I D The Design Information Transmittal contains information from a superseded calculation.

33. Does the calculation reference the latest revision of the design input source? I t8l I D I D The calculation uses information from a superseded calculation. This information is provided in a Design Information Transmittal.

34. Were all applicable plant operating modes considered? I t8l I D I D ASSUMPTIONS I I I Page 2 of 5

CALC NEE-323-CALC-002 Attachment 1 NO.

~ ENERCON CALCULATION PREPARATION CHECKLIST Excefl~flC~~Eve.ry p,ojec_t Every day.

REV. 0 CHECKLIST ITEMS 1 YES NO N/A

35. Are assumptions reasonable/appropriate to the objective? [gl D D

36. Is adequate justification/basis for all assumptions provided? [gl D D

37. Are any engineering judgments used? D 0 D Engineering judgement not used as design input.

38. Are engineering judgments clearly identified as such? D D [gl Engineering Judgement is not used as a design input.

39. If engineering judgments are utilized as design inputs, are they reasonable and can they be quantified or substantiated by reference to site or industry standards, D D [gl engineering principles, physical laws or other appropriate criteria?

Engineering Judgement is not used as a design input. METHODOLOGY

40. Is the methodology used in the calculation described or implied in the plant's

[gJ licensing basis? D D The scope of calculation is outside of plant licensing basis

41. If the methodology used differs from that described in the plant's licensing basis, has

[gJ the appropriate license document change notice been initiated? D D see above.

42. Is the methodology used consistent with the stated objective? [gl D D

43. Is the methodology used appropriate when considering the quality classification of the calculation and intended use bf the results?

[gl .P, D BODY OF CALCULATION

44. Are equations used in the calculation consistent with recognized engineering practice

[gJ and the plant's design and license basis? - D D

45. Is there reasonable justification provided for the use of equations not in common use? D D [gJ There are no uncommon equations used in the calculation.

46. Are the mathematical operations performed properly and documented in a logical fashion?

[gJ D D

47. Is the math performed correctly? [gJ D D

48. Have adjustment factors, uncertainties and empirical correlations used in the analysis been correctly applied?

[gl D D

49. Has proper consideration been given to results that may be overly sensitive to very

[gJ small changes in input? D D SOFTWARE/COMPUTER CODES

50. Are computer codes or software languages used in the preparation of the

[gJ calculation? D D MCNP and Scale are used Page 3 of 5

CALC NEE-323-CALC-002 Attachment 1 NO. a ENERCON

            ~cellence.-E~e_ry project EvtJ!Y r[ay.

CALCULATION PREPARATION CHECKLIST REV. 0 CHECKLIST ITEMS1 YES NO N/A

51. Have the requirements of CSP 3.09 for use of computer codes or software

[81 D D languages, including verification of accuracy and applicability been met?

52. Are the codes properly identified along with source vendor, organization, and revision

[81 D D level?

53. Is the computer code applicable for the analysis being performed? [81 D D

54. If applicable, does the computer model adequately consider actual plant conditions? [81 D D

55. Are the inputs to the computer code clearly identified and consistent with the inputs

[81 D D and assumptions documented in the calculation?

56. Is the computer output clearly identified? [81 D D

57. Does the computer ouJput clearly identify the appropriate units? [81 D D

58. Are the computer outputs reasonable when compared to the inputs and what was

[81 D D expected?

59. Was the computer output reviewed forERROR or WARNING messages that could

[81 D D invalidate the results? RESULTS AND CONCLUSIONS

60. Is adequate acceptance criteria specified? D D [81 There is no acceptance criteria as discussed in calc.

61. Are the stated acceptance criteria consistent with the purpose of the calculation, and D D [81 intended use?

See above

62. Are the stated acceptance criteria consistent with the plant's design basis, applicable D D [81 licensing commitments and industry codes, and standards?

See above

63. Do the calculation results and conclusions meet the stated acceptance criteria? D D [81 See above.

64. Are the results represented in the proper units with an appropriate tolerance, if

[81 D D applicable?

65. Are the calculation results and conclusions reasonable when considered against the

[81 D D stated inputs and objectives?

66. Is sufficient conservatism applied to the outputs and conclusions? [81 D D Page 4 of 5

CALC NEE-323-CALC-002 Attachment 1 NO. I CALCULATION PREPARATION ENERCON CHECKLIST Exceflence-Eve,y project. Every day. REV. 0 CHECKLIST ITEMS 1 YES NO N/A

67. Do the calculation results and conclusions affect any other calculations? D ~ D No other calculations are affected by this calculation.

68. If so, have the affected calculations been revised? D D ~

No other calculations are affected by this calculation.

69. Does the calculation contain any conceptual, unconfirmed or open assumptions requiring later confirmation? D ~ D There are no open assumptions requiring confirmation later.

70. If so, are they properly identified? D D ~

There are no open assumptions requiring confirmation later. DESIGN REVIEW 71 . Have alternate calculation methods been used to verify calculation results? D ~ D No a Design Review was performed . Note:

1. Where required , provide clarification/justification for answers to the questions in the space provided below each question. An explanation is required for any questions answered as "No' or "N/A".

Originator: Jay Bhatt Print Name and Sign Date Page 5 of 5

BASES-EOP 3 DAEC EOP BASES DOCUMENT Rev. 13 EOP 3 - SECONDARY CONTAINMENT CONTROL Page 27 of 29 GUIDELINE Spent Fuel Pool level cannot be maintained above 37 ft 1 in. SF/L-2 D Maintain Spent Fuel Pool level above 36 ft .

                               ..- If necessary, use alternate or external makeup sources (SEP 312) .
                               ..- Use only systems IlQ1 required for SF/L-3        adequate core cooling .

D DISCUSSION If spent fuel pool level cannot be restored and maintained above the low level alarm setpoint, an alternate control band is established above the higher of the spent fuel pool level LCO (36 ft.) or the Minimum Safe Operating Spent Fuel Pool Level (25.17 ft.). If necessary, normal spent fuel pool makeup may be augmented-by one or more of the alternate and external sources listed in SEP 312. The Minimum Safe Operating Spent Fuel Pool Level is generically defined to be the lowest water level providing adequate radiation shielding to (1) protect personnel performing local operations required by the EOPs and (2) allow unrestricted access to the main control room . At the DAEC , the Minimum Safe Operating Spent Fuel Pool Level is defined consistent with NEI 12-02 Level 2, described as the level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck. The corresponding spent fuel pool level at the DAEC is defined to e 25.17 ft., approximately 10 ft. above the top of the fuel racks.

Local Operations for Operating and Normal Shutdown/Cooldown Procedure Step Action If action not performed, does Building Elevation Room Mode Section this prevent shutdown or and Stec cooldown? IPOl3, Between 50% and 60% Reactor Power No. The Feed Pumps and NIA NIA NIA NIA Section 5, shutdown one Condensate and Reactor Condensate Pumps can be tripped step (9) Feed Pump per 01 644 unless otherwise from the Control Room if directed by CRS. necessary, and HPCI and/or RCIC can be used to maintain RPV Level. IPOl3, When turbine load is lowered to No. 2nd Stage Reheat can be left in N/A N/A NIA NIA Section 5, approximately 200 MWe, remove the 1E- se~,rvice and the turbine can be step (10) 18A[B] 2nd Stage Reheat System from tripped if necessary. service in accordance with 01 646, Extraction Steam. IPOl 4, Secure condensate demineralizers as No. Condensate Demineralizers N/A NIA NIA NIA Section 3 directed by 01 639, Section 5.1. will automatically go into the "hold" step (10) mode as power and flow are lowered. IP014, Commence primary containment purge No. This is only necessary if a NIA NIA NIA NIA Section 3 per 01573. Drywell entry is anticipated. steo (11) IPOl4, At the refueling bridge, verify that the Main No. Control rod insertion will not be NIA N/A N/A NIA Section 3 Disconnect is closed and that the inhibited. step (13) SYSTEM START pushbutton has been deoressed. IPOl 4, Prior to disconnecting the generator from No. Aux Boiler is not required to NIA NIA NIA NIA Section 3 the grid, perform the following: (a) If accomplish shutdown. step (14) needed, start up the Auxiliary Boiler per 01 727. IP014, Following Turbine Trip: (a) Verify that No. These systems can be left in NIA NIA N/A N/A Section 3 Reactor Coolant Chloride and service if necessary. step (22) Conductivity analyses have been performed. (b) Operate the Turbine Lube ~ Oil and Turning Gear System per 01  ; 693.3. (c) Shut down the generator per 01 698. (d) Shut down the turbine per 01 693.1.

Procedure Step Action If action not performed, does Building Elevation Room Mode Section this prevent shutdown or and Step .. cooldown? IP014, Shut down the following generator support No.c, These systems can be left in N/A N/A N/A N/A Section 3 systems, as desired: Isolated Phase Bus service if necessary. step (24) Cooling - OI 698, Stator Water Cooling - 01 697, H2 Seal Oil - 01 695.1, H2 and CO2 Gas - 01 695.2 IPOl4, Secure hydrogen, oxygen and/or air No. The Hydrogen Water N/A N/A N/A N/A Section 3 injection per 01 563, Hydrogen Water Chemistry System will secure itself step (26) Chemistrv. if left in service. IP014, As directed by the CRS, perform the No. The MSIVs can be closed if N/A N/A N/A N/A Section 3 following steps as necessary to liniit necessary to limit plant cooldown step (27) reactor vessel depressurization following rate. the reactor scram: (b) Start 1P32 Mechanical Vacuum Pump per 01 691. (c) Secure the SJAEs and Offgas per 01 691 and 01672. IPOl 4, For the remainder of this section use the (a) No. The MSIVs can be N/A N/A N/A N/A Section 4 following methods as necessary to closed if- necessary to limit step (6) cooldown and depressurize the reactor plant cooldown rate. vessel to maintain a controlled cooldown (b) No - operated from the rate less than the TS Limit of 100°F in any Control Room 1 hour period. (a) Use the Main Turbine (c) No - Operated from the Bypass Valve to control cooldown per 01 ._... Control Room 693.1 Section 4.5 if available, (b) If '(d) No. The MSIVs can be desired cooldown with RCIC per 01 150 closed if necessary to limit (preferred method if MSIVs are closed), plant cooldown rate. (c) If desired cooldown with HPCI per 01 (e) No. The MSIVs can be 152 (RCIC may become inadequate as closed if necessary to limit pressure lowers) (d) Control steam flow plant cooldown rate. from the reactor vessel to the main condenser through steam seals and steam drains, (e) Secure steam seals per 01 692 as required to limit cooldown after the turbine is on the jack and vacuum is broken.

Procedure Step Action If action not performed, does Building Elevation Room Mode Section this prevent shutdown or and Step cooldown? IPOl4, As plant cooldown continues perform the No. The MSIVs can be closed if N/A N/A N/A N/A Section 4 following: (NA if MSIVs are closed) (a) necessary to limit plant cooldown step (7) Control steam seal pressure 3 to 4 psig rate. using M0-1169, MAIN STEAM SUPPLY, M0-1170, REGULATOR BYPASS and/or M0-1171, M~NUAL UNLOADER on 1C07, (b) Start 1P-32 MECHANICAL VACUUM PUMP per Ol 691, (c) When reactor pressure approaches 500 psig or cooldown rate cannot be controlled within the limit, then secure SJAEs and Offgas System per 01 691 and 01 672, respectively, if not previously secured, (d) If not using EHC Pressure Set to control plant cooldown, then at 1C07, use the PRESSURE SET ADJUST pushbuttons to maintain A[B] PRESSURE SET DEMAND between 150 and 50 psig above reactor pressure as reactor pressure decreases. Otherwise, N/A. IPOl4, At approximately 400 psig, secure the No. The Feed Pumps and N/A N/A N/A N/A Section 4 operating feed pump per 01 644. Condensate Pumps can be tripped step (8) from the Control Room if necessarv. IPOl4, When RHR Shutdown Cooling Isolation No, this system can be placed in N/A N/A N/A N/A Section 4 Interlocks can be reset service from the Control Room if step (9) (approximately 100 psig), reset the netessary. isolation, then initiate Shutdown Cooling oer 01149. IPOl4, Perform the following after the turbine trip, No. These systems can be left in N/A N/A N/A N/A Section 4 if needed: (a) Verify that Reactor Coolant service if necessary. step (10) Chloride and Conductivity analysis has been performed, (b) Operate the Turbine Lube Oil and Turning Gear System per 01 693.3, (c) Shutdown the Main Generator per 01698, (d) Shutdown the Main Turbine per 01 693.1.

Procedure Step Action If action not performed, does Building Elevation Room Mode Section this prevent shutdown or and Step cooldown? IPOl4, Shutdown the following systems as No. These systems can be left in N/A N/A N/A N/A Section 4 directed by the CRS/OSM. service if necessary. step (11) (a) Isolated Phase Bus Cooling per 01 698, (b) Stator Water Cooling per 01697, (c) H2Seal Oil perOl 695.1, (d) H2and CO2 Gas per 01 695.2, (e) Secure SJAEs per 01691 and Offgas per 01 672 if not previously performed. IPOl4, Perform the following at approximately 50 No. The Feed Pumps and N/A N/A N/A N/A Section 4 psig: (a) Close the BYPASS VALVE . Condensate Pumps can be tripped step (12) OPENING JACK SELECTOR, (b) Line up from the Control Room if and place RFP Stuffing Box Pump 1P-134 necessary. in operation to maintain Seal Water Drain ,; Tank 1T-135 level. ' IPOl4, When steam seal pressure cannot be No. The MSIVs can be closed if N/A N/A N/A N/A Section 4 maintained or the turbine shaft has cooled necessary to limit plant cooldown step (13) per 01 693.3, open Condenser Vacuum rate. - Breaker valves V-03-67 and V-03-73. IPOl4, Secure MECHANICAL VACUUM PUMP No. The MSIVs can be closed if N/A N/A N/A N/A Section 4 1P-32 when no longer required per 01 necessary to limit plant cooldown step (14) 691. rate. IPOl4, When the condenser is at atmospheric No. The MSIVs can be closed if N/A N/A N/A N/A Section 4 pressure, secure the Turbine Steam Seal necessary to limit plant cooldown step (15) System per 01692. rate. IPOl 4, Shut down the operating condensate No. The Feed Pumps and N/A N/A N/A N/A Section 4 pump per OI 644 when no longer required Condensate Pumps can be tripped step (18) for RPV Level Control or Hotwell cleanup from the Control Room if recirculation. necessarv. Conclusion of manual action evaluation for EALs RA3 and HAS is shown below: EALs RA3 and HAS are not applicable to DAEC because the evaluation has shown that there are no rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. All areas outside th'l3 Control Room that contain equipment necessary for normal plant operation, cooldown and shutdown do not require physical access to operate.

Development of EAL Threshold values from NEE-323-CALC-005 Calculated values are provided in Calc-005 as shown below. Table 3- Recommended RA 1, RS1, and RG1 EAL Thresholds (Modes 1, 2, 3}

                    ,,   **.c- -- * - ~     .... ._ ..                                                                         ...

Release Point M.1 RS1 RG1

                                                               µd/cc.                      . pd/cc ,              µci/cc .

Turbine Building 1.58£-02 1.SSE-01 1.5BE+oo Reactor Building 1.22£:-02 1.22E-01. 1.llEi-00 Offgas Stack 4.39£..01 4.3'9Et02 4.39E+o3' LLRPSF 151E-02 1.51E-01. 151E+oo* Table 4- Recommended RA 1, RS1, and RG1 EAL 1hreshokis (Modes 4, 5}

                         *-q-**
                         *Release Point                RA1 ci/ct
                                                                       ..                        RS1 d/cc RG:l ci/cc Turbine Building                      1..30cf-02                     1.30E-01             1.5/JE+oo Reactor Building                      1.0lf-02                       1,0lE-01             1.0!E+oo Offgas Stack                      452f..01                       4.521:+0.2.          4.52E+o3' ll.RPSF                    1.25£0 02                      1.25E-Ol.           1.25!:+oo*

• Per Design Input 5.8 the results in EAL threshold values exceed the range of the monitor.

The following table of threshold values was developed for use in the DAEC EAL scheme by averaging the separate Mode 1-3 and Mode 4-5 thresholds from Calc-005, and then rounding the average values for ease of EAL evaluator use, as well as to provide a step-wise progression through the emergency classification. Resulting values are shown in the Alert, SAE, and GE columns below:

           .;,.\. ,,;:,r. ::                 . ,>,,,;, -T~ble'R-l*~*EffluentMdriitcir Cla~sific~tiol'I Thresliotas\'. '                      '   ..
         ':*0***   i";!' i;i.t /'i;i              Mon.i_torl      ,~.,  'It;\'

• j.::,* *: -;iGE{( . JI' s~~* ,::,;;~. '*f ,Al~rt J1J5 H~;- NOUEi*:

Reactor Building ventilation rad monitor 1.lE+OO uci/cc 1.lE-01 uci/cc 1.lE-02 uci/cc 8.0E-04 uci/cc (Kaman 3/4, 5/6, 7/8)

              <I)     Turbine Building ventilation rad monitor

J 1.4E+OO uci/cc 1.4E-Ol uci/cc 1.4E-02 uci/cc 8.0E-04 uci/cc 0 (Kaman 1/2)

QJ

              <I) ro      Offgas Stack rad monitor C)                                                                      4.5E+03 uci/cc      4.SE+02 uci/cc         4.5 E+Ol uci/cc     2.0E-01 uci/cc (Kaman 9/10)

LLRPSF rad monitor ... 1.4E-Ol uci/cc 1.4E-02 uci/cc 1.2E-03 uci/cc (Kaman 12) GSW rad monitor (RIS-4767)

                                                                                              -                *-                  1.7E+04 cps       1.5E+03 cps "C
             '::i     RHRSW & ESW rad monitor                                                 ...

C' (RM-1997)

                                                                                                               *-                  1.2E+04 cps       8.4E+02 cps RHRSW & ESW Rupture Disc rad monitor (RM-4268)
                                                                                              *-               -                   1.8E+04 cps       l.OE+03 cps

CALC NO. NEE-323-CALC-005 g ENERCON Excellence-Every project. Every day CALCULATION COVER SHEET REV. 00 PAGE NO. 1 of 34 Client: Duane Arnold Energy Center Revised Gaseous Radiological EALs per NEI

Title:

99-01 Rev. 06 Project Identifier: NEE-323 Item Cover Sheet Items Yes No 1 Does this calculation contain any open assumptions, including preliminary information, that require confirmation? (If YES, identify the assumptions.) D ~ 2 Does this calculation serve as an "Alternate Calculation"? (If YES, identify the design verified calculation.) D ~ Design Verified Calculation No. _ _ 3 Does this calculation supersede an existing Calculation? (If YES, identify the design verified calculation.) D [8l Superseded Calculation No. Scope of Revision: Initial Issue Revision Impact on Results: Initial Issue Study Calculation D Final Calculation ~ Safety-Related D Non-Safety-Related ~ (Print Name and Sign) Originator: Ryan Skaggs Date: 12/14/17 Design Verifier1 (Reviewer if NSR): Jay Bhatt Date: 12/14/17 Approver: Zachary Rose Date: 12/14/17 Note 1: For non-safety-related calculation, design verification can be substituted by review.

CALC NO. NEE-323-CALC-005 (J ENERCON CALCULATION E.x.cel(enc~-Every project. Every dpy. REVISION STATUS SHEET REV. 00 CALCULATION REVISION STATUS REVISION DATE DESCRIPTION 00 12/14/17 Initial Issue PAGE REVISION STATUS PAGE NO. REVISION PAGE NO. REVISION All 00 APPENDIX/ATTACHMENT REVISION STATUS APPENDtX NO. NO.OF REVISION ATTACHMENT NO.OF REVISION PAGES NO. NO. PAGES NO. A 8 00 1 4 00 Page 2 of34

CALC NO. NEE-323-CALC-005 lyl. ENERCON TABLE OF CONTENTS Excellen~e_;,Every project. Every day. REV. 00 Section Page No.

1.0 Purpose and Scope

4 2.0 Summary of Results and Conclusions 4 3.0 References 6 4.0 Assumptions 7 5.0 Design Inputs 8 6.0 Methodology 13 7.0 Calculation 19 8.0 Computer Software 33 9.0 Impact Assessment 34

                                                                                    #of List of Appendices                                                                 Pages Appendix A - Dose Spreadsheet Output"                                          8
                                                                                    #of List of Attachments                                                                 Pages Attachment 1 - Calculation Preparatiot:t Checklist                             4 Page 3 of34

CALC NEE-323-CALC-005 ENERCON Revised Gaseous Radiological NO. Excellence-Every projecr. Every day. EALs per NEI 99-01 Rev. 06 REV. 00

1.0 Purpose and Scope

The DAEC site is implementing new requirements of Revision 6 to the Document NEI 99-01, "Development of Emergency Action Levels for Non-Passive Reactors. " One of the changes included in Revision 6 to NEI 99-01 is a new basis for the Emergency Action Level (EAL) RA1 . The requirements for RS1 and RG1 did not change from NEI 99-01 Rev. 05 with the implementation of NEI 99-01 , Rev. 06. The following table is extracted from Section 6 of Revision 6 to NEI 99-01: ALERT SITE AREA EMER- GENERAL EMER-GENCY GENCY AA1 Release of gaseous AS1 Release of gaseous AG1 Release of gase-or liquid radioactivity re- radioactivity resulting in ous radioactivity result-suiting in offsite dose offsite dose greater than ing in offsite dose greater than 10 mrem 100 mrem TEOE or 500 greater than 1,000 mrem TEOE or 50 mrem thy- mrem thyroid COE. TEOE or 5,000 mrem roid COE. Op. Modes: All thyroid COE. Op. Modes: All Oo. Modes: All AA1, AS1, AG1 compares to DAEC terminology RA1, RS1, RG1, respectively. This calculation determines the effluent radiation monitor readings that correspond to the RA 1, RS 1, and RG 1 thresholds . 2.0 Summary of Results and Conclusions The results below show the RA 1 EAL release concentration thresholds and associated dose rates for each release point for a decay time of five hours and 36 hours. The highlighted dose indicates which threshold was met at the release concentration. Table 1- RA1 EAL Release Concentration Thresholds (Decay= 5 hours (Mode 1, 2, 3)) Release Point Release Concentration CEDE EDE TEDE CDEThyrold

                                              µCi/cc                 mrem         mrem              mrem           mrem Turbine Building                         l.58E-02                  2.38         0.39               2.77         50.0 Reactor Building                         1.22E-02                2.37**         0.39               2.76        49.8**

Offgas Stack 4 .39E+Ol 1.96 8.05* 10.00 41.1 Low-Level Radwaste Processing and 1.SlE-02 2.37 0.39 2.76 49.7 Storage Facility (LLRPSF)

• Calculation of this value was demonstrated in Section 7.3
  • Calculation of this value was demonstrated in Section 7.4 Table 2 - RA 1 EAL Release Concentration Thresholds (Decay= 36 hours (Mode 4, 5))

Release Point Release Concentration CEDE EDE TEDE CDEThyrold

                                               µCl/cc                mrem         mrem               mrem          mrem Turbine Building                          1.30E-02                 2.59         0 .07               2.67         49.7 Reactor Building                         l.OlE-02                  2.60        0 .07               2.68         49.9 Offga s Sta ck                         4.52E+Ol                  2.61        1.41               4.02          50.0 LLRPSF                              1.25E-02                  2.60        0 .07               2.67         49.8 Page 4 of 34

CALC NEE-323-CALC-005 ENERCON Revised Gaseous Radiological NO. Excellence-Every project Every day. EALs per NEI 99-01 Rev. 06 REV. 00 Resultant EAL thresholds: The tables below show the release concentration threshold for RA 1, RS 1, and RG 1 based on the results above for both a decay time of five hours and a decay time of 36 hours. From Section 1.0: RS1 thresholds are 10 times larger than those for RA1 RG 1 thresholds are 100 times larger than those for RA 1 Table 3- Recommended RA 1, RS1, and RG1 EAL Thresholds (Modes 1, 2, 3) Release Point RAl RSl RGl

                                                   µCl/cc             µCi/cc        µCi/cc Turbine Building       1.SBE-02          1.SBE-01     1.58E+OO Reactor Building       1.22E-02          1.22E-01     l.22E+OO Offgas Stack        4.39E+Ol          4.39E+02     4.39E+03 LLRPSF         1.SlE-02          1.SlE-01    1.s1E+oo*

Table 4 - Recommended RA 1, RS1, and RG1 EAL Thresholds (Modes 4, 5) Release Point RAl RS1 RGl

                                                    µCl/cc            µCl/cc        µCl/cc Turbine Building       l.30E-02          1.30E-01      l.30E+OO Reactor Building       l.OlE-02          1.0lE-01      l.OlE+OO Offgas Sta ck       4.52E+Ol          4.52E+02     4.52E+03 LLRPSF         1.25E-02           1.25E-01   1.2sE+oo*

• Per Design Input 5.8 the results in EAL threshold values exceed the range of the monitor.

Page 5 of 34

CALC

a. ENE'flCON E_~l{enc~~~very project. *Every (j<iy.

Revised Gaseous Radiological EALs per NEI 99-01 Rev. 06 NO. REV. NEE-323-CALC-005 00 3.0 References 3.1 NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors", Nuclear Energy Institute, November 2012. 3.2 NUREG-1940, RASCAL 4: Description of Models and Methods, United States Nuclear Regulatory Commission, Office of Nuclear Security and Incident Response, 2012. 3.3 NUREG-1940 Supplement 1, RASCAL 4.3: Description of Models and Methods, United States Nuclear Regulatory Commission, Office of Nuclear Security and Incident Response, 2015.

  - 3.4 NUREG-1228, Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents, United States Nuclear Regulatory Commission, Division of Operational Assessment, 1988.

3.5 NUREG-1465, Accident Source Terms for Light-Water Nuclear Power Plants, United States Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, 1995. 3.6 DAEC UFSAR, Chapter 15-0. 3.7 DAEC UFSAR, Chapter 15-2. 3.8 DAEC Offsite Dose Assessment Manual (ODAM). 3.9 Plant Chemistry Procedure PCP 8.3, Alarm Setpoints and Background Determination for KAMAN Normal Range Monitors. 3.1 oDAEC Nuclear Station HRN-HRH Radiation Monitor Operation, Maintenance and Troubleshooting Manual, ©2000, by Engineering Solutions, 310 Luchana Drive, Litchfield Park, Arizona. 3:11 DAEC Emergency Plan, Section 'I', Rev. 27. 3.12 Federal Guidance Report No. 11, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion and Ingestion Office of Radiation and Indoor Air, 1999. 3.13 Federal Guidance Report No. 12, External Exposure to Radionuclides in Air, Water, and Soil, 1993 .

  . 3.14 Table of                           Nuclides,   http:llatom.kaeri.re.kr:BOBO!tonlindex.html, retrieved 10/10117.

Page 6 of34

CALC NEE-323-CALC-005 0 , ENERC*ON Exa,lienci:--,-;Evi,ry proje~t. Every day. Revised Gaseous Radiological EALs per NEI 99-01 Rev. 06 NO. REV. 00 4.0 Assumptions The following are assumptions about the receptor:

• No credit is taken for radiation shielding provided by structures.
• No decay in-transit is assumed during the time elapsed between the release point and the receptor.

Both of the above assumptions are acceptable because they will result in a higher dose to the receptor and conservatively lower thresholds. Page 7 of 34

CALC NEE-323-CALC-005 0 .

 . ENERCON Excellence~Everyproject, Every day.

Revised Gaseous Radiological EALs per NEI 99-01 Rev. 06 NO. REV. 00 5.0 Design Inputs 5.1 Core Inventory The assumed isotopic mixture in Table 5 is taken from Table 1-1 of NUREG-1940. The core inventory (curies per megawatts thermal) in the table is based on calculations made by the NRC staff in December 2003 using the SAS2H control module of SCALE (Standardized Computer Analyses for Licensing Evaluation), Version 4.4a. Table 5 - Isotopic Mixture NUCLIDE CORE INVENTORY NUCLIDE CORE INVENTORY NUCLIDE CORE INVENTORY (Ci/MWt) {CUMWt) (Ci/MWt) Ba-139 4.74E+04 La-141 4.33E+04 Te-127 2.36E+03 Ba-140 4.76E+04 La-142 4.21E+04 Te-127m 3.97E+02 Ce-141 4.39E+04 Mo-99 5.30E+04 Te-129 8.26E+03 Ce-143 4.00E+04 Nb-95 4.50E+04 Te-129m 1.68E+03 Ce-144* 3.54E+04 Nd-147 1.75E+04 Te-131m 5.41E+03 Cm-242 1.12E+03 Np-239 5.69E+05 Te-132 3.81E+04 Cs-134 4.70E+03 Pr-143 3.96E+04 Xe-131m 3.65E+02 Cs-136 1.49E+03 Pu-241 4.26E+03 Xe-133 5.43E+04 Cs-137* 3.25E+03 , Rb-86 5.29E+01 Xe-133m 1.72E+03 ,., 1-131 2.67E+04 Rh-105 2.81E+04 Xe-135 1.42E+04 1-132 3.88E+04 ' Ru-103 4.34E+04 Xe-135m 1.15E+04 1-133 5.42E+04 Ru-105 3.06E+04 Xe-138 4.56E+04 1-134 5.98E+04 Ru-106* 1.55E+04 Y-90 2.45E+03 1-135 5.18E+04 Sb-127 2.39E+03 Y-91 3.17E+04 Kr-83m 3.05E+03 Sb-129 8.68E+03 Y-92 3.26E+04 Kr-85 2.78E+02 Sr-89 2.41E+04 Y-93 2.52E+04 Kr-85m 6.17E+03 Sr-90 2.39E+03 Zr-95 4.44E+04 Kr-87 1.23E+04 Sr-91 3.01E+04 Zr-97* 4.23E+04 Kr-88 1.70E+04 Sr-92 3.24E+04 La-140 4.91E+04

• Tc-99m 4.37E+04 Page 8 of34

CALC a ENERCON f!ceflenc~-Eve_ry proja;t. E~ery day. Revised Gaseous Radiological EALs per NEI 99-01 Rev. 06 NO. REV. NEE-323-CALC-005 00 5.2 Release Fraction Table 6 displays release fractions as a function of time taken from Table 1-4 which references Table 3-12 of NUREG-1465. Table 6 - Release Fraction NUCLIDE GROUP BWR CORE INVENTORY RELEASE FRACTION Cladding Failure Core Melt Phase Postvessel (Gap Release (In-Vessel Phase) Melt-Through Phase Phase) (Ex-Vessel Phase) (1.5-hour duration) (0.5-hour duration) (3.0-hour duration) Noble gases (Kr, Xe) 0.05 0.95 0 Halogens (I, Br) 0.05 0.25 0.30 Alkali metals (Cs, Rb) 0.05 0.20 0.35 Tellurium group (Te, Sb, Se) 0 0.05 0.25 Barium, strontium (Ba, Sr) 0 0.02 0.1 Noble metals (Ru, Rh, Pd, Mo, 0 0.0025 0.0025 Tc,Co) Cerium group (Ce, Pu, Np) 0 0.0005 0.005 Lanthanides (La, Zr, Nd, Eu, Nb, 0 0.0002. 0.005 Pm, Pr, Sm, Y, Cm, Am)

               *Reforence: Table 3-12 from NUREG-1465.

5:3 Gaseous Dispersion Factors The dispersion factors are taken from the ODAM Section 3. Table 7 - Dispersion Factors Dose due to Organ Dose Due to Plume/Submersion Particulates and Iodine ODAM Sections 3.5.2.1 and ODAM Section 3.8 3.9 Offgas Stack 2.8E-7 sec/m 3 3.1 E-7 sec/m3 Building Vents 4.3E-6 sec/m 3 3.9E-6 sec/m3 5.4 Isotopic half-lives Isotopic half-lives are taken from NUREG-1940,- Supplement 1. For those isotopes missing from that list, denoted by*, half-lives were obtained from the following website which is maintained by the Korea Atomic Energy Research Institute: http://atom.kaeri.re.kr:BOBO/ton/index. html Page 9 of34

CALC a ENE RC.ON Exce/lence~Every_ project Every_ day. Revised Gaseous Radiological EALs per NEI 99-01 Rev. 06 NO. REV. NEE-323-CALC-005 00 Table 8 contains the half-lives and calculated A (lambda) values. Table 8- Half-lives and Decay Constants Decay Tl/2 Tl/2 Isotope Tl/2 Lambda units Hours hrs*1

                                  ' .-.~-Mo-99 2.75                ~iJxs.,_  6.60E+Ol                                        1.05E-02 Nb-95                35.2            . ~iJY~-       8.45E+02                                        8.20E-04 Nd-147                   il                 days _    2.64E+02

• 2.63E-03

                                                                                                                                           . . ... ~ -  '  * "*? * * - .-,.-.,~
                                      ~p~23~                  2.36            . ciiJy~      5.66E+Ol                                        1.22E-02 Pr-143                 ~}-,~. _ _          days      3.26E+02                                        2.12E-03
                                                                                                                                 - . . . ,....,._.., .... --~? .,.~ ...

Pu-241 5260 . . . -~iJY~. l.26E+OS 5.49E-06 Rb-86 1,8-7, _ __days_ 4.4.9E+02 . 1.54E-03 .

                                                                                                                                           ..,.,. ..., ,,.~,.-. ,-~-

Rh-105 1.47 ciiJy~ _ 3.53E+Ol 1.96E-02 Ru-103 39.3 _ciays 9.43E+02 7.35E-04 Ru-105 0.185 ___ ciiJys_ 4.44E+OO 1.56E-01 Ru-106 368 ciiJy~ 8.83E+03 7.85E-05 Sb-127 3.85 ciiJys. 9.24E+Ol 7.SOE-03 Sb-129* 4.4 hours 4.40E+OO 1.58E-Ol Sr-89 50.5 ciiJys 1.21E+03 5.72E-04 Sr-90 10600 ~ays 2.54E+05 2.72E-06 Sr-91 0.396 days_ 9.SOE+OO 7.29E-02 Sr-92 0.113 ciiJys. 2.71E+OO 2.56E-01 Tc-99m 0.251 . ci1Jys 6.02E+OO 1.15E-01 Te-127 0.39 _ciays 9.36E+OO 7.41E-02 Te-127m 109 ___ci_iJ}'S 2.62E+03 2.65E-04 Te-129 0.0483 ciay~,- 1.16E+OO 5.98E-01 Te-129m 33.6 ciiJys 8.06E+02 8.60E-04 Page 10 of 34

CALC NEE-323-CALC-005 CJ ENERCON Ex,e/lence~Every project Ivery day. Revised Gaseous Radiological EALs per NEI 99-01 Rev. 06 NO. REV. 00 Decay Tl/2 Tl/2 Isotope Tl/2 Lambda units Hours Te-131m 1.25 ~ays 3.00E+Ol 2.31E-02

                                                                                                           .. ,..----~-- ...   *****-

Te-132

                                -.-....* *.~

3.26 _days 7.82E+01 8.86E-03 Xe-131m* 11.934 ~i!'{S 2.86E+02 2.42E-03 Xe-133 5.25 ~~ys 1.26E+02 5.SOE-03 Xe-133m* 2.19 ~,ay~. 5.26E+01 1.32E-02 Xe-135 0.379 ..dax~. 9.10E+OO 7.62E-02 Xe-135m* 15.29

                                                            .., *....... minutes                2.SSE-01   2.72E+OO
                                                                                                            ** ******--v. *-*.-.~..,

Xe-138* 14.08 minutes 2.35E-01 2.95E+OO

                                                                         "*"~ -**.$.-.,,,_-..,.

Y-90 2.67 d~ys 6.41E+01 1.08E-02

                                                                                                           .-.,, .*. ,... ,  ,.-- ~.

Y-91 58.5 days 1.40E+03 4.94E-04 Y-92 0.148 ~avs . 3.SSE+OO 1.95E-01 Y-93 0.421 ~~_ys_ . 1.01E+01 6.86E-02 Zr-95 64 d~y~. 1.54E+03 4.SlE-04 Zr-97 0.704 days 1.69E+01 4.lOE-02 5.5 Reduction Factor for Sprays NUREG-1940 Table 1-11 states that when sprays are used for longer than 1. 75 hours (but less than 2.25 hou*rs), the following factor is applied to reduce all of the particulate and iodine species. RFs =Exp(-o.s4t) Where t = the amount of tim_es sprays are in service. Note: This reduction factor does not apply to the noble gas~s. For this. calculation, sprays are used for a total of 2 hours as described in Section 6.1. The reduction factor is: RFs = e(-o.54*2l = 0.278 5.6 Standby Gas Treatment Filters NUREG.:1940 allows a reduction factor of 0.01 for filters like the standby gas treatment (SBGT) system. This factor is only applied to releases from the Offgas Stack. RFF= 0.01

5. 7 Secondary Containment NUREG-1228 provides a reduction factor for natural removal through settling and plate-out in the secondary containment. For a 0.5 hour holdup period, that reduction factor is 0.4. This factor is applied to the building vent releases but not the release from the Offgas Stack.

RFsc= 0.4 Page 11 of34

CALC NEE-323-CALC-005 g ENERCON. Exce/tence-~very project. Every day. Revised Gaseous Radiological EALs per NEI 99-01 Rev. 06 NO. REV. 00 5.8 Monitor Range and Exhaust Flow Rates Table 9 is developed from the DAEC Emergency Plan Section "I", ODAM Figure 3-1, and Procedure PCP 8.3. Table 9 - Monitor Range and Exhaust Flow Rates

                              .,                     *.Mo~itof             ,,             Monitor          '.Release,:

Release Point Common. Equipment ID Range .Flow

• Nam~,

                                                                              *.           µCi/cc

• CFM

                                 '. :      ..    . t;        ~ ~_;<'
                                                                        . ,i;,        .**

KAMAN RE-5945 / RE- lE-7 to Turbine Building 72,000 1/2 5946 lE+S KAMAN RE-7645, RE-3/4 7644 KAMAN RE-7647, RE- lE-7 to Reactor Building 93,000 5/6 7646 lE+S KAMAN RE-7649, RE-7/8 7648 KAMAN RE-4176, RE- lE-7 to Offgas Stack 10,000 9/10 4175 lE+S lE-7 to LLRPSF KAMAN 12 RE-8801 75,000 3E-1 5.9 Breathing Rate From NUREG-1940 and FG~11, the breathing rate is 3.33E-4 m3/second. 5.1 OExposure-to-Dose Conversion Factors for Inhalation The "Exposure-to-Dose Conversion Factors for Inhalation" by radionuclide provided in FGR11 _Table 2.1 allow the determination of the committed dose equivalent to the thyroid and the effective dose equivalent per unit per unit intake, and are shown in Table 11. 5.11 Dose Coefficients for Air Submersion The dose coefficients in Sv/Bq*s*m-3 from being submersed in air for each radionuclide to an effective dose are taken from Table 111.1 of FGR12, and are shown in Table 11. Page 12 of 34

CALC NEE-323-CALC-005 lg ENERCO.N Revised Gaseous Radiological EALs per NEI 99-01 Rev. 06 NO. E~l(ence~E.ve_ry projec;t. E't'.ery d~y. REV. 00 6.0 Methodology This calculation will equate a radioactive material release rate as measured at the gaseous effluent radiation monitors with the dose received to a member of the public at an offsite location. The relationship is highly influenced by the mixture of radioisotopes in the effluent and the dispersion of gases after they have left the facility. Primary guidance is provided by NUREG-1940 and NUREG-1228. 6.1 Scenario The following generalized timeline is used to determine the phenomenon that can affect the mixture of radioisotopes in effluent. This scenario is realistic, but bounds an event that could occur in a shorter total time frame:

• T= O hr. Major recirculating system line break occurs. Reactor is shut down.
• T= 1 hr. Core is uncovered.
• T= 1 hr. Sprays are initiated.
• T = 2 hrs. Core is covered.
• T= 4.5 hrs. A catastrophic event causes damage to the drywall and the sec-ondary containment.

o The gaseous mixture from the Drywall spreads into the Reactor Build-ing, Turbine Building, and LLRPSF. o Mean average holdup time of the gas in these buildings is 0.5 hours. Scenario timing will affect the mixture of radioisotopes and is summarized here:

• The core is uncovered for 1 hour. 1
• Core/Drywall Sprays are running for a total of 2 hours. *
• Primary Containment integrity is maintained for 4 hours.
• Source holdup time in secondary containment is 0.5 hours.

        ~      Source decay time from shutdown to the release point is 5 hours.

• When the reactor.is in mode 4 or 5, the total decay time is 36 hours.

Other Factors:

• The flow rates from the effluent exhaust points are listed in Design Input 5.8.
• The gaseous effluent radiation monitors are equally efficient for the monitor-ing of noble gases, particulates, and iodines.
• All releases from the Offgas Stack are filtered by the Standby Gas Treatment system.
• Removal of particulates and iodines by natural process during holdup in sec-ondary containment are credited for releases from the building vents only.

6.2 Receptor The receptor is an adult located at the ODAM-described location of minimal dispersion who is exposed to the radioactive release for one hour. Due to this relatively short duration, the only exposure pathways are inhalation and submersion. Assumptions related to the receptor are found in Section 4.0. Page 13 of 34

CALC a ENERCON Excef(ence-::Eve_ry projec;t. .Every ddy. Revised Gaseous Radiological EALs per NEI 99-01 Rev. 06 NO. REV. NEE-323-CALC-005 00 6.3 General Approach With a given mixture of radionuclides, the dose received by an individual offsite is a function of the gross activity present in the gaseous mixture. The resultant dose received by an offsite receptor is dependent not only on the gross radioactivity levels of the effluent but also upon the isotopic mixture present in the gas. This calculation predicts the relative contribution of each radionuclide to the gross radiation monitored by the effluent monitor. With the fractionation of the mixture of radionuclides understood, a given gross output reading (µCi/cm 3 ) from an effluent radiation monitor can be scaled to determine the concentration of each isotope present in the effluent. The calculation then uses default dispersion factors described in the Offsite Dose Assessment Manual to determine the resultant concentration of radionuclides to which an individual offsite would be exposed. Dose conversion factors provided in Federal Guidance Report 11 (FGR11) and 12 (FGR12) are used to determine the dose (mrem) to an individual offsite due to their exposure to the gaseous mixture of radionuclides. With the given radionuclide mixture and dispersion factors understood, an iterative process can be used to relate the effluent monitor reading to a target offsite dose. Two types of radiation dose are calculated: 1) TEDE and 2) CDE Thyroid. CDE or Committed Dose Equivalent is the radiation dose to a specific organ due to an uptake of radioactive material. In this case, the uptake is limited to inhalation of radioactive material in the plume. TEDE or Total Effective Dose Equivalent is the summation of the Effective Dose Equivalent (EDE) and the Committed Dose Equivalent (CEDE). TEDE = EDE + CEDE. EDE is the dose due to an individual being directly exposed (by submersion) to the radiation present in the gaseous release (shine). CEDE is the sum of the CDE for each organ of the body with weighting factors applied 1 for each organ. In this calculation, only contributions from the inhalation pathway are considered. An iterative process is used to determine the gross radiation monitored by the effluent monitors that correspond to the threshold doses. 6.4 Source Term This calculation will not analyze for the total activity released from the core. It will only analyze for the ratios of the isotopic species that are released from the core. Various phenomena will act to change the composition of the isotopic mixture in the time between reactor shutdown and release from the facility. In summary the removal phenomena addressed here include: Page 14 of 34

CALC NEE-323-CALC-005 ~ ENERCON Exc<:llence-;-:Every project. Every day. Revised Gaseous Radiological EALs per NEI 99-01 Rev. 06 NO. REV. 00 RF1 = Fraction of the activity released from the inventory of damaged fuel described in Section 6.5. RFs = Fraction of the activity remaining after reduction by containment spray from Section 5.5. RFR = Fraction of activity remaining after 5 hours or 36 hours of radioactive decay described in Section 6.6. RFF = Fraction of the activity remaining after filter by SBGT filters from Section 5.6. RFsc = Fraction of activity remaining after natural removal processes in secondary containment from Section 5.7. Combining these factors provides a single fraction to derive a depleted source: RFTotal = RF1

• RFs *RFR
• RFF
• RFsc 6.5 Fuel Damage Release Fractions Table 6 contains release fractions for three time periods representing the total amount of time the core has been assumed to be uncovered. They are: Oto 0.5 hours, 0.5 to 2 hours, and 2 to 5 hours. For this calculation, the core is assumed to be uncovered for one hour. A spreadsheet is used to scale the release fraction between the 0.5 hour point and the 2 hour point.
• The Reduction Factor, RF1, due to the release fraction is 100% of the release expected in the first 0.5 hour PLUS t/3 of the amount released as expected in the period between 0.5 hours and 2 hou'rs.

Example for Alkali Metals: 0.05 .

                                                              * (00.5*5 hhr) r
                                                                             +  0.2  * (05

• hhr) = 0.116 7 1.5 r Table 10- Release Fractions by Time Step (Hours)

Group Time (h) by step 0.5 1.5 Cumulative Alkali Metals 0.050 0.2000 0.1167 Barium Group 0.000 0.0200 0.0067 Cerium Group 0.000 0.0005 0.0002 Halogen 0.050 0.2500 0.1333 Lanthanides 0.000 0.0002 0.0001 Noble Gas 0.050 0.9500 0.3667 Noble Metals 0.000 0.0025 0.0008 Tellurium group 0.000 0.0500 0.0167 Page 15 of 34

CALC NEE-323-CALC-005 lg ENERC.ON Revised Gaseous Radiological NO. Excellence--::-Every projec;t. Every day. EALs per NEI 99-01 Rev. 06 REV. 00 6.6 Radioactive Decay The total amount of time the radioactive source is allowed to decay before being exhausted as an effluent is 5 hours or 36 hours depending on the reactor mode per Section 6.1. The generalized equation for radioactive decay is: A= Aoe<-At> Where: A = decayed activity Ao= initial activity A = isotopic decay constant t = elapsed time and A= ln2 / t% With an end goal of a total reduction factor RFTota1, a radiation decay factor RFR is derived from the general equation above: RFR = e<-At>

6. 7 Effective Dose Equivalent - Noble Gas Submersion Submersion dose from noble gases is.calculated with guidance provided in FGR12.

The concentration of an isotope i present in the plume at the receptor is calculated: Xir= Xiv* v* (~) With the isotopic concentration at the receptor known, the dose (mrem) at the receptor is calculated: Dose= Li(Xir

• hEsoa Where concentration of radionuclide ; present at the receptor (Ci/m3)

Note: Ci/m 3 = JJCi/cc volume of gas released (m3 ) concentration of radionuclide ; released from the stack or building vent. (Ci/m 3 ) i= each isotope present in the gaseous release (~)= dispersion factor for that release point (sec/m 3 ) hEsoi= factor converting the gas concentration to effective dose equivalent. 3 mre7:1 crn ) ( µCi sec Page 16 of34

CALC NEE-323-CALC-005 0 ENERCON E_xcel(ence:~ve/y projecr. fll'.ef)' defy. Revised Gaseous Radiological EALs per NEI 99-01 Rev. 06 NO. REV. 00 As described in Section 7 .3, a spreadsheet is used to determine the EDE dose contribution for each isotope in the mixture. 6.8 Committed Dose Equivalent: Thyroid Organ dose from airborne particulates and iodines is calculated with guidance provided in FGR11. The concentration of an isotope i present in the plume at the receptor is calculated: Xir= Xiv* v* (~) With the isotopic concentration at the receptor known, the dose (mrem) at the receptor can be calculated: Dose = Li(Xir

• B
• t
• hrsoD Where concentration of radionuclide ; present at the receptor (Ci/m 3 )

Note: Ci/m 3 = µCi/ cm 3 v= volume of gas released (m 3 )

 *xiv=             concentration of radionuclide ; released from the stack or building vent.

(Ci/m 3 ) i= each isotope present in the gaseous release (~}= disp~rsion factor for that release point (sec/m 3 ) B= breathing Rate (cm 3/sec) hrsoi= factor converting the gas concentration to effective dose equivalent. (mrem/µCi) t= " time the dose is to be integrated (sec) As described in Section 7.4, a spreadsheet is used to determine the thyroid CDE dose contribution for each isotope in the mixture. 6.9 Committed Effective Dose Equivalent Committed Effective Dose Equivalent from airborne particulates and iodines is calculated with guidance provided in FGR11. The concentration XirOf an isotope i present in the plume at the receptor is calculated: Xir= Xiv* v* (~) With the isotopic concentration at the receptor known, the dose (mrem) at the receptor can be calculated: Dose = LiCXir

• B
• t
• hEsOi)

Where Page 17 of 34

CALC NEE-323-CALC-005 a ENEI{CON Excellence-Every project. Every day. Revised Gaseous Radiological EALs per NEI 99-01 Rev. 06 NO. REV. 00 concentration of radionuclide ; present at the receptor (Ci/m3 ) Note: Ci/m3 = µCi/ cm 3 v= volume of gas released (m 3) Xiv= concentration of radionuclide i released from the stack or building vent. (Ci/m 3 ) i= each isotope present in the gaseous release (~)= dispersion factor for that release point (sec/m 3 ) B= breathing Rate (cm 3/sec) factor converting the gas concentration to effective dose equivalent. (mrem/µCi) t= time the dose is to be integrated (sec) As described in Section 7.4, a spreadsheet is used to determine the CEDE dose contribution for each isotope in the mixture. Page 18 of 34

CALC

.                                                                                                       NEE-323-CALC-005

~ ENERCON Exce/lence~Every project. Every ddy. Revised Gaseous Radiological EALs per NEI 99-01 Rev. 06 NO. REV. 00 7 .0 Calculation All calculations were completed using Microsoft Excel. Sample calculations are shown in the subsections that follow. 7.1 Dose Factors FGR11 and FGR 12 display dose factors in the SI units of Sv/Bq and Sv m3/ Bq sec, respectively. Traditional units of mrem/µCi and mrem cm 3/µCi sec are desired. FGR11: 1 &¥ 1E+05 mrem 3.70E+09 mrem

                 --t--+----+---+----+---+-------=----+---

2.7E-11 Q 1.00E+G µCi µCi The conversion factor from Sv/Bq to mrem/µCi is 3. 70E+09. FGR 12: 1 s,.. ffia. 1E+05 mrem 1 ~ lE+OG ml Q 3.70E+l5 mrem cm3

                                                                                                     =
    ~        sec                    &¥           2.7E-11    Q                 ,ffia. lE+OG µCl                           µCi sec The conversion factor from Sv m 3/Bq sec to mrem cm3/µCi sec is 3.70E+15.

The thyroid, CEDE, and submersion dose factors in the traditional units for each isotope are calculated in the table below. Column C, D, and H are dose factors from Sections 5.10 and 5.11 and Columns E and I are the conversion factors from above. ~ Column F, G, and J are the hT50i, hE5oi, and hEsoi factors as described in Sections 6.8, 6.9, and 6.7, respectively. Line 6 of Table 11 illustrates the formulas for Ba-139. Table 11- Isotopic Dose Factors FGRll FGRll Units Thyroid CEDE FGR 12: Units Submersion Isotope Thyroid CEDE Conversion mrem Svm3 Conversion mrem cc Sv Sv IDifil!l Bq Bq Factor µCi µCi Bqsec Factor µCi sec 2.40E-12 4.64E-11 3.70E+09 =E6*C6 =E6*D6 2.llE-15 3.70E+15 =l6*H6 Ba-139 2.40E-12 4.64E-ll 3.70E+09 8.88E-03 1.72E-Ol 2.17E-15 3.7E+l5 8.03E+OO Ba-140 2.SGE-10 1.0lE-09 3.70E+09 9.47E-01 3.74E+OO 8.SBE-15 3.7E+l5 3.17E+Ol Ce-141 4.GlE-11 2.42E-09 3.70E+09 1.71E-01 8.95E+OO 3.43E-15 3.7E+l5 l.27E+Ol Ce-143 1.21E-11 9.lGE-10 3.70E+09 4.48E-02 3.39E+OO 1.29E-14 3.7E+15 4.77E+Ol Ce-144 1.88E-09 1.0lE-07 3.70E+09 6.96E+OO 3.74E+02 8.53E-16 3.7E+15 3.16E+OO Cm-242 9.41E-10 4.67E-06 3.70E+09 3.48E+OO 1.73E+04 5.69E-18 3.7E+l5 2.llE-02 Cs-134 1.llE-08 1.25E-08 3.70E+09 4.llE+Ol 4.63E+Ol 7.57E-14 3.7E+15 2.80E+02 Cs-136 1.73E-09 1.98E-09 3.70E+09 6.40E+OO 7.33E+OO 1.0GE-13 3.7E+l5 3.92E+02 Cs-137 7.93E-09 8.63E-09 3.70E+09 2.93E+Ol 3.19E+Ol 7.74E-18 3.7E+l5 2.BGE-02 1-131 2.92E-07 8.89E-09 3.70E+09 l.08E+03 3.29E+Ol 1.82E-14 3.7E+l5 6.73E+Ol 1-132 1.74E-09 1.03E-10 3.70E+09 6.44E+OO 3.81E-01 1.12E-13 3.7E+15 4.14E+02 1-133 4.86E-08 1.58E-09 3.70E+09 1.80E+02 5.85E+OO 2.94E-14 3.7E+l5 1.09E+02 Page 19 of 34

CALC NEE-323-CALC-005 ly ENERCON Excellence-Every projecr. Every day. Revised Gaseous Radiological EALs per NEI 99-01 Rev. 06 NO. REV. 00 FGR11 FGR11 Units Thyroid CEDE FGR 12: Units Submersion Thyroid CEDE Isotope Conversion .!!llfilD. .!!llfilD. Svm 3 Conversion *mrem cc Sv Sv Factor µCi µCi Bq sec Factor µCi sec Bq Bq 1-134 2.88E-10 3.SSE-11 3.70E+09 1.07E+OO 1.31E-01 1.3E-13 3.7E+15 4.81E+02 1-135 8.46E-09 3.32E-10 3.70E+09 3.13E+Ol 1.23E+OO 7.98E-14 3.7E+15 2.95E+02 Kr-83m 1.SE-18 3.7E+15 5.SSE-03 Kr-85 1.19E-16 3.7E+15 4.40E-01 Kr-8Sm 7.48E-15 3.7E+15 2.77E+01 Kr-87 4.12E-14 3.7E+15 1.52E+02 Kr-88 1.02E-13 3.7E+15 3.77E+02 La-140 1.22E-10 1.31E-09 3.70E+09 4.SlE-01 4.8SE+OO 1.17E-13 3.7E+15 4.33E+02 La-141 9.40E-12 1.57E-10 3.70E+09 3.48E-02 5.81E-01 2.39E-15 3.7E+15 8.84E+OO La-142 8.74E-12 6.84E-11 3.70E+09 3.23E-02 2.53E-01 1.44E-13 3.7E+15 5.33E+02 Mo-99 1.17E-10 1.07E-09 3.70E+09 4.33E-01 3.96E+OO 7.28E-15 3.7E+15 2.69E+Ol Nb-95 3.58E-10 1.57E-09 3.70E+09 1.32E+OO 5.81E+OO 3.74E-14 3.7E+15 1.38E+02 Nd-147 1.94E-11 1.85E-09 3.70E+09 7.18E-02 6.85E+OO 6.19E-15 3.7E+15 2.29E+Ol Np-239 7.62E-12 6.78E-10 3.70E+09 2.82E-02 2.51E+OO 7.69E-15 3.7E+15 2.85E+Ol Pr-143 1.68E-18 2.19E-09 3.70E+09 6.22E-09 8.lOE+OO 2.lE-17 3.7E+15 7.77E-02 Pu-241 1.24E-11 2.23E-06 3.70E+09 4.59E-02 8.25E+03 7.25E-20 3.7E+15 2.68E-04 Rb-86 1.33E-09 1.79E-09 3.70E+09 4.92E+OO 6.62E+OO 4.81E-15 3.7E+15 1.78E+Ol Rh-105 2.57E-11 2.58E-10 3.70E+09 9.SlE-02 9.SSE-01 3.72E-15 3.7E+15 1.38E+Ol Ru-103 5.97E-10 2.42E-09 3.70E+09 2.21E+OO 8.95E+OO 2.25E-14 3.7E+15 8.33E+01 Ru-105 1.SOE-11 1.23E-10 3.70E+09 5.SSE-02 4.SSE-01 3.81E-14 3.7E+15 1.41E+02 Ru-106 1.37E-08 1.29E-07 3.70E+09 5.07E+01 4.77E+02 0 3.7E+15 O.OOE+OO Sb-127 1.SOE-10 1.63E-09 3.70E+09 5.SSE-01 6.03E+OO 3.33E-14 3.7E+15 1.23E+02 Sb-129 2.07E-11 1.74E-10 3.70E+09 7.66E-02 6.44E-01 7.14E-14 3.7E+15 2.64E+02 Sr-89 4.lGE-10 1.12E-08 3.70E+09 1.54E+OO 4.14E+Ol 7.73E-17 3.7E+15 2.86E-01 Sr-90 2.64E-09 3.SlE-07 3.70E+09 9.77E+OO 1.30E+03 7.53E-18 3.7E+15 2.79E-02 Sr-91 4.08E-11 4.49E-10 3.70E+09 1.SlE-01 1.66E+OO 3.4SE-14 3.7E+15 1.28E+02 Sr-92 2.19E-11 2.18E-10 3.70E+09 8.lOE-02 8.07E-01 6.79E-14 3.7E+15 2.51E+02 Tc-99m S.OlE-11 8.80E-12 3.70E+09 1.85E-01 3.26E-02 5.89E-15 3.7E+15 2.18E+Ol Te-127 6.46E-12 8.60E-11 3.70E+09 2.39E-02 3.18E-01 2.42E-16 3.7E+15 8.95E-01 Te-127m 2.39E-10 5.81E-09 3.70E+09 8.84E-01 2.lSE+Ol 1.47E-16 3.7E+15 5.44E-01 Te-129 1.63E-12 2.42E-11 3.70E+09 6.03E-03 8.95E-02 2.75E-15 3.7E+15 1.02E+01 Te-129m 3.95E-10 6.47E-09 3.70E+09 1.46E+OO 2.39E+Ol 1.SSE-15 3.7E+15 5.74E+OO Te-131m 3.61E-08 1.73E-09 3.70E+09 1.34E+02 6.40E+OO 7.0lE-14 3.7E+15 2.59E+02 Te-132 6.28E-08 2.SSE-09 3.70E+09 2.32E+02 9.44E+OO 1.03E-14 3.7E+15 3.81E+01 Xe-131m 3.89E-16 3.,7E+15 1.44E+OO Xe-133 1.56E-15 3.7E+15 5.77E+OO Xe-133m 1.37E-15 3.7E+15 S.07E+OO Xe-135 1.19E-14 3.7E+15 4.40E+Ol Xe-135m 2.04E-14 3.7E+15 7.SSE+Ol Xe-138 5.77E-14 3.7E+15 2.13E+02 Y-90 9.52E-12 2.28E-09 3.70E+09 3.52E-02 8.44E+OO 1.9E-16 3.7E+15 7.03E-01 Y-91 1.lOE-10 1.32E-08 3.70E+09 4.07E-01 4.88E+Ol 2.6E-16 3.7E+15 9.62E-01 Y-92 3.69E-12 2.llE-10 3.70E+09 1.37E-02 7.81E-01 1.3E-14 3.7E+15 4.81E+01 Y-93 5.06E-12 S.82E-10 3.70E+09 1.87E-02 2.15E+OO 4.8E-15 3.7E+15 1.78E+Ol Zr-95 1.44E-09 6.39E-09 3.70E+09 5.33E+OO 2.36E+01 3.6E-14 3.7E+15 1.33E+02 Zr-97 9.56E-11 1.17E-09 3.70E+09 3.54E-01 4.33E+OO 9.02E-15 3.7E+15 3.34E+Ol Page 20 of34

CALC a* EI\JERCON It Excellence...:,Ev~ry project, Every day. Revised Gaseous Radiological EALs per NEI 99-01 Rev. 06 NO. REV. NEE-323-CALC-005 00 7.2 Source Term A spreadsheet is used to determine the total reduction factor RFTota1 for each isotope present in the source term as described in Section 6.4. The activity per megawatt thermal from Section 5.1 is multiplied by RFTota1 to find the* source term for each isotope. The spreadsheet for the Offgas Stack release is presented in Table 13. The relative activity released from damaged fuel (RF1) was determined in Section 6.5. Table 12- 2 Hours Reduction Factor (RF,) Cumulative 2 Hour Alkali Metals 0.1167 Barium Group 0.0067 Cerium Group 0.0002 Halogen 0.1333 Lanthanides 0.0001 Noble Gas 0.3667 Noble Metals 0.0008 Tellurium group 0.0167 A Spray Reduction factor of 0.278 for primary containment sprays (RFs> was derived in Section 5.5. Determination of the Radiation Decay fractions (RFR) was demonstrated in Section 6.6. In the spreadsheets below, the source decay time is 5 hours. Page 21 of34

CALC NEE-323-CALC-005 0- ENERCON Exceltence-~very project Every day. Revised Gaseous Radiological EALs per NEI 99-01 Rev. 06 NO. REV. 00 7 .2.1 Offgas Stack For the Offgas Stack release, credit is taken for filtering (RFF) by the Standby Gas Treatment system but not for the natural removal processes that occur in secondary containment (Rfsc). Table 13- Jsotopic Depletion and Release for Offgas Stack RF1 RFs Rfsc RFF RFR RFTotal

                                                                           +0.25 hr g      Release                 Secondary  SBGT          Decay      Total    Release Form                Isotope                                      Sprays MWTh    Fraction                    Con-    Filter       Fraction  Depletion Ci/MWTh Reduction tainment Barium Group                 Ba-139          4.74E+04     0.0067         0.2780   1.0000     0.01        0.0808   1.50E-06   7.lOE-02 Barium Group                 Ba-140          4.76E+04     0.0067         0.2780   1.0000     0.01        0.9887   l.83E-05   8.72E-Ol Cerium Group                 Ce-141          4.39E+04     0.0002         0.2780   1.0000     0.01        0.9956   4.61E-07   2.03E-02 Cerium Group                 Ce-143          4.00E+04     0.0002         0.2780   1.0000     0.01        0.9006   4.17E-07   1.67E-02 Cerium Group                 Ce-144          3.54E+04     0.0002         0.2780    1.0000    0.01        0.9995   4.63E-07   1.64E-02 Lanthanides               Cm-242           1.12E+03     0.0001         0.2780    1.0000    0.01        0.9991   1.85E-07   2.07E-04 Alkali Metals               Cs-134          4.70E+03     0.1167         0.2780    1.0000    0.01        0.9998   3.24E-04   1.52E+OO Alkali Metals               Cs-136          1.49E+03     0.1167         0.2780    1.0000    0.01        0.9890   3.21E-04   4.78E-01 Alkali Metals               Cs-137          3.25E+03     0.1167         0.2780    1.0000    0.01         1.0000  3.24E-04   1.05E+OO Halogen                1-131          2.67E+04     0.1333         0.2780    1.0000    0.01        0.9822   3.64E-04   9.72E+OO Halogen                1-132          3.88E+04     0.1333         0.2780    1.0000    0.01        0.2215   8.21E-05   3.19E+OO Halogen                1-133          5.42E+04     0.1333         0.2780    1.0000    0.01        0.8466   3.14E-04   1.70E+Ol Halogen                1-134          5.98E+04     0.1333         0.2780    1.0000    0.01        0.0191   7.09E-06   4.24E-01
        . Halog~n_               1-135          5.18E+04     0.1333         0.2780    1.0000    0.01
                                                                                                 ,~ ., .

0.5915 2.19E-04 1.14E+Ol Noble Gas* Kr-83m 3.05E+03 0.367 1.0 1.0 1.0 0.1505 5.52E-02 1.68E+02 Noble Gas Kr-85 2.78E+02 0.361 LO. 1.0 1.() 1.0000 3.67E-01 1.02E+02

                                                                ~

N~bleG~s Kr-85m 6.17E+03 0.367 1.0 1.0 1.0 0.4620 1.69E-Ol l.05E+03

     *Nobie Gas                  Kr-87           l.23E+04     0.367           1.0       Lb      ,1.0 0.0656   2.40E-02   2.96E+02
    ***~oble Gas                 Kr-88           1.70E+04     0.367           1.0       1.0      1.0   _,,_

0.2941 l.08E-01 1.83E+03 Lanthanides La-140 4.91E+04 0.0001 0.2780 1.0000 0.01 0.9176 1.70E-07 8.35E-03 Lanthanides La-141 4.33E+04 0.0001 0.2780 1.0000 0.01 0.4146 7.68E-08 3.33E-03 Lanthanides La-142 4.21E+04 0.0001 0.2780 1.0000 0.01 0.1055 1.96E-08 8.23E-04 Noble Metals Mo-99 5.30E+04 0.0008 0.2780 1.0000 0.01 0.9488 2.20E~06 1.17E-Ol Lanthanides Nb-95 4.50E+04 0.0001 0.2780 1.0000 0.01 0.9959 1.85E-07 8.31E-03 Lanthanides Nd-147 1.75E+04 0.0001 0.2780 1.0000 0.01 0.9870 1.83E-07 3.20E-03 Cerium Group Np-239 5.69E+05 0.0002 0.2780 1.0000 0.01 0.9406 4.36E-07 2.48E-Ol Lanthanides Pr-143 3.96E+04 0.0001 0.2780 1.0000 0.01 0.9894 l.83E-07 7.26E-03 Cerium Group Pu-241 4.26E+03 0.0002 0.2780 1.0000 0.01 1.0000 4.63E-07 1.97E-03 Alkali Metals Rb-86 5.29E+Ol 0.1167 0.2780 1.0000 0.01 0.9923 3.22E-04 1.70E-02 Noble Metals Rh-105 2.81E+04 0.0008 0.2780 1.0000 0.01 0.9064 2.lOE-06 5.90E-02 Noble Metals Ru-103 4.34E+04 0.0008 0.2780 1.0000 0.01 0.9963 2.31E-06 1.00E-01 Noble Metals Ru-105 3.06E+04 0.0008 0.2780 1.0000 0.01 0.4581 1.06E-06 3.25E-02 Noble Metals Ru-106 1.55E+04 0.0008 0.2780 1.0000 0.01 0.9996 2.32E-06 3.59E-02 Tellurium group Sb-127 2.39E+03 0.0167 0.2780 1.0000 0.01 0.9632 4.46E-05 1.07E-01 Tellurium group Sb-129 8.68E+03 0.0167 0.2780 1.0000 0.01 0.4549 2.llE-05 l.83E-01 Barium Group Sr-89 2.41E+04 0.0067 0.2780 1.0000 0.01 0.9971 1.85E-05 4.45E-01

  . Barium Group                 Sr-90           2.39E+03    0.0067         0.2780    1.0000    0.01         1.0000  l.85E-05   4.43E-02 Page 22 of34

CALC NEE-323-CA.LC-005 g ENERCON Excellence.,.,.Every project. Every day. Revised Gaseous Radiological EALs per NEI 99-01 Rev. 06 NO. REV. 00 RF1 RFs RFsc RFF RFR RFTotal

                                                                           + 0.25 hr g      Release                   Secondary            SBGT            Decay      Total    Release Form                     Isotope                                 Sprays MWTh      Fraction                     Con-               Filter        Fraction  Depletion Ci/MWTh Reduction tainment Barium Group                     Sr-91          3.01E+04   0.0067      0.2780         1.0000               0.01          0.6944   l.29E-05   3.87E-01 Barium Group                     Sr-92          3.24E+04   0.0067      0.2780         1.0000               0.01          0.2786   5.16E-06   1.67E-01 Noble Metals                  Tc-99m           4.37E+04   0.0008      0.2780         1.0000               0.01          0.5625   1.30E-06   5.70E-02 Tellurium group                   Te-127          2.36E+03   0.0167      0.2780         1.0000               0.01          0.6905   3.20E-05   7.55E-02 Tellurium group                 Te-127m           3.97E+02   0.0167      0.2780         1.0000               0.01          0.9987   4.63E-05   1.84E-02 Tellurium group                   Te-129          8.26E+03   0.0167      0.2780         1.0000               0.01          0.0503   2.33E-06   1.93E-02 Tellurium group                 Te-129m           1.68E+03   0.0167      0.2780         1.0000               0.01          0.9957   4.61E-05   7.75E-02 Tellurium group                 Te-13lm           5.41E+03   0.0167      0.2780         1.0000               0.01          0.8909   4.13E-05   2.23E-Ol Tellurium group                   Te-132          3.81E+04    0.0167     0.2780         1.0000               0.01          0.9567   4.43E-05   1.69E+OO
    * ** Nbbl~Gas

• Xe-131m 3.65E+02 0.367 1.0. .1.0
• 1.0 0.9880 3.62E-01 1.32E+02 Noble Gas Xe-133 Xe-133m 5.43E+04 1.72E+03 0.367 *1.0

                                                                              'LO**
                                                                                             . i.o to**
                                                                                                                . 1.0 i.o 0.9729 0.9362 3.57E-01 3.43E-01 1.94E+04 5.90E+02 N6bleGa~                                                  0\3157 Noble Ga}                    Xe-135          1.42E+04    .0.367        1,0*          i.o .                1.0. -    '  0.6832  2.SOE-01   3.56E+03
                                                                                                '. '~
        'No61e G~~*                 Xe-135m           1.15E+04     0.367   '_ 1.0            . i:o.      ,....
                                                                                                                  '1.0 *'       0.0000  4.55E-07   5.23E-03
 *'      Nob.!eGas                    Xe-138          4.56E+04    '(),36?    _).o              1.0 ..              1.0 .        0.0000  l.41E-07   6.45E-03 Lanthanides                     Y-90          2.45E+03    0.0001     0.2780          1.0000              0.01           0.9474  1.76E-07   4.30E-04 Lanthanides                     Y-91          3.17E+04    0.0001     0.2780          1.0000              0.01           0.9975  1.85E-07   5.86E-03
.       Lanthanides                     Y-92          3.26E+04    0.0001     0.2780          1.0000              0.01           0.3769  6.99E-08   2.28E-03 Lanthanides                     Y-93          2.52E+04    0.0001     0.2780          1.0000              0.01           0.7096  1.32E-07   3.31E-03 Lanthanides                    Zr-95          4.44E+04    0.0001     0.2780          1.0000              0.01           0.9977  l.85E-07   8.21E-03 Lanthanides                    Zr-97          4.23E+04    0.0001     0.2780          1.0000              0.01           0.8145  l.SlE-07   6.39E-03 Page 23 of34

CALC NEE-323-CALC-005 ~ ENERCON Revised Gaseous Radiological NO.

     ~llenc~-Every project. Every day.       EALs per NEI 99-01 Rev. 06 REV.        00 7.2.2       Building Vents For releases from Building Vents, no credit is taken for filtering (RFF) by the Standby Gas Treatment system. Credit is taken for the natural removal processes that occurs in secondary containment (RFsc). This source term also has radioactive decay occurring for 5 hours.

Table 14- Isotopic Depletion and Release for Building Vents RF1 RFs RFsc RFF RFR RFrotal

                                                           +0.25 hr g     Release             Secondary   SBGT      Decay     Total    Release Form              Isotope                             Sprays MWTh     Fraction               Con-     Filter   Fraction Depletion Ci/MWTh Reduction tainment Barium Group            Ba-139        4.74E+04   0.0067    0.2780     0.4000    1.00     0.0808   5.99E-05  2.84E+OO Barium Group            Ba-140        4.76E+04   0.0067    0.2780     0.4000    1.00     0.9887   7.33E-04  3.49E+01 Cerium Group            Ce-141        4.39E+04   0.0002    0.2780     0.4000    1.00     0.9956   1.85E-05  8.lOE-01 Cerium Group            Ce-143        4.00E+04   0.0002    0.2780     0.4000    1.00     0.9006   1.67E-05  6.68E-01 Cerium Group            Ce-144        3.54E+04   0.0002    0.2780     0.4000    1.00     0.9995   1.85E-05  6.56E-01 Lanthanides           Cm-242         1.12E+03   0.0001    0.2780     0.4000    1.00     0.9991   7.41E-06  8.30E-03 Alkali Metals           Cs-134        4.70E+03   0.1167    0.2780     0.4000    1.00     0.9998   1.30E-02  6.10E+01 Alkali Metals           Cs-136        1.49E+03   0.1167    0.2780     0.4000    1.00     0.9890   1.28E-02  1.91E+01 Alkali Metals           Cs-137        3.25E+03   0.1167    0.2780     0.4000    1.00      1.0000  1.30E-02  4.22E+01 Halogen                1-131 **    2.67E+04   0.1333    0.2780     0.4000    1.00     0.9822

• 1.46E-02 3.89E+02 Halogen 1-132 3.88E+04 0.1333 0.2780 0.4000 1.00 0.2215 3.28E-03 1.27E+02 Halogen 1-133 5.42E+04 0.1333 0.2780 0.4000 1.00 0.8466 1.26E-02 6.80E+02 Halogen 1-134 5.98E+04 0.1333 0.2780 0.4000 1.00 0.0191 2.84E-04 1.70E+01 Halogen 1-135 5.18E+04 0.1333 0.2780 0.4000 1.00 0.5915 8.77E-03 4.54E+02 Noble Gas Kr-83m 3.05E+03 0.367 i.o 1.0 1.0 0.1505 5.52E-02 1.68E+02 Noble Gas Kr-85 2.78E+02 0.367 1.0 1.0 1.0 1.0000 3.67E-01 1.02E+02 Noble Gas Kr-85m 6.17E+03 0.367 i.o 1.0 1.0 0.4620 1.69E-01 1.05E+03 Noble Gas Kr-87 1.23E+04 0.367 1.0 1.0 1.0 0.0656 2.40E-02 2.96E+02

  *Noble Gas               Kr-88       1.70E+04    0.367      1.0        1.0      LO       0.2941  1.08E-01  1.83E+03 Lanthanides            La-140        4.91E+04   0.0001    0.2780     0.4000    1.00      0.9176  6.80E-06  3.34E-01 Lanthanides            La-141        4.33E+04   0.0001    0.2780     0.4000    1.00      0.4146  3.07E-06  1.33E-01 La nth a nides         La-142        4.21E+04   0.0001    0.2780     0.4000    1.00     0.1055   7.82E-07  3.29E-02 Noble Metals             Mo-99        5.30E+04   0.0008    0.2780     0.4000    1.00      0.9488  8.79E-05  4.66E+OO Lanthanides             Nb-95        4.50E+04   0.0001    0.2780     0.4000    1.00     0.9959   7.38E-06  3.32E-01 Lanthanides            Nd-147        1.75E+04   0.0001    0.2780     0.4000    1.00     0.9870   7.32E-06  1.28E-01 Cerium Group            Np-239        5.69E+05   0.0002    0.2780     0.4000    1.00     0.9406   1.74E-05  9.92E+OO Lanthanides            Pr-143        3.96E+04   0.0001    0.2780     0.4000    1.00     0.9894   7.34E-06  2.91E-01 Cerium Group            Pu-241        4.26E+03   0.0002    0.2780     0.4000    1.00     1.0000   1.85E-05  7.90E-02 Alkali Metals            Rb-86        5.29E+01   0.1167    0.2780     0.4000    1.00      0.9923  1.29E-02  6.81E-01 Noble Metals            Rh-105        2.81E+04   0.0008    0.2780     0.4000    1.00      0.9064  8.40E-05  2.36E+OO Noble Metals            Ru-103        4.34E+04   0.0008    0.2780     0.4000    1.00      0.9963  9.23E-05  4.01E+OO Noble Metals            Ru-105        3.06E+04   0.0008    0.2780     0.4000    1.00      0.4581  4.25E-05  1.30E+OO Noble Metals            Ru-106        1.55E+04   0.0008    0.2780     0.4000    1.00     0.9996   9.26E-05  1.44E+OO Tellurium group          Sb-127        2.39E+03   0.0167    0.2780     0.4000    1.00     0.9632   1.79E-03  4.27E+OO Tellurium group          Sb-129        8.68E+03   0.0167    0.2780     0.4000    1.00     0.4549   8.43E-04  7.32E+OO Barium Group              Sr-89       2.41E+04   0.0067    0.2780     0.4000    1.00     0.9971   7.39E-04  1.78E+01 Page 24 of34

CALC NEE-323-CALC-005 QI ENERCONmellence~Every project. Every day. Revised Gaseous Radiological EALs per NEI 99-01 Rev. 06 NO. REV. 00 RF1 RFs RFsc RFF RFR RFTotal

                                                                        +0.25 hr Q        Release                        Secondary      SBGT     Decay     Total    Release Form                Isotope                                    Sprays MWTh      Fraction                            Con-       Filter* Fraction Depletion Ci/MWTh Reduction tainment Barium Group               Sr-90       2.39E+03    0.0067           0.2780           0.4000      1.00    1.0000   7.41E-04  1.77E+OO Barium Group               Sr-91       3.01E+04    0.0067           0.2780           0.4000      1.00    0.6944   5.15E-04  1.55E+01 Barium Group               Sr-92       3.24E+04    0.0067           0.2780           0.4000      1.00    0.2786   2.07E-04  6.69E+OO Noble Metals            Tc-99m         4.37E+04    0.0008           0.2780           0.4000      1.00    0.5625   5.21E-05  2.28E+OO Tellurium group           Te-127        2.36E+03    0.0167           0.2780           0.4000      1.00    0.6905   1.28E-03  3.02E+OO Tellurium group         Te-127m         3.97E+02    0.0167           0.2780           0.4000      1.00    0.9987   1.85E-03  7.35E-01 Tellurium group           Te-129        8.26E+03    0.0167           0.2780           0.4000       1.00   0.0503   9.32E-05  7.70E-01 Tellurium group         Te-129m         1.68E+03    0.0167           0.2780           0.4000       1.00   0.9957   1.85E-03  3.10E+OO Tellurium group         Te-131m         5.41E+03    0.0167           0.2780           0.4000       1.00    0.8909  1.65E-03  8.93E+OO Tellurium group           Te-132        3.81E+04    0.0167           0.2780           0.4000       1.00    0.9567  1.77E-03  6.76E+01 N~bl~**Gas           Xe-131m        3.65E+02     0;367              1;0             1.c:i       rn     0.9880  3.62E-01  1.32E+02
 .. Noble Gas             Xe-133        5.43E+04     9.367              i:o             1:0 "       i,o    0.9729  3.57E-01  1.94E+04

• Ntibi{Gas. Xe-133m 1.72E+03 *cB51** .**,~1.b .. :1.0\.* :i.1:0 0.9362 3.43E-01 5.90E+02
• Noble,. ' Gas

                   ~
                         .. Xe-135        1.42E+04    .0.367         . *. i.o             i.o .       rn*    0.6832  2.50E-01  3.56E+03 Nobli!Gas .          Xe-135m        1.15E+04     0.~3(57           *1.0

• 1.0
• io 0.0000 4.55E-07 5.23E-03
• Nbt>iJda~*

            *~ . --- - '

Xe-138 4.56E+04 *. :0:367.

                                                         ~"'-**** ..
                                                                           *1.0.......

1.() 1.0. 0.0000 1.41E-07 6.45E-03 Lanthanides Y-90 2.45E+03 0.0001 0.2780 0.4000 1.00 0.9474 7.02E-06 1.72E-02 Lanthanides Y-91 3.17E+04 0.0001 0.2780 0.4000 1.00 0.9975 7.40E-06 2.34E-01 Lanthanides Y-92 3.26E+04 0.0001 0.2780 0.4000 1.00 0.3769 2.79E-06 9.11E-02 Lanthanides Y-93 2.52E+04 0.0001 0.2780 0.4000 1.00 0.7096 5.26E-06 1.33E-01 Lanthanides Zr-95 4.44E+04 o.oq_oi 0.2780 0.4000 1.00 0.9977 7.40E-06 3.28E-01 Lanthanides Zr-97 4.23E+04 0:0001 0.2780 0.4000 1.00 0.8145 6.04E-06 2.55E-01 7.3 Effective Dose Equi~alent - Noble Gas Submersion Spreadsheets are used to calculate isotopic concentration at the receptor and the resultant radiation dose to the receptor for each of the isotopes in the mixture. For the example Effective Dose Equivalent calculation, the release point is the Offgas Stack at five hours since shutdown, and a gross concentration of 43.9 µCi/cm 3 (this concentration was determined iteratively to produce 1O mrem TEDE). The secondary containment holdup hours is set at <0.5 because the natural removal process in the Secondary Containment does not occur with the Offgas Stack. In Table 15, the column labeled "hEsoi Submersion mrem cm 3/µCi sec," is the dose factor for air submersion dose and is calculated in Section 7.1. The column labeled "Depleted Mix Ci/MWTh" is the "Release Ci/MWTh" calculated in Section 7 .2 for each isotope. The "Fraction" column determines the fraction each isotope contributes to the gross activity and is used to scale the activity for each isotope. Page 25 of34

CALC NEE-323-CALC-005 0 ENERCON Exc_eflence-:-:~very projet;t*.£very day. Revised Gaseous Radiological EALs per NEI 99-01 Rev. 06 NO. REV. 00 The column "Xiv Release Cone. µCi/cm 3" contains a calculation that scales the "Depleted Mix Ci/MWTh" column to a user entered gross concentration based on the "Fraction". In this case, the gross concentration entered was 43.9 µCi/cm 3 (4.39E+1 ). Values in the "Xir Receptor Cone. µCi/cm 3" column are calculated by multiplying the release concentration by the applicable dispersion factor, the volume of the release, and requisite conversion factors. The basic equation is from Section 6.7: Xir= XtJV

• v* (~)

Q For isotope 1-131, an example is presented: Xiv Release Cone. Flow (X/Q) 1.57E-02 µCi 10,000 #~ 2.83E-02 ff+~ 1 ~ 2.SOE-07 5e6 µCi cm 3 ~ 1 w 60 5e6 ff+.. cm 3 Where: v = 10,000 ft3/min is the rated flow from the Offgas Stack from Design Input 5.8. (XIQ) = 2.BOE-07 is the Noble Gas Dispersion coefficient (XIQ) for the Offgas Stack from Design Input 5.3. 2.83E-2 converts ft 3 to m3 Values in the "Submersion Dose mrem" (hEsoi) column are calculated by multiplying the factors "Xir Receptor Cone. µCi/cm 3", a time-units conversion factor, and the dose conversion factor calculated in Section 7.1. The basic equation for a one hour time period is shown in Section 6.7. Dose= Li(Xir

• hEsoa For isotope 1-131, an example is presented:

         "  .-    ~ .*

Submersion itt;;t;t hESOi Submersion Dose mrem 1.;* ': (-C~nc'. *-* I~ Ii ?i08E~O( i!8 6.73E+01 mrem Effl 3600 5e 5.04E-03 mrem I

                                                                                      =

Page 26 of34

CALC NEE-323-CALC-005 lyl ENERCON Revised Gaseous Radiological NO.

    *~llenre~Every project. .Ev'ery ~a"y.          EALs per NEI 99-01 Rev. 06 REV.                      00 For ease of comparison, the spreadsheet row for 1-131 is shown here:

Xiv Xir hESOi Depleted Release Receptor Submersion Submersion Mix Nuclide Fraction Cone. Cone. Dose mrem cc g

                                                                                       .lill                .lill                    mrem
                                  µCi sec            MWTh cm 3                 cm 3
            . ,~131               6.73E-1)        , 9.72E+O     . 3.582F4       . l.5?:E-2 .. *'*** ..;z.osE~~          **.':*  5,.04E~3*.              .

Table 15- Submersion Dose for Offgas Stack Xiv Xir hESOi Depleted Release Receptor Submersion Submersion Mix Nuclide Fraction Cone. Cone. Dose mrem cm 3 Ci

                                     µCi sec           MWTh                            J&i                  J&i                       mrem cm 3                cm 3 Ba-139                 8.03E+O          7.lOE-2         2.62E-6     1.lSE-4             1.52E-10                     4.39E-6 Ba-140                 3.17E+l          8.72E-1         3.21E-5     1.41E-3              1.86E-9                     2.13E-4 Ce-141                 1.27E+l          2.03E-2         7.46E-7     3.28E-5             4.33E-ll                     1.98E-6 Ce-143                 4.77E+l          1.67E-2         6.lSE-7     2.70E-5             3.57E-ll                     6.13E-6 Ce-144                 3.16E+o          1.64E-2         6.04E-7     2.65E-5             3.SOE-11                     3.98E-7 Cm-242                  2.llE-2         2.07E-4         7.64E-9     3.35E-7             4.43E-13                   3.36E-11 Cs-134                 2.80E+2         1.52E+O          5.62E-5     2.47E-3              3.26E-9                     3.28E-3 Cs-136                 3.92E+2          4.78E-1         1.76E-5     7.73E-4              1.02E-9                     1.44E-3 Cs-137                 2.86E-2          1.0SE+o         3.88E-5     1.70E-3              2.25E-9                     2.32E-7
                 *1-131'               6.731:+l

• s:i2e+o **iss2E~4 1.571:'.-2

                                                                                   --~. ... ,_.
                                                                                        ~

2.08E-8 5.04E-3

                                                                                                                                  .:r, .* ;...... ,,_ ..

1-132 4.14E+2 3.19E+O 1.17E-4 5.lSE-3 6.81E-9 1.02E-2 1-133 l.09E+2 1.70E+l 6.27E-4 2.75E-2 3.64E-8 1.42E-2 1-134 4.81E+2 4.24E-1 1.56E-5 6.86E-4 9.07E-10 1.57E-3 1-135 2.95E+2 1.14E+l 4.18E-4 1.84E-2 2.43E-!1 2.58E-2 Kr-83m 5.SSE-3 1.68E+2 6.20E-3 2.72E-1 3.GOE-7 7.19E-6 Kr-85 4.40E-1 1.02E+2 3.76E-3 1.65E-1 2.18E-7 3.45E-4 Kr-85m 2.77E+l 1.05E+3 3.85E-2 1.69E+O 2.23E-6 2.23E-1 Kr-87 1.52E+2 2.96E+2 1.09E-2 4.78E-1 6.32E-7 3.47E-1 Kr-88

• 3.77E+2 1.83E+3 6.75E-2 2.97E+o 3.92E-6 5.32E+O La-140 4.33E+2 8.35E-3 3.08E-7 1.35E-5 1.78E-ll 2.78E-5 La-141 8.84E+O 3.33E-3 l.23E-7 5.38E-6 7.llE-12 2.26E-7 La-142 5.33E+2 8.23E-4 3.03E-8 1.33E-6 1.76E-12 3.37E-6 Mo-99 2.69E+l 1.17E-1 4.29E-6 1.88E-4 2.49E-10 2.41E-5 Nb-95 1.38E+2 8.31E-3 3.0GE-7 l.34E-5 1.78E-ll 8.84E-6 Nd-147 2.29E+l 3.20E-3 1.lSE-7 5.lSE-6 6.84E-12 5.64E-7 Np-239 2.85E+l 2.48E-1 9.14E-6 4.0lE-4 5.30E-10 5.43E-5 Pr-143 7.77E-2 7.26E-3 2.68E-7 1.17E-5 1.SSE-11 4.34E-9 Pu-241 2.68E-4 1.97E-3 7.27E-8 3.19E-6 4.22E-12 4.07E-12 Rb-86 1.78E+l 1.70E-2 6.27E-7 2.75E-5 3.64E-ll 2.33E-6 Rh-105 1.38E+l . 5.90E-2 2.17E-6 9.54E-5 1.26E-10 6.25E-6 Ru-103 8.33E+l 1.00E-1 3.69E-6 1.62E-4 2.14E-10 6.42E-5 Ru-105 1.41E+2 3.25E-2 1.20E-6 5.25E-5 6.94E-ll 3.52E-5 Ru-106 O.OOE+O 3.59E-2 1.32E-6 5.81E-5 7.67E-11 O.OOE+O Sb-127 1.23E+2 1.07E-1 3.93E-6 1.73E-4 2.28E-10 1.0lE-4 Sb-129 2.64E+2 1.83E-1 6.74E-6 2.96E-4 3.91E-10 3.72E-4 Sr-89 2.86E-1 4.45E-1 l.64E-5 7.20E-4 9.52E-10 9.SOE-7 Sr-90 2.79E-2 4.43E-2 1.63E-6 7.16E-5 9.47E-11 9.SOE-9 Sr-91 1.28E+2 3.87E-1 1.43E-5 6.27E-4 8.28E-10 3.81E-4 Sr-92 2.51E+2 1.67E-1 6.16E-6 2.71E-4 3.58E-10 3.23E-4 Tc-99m 2.18E+l 5.70E-2 2.lOE-6 9.21E-5 1.22E-10 9.SSE-6 Page 27 of34

CALC a- ENERCON Excellence-,Every project. Every day. Revised Gaseous Radiological EALs per NEI 99-01 Rev. 06 NO. REV. NEE-323-CALC-005 00 Xiv Xir* hESOi Depleted Release Receptor Submersion Submersion Mix Nuclide Fraction Cone. Cone. Dose mrem cm 3 Ci 1!Q1 1!Q1 mrem

                                      µCi sec       MWfh cm3        cm 3 Te-127                8.95E-1       7.SSE-2   2.78E-6    1.22E-4    1.61E-10   5.20E-7 Te-127m                S.44E-1       1.84E-2   6.77E-7    2.97E-5    3.93E-11   7.69E-8 Te-129                1.02E+l       1.93E-2   7.09E-7    3.11E-5    4.11E-11   1.51E-6 Te-129m                5.74E+O       7.75E-2   2.86E-6    1.25E-4    1.66E-10   3.42E-6 Te-131m                2.59E+2       2.23E-1   8.23E-6    3.61E-4    4.77E-10   4.46E-4 Te-132                3.81E+l       1.69E+O   6.22E-5    2.73E-3     3.61E-9   4.95E-4 Xe-13lm                1.44E+O       1.32E+2   4.87E-3    2.14E-1     2.83E-7   1.46E-3 Xe-133                5.77E+O       1.94E+4   7.14E-1    3.13E+l     4.14E-5   8.60E-1 Xe-133m                S.07E+O      S.90E+2    2.18E-2    9.SSE-1     1.26E-6   2.30E-2 Xe-135                4.40E+l       3.56E+3   1.31E-1    S.75E+O     7.60E-6   1.20E+O Xe-135m                7.SSE+l       5.23E-3   1.93E-7    8.46E-6    1.12E-11   3.04E-6 Xe-138                2.13E+2       6.45E-3   2.37E-7    1.04E-5    1.38E-11   1.06E-5 Y-90                7.03E-1       4.30E-4   1.SSE-8    6.96E-7    9.19E-13   2.33E-9 Y-91                9.62E-1       S.86E-3   2.16E-7    9.48E-6    1.25E-11   4.34E-8 Y-92                4.81E+l       2.28E-3   8.39E-8    3.68E-6    4.87E-12   8.43E-7 Y-93                1.78E+l       3.31E-3   1.22E-7    5.36E-6    7.0SE-12   4.53E-7 Zr-95                1.33E+2       8.21E-3   3.03E-7    1.33E-5    1.75E-11   8.42E-6 Zr-97                3.34E+l       6.39E-3   2.35E-7    1.03E-5    1.36E-11   1.64E-6 2.71E+04  100.00%    4.39E+ol    S.SOE-5      8.05
                                                                       ! ~-~!1~;1 :              mtem Given a radiation effluent monitor reading of 43.9 µCi/cm 3 , and the assumptions of the scenario, the EDE value is_,8.05 mrem.

Spreadsheet cases are run for all four release points. See Section 2.0 for results. Page 28 of34

CALC NEE-323-CALC-005 0 ENERCON Exce/lence---;-Every p,oject. E.very day. Revised Gaseous Radiological EALs per NEI 99-01 Rev. 06 NO. REV. 00 7.4 CEDE and COE Thyroid For the example CEDE and COE Thyroid calculation, the release point is the Reactor Building at five hours since shutdown, and a gross concentration of .1.22E-2 µCi/cc, with a Secondary Containment Holdup time of 0.5 hours per Design Input 5.7 (this concentration was determined iteratively to produce 49.8 mrem Thyroid COE). In Table 16, the columns labeled "hrsoiThyroid mrem/µCi" and "hEsoi CEDE mrem/µCi" are the dose factors developed in Section 7.1. The column labeled "Depleted Mix Ci/MWTh" is the "Release Ci/MWTh" calculated above in Section 7 .2 for each isotope. The "Fraction" column determines the fraction each isotope contributes to the gross activity, and is used to scale the activity for each isotope. The column "xiv Release Cone. µCi/cm 3" contains a calculation that scales the "Depleted Mix" column to a user entered gross concentration based on the "Fraction" and is the variable Xiv in the equation below. In this case, the gross concentration entered was 1.22E-2 µCi/cc. Values in the "x;r Receptor Cone. µCi/cm 3" column are calculated by multiplying the release concentration by the applicable dispersion factor, the volume of the release, and requisite conversion factors. The basic equation from Section 6.8: Xir= Xiv* v* (~)

 . For isotope 1-131, an example is presented:

f .. ,, Xiv Release Cone. Flow (X/Q) 1* Re~~P!~::

                                                                                                             *'<;:one;*

1' r* ~ , ' '

                                                                                                           ',*C;'"'.,,'.C,,,*

1.63E-04 µCi 93,000 w 2.83E-02 ffi6 1 ffiffi 3.90E-06 6e6 [ ** ~.79E-?~ * µCi cm 3 ffiffi 1 #~ 60 6e6 ffi~ cm3 Where: v = 93,000 ft3/min is the rated flow from the Reactor Building from Design Input 5.8. (XIQ) = 3.90E-06 is the Particulate and Iodine dispersion coefficient for the Reactor Building from Design Input 5.8. Values in the column labeled "Inhalation Thyroid Dose mrem" are calculated by multiplying the following factors: concentration at the receptor, the breathing rate, the time, and the dose conversion factor. The basic equation is shown in Section 6.8. Dose = Li(Xir

• B
• t
• hrsoi)

For isotope 1-131, an example is presented: Page 29 of34

CALC NEE-323-CALC-005 ~ ENERCON Revised Gaseous Radiological EALs per NEI 99-01 Rev. 06 NO. Excellenc,;-Every projer;t. Every day. REV. 00 1* X'ir B

           ..                                                                                                   Inhalation
            -'\*

hrsoi Breathing Thyroid t: Rec.ep\or Cone. Time Rate Thyroid Dose I 2.7QE-08

                          -~

j,IG 1 flf 1.20E+06 ffl 1.08E+03 mrem 3.62E+Ol mrem I = -----1--- flf j2fi. Where: hrso; is the thyroid dose factor for each isotope from Section 7.1. B = 1.20E+06 cm 3/hr is the breathing rate. This value is equal to 3.33E-4 m3/sec from Design Input 5.9. Values in the "Inhalation CEDE Dose mrem" column are calculated by multiplying the following factors: concentration at the receptor, the breathing rate, the time, and the dose conversion factor. The basic equation comes from Section 6.9. Dose= Li(Xir

• B
• t
• hEsoa For isotope 1-131, an example is presented:

xir B Inhalation hESOi Receptor Breathing CEDE Cone Time Rate CEDE Dose 2.79E-08 j2fi. 1 flf 1.20E+06 3.29E+Ol mrem 1.lOE+OO mrem

                                                                                                              =----+-----

flf j2fi. For ease of comparison, the table row for 1-131 is shown here: Depleted Xiv Xir Inhalation Inhalation hrso! hESO! Release Receptor Thyroid CEDE Mix Thyroid CEDE Nuclide Fraction Cone. Cone. mrem mrem Ci Dose Dose

                                     µCi                 µCi            MWTh                           .b!.Q           1!Qi          mrem           mrem cm 3            cm 3 1-131               1.08E+3            . 3.29E+l       . 3.89E+2    . 1.34E*2 .. L63E-4           2,79E-8         3:62E+l       .l.lOE+O /

Table 16- Inhalation Thyroid and CEDE Dose for Reactor Building Depleted Xiv Xir Inhalation Inhalation hTSOI hEsot Release Receptor CEDE Thyroid CEDE Mix Thyroid Nuclide Fraction Cone. Cone. mrem .!DLfil!!. Ci Dose Dose

                                     µCi                 µCi            MWTh                           .b!.Q           1!Qi          mrem            mrem cm 3            cm3 Ba-139                8.88E-3              1.72E-1           2.84E+O      9.76E-5         1.19E-6        2.04E-10         2.17E-6        4.20E-5 Ba-140                9.47E-1              3.74E+O           3.49E+l      l.20E-3         1.46E-5         2.SOE-9         2.85E-3        1.12E-2 Ce-141                1.71E-1              8.95E+O           8.lOE-1      2.78E-5         3.40E-7        S.82E-11         1.19E-5        6.25E-4 Ce-143                4.48E-2              3.39E+O           6.68E-1      2.30E-5         2.SOE-7        4.79E-11         2.58E-6        1.95E-4 Ce-144                6.96E+O              3.74E+2           6.SGE-1      2.25E-5         2.75E-7        4.71E-11         3.93E-4        2.llE-2 Page 30 of34

CALC

                                                            '                                                       NEE-323-CALC-005 0 ENERCON Exce/lence--,Every project. :£very day.

Revised Gaseous Radiological EALs per NEI 99-01 Rev. 06 NO. REV. 00 Xiv Xir hrsot hESO! Depleted Inhalation Inhalation Thyroid CEDE Release Receptor Mix Thyroid CEDE Nuclide Fraction Cone. Cone.

                                 .!!!.@!!!.          !I!..Cfil!!. Ci                                                     Dose        Dose
                                    µCi                 µCi        MWTh                      l&1             l&1            mrem        mrem cm 3             cm 3 Cm-242                 3.48E+O              1.73E+4       8.30E-3     2.85E-7    3.48E-9         5.96E-13         2.49E-6      1.23E-2 Cs-134                 4.llE+l              4.63E+l       6.lOE+l     2.lOE-3    2.56E-5         4.38E-9          2.16E-1      2.43E-1 Cs-136                 6.40E+O              7.33E+O       1.91E+l     6.57E-4    8.02E-6          l.37E-9         1.0SE-2      1.21E-2 Cs-137                 2.93E+l              3.19E+l       4.22E+l     l.45E-3    1.77E-5          3.03E-9         1.07E-1      l.16E-1
                              --- ~:*..'*:*.* -::~**

1-131 l.08Et3 3:~~E-t;l' 3.89E+i l.34E-2 *1.63E-4 *2.79E:-8* .* 3:6.2.~+;L 1.10~:0 . 1-132 6.44E+O 3.81E-1 1.27E+2 4.38E-3 S.34E-S 9.lSE-9 7.07E-2 4.lSE-3 1-133 1.80E+2 S.85E+O 6.80E+2 2.34E-2 2.85E-4 4.88E-8 1.0SE+l 3.43E-1 1-134 l.07E+O 1.31E-1 1.70E+l S.83E-4 7.12E-6 1.22E-9 1.56E-3 1.92E-4 1-135 3.13E+l 1.23E+O 4.54E+2 1.56E-2 1.91E-4 3.26E-8 l.23E+O 4.SlE-2 Kr-83m O.OOE+O O.OOE+O 1.68E+2 S.79E-3 7.06E-5 l.21E-8 O.OOE+O O.OOE+O Kr-85 O.OOE+O O.OOE+O 1.02E+2 3.SOE-3 4.27E-5 7.32E-9 O.OOE+O O.OOE+O Kr-85m O.OOE+O O.OOE+O 1.05E+3 3.59E-2 4.3SE-4 7.SOE-8 O.OOE+O O.OOE+O Kr-87 O.OOE+O O.OOE+O 2.96E+2 1.02E-2 1.24E-4 2.12E-8 O.OOE+O O.OOE+O Kr-88 O.OOE+O O.OOE+O 1.83E+3 6.30E-2 7.69E-4 1.32E-7 O.OOE+O O.OOE+O La-140 4.51E-1 4.85E+O 3.34E-1 1.15E-5 1.40E-7 2.40E-11 1.30E-5 1.39E-4 la-141 3.48E-2 S.SlE-1 1.33E-1 4.SSE-6 S.5SE-8 9.55E-12 3.99E-7 6.66E-6 La-142 3.23E-2 2.53E-1 3.29E-2 1.13E-6 1.38E-8 2.36E-12 9.17E-8 7.lSE-7 Mo-99 4.33E-1 3.96E+O 4.66E+O 1.60E-4 1.95E-6 3.35E-10 1.74E-4 1.59E-3 Nb-95 1.32E+O 5.81E+O 3.32E-1 l.14E-5 1.39E-7 2.39E-11 3.79E-5 1.66E-4 Nd-147 7.18E-2 6.85E+O 1.28E-1 4.40E-6 S.37E-8 9.19E-12 7.92E-7 7.SSE-5 Np-239 2.82E-2 2.51E+O 9.92E+O 3.41E-4 4.16E-6 7.12E-10 2.41E-5 2.14E-3 Pr-143 6.22E-9 8.lOE+O 2.91E-1 9.99E-6 1.22E-7 2.09E-11 1.56E-13 2.03E-4 Pu-241 4.59E-2 8.25E+3 7.90E-2 2.71E-6 3.31E-8 S.67E-12 3.12E-7 5.61E-2 Rb-86 4.92E+O 6.62E+O 6.81E-1 2.34E-5 2.86E-7 4.89E-11 2.89E-4 3.89E-4 "l.\ Rh-105 9.51E-2 9.55E-1 2.36E+o 8.llE-5 9.90E-7 1.69E-10 . l.93E-5 1.94E-4 Ru-103 2.21E+O 8.95E+O 4.0lE+O l.38E-4 1.68E-6 2.88E-10 7.63E-4 3.09E-3 Ru-105 5.55E-2 4.55E-1 1.30E+O 4.47E-5 5.45E-7 9.33E-11 6.21E-6 5.09E-5 Ru-106 S.07E+l 4.77E+2 1.44E+O 4.94E-5 6.02E-7 1.03E-10 6.27E-3 5.90E-2 Sb-127 S.SSE-1 6.03E+O 4.27E+O 1.47E-4 l.79E-6 3.0GE-10 2.04E-4 2.22E-3 Sb-129 7.66E-2 6.44E-1 7.32E+O 2.52E-4 3.07E-6 S.25E-10 4.83E-5 4.06E-4 Sr-89 1.54E+O 4.14E+l 1.78E+l 6.12E-4 7.47E-6 1.28E-9 2.36E-3 6.36E-2 Sr-90 9.77E+O 1.30E+3 l.77E+O 6.09E-5 7.43E-7 l.27E-10 l.49E-3 l.98E-l Sr-91 1.SlE-1 1.66E+O 1.SSE+l S.33E-4 6.SOE-6 1.llE-9 2.02E-4 2.22E-3 Sr-92 8.lOE-2 8.07E-l 6.69E+O 2.30E-4 2.81E-6 4.SOE-10 4.67E-5 4.6SE-4 Tc-99m l.SSE-1 3.26E-2 2.28E+O 7.83E-5 9.SSE-7 1.64E-10 3.64E-S 6.39E-6 Te-127 2.39E-2 3.lSE-1 3.02E+O 1.04E-4 1.27E-6 2.17E-10 6.22E-6 8.28E-5 Te-127m 8.84E-1 2.lSE+l 7.35E-1 2.53E-S 3.0SE-7 5.28E-11 S.60E-5 l.36E-3 Te-129 6.03E-3 8.95E-2 7.70E-l 2.65E-S 3.23E-7 5.53E-11 4.00E-7 S.94E-6 Te-129m 1.46E+O 2.39E+l 3.lOE+O 1.07E-4 1.30E-6 2.23E-10 3.90E-4 6.39E-3 Te-13lm 1.34E+2 6.40E+O 8.93E+O 3.07E-4 3.7SE-6 6.41E-10 l.03E-1 4.93E-3 Te-132 2.32E+2 9.44E+O 6.76E+l 2.32E-3 2.83E-5 4.85E-9 l.3SE+O S.49E-2 Xe-131m O.OOE+O O.OOE+O 1.32E+2 4.SSE-3 S.SSE-5 9.49E-9 O.OOE+O O.OOE+O Xe-133 O.OOE+O O.OOE+O 1.94E+4 6.66E-1 8.12E-3 1.39E-6 O.OOE+O O.OOE+O Xe-133m O.OOE+O O.OOE+O 5.90E+2 2.03E-2 2.48E-4 4.24E-8 O.OOE+O O.OOE+O Xe-135 O.OOE+O O.OOE+O 3.56E+3 1.22E-1 l.49E-3 2.SSE-7 O.OOE+O O.OOE+O Xe-135m O.OOE+O O.OOE+O S.23E-3 1.80E-7 2.19E-9 3.7SE-13 O.OOE+O O.OOE+O Xe-138 O.OOE+O O.OOE+O 6.4SE-3 2.22E-7 2.70E-9 4.63E-13 O.OOE+O O.OOE+O Y-90 3.52E-2 8.44E+O 1.72E-2 S.92E-7 7.22E-9 l.24E-12 S.22E-8 1.25E-5 Y-91 4.07E-1 4.88E+l 2.34E-l 8.0GE-6 9.83E-8 1.68E-11 8.22E-6 9.86E-4 Y-92 l.37E-2 7.81E-1 9.llE-2 3.13E-6 3.82E-8 6.54E-12 l.07E-7 6.13E-6 Page 31 of 34

CALC NEE-323-CALC-005 ENERCON Revised Gaseous Radiological NO.

  ~I/once-Every proi<<r. Every day     EALs per NEI 99-01 Rev. 06 REV.          00 Depleted               Xiv            Xir h rsot       hesot                                                     Inhalation  Inhalation CEDE                           Release       Receptor Thyroid                      Mix                                          Thyroid      CEDE Nuclide                                               Fraction  Cone.          Cone.
                     .!!lUllil    .!!!.Ifil!l     Ci                                            Dose        Dose
                       µCi           µ Ci      MWTh                  QQl           QQl         mrem        mrem cm 3          cm 3 Y-93           1.87E-2      2.lSE+O      1.33E-1    4.SGE-6  S.SGE-8       9.52E-12      2.14E-7     2.46E-5 Zr-95          S.33E+O       2.36E+l      3.28E-1    1.13E-5  1.38 E-7      2.36E-11      l .Sl E-4   6.69E-4 Zr-97           3.54E-1      4.33E+O      2.SSE-1    8.78E-6  1.07E-7       1.83E-11      7.78E-6     9.53E-5 2.91E+4    100.00%  l.22E-2        2.09E-6         50         2.37 1.22E-2                       mrem        mrem Thyroid      CEDE Given a radiation effluent monitor reading of 1.22E-2 µCi/cm 3 , and the assumptions of the scenario, the COE thyroid value is 50 mrem and the CEDE is 2.37 mrem.

Spreadsheet cases are run for all four release points. See Section 2.0 for results. Page 32 of 34

CALC NEE-323-CALC-005 ENERCON Revised Gaseous Radiological NO. Excell~nc~-Every projecr. Ev~ry day. EALs per NEI 99-01 Rev. 06 REV. 00 7.5 Resultant Dose Summary A single spreadsheet was used to calculate EDE, CEDE, and thyroid COE. With the given source term, when the user changes the effluent gross concentration value, the spreadsheet calculates resultant doses. Results and variables for the reactor building case are shown below. As can be seen here, an effluent release rate of 1.22E-02 µCi/cc at the Reactor Building will result in an offsite dose of approximately 50 mrem COE thyroid . This value corresponds to the new RA 1 EAL entry threshold of 50 mrem COE thyroid. Dose totals are taken from the tabular spreadsheet data presented on the preceding pages. Inhalation CEDE: 2.37 mrem Submersion EDE: 0.39 mrem TEDE:

                                                                          =====2.76         mrem Inhalation Thyroid COE:       49.8         mrem Release Point:           Reactor Building                                 SBGT ?:    off Effluent Cone.:                            1.22E-02               µCi/cc      Release Flow CFM:      93,000 Release: Hrs. since Rx. Shutdown:     5                Hrs. Core Uncovered:        1 Exposure Time (hrs.):                             1                    Secondary Containment Holdup Hrs.:      0.5 Hours w/ Sprays On:                              2                                           cm3 per ft3 : 0.0283168 Submersion X/Q:                           4.30E-06            sec/m3                  Inhalation X/Q:    3.90E-06  sec/m3 Breathing Rate                                   3.33E-4            m3/sec            1.20E+6     cm 3/hr Spreadsheet cases are run for all four release points and for decay times of five hours.

Cases were also run for all four release points for decay times of 36 hours in consideration of EAL entry thresholds that are mode dependent. The output for all release points and decay times are shown in Appendix 1. See Section 2.0 for results. 8.0 Computer Software No computer software is used in this calculation. Page 33 of 34

CALC NEE-323-CALC-005 ENERCON Revised Gaseous Rad iolog ical NO.

     &cellt!nce-Every project. Every day. EALs per NEI 99-01 Rev. 06 REV. 00 9.0 Impact Assessment This calculation is based on "realistic" assumptions for the purpose of declaring EALs, rather than typical conservative "bounding" type design basis analyses. The calculation documents the EAL threshold values for specific plan monitors to assist Operations and Emergency Response personnel in determining the new basis for EALs RA1 , RS1 , and RG1 in accordance with NEI 99-01 Rev. 6.

Page 34 of 34

CALC NEE-323-CALC-005 ENERCON Appendix A NO.

      &ctl~na-Every project. Ev~ry day..         Dose Spreadsheet Outputs REV.            00 Turbine Building: Modes 1, 2, and 3 Inhalation CEDE:               2.38           mRem Submersion EDE:                0.39           mRem TEDE :           2.77           mRem Inhalation Thyroid COE :                   50.0           mRem Release Point:                       Turbine Building                        SBGT ?:    off Effluent Cone.:           I       l.58E-02        I uCi/cc              Release Flow CFM :        72,000 Ill a,       Release: Hrs. since Rx. Shutdown:            I     5                        Hrs. Core Uncovered:    I      1

.c (ti "i: (ti Exposure Time (hrs.): I 1 Secondary Containment Holdup Hrs.: I 0.5 Hours w/ Sprays On: I 2 cm3 per ft3: 0.0283168 Submersion X/Q: 4.30E-06 sec/m 3 Inhalation X/Q: 3.90E-06 sec/m3 Breathing Rate 3.33E-4 m3/sec = l .20E+6 cm3/hr Page 1 of 8

CALC NEE-323-CALC-005 ENERCON Appendix A NO.

      &ctll<nct-Evrry projtt:t Every do1         Dose Spreadsheet Outputs REV.             00 Turbine Building: Modes 4 and 5 Inhalation CEDE:           2.59           mRem Submersion EDE:             0.07           mRem TEDE :          2.67           mRem Inhalation Thyroid COE:                 49.7           mRem Release Point:                       Turbine Building                      SBGT ?:    off Effluent Cone.:          I       1.30E-02       I uCi/cc             Release Flow CFM :         72,000 II)

(1' Release: Hrs. since Rx. Shutdown: I 36 Hrs. Core Uncovered: I 1 JS C'U "i: C'U Exposure Time (hrs.): I 1 Secondary Containment Holdup Hrs.: I 0.5 Hours w/ Sprays On: I 2 cm3 per ft3 : 0.0283168 Submersion X/Q: 4.30E-06 sec/m3 Inhalation X/Q: 3.90E-06 sec/m3 Breathing Rate 3.33E-4 m3/sec = 1.20E+6 cm3/hr Page 2 of 8

CALC NEE-323-CALC-005 ENERCON Appendix A NO. fxctllenct-Every proj<<:t Ev~ry doy. Dose Spreadsheet Outputs REV. 00 Reactor Building: Modes 1, 2, and 3 Inhalation CEDE: 2.37 mRem Submersion EDE: 0.39 mRem

                                                       ========

TEDE : 2.76 mRem Inhalation Thyroid CDE: 49.8 mRem Release Point: Reactor Building SBGT ?: off Effluent Cone.: I 1.22E-02 I uCi/cc Release Flow CFM: 93,000 Ill a, Release: Hrs. since Rx. Shutdown: I 5 Hrs. Core Uncovered: I 1 .c n, 'i: n, Exposure Time (hrs.): I 1 Secondary Containment Holdup Hrs.: I 0.5 Hours w/ Sprays On: I 2 cm3 per ft3: 0.0283168 Submersion X/Q: 4.30E-06 sec/m3 Inhalation X/Q: 3.90E-06 sec/m3 Breathing Rate 3.33E-4 m3/sec = 1.20E+6 cm3/hr Page 3 of 8

CALC NEE-323-CALC-005 ENERCON Appendix A NO. Exull<nc,-Every pro;ec:r Every day Dose Spreadsheet Outputs REV. 00 Reactor Building: Modes 4 and 5 Inhalation CEDE: 2.60 mRem Submersion EDE: 0.07 mRem

                                                     ========

TEDE: 2.68 mRem Inhalation Thyroid COE: 49.9 mRem Release Point: Reactor Building SBGT ?: off Effluent Cone.: I 1.0lE-02 I uCi/cc Release Flow CFM: 93,000 VI w Release: Hrs. since Rx. Shutdown: I 36 . Hrs. Core Uncovered: I 1 ..0 (0 'i: (0 Exposure Time (hrs.): I 1 Secondary Containment Holdup Hrs.: I 0.5 Hours w/ Sprays On: I 2 cm3 per ft3: 0.0283168 Submersion X/Q: 4.30E-06 sec/m3 Inhalation X/Q: 3.90E-06 sec/m3 Breathing Rate 3.33E-4 m3/sec = 1.20E+6 cm3/h r Page 4 of 8

CALC NEE-323-CALC-005 ENERCON Appendix A NO. Excel~nce-E~ry proJ<<t. Every day. Dose Spreadsheet Outputs REV. 00 Offgas Stack: Modes 1, 2, and 3 Inhalation CEDE: 1.96 mRem Submersion EDE: 8.05 mRem

                                                     ========

TEDE : 10.00 mRem Inhalation Thyroid COE: 41.1 mRem Release Point: IOffgas Stack SBGT ?: on Effluent Cone.: I 4.39E+ol I uCi/cc Release Flow CFM: 10,000 Ill (1) Release: Hrs. since Rx. Shutdown: I 5 Hrs. Core Uncovered: I 1 ..c ra ra Exposure Time (hrs.): I 1 Secondary Containment Holdup Hrs.: I <0.5 Hours w/ Sprays On: I 2 cm3 per ft3 : 0.0283168 Submersion X/Q: 2.SOE-07 sec/m3 Inhalation X/Q: 3.lOE-07 sec/ m3 Breathing Rate 3.33E-4 m3/sec = l.20E+6 cm3/hr Page 5 of 8

CALC NEE-323-CALC-005 ENERCON Appendix A NO. Excellenct:-Every proj<<t. Every day. Dose Spreadsheet Outputs REV. 00 Offgas Stack: Modes 4 and 5 Inhalation CEDE: 2.61 mRem Submersion EDE: 1.41 mRem

                                                        ========

TEDE: 4.02 mRem Inhalation Thyroid COE: 50.0 mRem Release Point: IOffgas Stack SBGT ?: on Effluent Cone.: I 4.52E+Ol I uCi/cc Release Flow CFM : 10,000 VI - a, Release: Hrs. since Rx. Shutdown: I 36 Hrs. Core Uncovered: I 1 ..0 n, n, Exposure Time (hrs.): I 1 Secondary Containment Holdup Hrs.: I <0.5 Hours~/ Sprays On: I 2 cm3 per ft3: 0.0283168 Submersion X/Q: 2.SOE-07 sec/m 3 Inhalation X/Q: 3.lOE-07 sec/m3 Breathing Rate 3.33E-4 m3/ sec = l.20E+6 cm3/hr Page 6 of 8

CALC NEE-323-CALC-005 ENERCON Appendix A NO. Exctll~nc,:-Every projt!Ct. Ev~ry day. Dose Spreadsheet Outputs REV. 00 LLRPSF: Modes 1, 2, and 3 Inhalation CEDE: 2.37 mRem Submersion EDE: 0.39 mRem

                                                           ========

TEDE : 2.76 mRem Inhalation Thyroid CDE: 49.7 mRem Release Point: ILLRPSF SBGT ?: off Effluent Cone.: I 1.SlE-02 I uCi/cc Release Flow CFM : 75,000 11'1 cu Release: Hrs. since Rx. Shutdown: I 5 Hrs. Core Uncovered: I 1 .c ttl "i: ttl Exposure Time (hrs.): I 1 Secondary Containment Holdup Hrs.: I 0.5 Hours w/ Sprays On : I 2 cm3 per ft3 : 0.0283168 Submersion X/Q: 4.30E-06 sec/m3 Inhalation X/Q: 3.90E-06 sec/m3 Breathing Rate 3.33E-4 m3/sec = 1.20E+6 cm3/hr Page 7 of 8

CALC NEE-323-CALC-005 ENERCON Appendix A NO. Exalknc~-Every projtct Ev~ry day. Dose Spreadsheet Outputs REV. 00 LLRPSH: Modes 4 and 5 Inhalation CEDE: 2.60 mRem Submersion EDE: 0.07 mRem

                                                     ========

TEDE : 2.67 mRem Inhalation Thyroid CDE : 49.8 mRem Release Point: ILLRPSF SBGT ?: off Effluent Cone.: I 1.25E-02 I uCi/cc Release Flow CFM : 75,000 II) QI Release : Hrs. since Rx. Shutdown: I 36 Hrs. Core Uncovered: I 1 .1 .c n, n, Exposure Time (hrs.): I 1 Secondary Containment Holdup Hrs.: I 0.5 Hours w/ Sprays On: I 2 cm3 per ft3 : 0.0283168 Submersion X/Q: 4.30E-06 sec/m 3 Inhalation X/Q: 3.90E-06 sec/ m3 Breathing Rate 3.33E-4 m3/ sec = 1.20E+6 cm3/hr Page 8 of 8

CALC Attachment 1 NEE-323-CALC-005 ENERCON NO. CALCULATION PREPARATION

              &cell~nu- Every proJ<<t. E~ry day.

CHECKLIST REV. 00 CHECKLIST ITEMS 1 YES NO N/A GENERAL REQUIREMENTS

1. If the calculation is being performed to a client procedure, is the procedure being used the latest revision? D D ~

The calculation is being prepared to ENERCON's procedures.

2. Are the proper forms being used and are they the latest revision? ~ D D

3. Have the appropriate client review forms/checklists been completed? D D ~

The calculation is being prepared to ENERCON's procedures.

4. Are all pages properly identified with a calculation number, calculation revision and page number consistent with the requirements of the client's procedure? ~ D D

5. Is all information legible and reproducible? ~ D D

6. Is the calculation presented in a logical and orderly manner? ~ D D

7. Is there an existing calculation that should be revised or voided? D ~ D This is a new calculation to support implementing NEI 99-01 Rev. 6

8. Is it possible to alter an existing calculation instead of preparing a new calculation for this situation? D ~ D

9. If an existing calculation is being used for design inputs, are the key design inputs, assumptions and engineering judgments used in that calculation valid and do they D [8] D apply to the calculation revision being performed .

10. Is the format of the calculation consistent with applicable procedures and expectations? ~ D D 11 . Were design input/output documents properly updated to reference this calculation? D D ~

12. Can the calculation logic, methodology and presentation be properly understood without referring back to the originator for clarification? ~ D D OBJECTIVE AND SCOPE

13. Does the calculation provide a clear concise statement of the problem and objective of the calculation? ~ D D

14. Does the calculation provide a clear statement of quality classification? ~ D D

15. Is the reason for performing and the end use of the calculation understood? ~ D D

16. Does the calculation provide the basis for information found in the plant's license basis? ~ D D

17. If so, is this documented in the calculation?

                                                                                                       ~     D    D

18. Does the calculation provide the basis for information found in the plant's design basis documentation? D [8] D Page 1 of 4

CALC Attachment 1 NEE-323-CALC-005 ENERCON NO. CALCULATION PREPARATION Exctllenct-EVf!ry project Evt:ry day. CHECKLIST REV. 00 CHECKLIST ITEMS 1 YES NO N/A

19. If so, is this documented in the calculation? D D 0 20 . Does the calculation otherwise support information found in the plant's design basis documentation? D 0 D 21 . If so, is this documented in the calculation? D D 0

22. Has the appropriate design or license basis documentation been revised , or has the change notice or change request documents being prepared for submittal? D D 0 DESIGN INPUTS

23. Are design inputs clearly identified? 0 D D

24. Are design inputs retrievable or have they been added as attachments? 0 D D

25. If Attachments are used as design inputs or assumptions are the Attachments traceable and verifiable? D D 0 26 . Are design inputs clearly distinguished from assumptions? 0 D D

27. Does the calculation rely on Attachments for design inputs or assumptions? If yes, are the attachments properl y referenced in the calculation? - D 0 D

28. Are input sources (including industry codes and standards) appropriately selected and are they consistent with the quality classification and objective of th e calculation? 0 D D 29 . Are input sources (including industry codes and standards) consistent with the plant's design and license basis? 0 D D

30. If applicable , do design inputs adequately address actual plant conditions? 0 D D 31 . Are input values reasonable and correctly applied? 0 D D

32. Are design input sources approved? 0 D D

33. Does the calculation reference the latest revision of the design input source? 0 D D

34. Were all applicable plant operating modes considered? 0 D D ASSUMPTIONS

35. Are assumptions reasonable/appropriate to the objective? 0 D D

36. Is adequate justification/basis for all assumptions provided? 0 D D

37. Are any eng ineering judgments used? D 0 D

38. Are engineering judgments clearly identified as such? D D 0

39. If engineering judgments are utilized as design inputs, are they reasonable and can they be quantified or substantiated by reference to site or industry standards, D D 0 engineering principles, physical laws or other appropriate criteria?

Page 2 of 4

CALC Attachment 1 NEE-323-CALC-005 ENERCON NO. CALCULATION PREPARATION Excellence-Every proj~r. Every doy. CHECKLIST REV. 00 CHECKLIST ITEMS 1 YES NO NIA METHODOLOGY 40 . Is the methodology used in the calculation described or implied in the plant's licensing basis? D 181 D 41 . If the methodology used differs from that described in the plant's licensing basis, has the appropriate license document change notice been initiated? D D 181

42. Is the methodology used consistent with the stated objective? 181 D D

43. Is the methodology used appropriate when considering the quality classification of the calculation and intended use of the results? 181 D D BODY OF CALCULATION

44. Are equations used in the calculation consistent with recognized engineering practice and the plant's design and license basis? 181 D D

45. Is there reasonable justification provided for the use of equations not in common use? D D 181

46. Are the mathematical operations performed properly and documented in a logical fashion? 181 D D

47. Is the math performed correctly? 181 D D

48. Have adjustment factors, uncertainties and empirical correlations used in the analysis been correctly applied? 181 D D 49 . Has proper consideration been given to results that may be overly sensitive to very small changes in input? 181 D D SOFTWARE/COMPUTER CODES

50. Are computer codes or software languages used in the preparation of the calculation? D 181 D

51. Have the requirements of CSP 3.09 for use of computer codes or software languages, includ ing verification of accuracy and applicability been met? D D 181

52. Are the codes properly identified along with source vendor, organization , and revision level? D D 181

53. Is the computer code applicable for the analysis being performed? D D 181

54. If applicable, does the computer model adequately consider actual plant conditions? D D 181

55. Are the inputs to the computer code clearly identified and consistent with the inputs and assumptions documented in the calculation? D D 181 56 . Is the computer output clearly identified? D D 181

57. Does the computer output clearly identify the appropriate units? D D 181 Page 3 of 4

CALC Attachment 1 NEE-323-CALC-005 ENERCON NO. CALCULATION PREPARATION Exceflena-Every projtct. E'llt:ty day CHECKLIST REV. 00 CHECKLIST ITEMS 1 YES NO N/A

58. Are the computer outputs reasonable when compared to the inputs and what was expected? D D [83

59. Was the computer output reviewed for ERROR or WARNING messages that could invalidate the results? D D [83 RESULTS AND CONCLUSIONS

60. Is adequate acceptance criteria specified? D D [83

61. Are the stated acceptance criteria consistent with the purpose of the calculation , and intended use?

[83 D D

62. Are the stated acceptance criteria consistent with the plant's design basis, applicable licensing commitments and industry codes, and standards?

[gl D D

63. Do the calculation results and conclusions meet the stated acceptance criteria? [gl D D

64. Are the results represented in the proper units with an appropriate tolerance, if applicable?

[83 D D

65. Are the calculation results and conclusions reasonable when considered against the stated inputs and objectives?

[gl D D

66. Is sufficient conservatism applied to the outputs and conclusions? [gl D D

67. Do the calculation results and conclusions affect any other calculations? D [gl D

68. If so, have the affected calculations been revised? D D [83

69. Does the calculation contain any conceptual, unconfirmed or open assumptions requiring later confirmation? D [gl D

70. If so, are they properly identified? D D [gl DESIGN REVI EW 71 . Have alternate calculation methods been used to verify calculation results? D D [gl No, a Design Review was performed .

Note:

1. Where required , provide clarification/justification for answers to the questions in the space provided below each question. An explanation is required for any questions answered as "No' or "N/A".

Originator: Ryan Skaggs 12/14/17 Print Name and Sign Date Page 4 of 4

BASES-EOP 3 DAEC EOP BASES DOCUMENT Rev. 13 EOP 3 - SECONDARY CONTAINMENT CONTROL Page 29 of 29 GUIDELINE SF/L-4 Spent Fuel Pool level drops to 16.36 ft 0 Operate Spent Fuel Pool sprays (SAMP 712) .

                                ..- Use on ly systems not required for SF/L-5 adequate core cooling .

0 DISCUSSION If spent fuel pool level cannot be controlled using alternate or external makeup sources, sprays are used to add water to the spent fuel pool, co_ ol exposed bundles, and reduce radioactivity releases. However, spray operation may damage electrical equipment and flood lower elevations of the secondary containment, complicating implementation of other emergency response strategies, and runoff from sprays could spread radioactivity release. Use'of sprays is therefore delayed until it is determined that spent fuel pool level cannot be maintained above the top of the fuel racks. As long as the spent fuel assemblies are covered with water , the fuel will not overheat and efforts should focus on providing sufficient makeup flow to keep the assemblies submerged . The lowest measurable spent fuel pool level using the wide range instrument is 16.16 ft., approximately one foot above the top of the spent fuel racks. The action level in SF/L-4 corresponds to NEI 12-02 Level 3, the level at which fuel remains covered but actions to implement make-up water addition should no longer be deferred. The "before" condition permits appropriate anticipatory action based on the spent fuel pool leakage rate , radiation levels, available resources, and the time required to place sprays in service. Steps to prepare spray equipment for use should be initiated while radiation levels permit access to the refueling floor and timed to optimize use of available resources. As in Steps SF/T-3 and SF/L-3, available spray sources may be alternated between RPV injection and spent fuel pool spray modes as long as adequate core cooling can be maintained, but maintaining adequate core cooling takes precedence over spent fuel pool cooling (refer to the discussions of Steps SF/T-3 and SF/L-3 above).

AC Sources - Operating B 3.8.1 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.1 AC Sources - Operating BASES BACKGROUND The unit Class 1E AC Electrical Power Distribution System AC sources consist of the offsite power sources (preferred and alternate preferred), and the onsite standby power sources (Diesel I Generators (DGs) 1G-31 and 1G-21). As discussed in UFSAR Section 3.1 .2.2.8 (Ref. 1), the design of the AC Electrical Power System provides independence and redundancy to ensure an available source of power to the Engineered Safety Feature (ESF) Systems via essential buses 1A3 and 1A4. The Class 1E AC Distribution System is divided into redundant load groups, so loss of any one group does not prevent the minimum safety functions from being performed. Each load group has connections to two preferred offsite power supplies and a single DG. Offsite power is supplied to the 161 kV and 345 kV switchyards from the transmission network by six transmission lines. The 345 kV switchyard and the 161 kV switchyard are connected via the autotransformer, and both sections of the switchyard are connected to the transmission grid by at least.two independent lines. From the 161 kV switchyard (the preferred power source), a single overhead transmission line feeds the startup transformer. From the startup transformer, dual isolated secondary windings provide feeds to the 4160 volt essential buses, 1A3 and 1A4, through separate bus supply lines and circuit breakers. The startup transformer is sized to supply all plant power (both essential and non-essential loads) during unit startup. From the tertiary winding on the autotransformer (the alternate preferred power source), a single 34.5 kV underground line feeds the standby transformer. From the standby transformer, a single 4160 volt line feeds both essential buses through separate bus supply circuit breakers. A detailed description of the offsite power network and circuits to the onsite Class 1E essential buses is found in the UFSAR, Sections 8.2.1.3 and 8.3.1.1.5 (Ref. 2). An offsite circuit consists of all breakers, transformers, switches, interrupting devices, cabling, and controls (continued) DAEC B 3.8-1 TSCR-044A

AC Sources - Operating B 3.8.1 BASES BACKGROUND required to transmit power from the offsite transmission network to (continued) the onsite Class 1 E essential bus or buses. Startup transformer (1X3) provides the normal source of power to the essential buses 1A3 and 1A4. If either 4.16 kV essential bus loses power, an automatic transfer from the startup transformer to the standby transformer (1 X4) occurs. The startup transformer and standby transformer are both sized to accommodate the starting of all ESF loads on receipt of an accident signal. Emergency loads are sequenced onto the essential buses regardless of the source of power (onsite or offsite). The onsite standby power source for 4.16 kV essential buses 1A3 and 1A4 consists of two DGs. DGs 1G-31 and 1G-21 are dedicated to essential buses 1A3 and 1A4, resr,>ectively. A DG s arts automatically on a Loss of Coolant Accident (LOCA) signal (i.e., low reactor water level signal or high drywall pressure signal) or on an essential bus degraded voltage or undervoltage signal. After the DG has started, it automatically ties to its respective bus after offsite power is tripped as a consequence of essential bus undervoltage or degraded voltage, independent of or coincident with a LOCA signal. The DGs also start and operate in the standby mode without tying to the essential bus on a LOCA signal alone. Following the trip.of offsite power, non emergency loads powered from essential buses are load shed. When the DG is tied to the essential bus, loads are then sequentially connected to its respective essential bus. The sequencing logic controls the permissive and starting signals to motor breakers to prevent overloading the DG. In the event of a loss of both the preferred power source and the alternate preferred power source, the ESF electrical loads are automatically connected to the DGs in sufficient time to provide for safe reactor shutdown and to mitigate the consequences of a Design Basis Accident (OBA) such as a LOCA. Certain required plant loads are returned to service in a predetermined sequence in order to prevent overloading of the DGs in the process. Within 25 seconds after the initiating signal is received, all automatic and permanently (continued) DAEC B 3.8-2 TSCR-111

ANNUN CIATOR RESPONSE PROCEDURE ARP 1C08A GENE RATOR AND AUXILIARY POWER Usage Level Reference Use Record the followi ng : Date/Time: I Initials: NOTE: User shall p erform and document a Temp Issue/Rev. Check to ensure revision is current, in accordance with procedure use and adherence requirements. Prepared By: I Date: Print Signature CRO SS-DISCIPLINE REVIEW (AS REQUIRED) Reviewed By: I Date: Print Signature Reviewed By: I Date: Print Signature PROCEDURE APPROVAL Approved By

            - - - -Print   - - - - -I - - - -Signature -------               Date:
                                                                                   ---~

Page 1 of 140 Rev. 89

ALARM WINDOW ENGRAVINGS AND GRID LAYOUT 1C08A Same on annunciator panel I COBB for Division 2 1 2 3 4 5 6 7 8 9 10 11 12 BUS 1A1 UNIITTERRUPTIBLE AC 125 voe AUXXFMRT01A1 SIU XFMR TO 1A1 STBY XFMR TO 1A3 SIU XFMR TO 1A3 STARTUP XFMR "A" DIESEL GEN A DG TO BUS 1A3 "A" DIESEL GEN A BREAKER 1A101 LOCKOUT TRIP BREAKER 1A102 BREAKER 1A301 BUS 1A3 BREAKER 1A302 1X3 1Y23 UNOERVOLTAGE OR SYSTEM 1 1G-31 BREAKER 1A311 1G-31 OR LOCKOUT TRIP TROU BLE TRIP TRIP TRIP TRIP TROUBLE INVERTER TROUBLE RUNNING TRIP LOCKOUT TRIP LOSS OF VOLTAGE LCXR,AR 1X91 INSTRUMENT AC 1A1 TOXFMR 1X11 1A1 TOXFMR 1X71 1A1 TOXFMR 1X51 SWITCHYARO SUPPLY LCXFMR 1X3f MAIN GENERATOR 125 voe "A" DIESEL GEN "A" DIESEL GEN "A" DIESEL GEN B BREAKER 1A107 BREAKER 1A108 BREAKER 1A109 BREAKER 1.A.110 BREAKER 1A303 BREAKER 1A312 OR IMPROPER PHASE 1Y21 UNOERVOLTAGE CHARGER 1012 FUEL OIL DAY TANK 1T-37A 1G-31 PHASE OVERCURREITT 1G-31 TRIP TRIP TRIP TRIP TRIP MCC 1 B91 BKR 1B903 SEQUENCE OR TROUBLE OVERSPEED TRIP LO-L=EL OR GROUND FAULT TRIP INVERTER TROUBLE LC 1B1/1B2 125 voe LC 1 B3 BREAKER INSTRUMENT AC XFMR 1X11 TO LC1B1 XFMR 1X51 TO LC 1B5 1Y11 125VDC AUX BOILER "A" DIESEL GEN "A" DIESEL GEN C BREAKER 18101 CROSS TIE BREAKER BREAKER 1BS01 SYSTEM 1 16301, 16302 BUS 1A3 STARTUP XFMR UNOERVOLTAGE CHARGER 10120 FUEL TANK H -34 PANEL 1C-93 1G-31 TRIP 1B107 TRP BATTERY 101 16303 OR 16304 LOSS OF VOLTAGE LOCKOUT TRIP OR TRIP TROUBLE LO LEVEL TROUBLE ENGINE CRANKING DISCONNECTED TRIP INVERTER TROUBLE LC 1651166 LC 1B 1 BREAKER LOAD CENTER 165 MCC 1 B34N1 B44A "A" DIESEL GEN "A" DIESEL GEN CROSSTIE MCC 1B34A 4KVBUS DIESEL FUEL OIL "A" DIESEL GEN "A" DIESEL GEN D 16102, 16103 BREAKER BREAKER 16502 TIE BKR 1 B3401 TIE BREAKER AUTO TRANSFER STORAGE TANK 1T-35 1G-31 1G-31 1G-31 1G-31 16104 OR 1 B105 16503 OR 1B504 1B3402 OR 1 B4402 LO LEVEL CONTROL POWER AUTOSTART ENGINE SHUTDOINN 19505 TRIP INOP START FAILURE TRIP TRIP TRIP FAILURE INHIBITED TRIP Page 2 of 140 Rev. 89

ANNUNCIATOR PANEL: 1C08A COORDINATES: A-9 PAGE: 1 of 3 125V DC SYSTEM 1 TROUBLE TITLE: 125 voe SYSTEM 1 TROUBLE (GROUND OR LOW VOLTAGE) 1.0 PROBABLE CAUSE{S) / INITIATING DEVICE{S) / SETPOINT{S) 1.1 Ground fault on 125 VDC Positive to ground Relay 64 9 Volt differential System 1 or negative to ground Relay 64 1.2 125 VDC System 1 low Relay 1010-27 105 voe (dee) voltage 1.3 Positive or negative metering FU 3 amp fuse blown blown fuse blown 2.0 AUTOMATIC ACTIONS 2.1 If due to a complete loss of 125 VDC System 1:

a. Various control systems half trip .

b. Scoop Tube for Recirc Pump A locks up.

c. Brealrer'1 B3401 auto trips open after 6 second time delay, 1B4401 auto closes.

d. Static Switch JS150 1 transfers from Inverter 1015 to Regulating Transformer 1Y1A.

2.2 If the 125 voe System 1 is not lost, no AUTOMATIC ACTIONS occur. Page 36 of 140 Rev. 89

ANNUNCIATOR PANEL: 1C08A COORDINATES: B-9 PAGE: 1 of 3 125V DC CHARGER 1D12 TROUBLE TITLE: 125 voe CHARGER 1012 TROUBLE NOTE This is a normal alarm anytime Charger 1012 is being changed over to Charger 1D120. 1.0 PROBABLE CAUSE(S) I INITIATING DEVICE(S) I SETPOINT(S) 1.1 AC Breaker 1012-01 open Relay K-8 OPEN position 1.2 DC Breaker 1012-02 open Relay K-7 OPEN position AND DC Breaker 1012-03 open Relay K-7 OPEN position 1.3 Charger Failure Relay K-5 < 5 Amps (dee) 40 second time delay 1.4 Reverse Current Relay K-4 Reverse Current Detected 1.5 DC Undervoltage Relay K-2 105 voe (dee) 1.6 AC Undervoltage Relay K-1 340 VAC (dee) 2.0 AUTOMATIC ACTIONS 2.1 If due to CAUSE 1.3, 1.4, or 1.5, Charger 1012 front panel trouble light illuminates. Page 69 of 140 Rev. 89

ANNUNCIATOR PANEL: 1C08A COORDINATES: C-9 PAGE: 1 of 3 125V DC CHARGER 1D120 TROUBLE TITLE : 125 voe CHARGER 10120 TROUBLE NOTE This is a normal alarm anytime 1D120 is being changed over to Chargers 1012 or 1022. 1.0 PROBABLE CAUSE(S) I INITIATING DEVICE(S) I SETPOINT(S) 1.1 AC Breaker 1D120-01 open Relay K-8 Open Position 1.2 DC Breaker 1D120-02 open Relay K-7 OPEN Position 1.3 DC Breaker 1D120-03 open Relay K-7 OPEN Position 1.4 Charger Failure Relay K5 < 5 Amps (dee) 40 Second TD 1.5 Reverse Current Relay K-4 Reverse Current Detected 1.6 DC Undervoltage Relay K-2 105 VDC (dee) 1.7 AC Undervoltage Relay K-1 340 VAC (dee) 2.0 AUTOMATIC ACTIONS 2.1 If due to CAUSE 1.4, 1.5, or 1.6, Charger 1D120 front panel trouble light illuminates. Page 102 of 140 Rev. 89

Distribution Systems - Operating B 3.8.7 Table B 3.8.7-1 (page 1 of 1) AC and DC Electrical Power Distribution Systems TYPE VOLTAGE DIVISION 1<al DIVISION 2<a> AC safety 4160 V Essential Bus 1A3 Essential Bus 1A4 buses 480V Load Centers Load Centers 183, 189 184, 1820 480V Motor Control Motor Control Centers Centers 1832, 1834 1842, 1844 125 voe buses 125 V Distribution Panels Distribution 1010, 1011, Panels 1013 1020,1021, RCIC Motor Control 1023 Center 1014 250 voe buses 250V NIA Distribution Panel 1040 Motor Control Centers

                        -.                                            1041 and 1042

<a> Each division of the AC and DC electrical power distribution systems is a subsystem. DAEC B 3.8-73 Amendment 223

IAOP 302 .1 LOSS OF 125 voe POWER ABNORMAL OPERATING PROCEDURE AOP 302.1 LOSS OF 125 VDC POWER Usage Level Reference Use NOTE This AOP is normally coordinated by the Reactor Operator. Record the following: Date/Time: _ _ _ _ _ I _ _ _ _ Initials: _ __ NOTE: User shall perform and document a Temp Issue/Rev. Check to ensure revision is current, in accordance with procedure use and adherence requirements. Enter the following as applicable: LOSS OF 125 voe 1011 PAGE 2 LOSS OF 125 voe 1013 PAGE 8 LOSS OF 125 voe 1014 PAGE 14 LOSS OF 125 voe DIV I PAGE 17 LOSS OF 125 voe 1021 PAGE 28 LOSS OF 125 voe 1023 PAGE 35 LOSS OF 125 voe DIV II PAGE 43 COMPLETE LOSS OF 125 voe PAGE 56 IAOP 302 .1 Page 1 of 63 Rev. 58

AOP 302 .1 LOSS OF 125 voe POWER LOSS OF 125 VDC DIV I IMMEDIATE ACTIONS

1. Place RX WATER LEVEL CONTROL INPUT SELECT HSS-4560 in A LEVEL at 1C05.

2. Place 1 OR 3 ELEMENT CONTROL SELECT HSS 4450 in 1 ELEMENT at 1C05.

AUTOMATIC ACTIONS 1G001 Voltage Regulator transfers to manual with no adjustment capability Loss of 1A 1/1 A3 breaker control MG Set A Scoop Tube Power Failure Lockout initiates (no Amber lockup light) SBGTS A SV-5801A fails closed, SV-5815A, SV-5817A and SV-5825A fail open MCC 1B34A/1 B44A will auto transfer to 1B44 JS1501 will transfer from Inverter 1015 to Reg Trans 1Y1A Group 3A Primary Containment Isolation (Lockout relay will not trip) Rx FEEDWATER FLOW B FEEDLINE Fl- 1626 fails downscale STEAM FLOW B STEAM LINE Fl-4409 fails downscale Recirc Lube Oil Pumps 1P202A & 1P202B trip, and 1P202C auto starts on low oil pressure (continued operation of the A Recirc MG Set is allowable in this condition). B GEMAC LEVEL Ll4560 fails low HPCI will not trip on high level Loss of 1010 with B LEVEL selected and no operator action results in : Feedwater control opens Feed Reg Valves Reactor water level goes high "B" Reactor Feed Pump and Main Turbine Trip on high level ("A" Reactor Feed Pump cannot be tripped remotely) Reactor Scram (Main turbine trip) Loss of 1A1 and 1A2 power (failure to auto transfer) If >26% power, Reactor Recirc. Pumps 1P-201 A and 1P-201 B trip (RPT) IAOP 302 .1 Page 17 of 63 Rev.58

AOP 302 .1 LOSS OF 125 voe POWER LOSS OF 125 voe DIV I FOLLOW-UP ACTIONS NOTE Follow up actions may be performed in any order.

1. Establish critical parameter monitoring of RPV Water Level, as priorities allow.

2. Stabilize reactor power level and maintain recirculation loop flows balanced.

Use manual control of the A MG Set and Recirc Pump from the MG Set Room in accordance with 01264. NOTE Buses 1A1 and 1A2 will not auto transfer to the startup transformer on a turbine trip.

3. Transfer Bus 1A2 to the startup transformer per 01 304 .1.

4. IF 1A4 has power available THEN verify TIE BREAKER 1B4401 MCC 1B44A/1 B34A is closed.

5. IF a Reactor Scram occurs: THEN perform the following:

a. Verify main turbine trip.

b. Verify the Hand I breakers open.

c. Only after Hand I are open, direct an operator to trip the GENERATOR EXCITER FIELD BREAKER locally.

6. Reference EPIP 1.1 for EAL assessment, for instrumentation, perform alarm panel checks as needed to confirm threshold is met.

7. Suspend all evolutions in progress associated with electrical switchgear and switching operations.

8. Locally operate affected switchgear to start and stop equipment as required.

IAOP 302.1 Page 18 of 63 Rev.58

AOP 302 .1 LOSS OF 125 voe POWER LOSS OF 125 voe DIV I FOLLOW-UP ACTIONS (continued) NOTE Loss of 125 VDC DIV 1 causes a loss of 1A 1 control power. If 1A 1 is on the aux transformer, all 1A1 loads will remain energized with no automatic trips or starts until the Main Generator is tripped . Likewise, when the main generator is tripped, All 1A 1 loads will be lost until manually restored. 1A1 loads:

• A Feed Pump (1A103)- no 211 " trip
• A Condensate Pump (1A 106)
• A Recirc MG Set (1A 104)
• A Circ Water Pump (1A105)
• Load Center 1B 1 (1A107)
• Load Center 1B5 (1A109)
• Load Center 1B7 (1A108)

9. IF 1A1 is deenergized THEN reenergize 1A1 locally:

a. Trip 1A101.

b. Strip 1A 1 loads.

c. Close 1A102.

d. Restart loads as required.

NOTE If operation of the RCIC System is required while the Division I 125 VDC System is deenergized, use the HPCI System in manual mode. 1O. Evaluate the status of the 125 VDC Electrical Distribution System to determine the cause of the malfunction. IAOP 302.1 Page 19 of 63 Rev. 58

AOP 302 .1 LOSS OF 125 VDC POWER LOSS OF 125 VDC DIV I FOLLOW-UP ACTIONS (continued) 11 . IF 1D10 is totally deenergized THEN perform the following :

a. Open all branch circuit breakers at Panels 1D10, 1D11 , 1D13, and 1D14 .

b. Verify MCC 1B32 energized .

c. Locally inspect circuit breakers at Panels 1D10, 1D11, 1D13, and 1D14.

d. Verify Battery Room ventilation .

e. Locally inspect Battery 1D1

f. Locally inspect Battery Chargers 1D12 and 1D120.

g. Restore power to 1D 1O and branch panels.

h. Comply with Tech Spec Requirements for "Distribution Systems - Operating" or "Distribution Systems - Shutdown", as applicable.

i. Comply with ACP 1412.4 Impairments to Fire Protection System.

12. IF 1D1 O is not lost, but one or THEN investigate and evaluate the status of more loads are lost the individually affected loads and comply with Tech Spec Requirements for "Distribution Systems - Operating" or "Distribution Systems - Shutdown",

as applicable.

13. IF 125 VDC Div I Battery 101 is THEN comply with Tech Spec Requirements made or found to be for "DC Sources - Operating" or "DC inoperable Sources - Shutdown", as applicable.

14. IF 125 VDC Div I Battery 1D 1 is THEN Verify battery cell parameters per

        < 110 VDC                                STP 3.8.4-02 within 24 hours.

15. WHEN Div I 125 voe System THEN send an operator to TIE BREAKER is restored and the use 1B3401 MCC 1B34A/44A to press of-TIE BREAKER reset button prior to closing.

1B3401 is desired

16. WHEN power is restored THEN reset A Scoop Tube lockout per 01264.

IAOP 302.1 Page 20 of 63 Rev.58

AOP 302 .1 LOSS OF 125 VDC POWER LOSS OF 125 VDC DIV I PROBABLE ANNUNCIATORS 1COBA, A-2 Bus 1A1 LOCKOUT TRIP OR LOSS OF VOLTAGE A-7 STARTUP XFMR 1X3 TROUBLE A-9 125 voe SYSTEM 1 TROUBLE B-9 125 voe CHARGER 1012 TROUBLE C-4 125 VDC SYSTEM 1 BATTERY 1D1 DISCONNECTED C-6 BUS 1A3 LOSS OF VOLTAGE C-B INSTRUMENT AC 1Y11 UNDERVOLTAGE OR INVERTER TROUBLE C-9 125 voe CHARGER 10120 TROUBLE D-5 MCC 1B34A TIE BKR 1B3401 TRIP D-9 A DIESEL GEN 1G-31 CONTROL POWER FAILURE 1COBC, A-3 MAIN GENERATOR LOCKOUT RELAY CKT LOSS OF 125 VDC B-3 MAIN GENERATOR VOLTAGE REGULATION IN MANUAL B-6 H2/STATOR COOLING PANEL 1CB3 LOSS OF DC PROBABLE INDICATIONS Annunciators - Loss of power to the following panels: 1C03 1C05 1C23A 1C25A 1C34 1C40 1C04 1COBB 1C24A 1C26A 1C35A 1C40A 1C14 1C09A 1C22 1COB - Loss of the following :

-  4160V BUS 1A1 switchgear control, indication and automatic trip functions
-  4BOV LC 1B 1 switchgear control and indication
-  4BOV LC 1B7 switchgear control and indication
-  4BOV LC 1B5 switchgear control and indication
-  4*1sov BUS 1A3 switchgear control, indication and automatic trip functions
-  4BOV LC 1B3 switchgear control and indication
-  4BOV LC 1B9 switchgear control and indication
-  TIE BREAKER 1B3401 MCC 1B34A/1B44A control and indication GENERATOR EXCITER FIELD BREAKER control and indication 1CO? - Loss of the following:

EMERGENCY BEARING OIL PUMP 1P-40 control and indication IAOP 302.1 Page 21 of 63 Rev.5B

AOP 302.1 LOSS OF 125 VDC POWER LOSS OF 125 VDC DIV I PROBABLE INDICATIONS (continued) 1C06 - Loss of the following:

-  A CIRC WATER PUMP 1P-4A control and indication
-  A CONDENSATE PUMP 1P-8A control and indication
-  A REACTOR FEEDWATER PUMP 1P-1Acontrol and indication
-  A GSW PUMP 1P-89A control and indication
-  A[C] RWS PUMP 1P-117A and C control and indication 1C05 - Loss of the following:
-  A WIDE RANGE LEVEL Ll-4539 indication fails low B GEMAC LEVEL Ll-4560 fails low B REACTOR PRESS Pl-4564 fails low
-  A CRD PUMP 1P-209A control and indication Rx FEEDWATER FLOW B FEEDLINE Fl-1626 fails downscale STEAM FLOW B STEAMLINE Fl-4409 fails downscale 1C04 - Loss of the following:

• MG SET LUBE OIL PUMP 1P-202A control and indication MG SET LUBE OIL PUMP 1P-202B control and indication

-  A MG SET SPEED CONTROL control and indication A MG SET EMERG AUX OIL PUMP 1P-204A control and indication RCIC TURBINE CONTROL VALVE HV-2406 position indication RCIC System Drain Valves RCIC STEAM LINE DRAIN ISOL CV-2410, and CLOSED RADWASTE DISCH ISOL CV-2435                                   **

RCIC System Condensate Pump Motor and Motor Operated Valves control and indication Inboard MSIVs indication and DC Solenoid Valve control 1C03 - Loss of the following:

-  A CORE SPRAY PUMP 1P-211A control and indication
-  A RHR PUMP 1P-229A and C RHR PUMP 1P-229C control and indication
-  A RHRSW PUMP 1P-22A and C RHRSW PUMP 1P-22C control and indication
-  A RHR HX SHELL OUTBD VENT M0-2044A and A RHR HX SHELL INBD VENT M0-2044B percent indication HPCI STEAM LINE DRAIN ISOL CV-2211 and CLOSED RADWASTE DISCH ISOL CV-2234 control and indication IAOP 302.1                             Page 22 of 63                             Rev.58

AOP 302.1 LOSS OF 125 VDC POWER LOSS OF 125 voe DIV I

***************************************************************INFORMATION***************************************************************

1D10 125 VDC Distribution Panel loads: 1D10 ckt 04 1D13 125 VDC Distribution Panel C 1D10 ckt 051D14 125 VDC RCIC MCC 1D10 ckt 061D11125 VDC Distribution Panel A 1D10 ckt 07 Instrument AC Inverter 1D15 Supply 1D11 125 VDC Distribution Panel A loads: 1D11 ckt 01 Load Center 1B1 switchgear control 1D11 ckt 02 RWCU F/D Panel 1C82 annunciators 1D11 ckt 03 Load Center 1B5 switchgear control 1D11 ckt 04 Load Center 1B7 switchgear control 1D 11 ckt 05 4160V Bus 1A 1 switchgear control 4KV BREAKER 1A101 AUX XFMR TO BUS 1A1 4KV BREAKER 1A102 STARTUP XFMR TO BUS 1A1 Reactor Feed Pump 1P-1A Supply Breaker 1A103 Reactor Recirculation MG Set 1G-201A Supply Breaker 1A104 Circulating Water Pump 1P-4A Supply Breaker 1A105 Condensate Pump 1P-8A Supply Breaker 1A106 FEEDER BREAKER 1A1071A1 TO LC XFMR 1X11 FEEDER BREAKER 1A108 1A1 TO LC XFMR 1X71 FEEDER BREAKER 1A109 1A1 TO LC XFMR 1X51 FEEDER BREAKER 1A110 1A1 TO SWYD LOAD CENTER Reactor pressure Channel B calculation and indication Reactor water level Channel B calculation and indication 1D11 ckt 06 Load Center 1B9 switchgear control 1D11 ckt 07 MCC breaker 1B3401 control (Normal) 1D11 ckt 08 Main Generator excitation control . Generator Exciter Field Breaker control and indication Motor Driven DC Regulator Setpoint Adjust (1 COB) Motor Driven AC Regulator Setpoint Adjust (1 COB) Exciter Field Flashing Regulator Transfer and Lockout Relay Exciter Field Bridge Overcurrent Alarm Generator Field Bridge Over temperature Alarm Exciter Field Current Limit Circuit Volts/Hertz Protective Panel 1D11 ckt 09 Generator H2 Cooling Panel 1C83 Associated Generator Trip and Load Runback Relays Annunciators IAOP 302.1 Page 23 of 63 Rev.58

AOP 302.1 LOSS OF 125 voe POWER LOSS OF 125 voe DIV I

* * **** .... * *** ................. ** ... * **.. * .. *............... *.. INFORMATION** .. * .. *** ....... *............ ** ........... * ....... *.. ***** ... .

1D11 125 VDC Distribution Panel A loads (continued}: 1D11 ckt 10 Main Transformer 1X1 control power 1D11 ckt 111G-31 Diesel Gen. Control Panel 1C117 1D11 ckt 12 1G-31 Diesel Gen. Control Panel 1C117 1D11 ckt 13 Startup Transformer 1X3 control power 1D11 ckt 15 Core Spray Channel A Relay Vertical Board 1C43 Core Spray System Loop A Logic 1D11 ckt 17 Radwaste Panel 1C84 annunciators 1D11 ckt 18 1G-31 Diesel Gen. Exciter Panel 1C93 1D11 ckt 19 Standby Gas Treatment System Control Panel 1C24A control PASS System Valves SV-4594A, SV-4595A, and SV-8772A (FC} A SBGTS valves SV-5801A, SV-5815A, SV-5817A, and SV-5825A (CV-5815A, CV-5817A, and CV-5825A (FO} (AV5801 (FC) A SBGTS vent shaft Rad Monitor Aux Relay 95-K134A (PCIS GP 3A input) A SBGTS Fire Deluge SV-5837A (CV-5837A (FC)) A SBGTS Preheater control (TORUS} EXTERNAL VACUUM BKR ISOL (CV-4304 (FO}) CAMS Loop A Isolation Valve control and indication Offgas Stack Exhaust Fan 1V-EF-18A remote control Panel 1C24A annunciators Panel 1C25A annunciators 1D11 ckt 20 Turbine Building and Control Room HVAC Panel 1C26 SFU Fire Deluge SV-7328A (CV-7328/l.. (FC}) A DIESEL GENERATOR 1G-31 Room Supply Fan 1V-SF-20 remote control A DIESEL GENERATOR 1G-31 Room Supply Fan 1V-SF-20 dampers D0-7001A and D0-7002A position indication 1V-SFU-30A valves CV-7301A and SV-7318A (AV-7301A and AV-7318A (FO}) Miscellaneous Reactor, Turbine, and Control Building isolation dampers 1C23A annunciators 1C26A annunciators and indications IAOP 302.1 Page 24 of 63 Rev. 58

AOP 302.1 LOSS OF 125 voe POWER LOSS OF 125 voe DIV I

..................................................... * ...... ***INFORMATION* ........ ** ................ *** ................................ .

1013125V DC Distribution Panel C loads: 1013 ckt 01 Reactor Recirculation Pump MG Set 1G-201A Control Panel 1C112A and Scoop Tube Power Failure Lock circuitry 1013 ckt 13 Recirculation Pump MG Set 1G-201A control (Normal and Standby Power) MG Set A Emergency Lube Oil Pump 1P-204A control and indication Loss of Division I ATWS/ARI/RPT Trip System (101313 only) 1013 ckt 02 Reactor Core Cooling Benchboard 1C03 RHR Heat Exchanger A Vent Valves M0-2044A and M0-2044B position indication (Zl-2044A and Zl-2044B) HPCI System Drain Valves SV-2211 and SV-2234 (CV-2211 and CV-2234) control and indication Position indication for CV-4309 (ZS-4309) 1013 ckt 03 Reactor Water Cleanup and Recirculation Benchboard 1C04 RCIC Inverter RCIC Governor Valve HV-2406 Position indication RCIC System Drain Valves SV-2410, and SV-2435 (CV-2410 and CV-2435) control and indication 1D13 ckt 04 Annunciator Power Panels 1C03, 1C04, 1C05, 1C08B, 1C34, Panel 1C22 Frequency Converter . 1D13 ckt 05 CAD Panel 1C35A, CAM Panel 1C09A SV-4332A, UP,per Drywell Spray CAD N2 Primary Containment lsolatio11, SV-4334A North Torus Spray Header Primary Containment lsol ~

• SV-4332B, and SV-4334B control and indication 1C35A Annunciators 1C09A Annunciators 1C014A EOP Annunciators 1D13 ckt 06 1C40 Annunciators 1D13 ckt 07 1C32 Channel A RHR Relay Vertical Board RHR Loop A Logic HPCI Low Water Level initiation signal HPCI Isolation Logic A HPCI Rx Hi-Level trip logic (1/2 of logic, HPCI will not trip on Hi-Level)

IAOP 302.1 Page 25 of 63 Rev.58

AOP 302.1 LOSS OF 125 VDC POWER LOSS OF 125 VDC DIV I

***************************************************************INFORMATION***************************************************************

1013 125 voe Distribution Panel C loads: 1013 ckt 08 Reactor Protection System Channel A Vertical Board 1C15 Recirculation Pump A Trip circuitry Backup Scram Valve SV-1840A (FC) 1013 ckt 09 Inboard Isolation Valve Relay Panel 1C41 Inboard MSIVs position indication and DC solenoid valve control 1013 ckt 10 1C40A Annunciators 1D13 ckt 11 EBO Pump 1P-40 Starter Control (HS-3151) & Indication Emergency Bearing Oil Pump 1P-40 control and indication 1013 ckt 12 Generator and Plant Relay Panel 1C31 Generator Primary Lockout Relay 286/P Startup Transformer- Bus 1A3 Breaker 1A302 Lockout Relay Essential Bus 1A3 Load Shedding Circuit Non-Essential Auto Transfer and Load Shed 1D 13 ckt 14 Auto Slowdown Panel 1C45 ADS main control power 1013 ckt 15 4160V Bus 1A3 switchgear control 4KV BREAKER 1A301 STANDBY TRANSFORMER TO BUS 1A3 4KV BREAKER 1A302 STARTUP TRANSFORMER TO BUS 1A3 FEEDER BREAKER 1A3031A3 TO LC XFMR 1X31 Core Spray Pump 1P-211 A Supply Breaker 1A304 RHR Pump 1P-229A Supply Breaker 1A305 RHR Pump 1P-229C Supply Breaker 1A306 RHR SW Pump 1P-22A Supply Breaker 1A307 RHR SW Pump 1P-22C Supply Breaker 1A308 GSW Pump 1P-89A Supply Breaker 1A309 CRD Pump 1P-209A Supply Breaker 1A310 A DG TO BUS 1A3 BREAKER 1A311 FEEDER BREAKER 1A312, 1A3 TO LC XFMR 1X91 Essential Bus 1A3 Degraded Voltage Detection Circuit 1D 13 ckt 16 Load Center 1B3 switchgear control 1D 13 ckt 17 RCIC Relay Vertical Board 1C30 RCIC Turbine Speed Controller RCIC Turbine Trip circuitry RCIC Initiation and Isolation Relay Logic A RCIC Instrumentation 1013 ckt 19 NSSS Temperature and Leak Detection Panel 1C21 RCIC Area Steam Leak Detection circuitry Division I RCIC Timer Logic Division I 1013 ckt 20 Remote Shutdown Panels 1C389 and 1C390 Transfer Switch Position Status Indication IAOP 302.1 Page 26 of 63 Rev.58

AOP 302.1 LOSS OF 125 voe POWER LOSS OF 125 VDC DIV I

***************************************************************INFORMATION***************************************************************

1D14 125 VDC RCIC MCC loads: 1D14 ckt 01 Steam Outboard Isolation Valve M0-2401 1D14 ckt 02 Steam Supply Valve M0-2404 1D14 ckt 03 Turbine Stop Valve M0-2405 1D14 ckt 04 Bypass to Condensate Valve M0-2426 1D14 ckt 05 Suction from Condensate Storage Tank Valve M0-2500 1D14 ckt 06 Minimum Flow Bypass Valve M0-251 O 1D14 ckt 07 Normally Open Discharge Valve M0-2511 1D14 ckt 08 Normally Closed Discharge Valve M0-2512 1D14 ckt 09 Test Discharge Valve M0-2515 1D14 ckt 10 Suction at Pool Valve M0-2516 1D14 ckt 11 Suppression Pool Suction Valve M0-2517 1D14 ckt 12 Gland Seal Vacuum Pump 1P-227 1D14 ckt 13 Gland Seal Vacuum Tank Condensate Pump 1P-228 IAOP 302.1 Page 27 of 63 Rev.58

SECTION 'F' Rev.29 EMERGENCY COMMUNICATIONS Page 15 of_ 17 FIGURE F-5 DAEC TELEPHONE SYSTEMS

                                                                                           -------------1 Ir-----,

I I I I Operational Shellsburg Satellite Support Center Communications (Access ControQ Cell Tower (1) (1) _ DAEC Offsite Laboratory and

                                                                            .;                                              PBX                              Decontamination Room                                  Center 4 emergency                                                  Microwave unlisted ines (Blue Phones) In                                          to Alliant Control Room, TSC, CAS, SAS                                                                                        Joint Public Tower Information Center Local                                                               'Alliant Telephone Company                                                                       Tower (1)*EOF                                    Central Office                                                                     Microwave toDAEC

'FPLE Duane Arnold Corpomle Offices

• County Sherttfs Offices
• Pelo Fire Department
• Mercy Hospital
• st.le Highway Patrol Qwest Emergency Operations
• State Emergency Management Division
• Unl'/mlly of Iowa (Cedar Rapids) EOF Facility

'NRC 'DOE 'FEMA Nannal Telephone

• Linn County Emergency Manngement (1) Denotes a Dedicated
• Benion County Emergency Management Services To other Beu Central Line Offices

       .::~~bAE,C :~MERGEN,GY..:RLAN
                   *-* < :  *~- 'L,,     ...--..~' § .:. '

SECTION 'F' Rev.29 EMERGENCY COMMUNICATIONS Page 16 of 17 FIGURE F-6 FEDERAL TELEPHONE SYSTEM (FTS-2001 l NRC EROS EOF TSC 1 © ENS © HPN @ RSCL © PMCL @ MCL

DAEC EMERGENCY PLAN SECTION 'F' Rev. 29 EMERGENCY COMMUNICATIONS Page 17 of 17 FIGURE F-7 ALL-CALL TELEPHONE SYSTEM IKtwpfadllty

DAEC EMERGENCY PLAN SECTION 'E' Rev.23 NOTIFICATION METHODS AND PROCEDURES Page 3 of 7 1.0 PURPOSE (1) This section describes the methods and procedures used by FPLE Duane Arnold to transmit emergency information to the Emergency Response Organization, local and state authorities, and subsequently, from such authorities to the public. Details required in the initial and follow-up message are described, along with a description of the types of news statements that will be used to provide the public with information and protective actions . 2.0 REQUIREMENTS (1) Methods used to accomplish notification of the Emergency Response Organization include the use of call lists contained in the Emergency Telephone Book, pager and automated telephone callout process. (2) The Emergency Telephone Book includes phone numbers and pager numbers (where applicable) of emergency response personnel who may be required to respond to an emergency condition. It also includes the 24-hour telephone numbers of local, state , and federal support agencies including the NRC. The NRC would normally be notified using the NRC ENS Telephone (FTS-2001 System) from the Control Room. The state and counties would normally be notified by dedicated microwave telecommunications link. 2.1 l~ITIAL NOTIFICATION (1) After declaration of an emergency cond ition , the Operations Shift Manager/ Supervisor will ensure that the following personnel and agencies are notified :

• Linn and Benton Counties State of Iowa NRC Operations Center
• Emergency Coordinator
• Emergency Response and Recovery Director
• NRC Resident Inspectors (2) Verification of Notification (a) The authenticity of initial notifications provided to Linn and Benton Counties and the State of Iowa do not require verification if the notification is made by the dedicated phone system .

(b) Local and state agencies notified by commercial communication system (telephone or facsimile) may require verification of the identity and authenticity of the caller and the message received.

BASES-DAEC EOP BASES DOCUMENT BREAKPOINTS Rev. 14 EOP BREAKPOINTS Page 7 of 14 BREAKPOINTS FOR REACTOR LEVEL CONTROL Page 1 of 2 RPV Level Item of Interest Significance (inches)

  +211       High Level Trip Setpoint,

• Loss of high pressure injection (FW, Main Turbine Trip HPCI, RCIC)
• Loss of 100% Heat Sink

  +170       Low Water Level Scram,

• RPS defeats needed in ATWS PCIS Groups 2, 3, 4 Isolations
• Containment Isolation,
• Shutdown Cooling Valves Close

(+119.5} High Pressure Injection,) * (HPCI/RCIC Auto Initiation} PCIS Group 5 Isolation, ARI

• RWCU Isolation
• ARI Initiation & Recirc Pump A TWS Trip

   +87 '     Two Feet Below              During ATWS if power >5% or unknown, Feedwater Sparger           lower level to +87 inches to reduce core inlet 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

• If no preferred Injection Subsystem is available, maximize injection with Alternate Injection Systems in EOP 1 when level < +15" Note 1: +15 inches is used for TAF than O 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.

Emergency Preparedness Program Frequently Asked Question (EPFAQ) EPFAQ Number: 2016-002 Originator: David Young Organization: NEI Relevant Guidance: NEI 99-01, Methodology for Development of Emergency Action Levels, Revisions 4 and 5; and NEI 99-01, Development of Emergency Action Levels for Non-Passive Reactors, Revision 6. NUMARC/NESP-007, Methodology for Development of Emergency Action Levels. Applicable Section{s): Initiating Condition (IC) HA2 in NEI 99-01, Revisions 4 and 5, and NUMARC/NESP-007, "FIRE or EXPLOSION Affecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown" ICs CA6 and SA9 in NEI 99-01, Revision 6: "Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode" Definition of VISIBLE DAMAGE in NEI 99-01, Revisions 4, 5 and 6, and NUMARC/NESP-007 Status: Complete NOTE: Based on NRG staff consideration of industry comments provided by letter dated February 16, 2017 (ADAMS Accession No. ML17079A228), a revision to these /Cs was proposed at the public meeting held on April 4, 2017. These changes were attached to the public meeting . notice (ADAMS Accession No. ML17089A458). Based on comments provided by the industry during the April 4, 2017 public meeting, the NRG staff revised the proposed revisions to these

 /Cs.

QUESTION OR COMMENT: A review of industry Operating Experience has identified a need to clarify an aspect of the definition of VISIBLE DAMAGE as it relates to the ICs cited above; adding this clarity is necessary to minimize the potential for an over-classification of an equipment failure. There may be cases where VISIBLE DAMAGE is the result of an equipment failure and limited to the failed component (i.e., the failure did not cause damage to any other component or a structure). The current definition of VISIBLE DAMAGE does not adequately differentiate between damage resulting from, and affecting only, the failed piece of equipment vs. an equipment failure causing damage to another component or a structure (e.g., by a failure-induced fire or explosion). Can the definition of VISIBLE DAMAGE be clarified to help ~void an inappropriate emergency declaration in cases where an equipment failure does not result in damage to another component or a structure (i.e., VISIBLE DAMAGE affects only the failed component)?

• A related question is also posed - Consistent with the approach used in other ICs, should a note be added to preclude an emergency declaration if the safety system affected by a hazard was not functional before the event occurred (e.g., tagged out for maintenance)?

PROPOSED SOLUTION: Yes; the sentence below may be added to the definition of VISIBLE DAMAGE [as defined in NEI 99-01, Revisions 4, 5, and 6]. Damage resulting from an equipment failure and limited to the failed component (i.e., the failure did not cause damage to a structure or any other equipment) is not VISIBLE DAMAGE. From a plant safety and change-in-risk perspective, the consequences from the failure of a 1 r

Emergency Preparedness Program Frequently Asked Question (EPFAQ) piece of equipment, accompanied by a hazard (e.g., a fire or explosion) that does not damage any other equipment or a structure, are essentially the same as the equipment failing with no attendant hazard. Neither event would appear to meet the definition of an Alert because the outcome does not involve an actual or potential substantial degradation of the level of safety of the plant (e.g., there has been no significant reduction in the margin to a loss or potential loss of a fission product barrier). Nuclear power plants are designed with redundant safety system trains that are required to be separated (i.e., installed in separate plant areas or have separation within an individual area). Absent any collateral damage to another component or a structure, a hazard associated with an equipment failure does not affect the ability to protect public health and safety, and there is no additional response benefit to be gained by declaring an emergency. The normal plant organization has sufficient resources and adequate guidance to respond to an equipment failure -guidance includes operating procedures and Technical Specifications; the fire protection [program], industrial safety and corrective action programs; and work management and maintenance requirements. Concerning the second question, an emergency declaration would not be appropriate in response to a hazard affecting a piece of equipment or system that was non-functional prior to the event (e.g., tagged out for maintenance). For this reason and consistent with the approach used in other ICs, the following note may be added to IC HA2 (NEI 99-01 R4 and RS), or ICs CA6 and SA9 (NEI 99-01 R6). Note: If the affected safety system (or component) was already non-functional before the event occurred, then no emergency classification is warranted. Consistent with the guidance in Regulatory Issue Summary (RIS) 2003-18, Supplement 2, Use of Nuclear Energy Institute (NE/) 99-01, "Methodology for Development of Emergency Action Levels," Revision 4, dated January 2003, it is reasonable to conclude that the changes proposed above would be considered as a "deviation." NRC .RESPONSE: The proposed guidance is intended to ensure that an Alert should be declared only when actual or potential performance issues with SAFETY SYSTEMS have*occurred as a result of a

• hazardous event. The occurrence of a hazardous event.will result in a Notification of Unusual Event (NOUE) classification at a minimum. In order to warrant escalation to the Alert classification, the hazardous event should cause indications of degraded performance to one train of a SAFETY SYSTEM with either indications of degraded performance on the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second SAFETY SYSTEM train, such that the operability or reliability of the second train is a concern. In addition, escalation to the Alert classification should not occur if the damage from the hazardous event is limited to a SAFETY SYSTEM that was inoperable, or out of service, prior to the event occurring. As such, the proposed guidance will reduce the potential of declaring an Alert when events are in progress that do not involve an actual or potential substantial degradation of the level of safety of the plant, i.e., does not cause significant concern with shutting down or cooling down the plant.

IC HA2 (NEI 99-01 R4 and RS; NUMARC/NESP-007), or ICs CA6 and SA9 (NEI 99-01 R6), do not directly escalate to a Site Area Emergency or a General Emergency due to a hazardous event. The Fission Product Barrier and/or Abnormal Radiation Levels/Radiological Effluent recognition categories would provide an escalation path to a Site Area Emergency or a General Emergency. The proposed addition of the following notes, applicable to ICs l-lA2 (NEI 99-01 R4 and R5; NUMARC/NESP-007), or ICs CA6 and SA9 (NEI 99-01 R6), provide further clarification as to how these Alert emergency classifications are considered. The revisions to these EALs, 2

Emergency Preparedness Program Frequently Asked Question (EPFAQ) including the addition of the notes, are consistent with the current NRC-endorsed Alert classification language.

1. Adding the following note to the applicable EALs, per this EPFAQ, is acceptable as it meets the intent of the EALs, is consistent with other EALs (e.g., EAL HAS from NEI 99-01, Revision 6; this revision was endorsed by the NRC in a letter dated March 28, 2013, available at ADAMS Accession No. ML12346A463), and ensures that declared emergencies are based upon unplanned events with the potential to pose a radiological risk to the public.

If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then this emergency classification is not warranted.

2. Adding the following note to help explain the EAL is reasonable to succinctly capture the more detailed information from the Basis section related to when conditions would require the declaration of an Alert.

If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted. Revising the EALs and the Basis sections to ensure potential escalations from a NOUE to an Alert, due to a hazardous event, is appropriate as the concern with these EALs is: (1) a hazardous event has occurred, (2) one SAFETY SYSTEM train is having performance issues as a result of the hazardous event, and (3) either the second SAFETY SYSTEM train is having performance issues or the VISIBLE DAMAGE is enough to be concerned that the second SAFETY SYSTEM train may have operability or reliability issues. Revising the definition for VISIBLE DAMAGE is appropriate as this definition is only used for these EALs and the revised EALs are based upon SAFETY SYSTEM trains rather than individual components or struc~ures. All of the changes discussed above are addressed in the attached markups to NEI 99-01, Revision 6. Licensees that use NESP-007, NEI 99-01 Revision 4, or NEI 99-01 Revision 5 EAL schemes can adopt this language in the relevant format the staff approved for their use. Consistent with the guidance in Regulatory Issue Summary (RIS) 2003-18, Supplement 2, Use of Nuclear Energy Institute (NE/) 99-01, "Methodology for Development of Emergency Action Levels," Revision 4, dated January 2003, a licensee's scheme change based on this EPFAQ should be considered as a "deviation" because a classification based on NRC-endorsed industry guidance in NEI 99-01, Revisions 4, 5 and 6, as well as in NUMARC/NESP-007, could be different from a classification based on this EPFAQ. RECOMMENDED FUTURE ACTION{S): 0 INFORMATION ONLY, MAINTAIN EPFAQ iZ] UPDATE GUIDANCE DURING NEXT REVISION 3

Emergency Preparedness Program Frequently Asked Question (EPFAQ) CA6 ECL: Alert Initiating Condition: Hazardous event affecting SAFETY SYSTEMS needed for the current operating mode. Operating Mode Applicability: Cold Shutdown, Refueling Example Emergency Action Levels: Notes:

• If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then this emergency classification is not warranted.
• If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted.

(I) a. The occurrence of ANY of the following hazardous events:

• Seismic event (earthquake)
• Internal or external flooding event
• High winds or tornado strike
• FIRE
• EXPLOSION

                 * (site-specific hazards)

• Other events with similar hazard character.istics as determined by the Shift Manager AND

b. 1. Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode.

AND

2. EITHER of the following:

• Event damage has caused indications of degraded performance to a second train of the SAFETY SYSTEM needed for the current operating mode, or
• Event damage has resulted in VISIBLE DAMAGE to the second train of a SAFETY SYSTEM needed for the current operating mode.

Basis: This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the current operating mode. In order to provide the appropriate context for consideration of an ALERT classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train, and there must be either indications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance, and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE 4

Emergency Preparedness Program Frequently Asked Question (EPFAQ) such that the potential exists for performance issues. Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has indications of degraded performance for criteria 1.b.1 of this EAL; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events. Indications of degraded performance address damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. Operators will make a determination of VISIBLE DAMAGE based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. Escalation of the emergency classification level would be via IC AS1. Developer Notes: For (site-specific hazards}, developers should consider including other significant, site-specific hazards to the bulleted list contained in EAL 1.a (e.g., a seiche). Nuclear power plant SAFETY SYSTEMS are comprised of two or more separate and redundant trains of equipment in accordance with site-specific design criteria. ECL Assignment Attributes: 3.1.2.B 5

Emergency Preparedness Program Frequently Asked Question (EPFAQ) SA9 ECL: Alert Initiating Condition: Hazardous event affecting SAFETY SYSTEMS needed for the current operating mode. Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Example Emergency Action Levels: Notes:

• If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then this emergency classification is not warranted.
• If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted.

(1) a. The occurrence of ANY of the following hazardous events:

• Seismic event (earthquake)
• Internal or external flooding event
• High winds or tornado strike
• FIRE
• EXPLOSION

                 * (site-specific hazards)

• Other events with. similar hazard characteristics as determined by the Shift Manager AND

b. 1. Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode.

AND

2. EITHER of the following:

• Event damage has caused indications of degraded performance to a second train of the SAFETY SYSTEM needed for the current operating mode, or
• Event damage has resulted in VISIBLE DAMAGE to the second train of a SAFETY SYSTEM needed for the current operating mode.

Basis: This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the current operating mode. In order to provide the appropriate context for consideration of an ALERT classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train, and there must be either indications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance, and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE 6

Emergency Preparedness Program Frequently Asked Question (EPFAQ) such that the potential exists for performance issues. Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has indications of degraded performance for criteria 1.b.1 of this EAL; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events. Indications of degraded performance address damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. Operators will make a determination of VISIBLE DAMAGE based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. Escalation of the emergency classification level would be via ICs FS1 or AS1. Developer Notes: For (site-specific hazards), developers should consider including other significant, site-specific hazards to the bulleted list contained in EAL 1.a (e.g., a seiche). Nuclear power plant SAFETY SYSTEMS are comprised of two or more separate and redundant trains of equipment in accordance with site-specific design criteria. ECL Assignment Attributes: 3.1.2.B 7

Emergency Preparedness Program Frequently Asked Question (EPFAQ) VISIBLE DAMAGE: Damage to a SAFETY SYSTEM train that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected SAFETY SYSTEM train. 8

EPFAQ Number: 2018-04 Originator: David Young Organization: NEI Relevant Guidance: This question concerns NEI 99-01, Development of Emergency Action Levels for Non-Passive Reactors, Revision 6 and EPFAQ 2016-002. Applicable Section(s): Initiating Conditions (ICs) CA6 and SA9, and the associated Emergency Action Levels (EALs) and Bases

• Date Accepted for Review: 5/31/2018 Status: Under Review QUESTION OR COMMENT:

Background

EPFAQ 2016-002 provided guidance intended to reduce the potential of declaring an Alert when events are in progress that do not involve an actual or potential substantial degradation of the level of safety of the plant, i.e., does not cause significant concern with shutting down or cooling down the plant. In responding to the EPFAQ, the staff determined that revising the EALs and the Basis sections of ICs CA6 and SA9 would be appropriate to ensure potential escalations from a NOUE to an Alert, due to a hazardous event, occur when there is: (1) a hazardous event, and (2) one SAFETY SYSTEM train having performance issues as a result of the hazardous event, and (3) either the second SAFETY SYSTEM train is having performance issues or VISIBLE DAMAGE sufficient to be concerned that the second SAFETY SYSTEM train may have operability or reliability issues. The response to EPFAQ 2016-002 works well for situations involving a safety system with two trains (a typical configuration); however, industry operating experience indicates that additional clarification is needed for three other cases as described in the questions below. Because this EPFAQ is based on material in EPFAQ 2016-002, the response to this EPFAQ may be considered only by sites that have implemented EPFAQ 2016-002 in a manner approved through an NRC Safety Evaluation Report (SER). Question Concerning ICs CA6 and SA9, how should an event leading to indications of degraded performance and/or VISIBLE DAMAGE be classified when:

1. The event affects equipment common to two or more safety systems or safety system trains? For example, a unit with a tank that is the water source for multiple safety injection systems or trains, such as a Refueling Water Storage Tank (RWST).

2. The event affects a safety system that has only one train. For example, a Boiling Water Reactor (BWR) unit with a single-train Reactor Core Isolation Cooling (RCIC) or High-Pressure Coolant Injection (HPCI) system.

3. The event affects two trains of a safety system having more than two trains. For example, a unit that has an Auxiliary/Emergency Feedwater system with three trains.

1 05/21/2018

Emergency Preparedness Program Frequently Asked Question (EPFAQ) PROPOSED SOLUTION: The following answers to the above questions are proposed:

1. An event affecting equipment common to two or more safety systems or safety system trains (i.e., there are indications of degraded performance and/or VISIBLE DAMAGE affecting the common equipment) should be classified as an Alert under CA6 or SA9, as appropriate to the plant mode. By affecting the operability or reliability of multiple system trains, the loss of the common equipment effectively meets the two-train impact criteria that underlie the EALs and Bases.

2. An event affecting a single-train safety system (i.e., there are indications of degraded performance and/or VISIBLE DAMAGE affecting the one train) would not be classified under CA6 or SA9 because the two-train impact criteria that underlie the EALs and Bases would not be met. If an event affects a single-train safety system, then the emergency classification should be made ~ased on plant parameters/symptoms meeting the EALs for another IC. Depending upon the circumstances, classification may also occur based on Shift Manager/Emergency Director judgement.

3. An event that affects two trains of a safety system (e.g., one train has indications of degraded performance and the other VISIBLE DAMAGE) that also has one or more additional trains should be classified as an Alert under CA6. or SA9, as appropriate to the plant mode. This approach maintains consistency with the two-train impact criteria that underlie the EALs and Bases, and is warranted because the event was severe enough to affect the operability or reliability of two trains of a safety system despite plant design criteria associated with system and system train separation and protection. Such an event may have caused other plant impacts that are not immediately apparent.

As stated above, this EPFAQ may be considered only by sites that have implemented EPFAQ 2016-002 in a manner approved through an NRC Safety Evaluation Report (SER). With this proviso met, the response to EPFAQ 2018-004 would then provide clarification of expected emergency classifications for cases not explicitly addressed by ICs CA6 and SA9 (from NEI 99- 01, Revision 6), and EPFAQ 2016-002; therefore, implementation of the guidance in this EPFAQ would improve the accuracy and timeliness of a classification following a hazardous event affecting a safety system. Moreover, the answers provided in EPFAQ 2018-004 would result in EAL interpretations that are consistent with the meaning and intent of NRG-approved EAL bases such that the classification of the event would not be different from that approved by the NRC in a site-specific application. For this reason, it is reasonable to conclude that incorporation of the guidance from this EPFAQ into an NRG-approved site-specific scheme reflecting the guidance in EPFAQ 2016-002 would be considered a "difference" in accordance with Regulatory Issue Summary (RIS) 2003-18, Supplerpent 2, Use of Nuclear Energy Institute (NE/) 99-01, "Methodology for Development of Emergency Action Levels," Revision 4, dated January 2003. This "difference" determination is contingent upon incorporating any or all of the three answer statements (as applicable to a facility) verbatim; any change to the scope or intent of the answers would make incorporation into a site-specific scheme a "deviation" per RIS 2003- 018, Supplement 2. 2 05/31/2018 I_

Emergency Preparedness Program Frequently Asked Question (EPFAQ) NRC RESPONSE: RECOMMENDED FUTURE ACTION(S): 0 INFORMAT ION ONLY, MAINTAIN EPFAQ [81 UPDATE GUIDANCE DURING NEXT REVISION 3 05/ 31/2018

BASES-DAEC EOP BASES DOCUMENT BREAKPOINTS Rev. 14 EOP BREAKPOINTS Page 7 of 14 I BREAKPOINTS FOR REACTOR LEVEL CONTROL Page 1 of 2 RPV Level Item of Interest Significance (inches)

    +211       High Level Trip Setpoint,

• Loss of high pressure injection (FW, Main Turbine Trip HPCI, RCIC)
• Loss of 100% Heat Sink

    +170       Low Water Level Scram,

• RPS defeats needed in A TWS PCIS Groups 2, 3, 4 Isolations
• Containment Isolation,
• Shutdown Cooling Valves Close

   +119.5      High Pressure Injection,

• HPCI/RCIC Auto Initiation PCIS Group 5 Isolation, ARI
• RWCU Isolation
• ARI Initiation & Recirc Pump ATWS Trip

     +87       Two Feet Below              During A TWS if power >5% or unknown, Feedwater Sparger           lower level to +87 inches to reduce core inlet 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

• If no preferred Injection Subsystem is available, maximize injection with Alternate Injection Systems in EOP 1 when level < +15" Note 1: +15 inches is used for TAF than O 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 .

SAG-3 HYDROGEN CONTROL ( START ) CAUTIONS 0 -------------- Operation of HPCI , RCIC, Cora Spray, or RHR with suction from the tol'\ls and pomp flow Atl.m!a the NPSH or \"Ortexllrnit may damage equipment. IF Whlle In this procedure : THEN 0 H2 and 02 insb'uments may indicate higher concentrations HJ QC D.z monitor is unavailable Notify Ch.mistry to manually sample the than actually exist inside the contairvnent following a large dtywell end IDNs le< H, end 0, (PASAP 2.6). the monitors sholld !!2! be independently used to operational decisions but may be used fof' trending break LOCA due to moisture c:ond.nsation In the sa mp* Hnes. During 1h11 24 hr f*iod follc,,Ning a large break LOCA. Drywall pr*nur* drops below 2.0 psig Verify containment sprays isolate. Torus p1"ass1.n drops below 2.0 psig Terminal* to,us aprays. I DRYWELL TORUS I

                                                                                                                                                                                                                                                                      -- 7 t---..----------1
                                                                                                                                                                                           £     f-- - + - - -+ ---'                                         (@)

IF ....... offsite release rate is expected to stay t>.kJw nonnal limits IF ....... offsite release rate Is exped9d to stay below normal limits (Oetai D), (Dete;I D), THEN .vent and ~ the primary containrMnt: THEN .vent and pu-ge the primary containment: Detail D I Normal Release Rate Limits OK to defeat lsoiatklns m;u4 high reclatlcn (OefHt 9). OK to det.at ISOlatklns nc;,tgl ~ radiation (Defeat 9).

                                                                                                       -     It pneunatlc Sl.4)pllts ar, unavallei.. UM SAMP 706,                                    -     If pneumatic aupphs ** 1Mmva!labla. use SAMP 706, A detwmnatlon that the: offsite release rate Is below                                                  V.ntng the Pmwy CorulrwMl"l Follo>,<wing Loss of                                              V.ntlng the Primary Corialrwnent Following loss of normal lnvts may be made by either:                                                                    PneunlldcSUpply.                                                                              Pnallllallc Sopply.

• Containment Atmosphere Racfiation Monitor 1. Ventasfolla,,,,s: 1. V.nt as follows; GASEOUS Channels on-scale IOd opelllhle.
• IF ........ toruswalfilevells~16ft,
• IF........ tcn.,swateri....elis~16fl,

              -   Monitofed locally on RIT -810:zAIB et 1C-219M3                                               THEN ..* vent tho drywoll lhrough tho!Slllll (SEP 301.1).                                     THEN ... vent dir'Ktfy from the 1QClA (SEP 301 .1).

ot RR-4379AIB ot lC-29 (blue channe().

• IF ....... toruswaterlavelis~16ft,
• IF ........ the torus c#lOOI: be vented,
• Containment ~ e (PASAP 7.4). OR .....the torus cannot be vent.ct, ANO .... toruswaterlewl i s ~ 13.5 ft, THEN . vent diredly from tho .._n (SEP 301.2). THEN ... vent tho 1oNs 1hrough tho !lll!Wlll (SEP 301 .2).

2. IF ........ the primary containment can be VWlt.d, 2. IF ........ the prim.ary eontalnm.nt can be vented, THEN ... purge ttMI ~ with nit[2Qt!'.l using N:2 purge THEN ... pur;. tti. l2!:Y). 'Mth n!'.t!2gtn using N2 purge (SEP 303.2). (SEP 303.2).

3. Stop the vent and pu-ge {If not required by other SAG 3. Stop the vent and pur;. (if not raquil'9d by otMr SAG steps) when: steps) when:

• Hydrogen Is no long.r ~ in the drywell,
• Hydrogen Is no k>nger detected in the torus, OR OR
• Offsite release rate raeches normal lirrits (Dela~ 0 ).
• Offsite .-.i.as. rate raeches nonnal limits (Detail 0 ).

H-3 ..., IF .... ... offsite r-..ase rate I s ~ to stay below G.n.ral IF ... .... offsitti .-.lease rate is expected to stay beiow General

                                                                                                     ~Lovols(EALRG1 ),                                                                             Emorgoncy l.avols (EAL RG1 ),

OR ..... RPV water lev.l ~ be main!aiMd abow +15 in. {TAF), OR ..... RPV water i.v.1 ~ be maintaNd abow +15 in. {TAF), Tl-lEN .vent and p.xge the prima,y containment Tl-lEN ... vent and pwge h primary containment:: OK to def.at All Isolations (Oefut 10). OK to defeat .111 llolatlons (Defut 10).

                                                                                                       -     If pnelfflatlc MWlff.,.. unavailable, use SAMP 706,                                     ... If pneumatic supples we unavallabla, use SAMP 706, Veotng the Primary Contalrment FotloY.1ng Loss of                                             Venting the Primary Conlalnment Followlng Loss of Pne<.matlcSUpply.                                                                             Pnallllatlc.Sopply.

1. Ventasfollows; 1. Vent ., fotlows:

• IF ....... toruswaterlev.lls~16ft,
• IF ........ toruswaterlevells~16ft, THEN .vent the drywell ttvough the klall (SEP 301 . 1). THEN ... vent directty from the!&!a.11 (SEP 301 .1).
• IF ....... toruswaterlevells~16ft,
• IF ........ the torus cwnot be vented, OR.... . the torus cannot be vented, ANO .... torus watw lewl i s ~ 13.5 fl, THEN .vent di redly from t h e ~ (SEP 301 .2). THEN ... vent tho ION> through tho !lll!Wlll (SEP 301.2).

2. IF ........ the pm,ary containment can be vent.ct, 2. IF ... ..... the prim.-y contaim.nt can be vented, THEN ... ~ t h e ~ wtth ~ at max flow THEN ... ""'116 tho 12M with !lfil2!I!!! at max flow using N, ""'116 (SEP 303.2). using N, ""'116 (SEP 303.2).

3. Stop tho vent and ""'116 [I not required b y - SAG 3. Stop the v.nt and pu-ge (If not r<<p.Jirad by Olhet SAG steps) when: st~s)when:

• Hydrogen is no k>nger' detected In eithw the dry-Nell or * ~ ia no longer detected in either 1h11 drywll or tholoNs, tho ION$,

OR OR

• Hjd<ogen is no longer dotoctod ;, tho drywall f!!!l
• Hydrogen is no k>nger detected in the torus~ torus dlyw9!1 oxygen is leu than 5%, oxygen is less then 5%,

OR OR

• RPI wai.r i.v.i can be ma.ltained ebove
• RPI water i.vel can be maintained abow

                                                                                                               +15 in. (TAF)IOd ofhite release rate reaches                                                  +15 ii. (TAF) aod of'fsite release rate reaches a General Emergency Level (EAL RG1 ).                                                         a General Emergency Level {EAL RG1 ).

H~ Vent and purge the primary contaimi.nt Vent and purge the primary contalM*rt:

                                                                                                 ..- OK to defut J!!!. lsolatiot'ls (Defeat 10).

Detail EI Spray Sources ..- OK to exceed release rate lnlts, OKtodefeatA!.isolations(Oefeat 10). OK to exCHd release rate limits.

                                                                                                 -     If pneumatic st"1Plles are Ul'\IValleble. UM SAMP 706, Ventng                           -     If pneumatic supplies are lNlvallabla, use SAMP 706, V.ntlng

• RHR (Ol 149) the PrfmMy Contalnmrent Following Losa or Pnet.matlc ~ - the Primary Contairment FollOWlng Loss of Pneu'natlc s ~.
• RHR Service Water (AIP 401 ) 1. Vent as follows: 1. IF........ pomwttodbySAG-1 ,
• IF ........ toruswaterlev-'*blbztt18ft, THEN ... - Jaus sprays (Detail E).
• Fn System (AJ P 404)

THEN..* vent tho drywol lhrough tho lo!J!t (SEP 301 .1). 2. Vent a, follows :

• Well Water (AIP 403)
• IF .....** toruswateclev.lis~16fl,
• IF ........ torutwaterlevells~1Sft,
• GSW(AJP403) OR ...... tti. torus CMnOt be wnted, THEN... vent directly from 1h11 mo.ti {SEP 301 .1).

THEN... vent <firocily tom tho <1oo,!11 (SEP 301.2).

• ESW (AIP 402)
• IF ........ the torus cannot be vented,

2. IF .......* the primary contairment can be vented, ANO .... toruswatert.vel I s ~ 13.5ft,

• Condenaate Servk:e Water (AJ P 405) THEN... pixve t h e ~ v.ith ~ at max flow. Use THEN ... vent tti. torus through the .dlXdf (SEP 301 .2).

whichever 1Mthod v.il reduce hydrogen below 6% or oxygen below 5% faster. 2. IF ........ the primary conuiiMlent can be v.nted,

• Airpurgo(SEP303.1) THEN ... purgeh.12Da.....;th~atmaxfla.¥.Usa w-hiCNver method wlR reduce hydrogen below 6% or
• N, ""'116 (SEP 3032) oxygen below 6% faster:

3. IF ..... .. pem,lttod by SAG-1 ,

• Airpurgo(SEP303.1)

THEN. op,,rate drywell spray. (Detaa E).

• N, purge (SEP 3032)

DUANE ARNOLD ENERGY CENTER .... .... Controlled Copy SAG-3 HYDROGEN CONTROi. REV. 6

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1.2. 7 HSM Dose Rates with a Loaded 24P, 528 or 61 BT DSC Limit/Specification : Dose rates at the following locations shall be limited to levels which are less than or equal to:

a. 400 mrem/hr at 3 feet from the HSM surface.

b. Outside of HSM door on center line of DSC 100 mrem/hr.

c. End shield wall exterior 20 mrem/hr.

Applicabil ity : This specification is applicable to all HSMs which contain a loaded 24P, 528 or 61BT DSC. Objective: The dose rate is limited to this value to ensure that the cask (DSC) has not been inadvertently loaded with fuel not meeting the specifications in Section 1.2 .1 and to maintain dose rates as-low-as-is-reasonably achievable (ALARA) at locations on the HSMs where surveillance is performed , and to reduce off-site exposures during storage. Action: a. If specified dose rates are exceeded, the following actions should be taken:

1. Ensure that the DSC is properly positioned on the support rails .

61BT DSC Dose Rate Thresholds = 2 X TS limits

2. Ensure proper installation of the HSM door.

Therefore: 3. Ensure that the required module spacing is maintained. 3 feet from HSM Surface 4. Confirm that the spent fuel assemblies contained in the DSC = 800 mrem/hr conform to the specifications of Section 1.2.1 .

5. Install temporary or permanent shielding to mitigate the dose to Outside HSM Door -

acceptable levels in accordance with 1 O CFR Part 20, 10 CFR Centerline of DSC 72 .104(a), and ALARA. = 200 mrem/hr

b. Submit a letter report to the NRC within 30 days summarizing the End Shield Wall Exterior action taken and the results of the surveillance, investigation and

= 40 mrem/hr findings. The report must be submitted using instructions in 10 CFR 72.4 with a copy sent to the administrator of the appropriate NRC regional office . Surveillance: The HSM and ISFSI shall be checked to verify that this specification has been met after the DSC is placed into storage and the HSM door is closed . Basis: The basis for this limit is the shielding analysis presented in Section 7.0, Appendix J, and Appendix K of the FSAR. The specified dose rates provide as-low-as-is-reasonably-achievable on-site and off-site doses in accordance w ith 10 CFR Part 20 and 10 CFR 72.104(a). Certificate of Compliance No. 1004 A-78 Amendment No. 9, Revision No. 1

D evelopment of Fission Product Barrier EAL Threshold Values from NEE-323-CALC-001 NOTE: Fuel Clad barrier LOSS 4.A(B) threshold values below are scaled from the 100% gap release instead of calculated based on 300uci/gm DE! as assumed in NE! 99-01 Revision 6 developer guidance. This variation from the NRC endorsed guidance is due to the calculated value not reflecting the intended 2-5% gap release threshold due to differences in plant design. The calculation will be formally revised to reflect this change in methodology. Drywell dose rate Torus dose rate Drywell dose rate Torus dose rate R/hr R/hr R/hr R/hr Values below are Values below are rounded for ease of rounded for ease of 1691 MWth 22700 2140 use, as well as to use, as well as to 100% Gap release provide a step-wise provide a step-wise progression progression Scaling factor to account for power 1.13 1.13 uprate to 1912 MWth Updated 100% Gap release 25667 2420 After application of 0.2 scaling factor for 20% Gap release 5133 484 5000 500 CTMNT barrier LOSS 4A(B) After application of 0.05 scaling factor for 5% Gap release 1283 121 1250 125 Fuel Clad barrier LOSS 4A(B)

CALC NO. NEE-323-CALC-OO 1 tr:, ENERCON

 .IM\;

Exci:llence-EVery project: Every day. CALCULATION COVER SHEET REV. 00 PAGE NO. 1 of 10 Client: Duane Arnold Energy Center Primary Containment Radiation EAL Threshold

Title:

Determination Project Identifier: NEE-323 Item Cover Sheet Items Yes No 1 Does this calculation contain any open assumptions, including preliminary information, that require confirmation? (If YES, identify the assumptions.) D IZI 2 Does this calculation serve as an "Alternate Calculation"? (If YES, identify the design verified calculation.) D IZI Design Verified Calculation No. -- 3 Does this calculation supersede an existing Calculation? (If YES, identify the design verified calculation.) D IZI Superseded Calculation No. -- Scope of Revision: Initial Issue Revision Impact on Results: Initial Issue 0 Study Calculation D Final Calculation ~ Safety-Related D Non-Safety-Related ~ (Print Name and Sign) Originator: Aaron Holloway Date: 12/12/17 Design Verifier1 (Reviewer if NSR): Jay Bhatt Date: 12/12/17 Approver: Zachary Rose Date: 12/12/17 Note 1: For non-safety-related calculation, design verification can be substituted by review.

CALC NO. NEE-323-CALC-001 Q ENERCON CALCULATION Exce/!ence---:E~ery project. Every day. REVISION STATUS SHEET REV. 00 CALCULATION REVISION STATUS REVISION DATE DESCRIPTION 00 12/12/17 Initial Issue PAGE REVISION STATUS PAGE NO. REVISION PAGE NO. REVISION All 00 APPENDIX/ATTACHMENT REVISION STATUS APPENDIX NO. NO.OF REVISION ATTACHMENT NO.OF REVISION PAGES NO. NO. PAGES NO. A 1 00 1 4 00 B 1 00

CALC NO. NEE-323-CALC-001 Q ENERCONExceilen~e-:Every project. £Very day. TABLE OF CONTENTS REV. 00 Section Page No.

1.0 Purpose and Scope

4 2.0 Summary of Results and Conclusions 4 3.0 References 5 4.0 Assumptions 5 5.0 Design Inputs 6 6.0 Methodology 8 7.0 Calculations 10 8.0 Impact Assessment 10

                                                                                  #of List of Appendices                                                                Pages Appendix A - Calculation Spreadsheet                                         1 Appendix B - Calculation Spreadsheet Formulas                                1
                                                                                  #of List of Attachments                                                               Pages Attachment 1 - Calculation Preparation Checklist                             4 Page 3 of9

a ENERCON

• Excellence-Every project E~ery day.

Primary Containment Radiation EAL Threshold Determination CALC NO. NEE-323-CALC-OO 1 REV. 00

1.0 Purpose and Scope

The purpose of this calculation is to determine the site-specific threshold for primary containment radiation in the event of a loss or potential loss of the three fission product barriers (fuel clad, Reactor Coolant System, containment). These site-specific values can be used to determine the Emergency Action Level (EAL) (FA1, FS1, or FG1) in accordance with Table 9-F-2 of NEI 99-01, Rev. 6. This calculation is nonsafety-related as it intended for emergency classification and not design basis purposes. There are no acceptance criteria associated with this calculation since the purpose is only to determine site-specific radiation thresholds. 2.0 Summary of Results and Conclusions The calculated primary containment radiation readings for each of the three fission product barriers are listed below. Note that the results presented below are calculated dose rates and do not account for background radiation or any installed detector check sources. Table 1 - Calculated Containment Atmospheric Monitoring System (CAMS) radiation readings for a release into the drywe/1 Failure Drywell Monitor (9184A/B) Reading (R/hr) rJR~~sWr G(ig!~iit!$rit~)v~~2,~,~>J0.;'.~;; *:ly**,fi~,y~:t: ;\:h ,*:;:~01 :e:;* t!*~~~r~:;0.~;;j;33~'.?'.ff:<i$~Wiir:~I1d~~¢iij[sfmXnimum readable 1 1 Fuel Clad (Loss) 2000 used as 2000 t;~9:ai~1~m~iitfOOqi~n:H~1i,~~s"s5:.;:~;J; ;*~t:tit\J p;r~'r:s tt~J,it:If~{'*':,;1)*trist;t,~1;~;1~JJ1iz;,;t<?~ijq~s!:t2:;~.Q~2::;':a 1 1

                        . Table 2 - Calculated CAMS radiation readings for a release into the torus Failure                      Torus Monitor (9185A/B) Reading (R/hr) fP/Ri,~~Jo~tJ&ij1t~t*;$y¥,t~ml<~((sJ:;.:f{~;:r;:is,Jr;~:;;:ss~~,~:**.\:~*i;f,;i).: "s*LQ:~rn~Et~iS\\P>>J~£'.~l§}:f0~:~/:ii:0:ii:~'f*~t,.l;~;~

Fuel Clad (Loss) 188 rounded to 200

  !;,*~.~i(a~miut***.,?~t~Att~fi:,;iJ~'~*,,i{\)]';*

Page 4 of 10

CALC fJ ENERCON

       &Cellence-:-Every project Every day.

Primary Containment Radiation NO. NEE-323-CALC-001 EAL Threshold Determination REV. 00 3.0 References 3.1 NG-88-0966, "Nuclear Generation Division Office Memo, G.E. Fuel Damage Documentation I Dose Rate Calculations", dated 03/18/88 3.2 IPOI 8, "Outage and Refueling Operations", Rev. 91 3.3 Bech-M115, "Reactor Vessel Instrumentation P&ID", Rev. 62 3.4 Duane Arnold Energy Center Facility Operating License Appendix A - Technical Specifications, as revised through Amendment No. 297 3.5 NEI 99-01, "Development of Emergency Action Levels for Non-Passive Reactors", Rev. 6 3.6 Shultis, J.K., "Fundamentals of Nuclear Science and Engineering", 2002 3.7 Lindeburg, M.R., "Mechanical Engineering Reference Manual for the PE Exam", Twelfth Edition, 2006 3.8 Federal Guidance Report (FGR) 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion", 1989 3.9 I.RIM-V115-01, "Victoreen Model 876A Containment Radiation Monitor Calibration", Rev. 10 Page 5 of 10

CALC 0- ENERCON Exci.:llence*7Every projeci. ~very day. Primary Containment Radiation EAL Threshold Determination NO. NEE-323-CALC-001 REV. 00 4.0 Assumptions 4.1 Reactor Pressure Vessel (RPV) water level is 535 inches above vessel zero for the purposes of calculating the total Reactor Coolant System (RCS) water volume. This corresponds to the middle of the band between the high and low RPV water level alarm points from Table 4 (Design Input 5.7) and represents the most realistic water inventory during normal operation. 4.2 The fission product isotopic distribution in the reactor coolant will be similar to that of the fission product gap inventory. This is reasonable since, in the event of a fuel cladding failure, the isotopes of concern (iodines) would be released to the reactor coolant at the same time and distribution. 4.3 All reactor coolant mass is assumed to be released into the primary containment. This is consistent with the NEI 99-01 Rev. 6 developer notes. 5.0 Design Inputs 5.1 Duane Arnold's license thermal power limit is 1912 MWth, taken from Reference 3.4. 5.2 The specific volume of saturated liquid water at 1000 psia is 0.02160 ft 3/lbm per Appendix 23.B of Reference 3.7. 5.3 The following unit conversions are used within this calculation: Table 3 -Applicable Unit Conversions _ Base Unit Equivalent Reference

                   , t)f;.$\~~~rt,($y}:':;,:0~i:,;l~,<;C: *i~::1JQ~~ilmt2w\:Z;,: .,. '.;*~; ;';c,Jt[ S:'.[~t:i':~:,~I<i:Z\'.*'~ . ~:;}l: *'.:S'{j:[i~;

1 1 1 Curie 3.7El0 3.6 5.4 The Technical Specifications limit for RCS activity is 0.2 µCi/gm Dose Equivalent 1-131 (DEi) per LCO 3.4.6 of Reference 3.4. 5.5 The RCS volume at the centerline of the Main Steam lines is 72,000 gallons per Reference 3.2. 5.6 The change in RCS volume per unit change in height is 100 gallons/inch per Reference 3.2.

5. 7 The following elevations are taken from Reference 3.3:

Table 4 - Pertinent RPVelevations relative to Vessel 0 Point Height above Vessel O(inches) Page 6 of 10

CALC Q ENERCON Primary Containment Radiation NO. NEE-323-CALC-001 Excellence-Every project Every day. EAL Threshold Determination 1------- - -------1 REV. 00 Point Height above Vessel O(inches) High Level Alarm 539.5

              ; LQ\V t~vel Al~rm : ,

5.8 The drywell dose rate at the CAMS monitor location for 100% gap release into the drywell of a 1691 MWth core'at 0.01 hours decay time is 2.27E4 R/hr, per Table 3 of Reference 3.1. 5.9 The torus dose rate at the CAMS monitor location for 100% gap release into the torus of a 1691 MWth core at 0.01 hours decay time is 2.14E3 R/hr, per Table 4 of Reference 3.1. 5.10 The drywall (9184 NB) and torus (9185 NB) radiation monitor ranges (1 to 107 R/hr) are taken from Reference 3.9. 5.11 The fission product gap inventories for Iodine isotopes are taken from Table 1 of Reference 3.1. These inventories correspond to a core with a rated thermal power of 1691 MW!h. Table 5 - Iodine Gap Inventories for 1691 MWth Core Nuclide 1691 MWth Gap Inventory (Ci)

                                  ' I~!~-~_______                        _____ ___ _ ___ :*._ 7.25E+03 _ _ ___ . ______ _

1-131 5.34E+05

                                  !:!':!-~I:_** , -,                                           ,____ ,}*::t1:?g+/-QJ____ .*,,_,*                                                                                 _,_:_.c:_

1-133 3.63E+05 l'!':!~-~--' -_*, ,_ - . '".8:19E+04:

                                                                                                           ----- _. __ ,_* ..**-*----*-**""- ... . ~~-~---~-- ---

1-135 l.93E+05 5.12 Dose conversion factors for effective dose due to inhalation are taken from Reference 3.8, Table 2.1. Table 6 - Dose Conversion Factors for Total Effective Dose from Inhalation Nuclide Dose Conversion Coefficient (Sv/Bq)

                                                                                                                          * * ""-\.

7.14E-lo *. .

                                                                                                                                    * * * ~ ' ' ' - " " - ' - * " " - * ~ " ' ' " ' - ' ,*  ~,{, * * * * - c,~ ***' ,_' --* * ** " *

• 1-131 8.89E-09 1-133 l.58E-09

                                         ] ..134:.,< i.\. (*
                                         '., ,,,,.'*.,*** '                  ~       '"***-*,~**"*"*-***-- *-----*-"-*'~---..

1-135 3.32E-10 Page 7 of 10

CALC 0 ENERCC)N Excellence-Every project Every day. Primary Containment Radiation NO. NEE-323-CALC-OO 1 EAL Threshold Determination REV. 00 6.0 Methodology The approach of this calculation is to scale the results of a previous calculation (NG-88-0966, Reference 3.1) based on the specific RCS activities as specified in NEI 99-01 Rev. 6. Scaling factors are determined for each of the fission product barrier failure thresholds specified in NEI 99-01 (i.e. Loss of RCS, Loss of Fuel Cladding, and Potential Loss of Primary Containment). The RCS activity concentrations are taken from the NEI 99-01 Rev. 6 developer notes and are listed below. Table 7 - RCS Activities for fission product barrier failures Failure RCS Activity

         !_ReaciQfCoolant SysienL(to*~s) .. ___ .* __ ** __ technical ~pecific!ttion tiwif', **-*-:1
          . F1:1~I fl~d (~oss). ....            *q m  . . ._..... ..,. _     30Q_~Ci/g ])o~e_Equivalenq-µ1 __ _

l Containment (Potential Loss) 20% fuel cladding failure * .. . These scaling factors are then applied to the CAMS radiation response from calculation NG-88-0966 to determine the site-specific values for the three thresholds. Additionally, the previous calculation determined the CAMS radiation monitor response for an assumed release of 100% gap activity from the core with a power level of 1691 MWth. However, Duane Arnold's licensed power level is now 1912 MWth. This does not impact the Reactor Coolant System and Fuel Clad barrier thresholds because the radiation responses are scaled based on the DEi levels. However, the difference in licensed power level will need to be accounted for in the Potential Loss of Containment threshold because this threshold is related to the total gap inventory. Scaling of the gap inventory based on power level is consistent with calculation NG-88-0966 as Table 1 of NG-88-0966 provides gap inventory per megawatt. 6.1 Determination of RCS water volume and mass The NEI 99-01 Rev. 6 primary containment radiation thresholds are based on specific RCS radioactivity concentrations. However, the corresponding total RCS activity must be known in order to compare these thresholds to the gap release assumed in calculation NG-88-0966. Therefore, the total mass of water in the RCS must first be determined so that the total RCS activity can be calculated for each threshold. The total RCS water volume at normal operation is determined by taking the RCS volume when filled to the centerline of the main steam lines, and then subtracting the difference in volume between the centerline of the main steam lines and the normal water level. This is presented in Equation 1 below: vnormal = VMSL - aH [Equation 1] Where: Vnormal = The RCS water volume at normal operation (gallons) Page 8 of 10

CALC a ENERCON Ex?;ellenc~EVery project Ev~ry day. Primary Containment Radiation EAL Threshold Determination NO. NEE-323-CALC-001 REV. 00 VMsL = The RCS water volume when filled to the centerline of the main steam lines (gallons) a = The change in RCS water volume per inch change in vessel height (gallons/inch) H = The distance between the centerline of the main steam lines and the normal RCS water level (inches) It should be noted that calculating the RCS volume as shown above does not include the volume of the steam lines. However, the volume of the steam lines filled to the centerline of the nozzle is very small compared to the total RCS volume, and therefore does not significantly impact the results of the calculation. The total mass of water in the RCS can then be determined based on the water density, as outlined in Equation 2 below: M norma.l.

                                                     = Vnonnal v              [Equation 2]

Where: Mnormal = The mass of water in the RCS at normal operation (grams) v = The specific volume of water at normal operation (grams/gallon) 6.2 Determination of Scaling Factors The scaling factors for the fuel clad and RCS barrier thresholds are determined by comparing the corresponding dose at eaeh RCS activity concentration to the DEi of the fission product gap inventory. This is presented in Equation 3 below: DCF1-131A1-131Mnormaz F = ~i=13s vcr.

                                       .l.ii=130 1i    I'i                       [Equation 3]

Where: F = The scaling factor for a given RCS activity concentration threshold. DCF,= The dose conversion factor for isotope "i" in mrem/Ci. These values are developed from

• Table 6 above.

A1-131 = The 1-131 concentration in the RCS for a given threshold in Ci/gram. These values are developed from Table 7 and Design Input 5.4 above. Mnormal = The mass of RCS water at normal operation in grams. This value is determined from Equation 2 above.

    /;=The gap inventory of iodine isotope "i" at a power level of 1691 MWth in Ci. These values are taken from Table 5 above.

Page 9 of 10

CALC 0 . ENERCON Ex~ellence-;-EveryprojeCr. Every day. Primary Containment Radiation EAL Threshold Determination NO. NEE-323-CALC-OO 1 REV. 00 For the potential loss of containment threshold, NEI 99-01 specifies that 20% of fuel cladding has failed, rather than giving a specific RCS activity concentration. Therefore, the scaling factor for this case is simply 0.2 (20% of the 100% gap release case) multiplied by the ratio of the new to previous licensed power levels (1912/1691) to account for the increased gap inventory. 6.3 Determination of CAMS Detector Response Once the scaling factors have been determined for each of the three RCS activity concentration thresholds, they can be . applied to the results of calculation NG-88-0966 to determine the CAMS detector response for each threshold. Specifically, the CAMS detector response can be obtained from Equation 3 below: [Equation 3] Where: Dj = The dose rate at the CAMS detector for an RCS activity concentration of "j" in R/hr Fj = The scaling factor for an RCS activity concentration of "j", determined from Equation 2 above DGAP = The dose rate at the CAMS monitor location for 100% gap release of a 1691 MWth core in R/hr 7 .0 Calculation All calculations were completed using Microsoft Excel. The calculation results and spreadsheet formulas are presented in Appendix A and B, respectively. 8.0 Impact Assessment This calculation is based on "realistic" assumptions for the purpose of declaring EALs, rather than typical conservative "bounding" type design basis analyses. The calculation results are intended to provide order of magnitude dose rates to assist Operations and Emergency Response personnel in determining the state of the three fission product barriers in accordance with NEI 99-01 Rev. 6. Page 10 of 10

CALC

                      ~ ENERCON                                             Appendix A                      NO.

NEE-323-CALC-001 Ercellence-Every project. Every day. Calculation Spreadsheet REV. 00 A _B C D E 1 1.Sievert 100000 mrem 1 1Ci 3.70E+10 Bq 3 4 Isotope DCF (Sv/Bq} DCF (rrirem/Ci} 1691.MW Gap Inventory (Ci} Dose (mrem} 5 l-i30 7;14E-10 2.64E+06 7.25E+03 1.92E+10 6 1-131 *8(89E-09 3:29E+07 5.34E+05 1.76E+13 7 1-132 1;03E-10 3:81E+05 8.45E"".04 3.22E+10 8 1-133 1.58E-09 5J~5E+06 3.63E+05 2.12E+12 9 1-134 " 3.55E-11 1.31E+05 8.19E+04 1.08E+10 10 1-135 3,32E-10 1.23E+0_6 1.93E+05 2.37E+11 11 12 Total 2.00E+13 13 VMSL 72000.0 gal 14 a 100;0 gal/inch 15 Elevation of the.Main Steam-Lines 620.5 inches above vessel 0 16 Elevation oftheNormal Water Level 535;0 i_nches above vessel 0 '17 H 85.5 inches 18 VNormal 63450 gal 19 VNormal 240184264;5 cc 20 Specific Volume.@ 1000 psia 0.0216 fti\3/lbm 21 water density@ 1000' psi a b:7416 ri,/cc 22 MNormal 178114424 grams 23 24 DrywellD_oseHate for 100%Gap Release (1691 MWth} 2.27E+04 R/hr -25 Torus Dose Rate for 100%Ga1tRelease (1691 MWth} 2.14E+03 R/hr 26 27 Threshold Scaling Factors {F) Drywell Dose Rate.{R/hr)

• Torus Dose:Rate {R/hr) 28 0.2 µCi/gm {TS Limit) 5.86E-05 1.33E+OO 1.25E-01 29 300 µCi/gm 30 20% failed Fuel

                                                                        --,       8.79E-02 2.26E-01
                                                                                                               .2.00E+03 S;13E+03 1.88E+02 4.84E+02 Page 1 of 1

CALC Q ENERC.ON Excellence-Every project Every day. Appendix B Calculation Spreadsheet NO. NEE-323-CALC-001 Formulas REV. 00 A B C o E 1 LSlevert lOOQOO

                                                                                     -                mrem 2 1Ci                                                                  37000000000                   Bq 3,

4 Isotope DCF (Sv/8cj) btF (mrem/Ci) 1691 MW. Gap Inventory (Ci) Dose (mrem) 5 1~130 0. 000000000714 =85*$B$1 *$B$2 7250 =D5"CS 6 1-131 0. 00000000889 =86"'$8$1 *$8$2 534000 . =D6*C6 7 1-132 0;000000000103 =87"'$8$1*$B$2 84500 =D7*C7 8 1-133 0.00000000158 =B8*$B$1"'$B$2 363000 =D8*C8 9 1-134 0.0000000000355 =89*$8$l*$B$2 819.00 =D9*C9 10 1-135 0.000000000332 =B10*$8$1*$8$2 193000 =D10*C10 11 12 Total =SUM(E5:El0) 13 VMSL. 7iOOO gal 14 a 100 gal/inch 15 Elevation of the Main Steam Lines' 620:s inches. abovttvessel 0 16 Elevatioi,ofthe Normal.Water Level 535 inches* above. vessel O 17 H =815-816 inches ta VNonnal =813-B14

• 817 gal 19 VNonnaf. =818*3785:41. cc 20 Specific Volume @ 100.0 psia 0.0216 ft"3/lbm 21 water density@lOOO psia =0.016018/820 g/cc

22 MNorrnal =B19*B2l grams 23 24 Drywell Dose Rate for 10()% Gap Release (1691 MWth) 22700 C R/hr 25 Torus Dose.Rate for 100% Gap Release,(1691MWth) 2140 R/hr 26 27 Threshold ScaHng Factors. (Fl Drywell Pose Rate (R/hr) Torus Dose Rate (R/hr) 28 0.2 µCl/gm (TS Limit) =$C$6*0.0000002*$B$22/($E$12J =B28*$B$24 =B28*$B$2S 29 300 µCl/gm. =$C$6*0.0003*$B$22/($E$12) =B29'!$B$24 . =B29~$B$2S .

30 20% Failed Fuel, . =0.2*1912/1691, =B30~$B$24 =B30*$B$25. Page 1 of 1

CALC

 <<:) ENERCON
             &cellence-Every project. Every day.

Attachment 1 CALCULATION PREPARATION NO. NEE-323-CALC-001 CHECKLIST REV. 00 CHECKLIST ITEMS1 YES NO N/A GENERAL REQUIREMENTS

1. If the calculation is being performed to a client procedure, is the procedure being used the latest revision? D D fgl The calculation is being prepared to ENERCON's procedures.

2. Are the proper forms being used and are they the latest revision? fgl D D

3. Have the appropriate client review forms/checklists been completed? D D fgl The calculation is being prepared to ENERCON's procedures.

4. Are all pages properly identified with a calculation number, calculation revision and page number consistent with the requirements of the client's procedure?

fgl D D

5. Is all information legible and reproducible? fgl D D

6. Is the calculation presented in a logical and orderly manner? fgl D D

7. Is there an existing calculation that should be revised or voided? D 181 D This is a new calculation to support implementing NEI 99-01 Rev. 6

8. Is it possible to alter an existing calculation instead of preparing a new calculation for this situation? D fgl D

9. If an existing calculation is being used for design inputs, are the key design inputs, assumptions and engineering judgments used in that calculation valid and do they 181 D D apply to the calculation revision being performed.

10. Is the format of the calculation consistent with applicable procedures and expectations?

fgl D D

11. Were design inpuUoutput documents properly updated to reference this calculation? D D fgl

12. Can the calculation logic, methodology and presentation be properly understood without referring back to the originator for clarification? 181 D D OBJECTIVE AND SCOPE

13. Does the calculation provide a clear concise statement of the problem and objective of the calculation? 181 D D

14. Does the calculation provide a clear statement of quality classification? fgl D D

15. Is the reason for performing and the end use of the calculation understood? 181 D D

16. Does the calculation provide the basis for information found in the plant's license fgl D D basis?

17. If so, is this documented in the calculation? fgl D D

18. Does the calculation provide the basis for information found in the plant's design basis documentation? D fgl D Page 1 of 4

CALC

<<:ii
 ....... ENERCON         .              .

Excellence-Every project. £very day. Attachment 1 CALCULATION PREPARATION NO. NEE-323-CALC-001 CHECKLIST REV. 00 CHECKLIST ITEMS1 YES NO N/A

19. If so, is this documented in the calculation? D D 181

20. Does the calculation otherwise support information found in the plant's design basis documentation? D 181 D

21. If so, is this documented in the calculation? D D 181

22. Has the appropriate design or license basis documentation been revised, or has the change notice or change request documents being prepared for submittal? D D 181 DESIGN INPUTS

23. Are design inputs clearly identified? 181 D D

24. Are design inputs retrievable or have they been added as attachments? 181 D D

25. If Attachments are used as design inputs or assumptions are the Attachments traceable and verifiable? D D 181

26. Are design Inputs clearly distinguished from assumptions? 181 D D

27. Does the calculation rely on Attachments for design inputs or assumptions? If yes, are the attachments properly referenced in the calculation? D 181 D

28. Are input sources (including industry codes and standards) appropriately selected and are they consistent with the quality classification and objective of the calculation? 181 D D

29. Are input sources (including industry_podes and standards) consistent with the plant's 181 D , ~D design and license basis?

30. If applicable, do design inputs adequately address actual plant conditions? 181 D D

31. Are input values reasonable-and correctly_ applied? 181 D D

32. Are design input sources approved? 181 D D

33. Does the calculation reference the latest revision of the design input source? 181 D D

34. Were all applicable plant operating modes considered? 181 D D ASSUMPTIONS

35. Are assumptions reasonable/appropriate to the objective? 181 D D

36. Is adequate justification/basis for all assumptions provided? 181 D D

37. Are any engineering judgments used? D [gl D

38. Are engineering judgments clearly identified as such? D D 181

39. If engineering judgments are utilized as design inputs, are they reasonable and can they be quantified or substantiated by reference to site or industry standards, D D 181 engineering principles, physical laws or other appropriate criteria?

Page 2 of 4

CALC

~ ENERCON                                               Attachment 1              NO.

NEE-323-CALC-001 Excellence-Every project. Every day. CALCULATION PREPARATION CHECKLIST REV. 00 CHECKLIST ITEMS 1 YES NO N/A METHODOLOGY

40. Is the methodology used in the calculation described or implied in the plant's licensing basis? D 1'81 D

41. If the methodology used differs from that described in the plant's licensing basis, has the appropriate license document change notice been initiated? D D 1:8:1

42. Is the methodology used consistent with the stated objective? [gl D D

43. Is the methodology used appropriate when considering the quality classification of the calculation and intended use of the results? 1:8:1 D D BODY OF CALCULATION

44. Are equations used in the calculation consistent with recognized engineering practice and the plant's design and license basis? 1:8:1 D D

45. Is there reasonable justification provided for the use of equations not in common use? D D 1:8:1

46. Are the mathematical operations performed properly and documented in a logical fashion?

1:8:1 D D

47. Is the math performed correctly? 1:8:1 D D

48. Have adjustment factors, uncertainties and empirical correlations used in the analysis been correctly applied? 1:8:1 D D

                                                                                                         ~, .

49. . Has proper consideration beaeri* given to results that may be overly sensitive to very L small changes in input? 1'81 D D SOFTWARE/COMPUTER CODES

50. Are computer codes or software languages used in the preparation of the calculation? D [gl D

51. Have the requirements of CSP 3.09 for use of computer codes or software languages, including verification of accuracy and applicability been met? D D 1:8:1

52. Are the codes properly identified along with source vendor, organization, and revision level? D D 1:8:1

53. Is the computer code applicable for the analysis being performed? D D 1:8:1

54. If applicable, does the computer model adequately consider actual plant conditions? D D 1:8:1

55. Are the inputs to the computer code clearly identified and consistent with the inputs and assumptions documented in the calculation? D D 1:8:1

56. Is the computer output clearly identified? D D 1:8:1

57. Does the computer output clearly identify the appropriate units? D D [gl Page 3 of4

CALC lgENERCON Attachment 1 NEE-323-CALC-001 NO. Excellence-;very project. Every day. CALCULATION PREPARATION CHECKLIST REV. 00 CHECKLIST ITEMS 1 YES NO N/A

58. Are the computer outputs reasonable when compared to the inputs and what was expected? D D 18]

59. Was the computer output reviewed for ERROR or WARNING messages that could invalidate the results? D D 18]

RES ULTS AND CONCLUSIONS

60. Is adequate acceptance criteria specified? 18] D D

61. Are the stated acceptance criteria consistent with the purpose of the calculation, and intended use? ~ D D

62. Are the stated acceptance criteria consistent with the plant's design basis, applicable licensing commitments and industry codes, and standards? ~ D D

63. Do the calculation results and conclusions meet the stated acceptance criteria? ~ D D

64. Are the results represented in the proper units with an appropriate tolerance, if applicable? ~ D D

65. Are the calculation results and conclusions reasonable when considered against the stated inputs and objectives?

18] D D

66. Is sufficient conservatism applied to the outputs and conclusions? 18] D D

67. Do the calculation results and conclusions affect any other calculations? D 18] D

68. If so, have the affected calculations been revised? D D ~

69.* Does the calculation contain any conceptual, unconfirmed or open assumptions requiring later confirmation? D 18] D

70. If so, are they properly identified? D D ~

DESIGN REVIEW

71. Have alternate calculation methods been used to verify calculation results? D D 18]

No, a Design Review was performed. Note:

1. Where required, provide clarification/justification for answers to the questions in the space provided below each question. An explanation is required for any questions answered as "No' or "N/A".

Originator: Aaron Holloway 12/12/17 Print Name and Sign Date Page 4 of4

BASES-DAEC EOP BASES DOCUMENT BREAKPOINTS Rev. 14 EOP BREAKPOINTS Page 12 of 14 BREAKPOINTS FOR PRIMARY CONTAINMENT PRESSURE CONTROL Pressure Item of Interest Significance (psig) 53 Primary Containment When PCPL is reached, containment venting (Torus) Pressure Limit (PCPL) is required.

   -21.4   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 (11.15)                           Initiation Limit and torus level is less than 13.5 feet 11.4   Drywell Spray Initiation Above 11.4 psig drywell pressure, drywell (Drywell) Limit (DWSIL) Break      spray initiation is unrestricted by the DWSIL.

Point " , <11 Torus Spray Initiation Start torus sprays prior to 11 psig, if possible. (Torus) Pressure If pressure is exceeded before torus sprays (11 .15) are initiated - initiate them anyway 2 Drywell High Pressure ECCS Initiation, Isolations and RPS defeats (Drywell) Scram Setpoint may be needed, EOP 1 and EOP 2 entry 1 Drywell N2 Makeup Drywell N2 makeup supply isolates if drywell (Drywell) Isolation pressure exceeds 1 psig

EOP3 EOP 4 SECONDARY CONTAINMENT CONTROL RADIOACTIVITY RELEASE CONTROL Fu.! Pool [Jeri-t 1'11 Vant Shatt RN Monitor AIM lem~ N luN ........ , . c l ~ level Are, waer ....... otfsbradlollctlvltyNINHrat:*

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tf"CIIIOOl,jDtff'EftENTW.. ftCICflee-lMOll2225,,tlrCfllCll4A __ 1n,_ ftCICEMl!ftcoo..EII .... IENT IICICII.OCNIIMINENT ~=~~1

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                                                                                                                                                                                                                     'IOftUI CAlWllt.K IWTli Dlf

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er:::.':!~ IC4 (T ..... ., SC.. ,r..~ D D CONTROLLED COPY EDI" 3 - lll(;()NCWIY C()NTl,NIEHT COHlAOI. l!OP 4 - MOIOr,lriCJMT'I' RELl!ASliCOHlltcll

BASES-DAEC EOP BASES DOCUMENT BREAKPOINTS Rev. 14 EOP BREAKPOINTS Page 12 of 14 BREAKPOINTS FOR PRIMARY CONTAINMENT PRESSURE CONTROL Pressure Item of Interest Significance (psig) 53 Primary Containment When PCPL is reached, containment venting {Torus) Pressure Limit (PCPL) is requirea.

  -21.4   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 (11.15)                           Initiation Limit and torus level is less than 13.5 feet 11.4   Drywell Spray Initiation Above 11.4 psig drywell pressure, drywell (Drywell) Limit (DWSIL) Break      spray initiation is unrestricted by the DWSIL.

Point

   <11    Torus Spray Initiation   Start torus sprays prior to .11 psig, if possible.

(Torus) Pressure If pressure is exceeded before torus sprays (11 .15) are initiated - initiate them anyway 2 Drywell High Pressure ECCS Initiation, Isolations and RPS defeats (Drywell) Scram Setpoint may be needed, EOP 1 and EOP 2 entry 1 Drywell N2 Makeup Drywell N2 makeup supply isolates if drywell (Drywell) Isolation pressure exceeds 1 psig

                                                                                                                                                                                                                                                                                   '--I0'_

EOP 2 - PRIMARY CONTAINMENT CONTROL

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EOP3 EOP4 SECONDARY CONTAINMENT CONTROL RADIOACTIVITY RELEASE CONTROL

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ltaltNI.AO,l,DACCENAREA. IOUTH CflO WCIDULl AREA 1111,.r 11111* 111111, 1191'1'-"'- H()ftn+ dlO MOOUlE

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IEOP , - IW:IOollCfMT'I' MUASE CONntOl

IAOP 901 EARTHQUAKE PROBABLE ANNUNCIATORS None PROBABLE INDICATIONS 1C35

-  The amber DESIGN BASIS EARTHQUAKE (DBE) light is ON .
-  The amber OPERATING BASIS EARTHQUAKE (OBE) light is ON .
-  The amber .01 G RECORDERS RUNNING light is ON .
-  The white CONTINUITY light is OFF.
-  The Seismic Wailing Alarm is sounding .
-  Building vibration .

A Cooling Tower Valve House

-  No power indicating light is operable.

IAOP 901 Page 11 of 16 Rev.30

IAOP 901 EARTHQUAKE

*** *** ** ******* ** * **** *** ** ** ** ** ******** ** ** ** ****** ** * **** **INFORMATION **** ******** ****** ** **** *************** ****** *** ******** ** * **

• Earthquake OBE DBE Ground Acceleration 0.06g 0.12g IAOP 901 Page 13 of 16 Rev. 30

AOP 915 SHUTDOWN OUTSIDE CONTROL ROOM SECTION 1 I TRANSFER OF CONTROL TO THE REMOTE SHUTDOWN PANEL CONDITIONAL STATEMENTS IF while performing this procedure: IF Control Room access is regained THEN when directed by the Emergency Response and Recovery Director AND resume control of unaffected personnel are available components from the Control Room AND maintain control of Division II components from 1C388 until operability of Control Room instruments, indications and controls has been verified. NOTE

• Operations personnel evacuate to the Remote Shutdown Panel except: the STA, Shift Communicator, and on-site personnel not on shift evacuate to the TSC.
• The preferred evacuation route to the Remote Shutdown Panel is out the back door df the Control Room, and down the stairs. Emergency lightfng is provided for this path.
• The alternate evacuation route to the Remote Shutdown Panel is out the front doo*r of the Control Room, and down the stairs to access control. Emergency lighting is provided for this path.
• Since fire induced failure in 1C05 could adversely affect manual scram circuits, the initiation of ATWS ARI/RPT provides a redundant and diverse means of control rod insertion.

CAUTION For Control Room evacuation as the result of a fire, transfer of control at panels 1C388, 1C389, 1C390, 1C391, 1C392 is required to be completed within 20 minutes. IAOP 915 Page 4 of 94 Rev. 571

RCS Specffic Activity 3.4.6 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.6 RCS Specific Activity LCO 3.4.6 The specific activ1ty of the reactor coolant shall be limited to DOSE EQUIVALENT I-131 specific activity~ 0.2

                        µCi/gm.

APPLICABILITY: MODE 1. MODES 2 and 3 with any main steam 11ne not isolated. ACTIONS CONDITION REQUIRED ACTION COMPLErION TIME A. Reactor coolant ----*--*-----NOTE----------- specific activity LCO 3.0.4.c is applicable.

           > 0.2 µCi/gm and         ----------------------------
           ~ 2.0 µCi/gm DOSE EQUIVALENT 1-131.        A.1      Determine DOSE          Once per 4 hours EQUIVALENT I-131.

V Mm A.2 Restore DOSE 48 hours EQUIVALENT I-131 to wfth1n limits. B. Required Action and B.l Determine DOSE Once per 4 hours associated Completion EQUIVALENT I-131. Time of Condition A not met. AtID

          .QB                       B.2.1 Isolate all main          12 hours steam 1i nes.

Reactor Coolant specific activity> 2.0 DB

          µCi/gm DOSE EQUIVALENT I-131.

(continued) 2.0 uci/gm chosen as EAL threashold since levels above that activity '\_) directly influence continued plant operation. OAEC 3.4-13 MO 255" I

RCS Operational LEAKAGE 3.4.4 \_) 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4 .4 RCS Operational LEAKAGE LCO 3.4.4 RCS operational LEAKAGE shall be limited to:

a. s 5 gpm unidentified LEAKAG8: *

b. s 25 gpm total LEAKAGE averaged over the previous 24 hour period: and

c. s 2 gpm increase in unidentified LEAKAGE within the previous 24 hour period in MODE 1.

APPLICABILITY: MODES 1. 2. and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Unidentified LEAKAGE A.l Reduce LEAKAGE to 4 hours not within limit. within 1imits. OR ' " Total LEAKAGE not within limit. . . B. Unidentified LEAKAGE B.l Reduce unidentified}}