View source for NG-17-0235, Attachment 1: Duane Arnold Energy Center, License Amendment Request TSCR-166, Redline Markup of NEI 99-01, Revision 6. Part 1 of 2

{{Adams
| number = ML17363A071
| issue date = 12/15/2017
| title = Attachment 1: Duane Arnold Energy Center, License Amendment Request TSCR-166, Redline Markup of NEI 99-01, Revision 6. Part 1 of 2
| author name = 
| author affiliation = NextEra Energy Duane Arnold, LLC
| addressee name = 
| addressee affiliation = NRC/NRR
| docket = 05000331
| license number = 
| contact person = 
| case reference number = NG-17-0235
| package number = ML17363A067
| document type = License-Application for Facility Operating License (Amend/Renewal) DKT 50
| page count = 151
}}

=Text=
{{#Wiki_filter:ENCLOSURE NEXTERA ENERGY DUANE ARNOLD, LLC DUANE ARNOLD ENERGY CENTER LICENSE AMENDMENT REQUEST TSCR-166 

==SUBJECT:==
Adoption of Emergency Action Level Scheme Pursuant to NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors" 1.0 

==SUMMARY==
DESCRIPTION 2.0 DETAILED DESCRIPTION 3.0 TECHNICAL EVALUATION 

==4.0 REGULATORY EVALUATION==

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 Significant Hazards Consideration 4.4 Conclusions 5.0 ENVIRONMENTAL CONSIDERATION 

==6.0 REFERENCES==

Attachment 1 -Redline Markup of NEI 99-01 Revision 6 Attachment 2 -Clean Copy of the Proposed DAEC EAL Scheme Attachment 3 -Deviations and Differences Matrix Attachment 4 -Supporting Technical Information Attachment 5 -DAEC EAL Scheme Wallboards Page 1 of 10 1.0 

==SUMMARY==
DESCRIPTION NextEra Energy Duane Arnold, LLC (NextEra) requests an amendment to the Duane Arnold Energy Center (DAEC) Emergency Plan to adopt the Nuclear Energy lnstitute's (NEl's) revised Emergency Action Level (EAL) scheme described in NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors" (Reference 1 ). Nuclear Energy Institute (NEI) 99-01 provides guidance to nuclear power plant operators for the development of a site-specific emergency classification scheme. The methodology described in this document is consistent with Federal regulations, and related US Nuclear Regulatory Commission (NRG) requirements and guidance.
In particular, this methodology has been endorsed by the NRG (Reference 
: 2) as an acceptable approach to meeting the requirements of 10 CFR 5Q.47(b)(4), related sections of 10 CFR 50, Appendix E, and the associated planning standard evaluation elements of NUREG-0654/
FEMA-REP-1, Rev. 1, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, November 1980. Revision 6 of NEI 99-01 addresses changes recommended by the NRC in a letter to NEI on October 12, 2010, (Reference 
: 7) along with other enhancements identified by the industry during implementation of Revisions 4 and 5 of the guidance.
Enhancements over prior versions of the guidance include: 1. Clarifying numerous EALs that have been typically misinterpreted by the industry in the development of their site-specific EAL scheme. 2. Clarifying the intent of EALs that have been historically misclassified. 
: 3. Incorporating lessons-learned from industry events and NUREG/CR-7154, "Risk Informing Emergency Preparedness Oversight:
Evaluation of Emergency Action Levels -A Pilot Study of Peach Bottom, Surry and Sequoyah." 4. Performing a detailed review of the guidance to re-validate that the EALs are appropriate and are at the necessary emergency classification level based upon 32 years of industry and NRC experience with EAL scheme development and implementation.
Page 2 of 10 2.0 DETAILED DESCRIPTION The proposed changes involve revising DAEC's current EALs to a scheme based on NEI 99-01, Revision 6 (Reference 1). The NRC endorsed NEI 99-01, Revision 6 in March 2013 (Reference 2). Attachment 1 of this evaluation contains a redline markup of the endorsed guidance *document showing all proposed changes. A clean copy of the resulting DAEC EAL scheme is presented in-Attachment 
: 2. This clean copy includes site-specific indications, parameters and information.
Data and information specific to DAEC are supported by references, calculations, procedures and drawings which are included in Attachment 
: 4. An operator aid (wallboard) was created in order to assist in the prompt assessment and classification of events. The wallboard is included as Attachment 5 for information only. Deviations and Differences Matrix A matrix of deviations and differences (Attachment 
: 3) has been developed that provides a tabular format of the Initiating Conditions (I Cs), Mode Applicability, and EALs (with threshold values) in NEI 99-01, Revision 6 alongside the proposed DAEC EALs. This matrix provides a means of assessing the proposed EAL in terms of "Deviations" and "Differences" from the NRG-endorsed guidance.
The proposed EAL changes were evaluated in accordance with applicable regulatory requirements (e.g., 10 CFR 50.54(q) and Appendix E, Section IV.8.1). The evaluation assessed the conformance of the proposed EAL changes to those described in the endorsed guidance of NEI 99-01, Revision 6 to determine if the proposed EAL wording change resulted in "No Change" to the guidance, a "Difference" in the wording provided, or a "Deviation" from the guidance.
Any items considered to be "Differences" or "Deviations" are based on the definitions provided in RIS 2003-18, "Use of NE/ 99-01, Methodology for Development of Emergency Action Levels," and supporting supplements (References 3, 4, and 5). The RIS and supporting supplements were issued to clarify technical positions regarding the revision of EALs. Specifically, the RIS documentation provides clarification on the level of detail licensees need to provide to support proposed changes to EALs. The RIS documents suggest that specific information be included with the EAL revision submittal to help facilitate the review process. The RIS information defines an EAL "Difference" and "Deviation" as follows:
* A "Difference" is an EAL change where the basis scheme guidance (e.g., NUREG, NUMARC, and NEI) differs in wording but agrees in meaning and intent, such that classification of an event would be the same, whether using the basis scheme guidance or the site-specific proposed EAL. Examples of "Differences" include the use of site-specific terminology or administrative reformatting of site-specific EALs.
* A "Deviation" is an EAL change where the basis scheme guidance differs in wording and is altered in meaning or intent, such that classification of the event could be different between the basis scheme guidance and the site-specific proposed EAL. Examples of "Deviations" include the use of altered mode applicability, altering key words or time limits, or changing words of physical reference (protected area, safety-related equipment, etc.). Page 3 of 10 L __ _ Any "Differences" identified between the NEI 99-01, Revision 6 EALs and the proposed EALs being developed by NextEra have been identified and are listed in Attachment 
: 3. Certain global differences were identified in this evaluation and are referenced by the matrix. Examples of global differences include: * "A" series Initiating Conditions (IC) were changed to "R" series to follow current convention.
* To the extent possible, IC and EAL identification numbering has been retained from the station's existing EALs.
* ICs and EALs which contain logical connectors (e.g., AND, OR, EITHER) are capitalized and bolded where appropriate to be consistent with the station's current EAL presentation scheme.
* Some parameters or indications listed in EALs were placed in tables or bulletized lists for ease of operator reference.
* Operating Mode Applicability lists mode numbers (i.e., Modes 1 and 2) versus mode names (i.e., Power Operation, Startup).
* Developer's Notes were deleted
* Defined terms are listed in ALL CAPS, and have been pulled forward into a section preceding the basis of each IC/EAL where the defined terms are used for ease of operator use. NextEra has determined that these "Differences" do not result in a reduction in effectiveness or constitute a change to the intent of the NEI 99-01, Revision 6 EALs, and the proposed EAL scheme is technically complete and consistent with EAL schemes implemented at similarly designed plants. Any EAL (Initiating Condition or Threshold Value) that does not meet the "intent" of the NEI 99-01, Revision 6 guidance, or which may result in an event being classified differently from the guidance, would be identified as a "Deviation." Based on this definition, NextEra is proposing an EAL scheme containing one "Deviation" from the endorsed guidance of NEI 99-01 Revision 6. This proposed "Deviation" is discussed and evaluated below. Proposed Deviation NextEra proposes to adopt the Deviation contained in NRG-approved EPFAQ 2016-02 (Reference 6). This EPFAQ was originated by NEI to clarify an aspect of the definition of VISIBLE DAMAGE as it relates to the Initiating Conditions CA6 and SA9 in order to minimize the potential for an over-classification due to an equipment failure. Based on industry comments and subsequent staff discussions, a proposed revision to these ICs was discussed in the public meeting held on April 4, 2017. Based on comments provided by the industry during the April 4, 2017 public meeting, NRC staff revised the NRC Response portion of the FAQ, including provision of draft replacement I Cs CA6 and SA9 for the NEI 99-01 Revision 6 scheme.
* Page 4 of 10 DAEC proposes to adopt the revised CA6 and SA9 Initiating Conditions, as well as the revised definition of VISIBLE DAMAGE provided by this FAQ. As noted in the NRC Response section of the FAQ: 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 structures.
Consistent with the guidance in Regulatory Issue Summary (RIS) 2003-18, Supplement 2, Use of Nuclear Energy Institute (NEI) 99-01, "Methodology for Development of Emergency Action Levels," Revision 4, dated January 2003, it is reasonable to conclude that the changes proposed to ICs CA6 and SA8, as well as the revised definition of VISIBLE DAMAGE would be considered a Deviation from the formally endorsed guidance of NEI 99-01 Revision 6. 3.0 TECHNICAL EVALUATION NextEra has evaluated the proposed EAL changes considering the requirements of 10 CFR 50.54(q), paragraph (b) of 10 CFR 50.47, "Emergency plans," 10 CFR 50, Appendix E, "Emergency Planning and Preparedness for Production and Utilization Facilities," Regulatory Issue Summary (RIS) 2003-18, "Use of NEI 99-01, Methodology fo,r Development of Emergency Action Levels" (including supporting supplements), and RIS 2005-02, Revision 1, "Clarifying the Process for Making Emergency Plan Changes." The proposed changes to the DAEC EAL scheme contained in this submittal do not reduce the capability to meet the applicable emergency planning requirements established in 10 CFR 50.47 and 10 CFR 50, Appendix E. In addition, by adopting the latest guidance provided in NEI 99-01, Revision 6, DAEC will continue to provide emergency classifications consistent with EAL schemes implemented at similarly designed plants. Page 5 of 10  

==4.0 REGULATORY EVALUATION==

4.1 Applicable Regulatory Requirements/Criteria 10 CFR 50.47, "Emergency plans," sets forth emergency plan requirements for nuclear power plant facilities.
The regulation in 10 CFR 50.47(a)(1 
)(i) states, in part, that: [ . .]no initial operating license for a nuclear power reactor will be issued unless a finding is made by the NRG that there is reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency.
10 CFR 50.47(b) establish_es the standards that the onsite and offsite emergency response plans must meet for NRC staff to make a positive finding that there is reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency.
Planning standard (4) of this section requires that onsite and offsite emergency response plans contain: A standard emergency classification and action level scheme, the bases of which include facility system and effluent parameters, is in use by the nuclear facility licensee, and State and local response plans call for reliance on information provided by facility licensees for determinations of minimum initial offsite response measures.
10 CFR 50.47(b)(4) specifies a standard emergency classification and action level scheme, assuring that implementation methods are relatively consistent throughout the industry for a given reactor and containment design while simultaneously providing an opportunity for a licensee to modify its EAL scheme as necessary to address plant-specific design considerations or preferences.
10 CFR 50, Appendix E, Section IV.B, Assessment Actions, states in subsection 1: The means to be used for determining the magnitude of, and for continually assessing the impact of, the release of radioactive materials shall be described, including emergency action levels that are to be used as criteria for determining the need for notification and participation of local and State agencies, the Commission, and other Federal agencies, and the emergency action levels that are to be used for determining when and what type of protective measures should be considered within and outside the site boundary to protect health and safety. The emergency action levels shall be based on in-plant conditions and instrumentation in addition to onsite and offsite monitoring.
By June 20, 2012, for nuclear power reactor licensees, these action levels must include hostile action that may adversely affect the nuclear power plant. The initial emergency action levels shall be discussed and agreed on by the applicant or licensee and state and local governmental authorities, and approved by the NRG. Thereafter, emergency action levels shall be reviewed with the State and local governmental authorities on an annual basis. Page 6 of 10 By means of a letter dated March 28, 2013 (Reference 2), the NRC completed its review of the draft version of NEI 99-01, Revision 6, dated November 2012, and found it acceptable for use by licensees seeking to upgrade their emergency action levels (EAL) in accordance with Appendix E to Part 50 of Title 10 of the Code of Federal Regulations (10 CFR). This endorsement letter also contained the following admonition related to use of NEI 99-01 Revision 6: " ... this is considered a significant change to the EAL scheme development methodology and licensees seeking to use this guidance in the development of their EAL scheme must adhere to the requirements of 10 CFR Part 50, Appendix E, Section IV.B.2." Accordingly, this change cannot be made by NextEra under the provisions of 10 CFR 50.54(q) and NextEra must submit an application for an amendment to its Emergency Plan and receive NRC approval before implementing the change. 4.2 Precedent The NRC has previously issued numerous license amendments for adopting NEI 99-01 Revision 6 EAL schemes. 4.3 Significant Hazards Consideration In accordance with 10 CFR 50.90, NextEra Energy Duane Arnold, LLC, (NextEra) requests amendment to the Duane Arnold Energy Center (DAEC) Emergency Plan to support the adoption of Emergency Action Level (EAL) schemes based on NEI 99-01, Revision 6, which has been endorsed by the NRC as documented in a letter dated March 28, 2013 (Reference 2). The proposed changes to NextEra's EAL scheme to adopt the guidance provided in NEI 99-01, Revision 6 do not reduce the capability to meet the emergency planning requirements established in 10 CFR 50.47 and 1 O CFR 50, Appendix E. The proposed change does not reduce the functionality, performance, or capability of NextEra's Emergency Response Organization (ERO) to respond in mitigating the consequences of an accident.
All NextEra ERO functions will continue to be performed as required.
The proposed changes have been reviewed considering the applicable requirements of 10 CFR 50.47, 10 CFR 50, Appendix E, and other applicable NRC documents.
NextEra has evaluated the proposed changes to the Duane Arnold Energy Center (DAEC) Emergency Plan and determined that the changes do not involve a Significant Hazards Consideration.
An analysis of the issue of no significant hazards consideration is presented below. Page 7 of 10  
: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response:
No The proposed change does not impact the physical configuration or function of plant structures, systems, or components (SSCs) or the manner in which SSCs are operated, maintained, modified, tested, or inspected.
No actual facility equipment or accident analyses are affected by the proposed changes. The change revises the NextEra Emergency Action Levels to be consistent with the NRC endorsed EAL scheme contained in NEI 99-01, Revision 6, "Methodology for Development of Emergency Action Levels," but does not alter any of the requirements of the Operating License or the Technical Specifications.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 
: 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
Response:
No The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed).
The proposed change does not create any new failure modes for existing equipment or any new limiting single failures.
Additionally, the proposed change does not involve a change in the methods governing normal plant operation, and all safety functions will continue to perform as previously assumed in the accident analyses.
Thus, the proposed change does not adversely affect the design function or operation of any structures, systems, and components important to safety. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed change. The proposed change does not challenge the performance or integrity of any safety-related system. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
Page 8 of 10  
: 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response:
No The margin of safety associated with the acceptance criteria of any accident is unchanged.
The proposed change will have no affect on the availability, operability, or performance of safety-related systems and components.
The proposed change will not adversely affect the operation of plant equipment or the function of equipment assumed in the accident analysis.
The proposed amendment does not involve changes to any safety analyses assumptions, safety limits, or limiting safety system settings.
The changes do not adversely impact plant operating margins or the reliability of equipment credited in the safety analyses.
Therefore, the proposed change does not involve a significant reduction in a margin of safety. Based upon the above analysis, NextEra concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), "Issuance of Amendment," and accordingly, a finding of "no significant hazards consideration" is justified.
4.4 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public. Page 9 of 10 5.0 ENVIRONMENTAL CONSIDERATION NextEra has evaluated the proposed amendment for environmental considerations.
The proposed change is applicable to emergency planning requirements involving the adoption of the NRG-endorsed EAL guidance as described in NEI 99-01, Revision 6, and does not reduce the capability to meet the emergency planning standards of 10 CFR 50.47(b) and the requirements of 10 CFR 50, Appendix E. The proposed change does not involve (i) a significant hazards consideration; (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite; or (iii) a significant increase in the individual or cumulative occupational radiation exposure.
Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Therefore, pursuant to 10 CFR 51.22(b), no environmental.
impact statement or environmental assessment needs to be prepared in connection with the proposed amendment. 

==6.0 REFERENCES==
: 1. NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors, dated November 2012 (ML 12326A789) 
: 2. Letter from Nuclear Regulatory Commission to Susan Perkins-Grew (Nuclear Energy Institute), U.S. Nuclear Regulatory Commission Review and Endorsement of NEI 99-01, Revision 6, November 2012, dated March 28, 2013 (ML 12346A463) 
: 3. NRC Regulatory Issue Summary 2003-18, Use of NEI 99-01, Methodology for Development of Emergency Action Levels Revision 4, Dated January 2003, dated October 8, 2003 (ML032580518) 
: 4. NRC Regulatory Issue Summary 2003-18, Supplement 1, Use of Nuclear Energy Institute (NEI) 99-01, Methodology for Development of Emergency Action Levels Revision 4, Dated January 2003, dated July 13, 2004 (ML041550395) 
: 5. NRC Regulatory Issue Summary 2003-18, Supplement 2, Use of Nuclear Energy Institute (NEI) 99-01, Methodology for Development of Emergency Action Levels Revision 4, Dated January 2003, dated December 12, 2005 (ML051450482) 
: 6. Emergency Preparedness Program Frequently Asked Question (EPFAQ) 2016-002, Clarification of Equipment Damage, (ML 17195A299) 
: 7. Letter from Nuclear Regulatory Commission to Susan Perkins-Grew (Nuclear Energy Institute), Proposed Changes for Consideration to the Nuclear Energy Institute (NEI) Report, NEI 99-01, "Methodology for the Development of Emergency Action Levels," dated October 12, 2010 (ML 102810390)
Page 10 of 10 ATTACHMENT 1 NEXTERA ENERGY DUANE ARNOLD, LLC DUANE ARNOLD ENERGY CENTER LICENSE AMENDMENT REQUEST TSCR-166 REDLINE MARKUP OF NEI 99-01 REVISION 6 302 pages follow \ ' \
NEI 88 01 [Revision 
&] Development of Emergency Action Levels for Non-Passive Reactors November 2012  
[THIS Pf ... GE IS LEFT BLANK INTENTION1A 
... LLY]
NEI 99 01 [Revision 
&] Nuclear Energy Institute Duane Arnold EnergJ Center (DAEC) Emergency Action Levels Technical Bases Document TBD, 2018 November 2012 Nu e l ea r E n e r gJ 1 In s ti lbt l e, 1 776 l Street N. W, S ui te 4 0 0 , WclS hin gte n D. C. (20 2. 7 39. 8 000)
ACKNOWLEDGMENTS This document v1as prepared by the Nuclear Energy Institute (NEI) Emergency Action Level (EAL) Task Force. NEI Chairpersen:
David Young Preparatien Team Larry Baker E>celon Nuclear/Corporate Craig Banner PSEG Nuclear/Salem and Hope Creek Nuclear Generating Stations/USA John Egdorf Dominion Generation/Kewaunee Power Station Jack Lev.ri.s Entergy Nuclear/Corporate C. Kelly Walker Operations Support Services, Inc. Review Team Chris Boone Southern Nuclear/Corporate John Callahan Xcel Energy/Corporate/USA Bill Chausse Enercon Services , Inc. Kent Crocker Progress Energy/Bruns,.vick Nuclear Plant Don Crowl Duke Energy/Corporate Roger Freeman Constellation Energy Nuclear Group/Corporate Walt Lee TVA Nuclear/Corporate Ken Meade FENOC/Corporate Don Mathena NeJctEra Energy/Corporate David Stobaugh EP Consulting , LLC Nick Turner Callav,'ay Plant/STARS Maureen Zawalick Diablo Canyon Pov,rer Plant/STARS NOTICE Neither NEI, nor any of its employees , members, Slrpporting organizations, contractors, or consultants make any \Varranty, expressed or implied, or assume any legal responsibility for the accuracy or completeness of, or assume any liability for damages resulting from any use of, any information apparatus, methods, or process disclosed in this report or that such may not infringe privately ovmed rights. Nuelear Energy b?stih:tle , 17761 Street N. W., Si1i!e 4QQ, Washingten D. C. (2Q2. 739.8(}(}0)
EXECUTIVE 

==SUMMARY==
NEI 99 01 (Revision 
: 6) Jl,Joyember 2012 Federal regulations require that a nuclear pov,rer plant operator develop a scheme for the classification of emergency events and conditions.
This scheme is a fundamental component of an emergency plan in that it provides the defined thresholds that will allow site personnel to rapidly implement a range of pre planned emergency response measures.
An emergency classification scheme also facilitates timely decision making by an Offsite Response Organization (ORO) concerning the implementation of precautionary or protective actions for the public. The purpose of Nuclear Energy Institute (1'1EI) 99 01 is to provide guidance to nuclear power plant operators for the development of a site specific emergency classification scheme. The methodology described in this document is consistent with Federal regulations, and related ug Nuclear Regulatory Commission 
(]'l'RC) requirements and guidance. In particular, this methodology has been endorsed by the NRG as an acceptable approach to meeting the requirements of 10 CFR &sect; 50.47(b)(4), related sections of 10 CFR &sect; 50, AppendiJ( E, and the associated plar~'ling standard evaluation elements ofNUREG 0654/ FEMA. REP 1, Re\'. 1 , Criteria for Preparation andEvaluetion C>}Rediologicel Emergency Response Plens end Prcperedness in Support C>}Nuclear Po,ver Plants, November 1980. NEI 99 01 contains a set of generic Initiating Conditions (ICs), Emergency Action Levels (EA.Ls) and fission product ba.rrier status thresholds.
It also includes supporting techaical basis information, developer notes and recommended classification instructions for users. Users should implement ICs, EA.Ls and thresholds that.are as close as possible to the generic material presented in this document with allovv<ance for changes necessary to address site specific considerations such as plant design, location, terminology, etc. Properly implemented, the guidance in NEI 99 01 v,rj.11 yield a site specific emergency classification scheme with clearly defined and readily observable EA.Ls and thresholds.
Other benefits include the development of a sound basis document, the adoption of industry standard instructions for emergency classification (e.g., transient events , classification of multiple e"''ents , upgrading, downgrading, etc.), and incorporation of features to improve human performance. An emergency classification using this scheme will be appropriate to the risk posed to plant workers and the public , and should be the same as that made by another NEI 99 01 user plant in response to a similar event. The individuals responsible for developing an emergency classification scheme are strongly encouraged to review all applicable NRG requirements and guidance prior to begin..'ling their efforts. Questions concerning this document may be directed to the NEI Emergency Preparedness staff, NEI EAL task force members or submitted to the Emergency Preparedness Frequently Asked Questions process. Finally, unique gtate and local requirements associated 1.vith an emergency classification scheme are not reflected in this guidance.
Incorporation of these requirements may be performed on a case by case basis in conjunction with the appropriate ORO agency. Any such changes 1 ,vill require a review under the applicable sections of 10 CFR 50.
l>lEI 99 01 (Re&#xa5;isiaR e) l>Ja;cember 2012 [THIS PA.GE IS LEFT BLi\NK INTENTIONA.LLY]
11 TABLE OF CONTENTS NE! 99 01 (Revision 
: 6) November 2012 1 BASIS FOR EMERGENCY ACTION LEVELS .................................................................
1 1.1 OPERA TING REACTORS ..................................................................................................
1 1.2 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) .....................................
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 THRESHOLD 
.....................................................................
8 3 DESIGN OF THE DAEC EMERGENCY CLASSIFICATION SCHEME ...........................
11 3.1 ASSIGNMENT OF EMERGENCY CLASSIFICATION LEVELS (ECLS) .............................
11 3.2 TYPES OF INITIATING CONDITIONS AND EMERGENCY ACTION LEVELS ....................
17 3.3 DAEC-SPECIFIC ORGANIZATION AND PRESENTATION OF GENERIC INFORMATION 18 3.4 IC AND EAL MODE APPLICABILITY 
............................................................................
20 4 DAEC SCHEME DEVELOPMENT 
................................................................................
22 4.1 GENERAL DEVELOPMENT PROCESS ............................................................................
22 4.2 CRITICAL CHARACTERISTICS 
......................................................................................
23 4.3 INSTRUMENTATION USED FOR EALS ..........................................................................
25 4.4 EAL/THRESHOLD REFERENCES TO AOP AND EOP SETPOINTS/CRITERIA 
..............
27 5 GUIDANCE ON USING THE DAEC EALS ....................................................................
29 5 .1 GENERAL CONSIDERATIONS 
........................................................................................
29 5.2 CLASSIFICATION METHODOLOGY 
...............................................................................
32 5.3 CLASSIFICATION OF MULTIPLE EVENTS AND CONDITIONS 
........................................
32 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 ...............................................................
34 5.8 CLASSIFICATION OF TRANSIENT CONDITIONS 
............................................................
34 5.9 AFTER-THE-FACT DISCOVERY OF AN EMERGENCY EVENT OR CONDITION 
..............
35 111 NEI 99 01 (Revision 
: 6) *November 2012 5.10 RETRACTION OF AN EMERGENCY DECLARATION 
.......................................................
35 6 ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT ICS/EALS ........................
36 7 COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTION ICS/EALS ....................
70 8 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS ............
112 9 FISSION PRODUCT BARRIER ICS/EALS ****************************************************************
115 10 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS ....... 173 11 SYSTEM MALFUNCTION ICS/EALS *************************************************************************
210 APPENDIX A -ACRONYMS AND ABBREVIATIONS 
........................................................
A-1 APPENDIX B -DEFINITIONS 
*******************************************************************************************
B-1 l V NEI 99 01 (Re11ision 
: 6) No11eFB.ber 2012 DEYElOMENT OF DUANE ARNOLD EMERGENCY ACTION LEVELS FOR NON PASSIVE REACTORS TECHNICAL BASIS DOCUMENT 1 BASIS FOR EMERGENCY ACTION LEVELSREGULATORY BACKGROUND 1.1 OPERATING REACTORS Title 10, Code of Federa l 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 list ed below will aid the reader in understanding the key terminology provided in Section 3.0 of this document.
* 10 CFR &sect; 50.47(a)(l)(i)
* 10 CFR &sect; 50.47(b)(4)
* 10 CFR &sect; 50.54(q)
* 10 CFR &sect; 50.72(a)
* 10 CFR &sect; 50, Appendix E, IV.B, Assessment Actions
* 10 CFR &sect; 50 , Append i x E, IV.C, Activa tion 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, Criteriafor Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, October 1980. [Refer to Appendix 1 , Emergency Action Level Guidelines for Nuclear Power Plants] NUREG-1022, Event Reporting Guidelines 10 CFR &sect; 50. 72 and&sect; 50. 73 Regulatory Guide 1.101, Emergency Response Planning and Preparedness for Nuclear Power Reactors 1 1.2 ~lei 99 01 (RevisioA 
: 6) ~Joyember 2012 The above list is not all inclusive and it is strongly recommended that scheme developers consult v,zith licensing/regulatory compliance personnel to identify and understand all applicable requirements and guidance.
Questions may also be directed to the NEI Emergency Preparedness staff. PER.\1:ANENTLY DEFUELED STATION NEI 99 01 provides guidance for an emergency classification scheme applicable to a permanently defueled station. This is a station that generated spent fuel under a 10 CFR &sect; 50 license, has permanently ceased operations and will store the spent fuel onsite for an extended period of time. The emergency classification levels applicable to this type of station are consistent v,zith the requirements of 10 CFR &sect; 50 and the guidance in J\illREG 065 4 /FEMA. REP 1. In order to relax the emergency plan requirements applicable to an operating station, the owner of a permanently defueled station must demonstrate that no credible event can result in a significant radiological release beyond the site boundary.
It is e>cpected that this verification 1 Nill confirm that the source term and motive force available in the permanently defueled condition are insufficient to warrant classifications of a Site ,t\rea Emergency or General Emergency.
Therefore, the generic Initiating Conditions (ICs) and Emergency A.ction Levels (EALs) applicable to a permanently defueled station may result in either a Notification of Unusual Event (NOUE) or an Alert classification.
The generic ICs and EALs are presented in Appendi>c C, Permanently Dcfaeied Station I-Cs/EALs.
8l.1__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 ,&sect;-50 and the guidance in NUREG 0654/FEMA-REP-l.
The initiating conditions germane to a 10 CFR-&sect; 72.32 emergency plan (as described in NUREG-1567) are subsumed within the classification scheme for a 10 CFR ,&sect;-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 oflC 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 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 2 l>JBI 99 01 (Revision 
: 6) l>lo1remaer 2012 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 -&sect;--72.32 emergency plan are generally consistent with those for a Notification of Unusual Event in a 10 CFR-&sect;--50.47 emergency plan (e.g., to provide assistance ifrequested).
Also , the licensee's Emergency Response Organization (ERO) required for 10 CFR -&sect;--72.32 emergency plan is different than that prescribed for a 10 CFR-&sect;--50.47 emergency plan (e.g., no emergency technical support function).
3  
-h4Ll_NRC ORDER EA-12-051 Ne! 99 01 (Rev i sioa 6) 1'foyember 2012 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 thre e reactors.
While the lo ss of power also impaired the s pent 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 neces sary to ens ure 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 eac h spent fuel pool with reliable lev e l instrumentation to significantly enhance the ab ilit y of key deci sion-mak ers to allocate resources effectively following a beyond design basis event. To thi s en d , th e NRC is s u e d Order EA-12-051, Issuance of Order to Modify Licenses with R egard to Reliabl e 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 , "A ll 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 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 w ith Regard to Reliable Sp e nt 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 existing IC~ AA+/-RA2 , and new ICs AS2 RS2, and AG2RG2. Associated EAL notes , bases and developer notes are also provided.
It is recommended that these EA.Ls be implemented when the enhanced spent fuel pool level instrumentation is available for use. 4 The reg,,late . NE! 99 O I (Revisiea O) ineludin r; preeess that licensees fell ,, . Novombe, >O I> g nen seheme eh e fr te make ehai: . reg,,latien, lieensees are aHges to EALs, is 10 CFR 50 5~0ges te their Olflergeney plan whethef er net i't I rnspens1ele feF e1raluat1* . (q). In aeeeFdanee mith ti.. , . resu ts i d . ' u ng a props d "' tt tliS heensee's detefm* t*n a Fe ttetwn in the effeet'se change and determ* . * . ma !OR tl,e liee . ,veaess of tl,e I A !R!Hg er pner review a,,,1 ' . 11See will either mak ti, P '"'* ',s a result efth apprnval m aeee ...I e e change ef s i. . . e I Aruanee with IO CFR ;,uffilt rt to the NRG .5r~~1: 1:PP:L:l:C:A:B:IL 
:I:T~\~' T:,AA~'4' IN<3El>*ND.&4Albb-~~==:
5: 0:*9 :o:.*DF""'=' 0, ,.D~ ANCED AND 8MALL MO II,
* DULAR RUE'!' n e ga1uanee in this d . ----OR uESIGNS the Uni~ed gtates, epe:~7:e;t pnmmily addresses eemmernia r-::eratlOR plant designs)* 
!,,,.~'.""'a,,eatly defueled, as 0f 2Q;2R(HeleaF pevref FeaetefS in e en FefeFFed te as 7fEI ' fr~,, eF, it may be adapted t d se ealled 1st --and-+/-oo I . , geflOfat I e a->>ans d e *s.S1fieati0n seh01H0 fer "' ~:~n p ant designs) as well o' '. e non passive designs dev1al!0ns from tl,e en ~A a . 'aneed non passive reac . e.elepers of an eme,gene . arull eriteria, attd Ojl!.u:;:r"danee te aeeeent fer the ~~i:a!lt may need to propose l pants. maetenstws and eapab'l't' Fenees m design;3arnm t 1 1 1es, between 2 ru1 ~1'4 e eFS geneFatien Them me signifieant desi . ?ewef plants (ef any en:n ~nd epeFatmg diffeFenees betrn ill S0'1f60 tem,). Fer !;s :::~:!!mall Modular Re.;:r-:(~i:~*r**ial R11elear eemnent is net applie~l e.g., d1ffernnees a e te gMRs. 5 NEJ 99 01 (Revision 
: 6) *November 2012 2 KEY TERMINOLOGY USED IN NEI 99-01DAEC EAL SCHEME There are several key terms that appear throughout the NEI 99 OlEAL 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. Unusual Event Initiating Condition Emergency Action Level (1)
* Operating Mode Applicability
* Notes
* Basis I Emergency Classification Level Alert I SAE Initiating Condition Emergency Action Level (1)
* Operating Mode Applicability
* Notes
* Basis Initiating Condition Emergency Action Level (1)
* Operating Mode Applicability
* Notes
* Basis l GE Initiating Condition Emergency Action Level (1)
* Operating Mode Applicability
* Notes
* 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 estab lished 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 level s, in ascending order of severity, are: Notification of Unusual Event (NOUE) Alert Site Area Emergency (SAE) General Emergency (GE) I 2.1.1 Notification of Unusual Event (NOUE}l-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. + This term is sometimes shortened to Unusual Event (UE) or other similar site speeifie terminology.
The terms l'lotifieation of Unusual Event, NOUE and Unusual Event are used rnterehangeably throughout this doeument 6 NEI 99 01 (Revision 
: 6) November 2012 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 ifrequired , 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 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 P AG 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 2.2 INITIATING CONDITION 
{IC) NE! 99 01 (Revision 
: 6) l>lo11ember 2012 An event or condition that aligns with th e definition of one of the four emergency classification l eve l s by virtue of th e potentia l or actual effects or consequences.
Discussion:
An IC describes an eve nt or condition , the severity or consequences of which meets the definition of an e m ergency classification level. An IC can b e expressed as a con tinuou s , measurable parameter (e.g., RCS l eakage), an event (e.g., an earthquake) or the status of one or more fission product barriers (e.g., lo ss of the RCS banier). Appe ndi x 1 ofNUREG-0654 does not co nt a in example Emergency Action Levels_ (EA.Ls) for eac h ECL, but rather Initiating Conditions (i.e., plant conditions that indicat e that a radiological emerge nc y , or events t hat could l ea d to a radiological e mergency , ha s occ urr e d). NUREG-0654 s t ates that the Initiating Conditions form th e b as is for esta blishm en t by a licensee of the specific plant in strume nt ation readings (as applicable) whic h , i f exceeded, wo uld initiate th e emerge nc y classification.
Thus, it is the specific instrument rea din gs that would b e the EA.Ls. Considerations for the assignment of a particular Initiating Condition to an emergency classification level are discussed in Section 3. 2.3 EMERGENCY ACTION LEVEL {EAL) A pre-determined , site-specific, observable threshold for an Initiating Condition that, whe n met or ex ceed e d , places the plant in a given emergency classific a tion level. Discussion:
EAL stateme nts may utili ze a variety of criteri a includin g instrument readings and status indications; observable events; results of calculations and anal yses; entry into particular procedures; and the occurrence of natural phenomena. 2.4 FISSION PRODUCT BARRIER THRESHOLD A pre-determined , site-specific , ob serva bl e threshold indicatin g th e lo ss or potential los s of a fission product barrier. Discussion:
Fission product b arr i er thr esho ld s repr ese nt thr eats to the defense in depth design concept that precludes the release of radioactive fission product s to the enviro nm ent. This concept relies on multiple phy sica l barri ers, any one of which , if maintained intact , precludes th e r e lea se of s i g nificant amounts of radioactive fission products to the environment.
The primar y fission product barriers are: Fuel Clad Reactor Coolant System (RCS) Containment 
---Upon determination that one or more fission product barrier thr es holds ha ve been exceeded, the combination of barrier loss and/or potential lo ss thresholds is compared to the fission product barrier IC/EAL criteria to determine the appropriate ECL. In some accident sequences, the !Cs and EA.Ls presented in th e Abnormal Radiation Levels/ Radiological Effluent (AR) Recognition Category will be exceeded at the same 8 l>JEI 99 01 (Revision 
: 6) "t-Jovember 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 NEI 99 01 (Revision 
: 6)
* No,*ember 2012 10 NEI 99 01 (ReY i s i on 6) }loYember 2012 3 DESIGN OF THE NEI 99 01 DAEC EMERGENCY CLA SSIFICATIO N S CH E ME 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-&sect;-50. 72. Guidance concerning these reporting requirements, and example events, are provided in NUREG-1022.
Certain events reportable under the provisions of 10 CFR f-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 ohhis 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 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 Calculation Assessment Manual (ODAM) radiologica l release l imits Review of selected Updated Final Safety Analysis Report (UFSAR) accident ana l yses 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 attrib u tes were-are created used by the Revision 6 Preparation Team to aid i n the development ofICs and Emergency Action Levels (EALs). The team decided to include the attributes in this revis i on since theyThe attributes may b e usefu l in briefing and training settings ( e.g., helping an Emergency Director understand why a particular cond i tion is classified as an Alert). It should be stressed that developers not attempt to redefine these attributes or apply them in any fashion that wou l d change the . "d . d. h" d + genenc gm ance contame m t 1socument 
-; + The us e of EGL attributes is at the diseretion ofa lieensee and is not a requirement of the }JRG. Ifa l ieensee ehooses in ineorporate the EGL attril3utes into their sehetlle basis doeurnent , it must 13e YeF)' elear that the NRG staff 11 NEI 99 01 (RevisioR 
: 6) November 2012 . . , r ose In articular , the staff does not coRsider th~ has not endorsed their acce13tability or a1313l_1Cat1on 
~z ::~:i~:n s .. As :result , the use of the attributes as a basis for attribute s tatemeRts to su13ersede the established E JUS-1 f b , t b . t'fl *iRa E 6 L chaRaes is unacce13table. 1 2 The attributes of each EGL are presented belov,r. 13 l>JEI 99 01 (RevisioR 
: 6) l>lovember 2012 3.1.1 Notification of Unusual Event (NOUE) NEJ 99 0 I (Revision 
: 6) J>lo11ember 20 I 2 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 l ocal, 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, including those directed at an Independent Spent Fuel Storage Installation (ISFSI). I 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 l ead to the l oss or potential loss of : multip l e fission product barriers within a relative l y 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 PAG at or beyond the site boundary. (D) A HOSTILE ACTION occurring within the plant PROTECTED AREA. 14 3.1.4 General Emergency (GE) NEI 99 01 (Revision 
: 6) *November 2012 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 mater ia ls 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. 15 3.1.5 Risk-Informed Insights ~!el 99 01 (Revision 
: 6) ~Joyemeer 
: 20) 2 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 I Cs and EALs also considered insights from several site-specific probabilistic safety assessments (PSA also known 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 
'Nater Reactors (PWRs) aoo-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 -&sect;--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 Coo lan t Accident (LOCA) with early containment failure , a Station Blackout l asting longer than the site specificDAEC 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 NEl 99 01 (Revision 
: 6) l>Jo;rember 2012 3.2 TYPES OF INITIATING CONDITIONS AND EMERGENCY ACTION LEVELS The NEI 99-01 methodology makes use of symptom-based, barrier-based and 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 normal, reactor operators will implement procedures to identify the probable cause(s) and take corrective action. Fission product barrier-based I Cs 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 I Cs and EALs cons i der the level of challenge to each individual barrier -potentially lost and lost -and the total number of barriers under challenge.
Event-based I Cs and EALs define a variety of specific occurrences that have potential or actual safety significance.
These include the failure of an automatic reactor scram/trip to shut down the reactor, natural phenomena (e.g., an earthquake), or man-made hazards such as a toxic gas release. 17 3.3 N888 DESIGN DIFFERENCES NEI 99 0 I (Revis i on 6) November 2012 The NEI 99 01 emergency classification scheme accounts for the design differences between P'.VR.s and BWRs by specifying EALs unique to each type of Nuclear Steam Supply System (NSSS). There are also significant design differences among P'.VR NSSSs; therefore, guidance is provided to aid in the development of El\Ls appropriate to different P'.VR NSSS types. 'Nhere necessary, deve l opment g ui dance also addresses unique considerations for advanced non passive reactor designs such as the /\.dvanced Boiling Water Reactor (AB'NR), the Advanced Pressurized 
'.Vater Reactor (APWR) and the Evolutionary Power Reactor (EPR). Developers will need to consider the relevant aspects of their plant's design and operating characteristics v~rhen converting the generic guidance of this document into a site specific classification scheme. The goal is to maintain as much fidelity as possib l e to the intent of generic ICs and EALs within the constraints imposed by the plant design and 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 document; these changes will be reviev,zed for acceptability by the NRG as part of the scheme approval process. The guidance in NEI 99 01 is not applicable to advanced passive light water reactor designs. An Emergency Classification Scheme for this type of plant should be developed in accordance with NEI 07 01, A1ethodoiogy for Development of Emergency Action Levels, Ad 1 ,1anced Passi>,1e Light Water Reactors. ~3.3 DAEC-SPECIFIC ORGANIZATION AND PRESENTATION OF GENERIC INFORMATION 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 I 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 Safet y -Section 10 S -System Malfunction 
-Section 11 PD Permanently Defueled Station Appendi)c 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 Nel 99 01 (Revision 
: 6) 't>foyember 2012 Example Emergency Action Level(s)-Provides examples ofreports and indications that are considered to meet the intent of the IC. Developers should address each e>mmple EAL. If the generic approach to the development of an example EAL cannot be used (e.g., an assumed instrumentation range is not available at the plant), the developer should attempt to specify an alternate means for identifying entry into the IC. For Recognition Category F , the fission product barrier thresholds are presented in tables applicable to BV/Rs 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 Ne! 99 01 (ReYisioA 
: 6) J!>loYemeer 2012 Developer Notes Information that supports the development of the site specific ICs and EALs. This may include clarifications, references, examples, instructions for calculations, etc. Developer notes should not be included in the site's emergency classification scheme basis document.
Developers may elect to include information resulting from a developer note action in a basis section. ECL Assignment A.ttributes Located within the Developer Notes section, specifies the attribute used for assigning the IC to a given EGL. H3.4 IC AND EAL MODE APPLICABILITY The NEI 99 01 DAEC emergency classification scheme was developed recognizing that the applicability of I Cs 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 EALs will come into play to reflect the opening of systems for routine maintenance, the 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 MA TRIX Recognition Category Mode AR C E F H s Power Operations X X X X X Startup X X X X X Hot StUHdby ' ' ' ' . . . Hot Shutdown X X X X X Cold Shutdown X X X X Refueling X X X X Defueled X X X X Permanently De:fueled 20 Power Operations (1): Startup (2): Hot Shutdown (3): Cold Shutdown (4): Refueling (5): Mode Switch in Run NB! 99 01 (Revision 
: 6) Noyember 2012 Mode Switch in Startup/Hot Standby or Refuel (with a ll vessel head closure bolts fully tensioned)
Mode Switch in Shutdown , Average Reactor Coolant Temperature 
>~212 &deg;P (with all vessel head closure bolts fully tensioned)
Mode Switch in Shutdown , Average Reactor Coolant Temperature~ 
~212 &deg;P (with all vessel head closure bolts fully tensioned)
Mode Switch in Shutdown or Refuel ,--aoo (with one or more vessel head closure bolts less than fully tensioned}-Typieal PWR OPERATING 1Vl0DE8 Povrer Operations 
(}):Reactor Power> 5%, Keff:?: 0.99 Startup (2): Reactor Power< 5%, Keff > 0.99 Hot Standby (3): RCS:?: 350 &deg;P, Keff < 0.99 Hot Shutdown (4): 200 &deg;P < RCS < 350 &deg;P, Keff < 0.99 Cold Shutdown (5): RCS< 200 &deg;P, Keff < 0.99 Refueling (6): One or more vessel head closure bolts less than fully tensioned Developers will need to incorporate the mode criteria from unit specific Technical Specifications into their emergency classification scheme. In addition, the scheme must also include the following mode designation specific to NEI 99 01: Defueled (+/-'fone):
All fuel removed from the reactor vessel (i.e., full core offload during refueling or eJctended outage). 21 NEI 99 01 (ReYision 
: 6) No\'ember 2012 4 SITE SPECIFIC SCHEME DE'JELOPMENT GUIDANCEDEVELOPMENT OF THE DAEC EMERGENCY CLASSIFICATION SCHEME This section provides detailed guidance for developing a site specific emergency classification scheme. Conceptually, the approach discussed here min-ors the approach used to prepare emergency operating procedures generic material prepared by reactor vendor ovmers groups is converted by each nuclear power plant into site specific emergency operating procedures. Likewise , the emergency classification scheme developer 
\Vill use the generic guidance in NEI 99 01 to prepare a site specific emergency classification scheme and the associated basis document.
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 v,'ill lead to an inconsistent or incomplete emergency classification scheme that will likely not receive the necessary regulatory approval.
4.1 GENERAL IMPLEMENTATION CUIDANCEDEVELOPMENT PROCESS The guidance in NEI 99 01 is not intended to be applied to plants "as is"; however, developers should attempt to keep their site specific schemes as close to the generic guidance as possible. The goal is to meet the intent of the generic Initiating Conditions (ICs) and Emergency Action Levels (EA.Ls) within the conte>ct of site specific characteristics locale , plant design, operating features, tenninology, etc. Meeting this goal will result in a shorter and less cumbersome NRG review and approval process, closer alignment 1.vith the schemes of other nuclear power plant sites and better positioning to adopt future industry 1.vide scheme enhancements.
When properly developed, the The DAEC ICs and EALs should were developed to be unambiguous and readily assessable.
l.,s discussed in Section 3 , the generic guidance includes ICs and mmmple EA.Ls. It is the intent of this guidance that QQ!h 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 not compatible 1 ,vith a generic IC or El.,L , efforts should be made to identify an alternate IC or EAL. If an IC or EAL includes an e>cplicit reference to a mode dependent technical specification limit that is not applicable to the plant, then that IC and/or EAL need not be included in the site specific scheme. In these cases , developers must provide adequate documentation to justify 1.vhy the IC and/or EAL \Vere not incorporated (i.e., sufficient detail to allow a third party to understand the decision not to incorporate the generic guidance).
Useful acronyms and abbreviations associated with the NEI 99 01DAEC emergency classification scheme are presented in Appendix A, Acronyms and Abbreviations. specific entries may be added if necessary.
22 Nm 99 0 I (Revision 
: 6) ~lovember 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 examples of acceptable modifications to the generic guidance. These may be incorporated depending upon site developer and user preferences.
The ICs within a R~cognition Category may be placed in reverse order for presentation purposes (e.g., start with a General Emergency at the left/top of a user aid, followed by Site Area Emergency, Alert and NOUE). The Initiating Condition numbering may be changed. The first letter of a Recognition Category designation may be changed, as follows, provided the change is carried through for all of the associated IC identifiers.
* R may be used in lieu of A
* M may be used in lieu of S For e)rnmple, the Abnormal Radiation Levels / Radiological Effluent category designator "A" (for A.bnormal) may be changed to " R" (for Radiation).
This means that the associated ICs would be changed to RU 1, RU2, RAJ, etc. The ICs and EALs from Recognition Categories S 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 and EALs for Emergency 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 Notes section is included to assist developers with crafting correct IC and EAL statements.
This material is not required to be in the fina l emergency classification scheme basis document.
4.2 CRITICAL CHARACTERISTICS As discussed above , developers are encouraged to keep their site specific schemes as close to the generic guidance 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, specific scheme mustDAEC include~ 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 NEI 99 01 (Re,*ision 
: 6) NoveFBl3er 
?O 12
* 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 u ser-fr i end l y.
* The scheme facilitates upgrading and downgrading of the emergency classification where necessary.
* The scheme facilitates cla s sification of multiple concurrent events or conditions. 24 1>rn1 99 0 I (Revision 
: 6) l>Jo;cem.ber 2012 4.3 INSTRUMENTATION USED FOREALS 4 .4 Instrumentation referenced in EAL statements should include 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 used for EALs need not be safety related , addressed by a Technical Specification or ODCM/RETS control requirement , nor powered from an emergency power source; however , EAL developers should strive to DAEC incorporate g instrumentation that is reliable and routinely maintained in accordance with site programs and procedures. Alarms refer e nced in EAL statements should beare those that are the most operationally significant for the described event or condition.
Scheme developers should ensure that specified values used 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 of reading may not b e readily differentiated from an instrument failure. Findings and violations related to EAL in s trumentation issues may be located on the NRG *.vebsite. PRESENTATION OF SCHEME INFORMATION TO USERS The US Nuclear Regulatory Commission (NRG) e)cpects licensees to establish and maintain the capability to assess , classify and declare an emergency condition promptly within 15 minutes after the availability of indications to plant operators that an emergency action level has been , or may be , e)cceeded.
When 1 t1t'Fiting an emergency classification procedure and creating r e lat e d user aids , the developer must determine the presentation method(s) that best supports the end users by facilitating accurate and timely emergency classification.
To this end , developers should con s ider the following points. The first users of an emergency cla ss ification procedure ar e the operators in the Control Room. During the allowable classification time period , they may have responsibility to perform other critical tasks , and will 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, wh e n warrant e d , recommend a different emergency classification level to the Technical Support Center (TSC) and/or Emergency Operations Facility (EOF). Emergency Directors in the TSC and/or EOF will have more opportunity to focus on making an emergency classification , and *;,,rill probably have advisors from Operations available to help them. EmergenC)' classification scheme information for end users should be presented in a manner \Vith v v hich licensed operators are most comfortable. Developers will need to v,rork closely with representatives from the Operations and Operations Training Departments to develop readily usable and easily understood classification tools (e.g., a procedure and related user aids). If n e cessary , an alternate method for presenting 25 4.5 NEI 99 0 I (Revision 
: 6) J>lovember 20 I 2 emergency classification scheme information may be developed for use by Emergency Directors and/or Offsite Response Organization personnel.
A wallboard is an acceptable presentation method provided that it contains all the information necessary to make a correct emergency classification.
This information includes the ICs, Operating Mode Applicability criteria, Ei\Ls and Notes. Notes may be kept with each applicable EAL or moved to a common area and referenced; a reference to a Note is acceptable as long as the information is adequately captured on the wallboard and pointed to by each applicable EAL+. Basis information need not be included on a wallboard but it should be readily available to emergency classification decision makers. In some cases, it may be advantageous to develop two v1allboards one for use during power operations, startup and hot conditions, and another for cold shutdown and refueling conditions.
Alternative presentation methods for the Recognition Category F ICs and fission product barrier thresholds are acceptable and include flov,r charts, block diagrams, and checklist type tables. Developers must ensure that the site specific method addresses all possible threshold combinations and classification outcomes shown in the BWR or P'."VR EAL fission product barrier tables. The NRG staff considers the presentation method of the Recognition Category F information to be an important user aid and may request a change to a particular proposed method if, among other reasons, the change is necessary to promote consistency across the industry.
INTEGRATION OF ICs/EALs 'I/ITH PLANT PROCEDURES A rigorous integration of IC and EAL references into plant operating procedures is not recommended.
This approach would greatly increase the administrative controls and workload for maintaining procedures.
On the other hand, performance challenges may occur if recognition of meeting an IC or EAL is based solely on the memory of a licensed operator or an Emergency Director , especially during periods of high stress. Developers should consider placing appropriate visual cues (e.g., a step , note, caution , etc.) in plant procedures alerting the reader/user to consult the site emergency classification procedure. Visual cues could be placed in emergency operating procedures , abnormal operating procedures , alarm response procedures , and normal operating procedures that apply to cold shutdown and refueling modes. As an e>mmple, a step , note or caution could be placed at th e beginning of an RC8 leak abnormal operating procedure that reminds the reader that an emergency classification assessment should be performed. 
+ Where appropriate , the Notes shown in the generie g1:1idanee ty-pieally inel1:1de the event/condition EGL and the d1:1ration time specified in the EAL. Jf developers prefer to have seYeral IGs reference a common J>IOTE on a wallboard display , it i s acceptable to remove the EGL and time criterion and 1:1se a generic statement.
For example , a common NOTE co1:1ld read " The Emergenc;' Director sho1:1ld declare the emergency promptly Hpon determining that the applicable EAL time has been ei(ceeded , or will likel;* be eireeeded." 26 4.6 BASIS DOCUMENT NEI 99 01 (Revision 
: 6) *November 2012 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 development information in a readily accessible format. It can be referred to in training situations and when making an actual emergency classification , if necessary.
The document is also useful for establishing configuration management controls for EP related equipment and e)(plaining an emergency classification to offsite authorities.
The content of the basis document should include, at a minimum, the follov,ring:
a i\. site specific Mode l\pplicability Matri)( and description of operating modes, similar to that presented in section 3.5. a A discussion of the emergency classification and declaration process reflecting the material presented in Section 5. This material may be edited as needed to align with site specific emergency plan and implementing procedure requirements.
a Each Initiating Condition along with the associated EA.Ls or fission product barrier thresholds , Operating Mode i\pplicability , Notes and Basis information.
a A listing of acronyms and defined terms, similar to that presented in Appendices l'~ and B, respectively.
This material may be edited as needed to align with site specific characteristics.
a Any site specific background or technical appendices that the developers believe 1 ,vould be useful in e)(plaining or using elements of the emergency classification scheme. A Basis section should not contain information that could modify the meaning or intent of the associated IC or EAL. Such information should be incorporated within the IC or EAL statements , or as an EAL Note. Information in the Basis should only clarify and inform decision making for an emergency classification.
Basis information should be readily available to be referenced , if necessary, by the Emergency Director.
For e)rnmple, a copy of the basis document could be maintained in the appropriate emergency response facilities.
Because the information in a basis document can affect emergency classification decision making (e.g., the Emergency Director refers to it during an event), the NRG staff e)cpects that changes to the basis document 1 ,vill be evaluated 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 plant'sDAEC AOPs and EOPs. This approach is intended to maintain good alignment between operational diagnoses and emergency classification assessments. Developers should verify that aA ppropriate administrative controls are in place to ensure that a subsequent change to an AOP or EOP is screened to determ i ne if an evaluation pursuant to 10 CFR 50.54(q) is required.
27 4.8 DEVELOPER AND USER FEEDBACK Nel 99 01 (Re,,*isioR 
: 6) }Joyemaer 2012 Questions or comments concerning the material in this document may be directed to the NEI Emergency Preparedness staff, NEI EAL task force members or submitted to the Emergency Preparedness Frequently Asked Questions process. 28 NE! 99 0 l (ReYision 
: 6) NoYember 2012 5 GUIDANCE ON MAKING EMERGENCY CLA&SIFICATIONSUSING 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 estab li sh 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 l eve l. The NRC staff has provided guidance on implementing this requirement in NSIR/DPR-ISG-01 , Interim Staff Guidance, Emergency Planning for Nuclear Pow er Plants. All emergency classification assessments sho uld 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 accomp li shed 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 I Cs 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 wi ll likely exceed , the applicable time. If an ongoing radiological release is detected and the release start time is unknown , it shou ld be assumed that the release duration specified in the IC/EAL has been exceeded, absent data to the contrary.
For EAL thresholds that specify a duration of the off-normal condition, the NRC expects that the emergency declaration process run concurrently with the specified threshold duration.
Once the off-normal condition has existed for the duration specified in the EAL, no further effort on this declaration is necessary-the EAL has been exceeded. Consider as an example, the EAL "fire which is not extinguished within 15 minutes of detection." On receipt of a fire alarm, the plant fire brigade is dispatched to the scene to begin fire suppression efforts.
* If the fire brigade reports that the fire can be extinguished before the specified duration, the emergency declaration is placed on hold while firefighting activities continue.
If the fire brigade is successful in extinguishing the fire within the specified duration from detection, no emergency declaration is warranted based on 29 that EAL. 30 l>JEI 99 01 (Revision 
: 6) l>Joyember 2012 NEI 99 01 (Revision 
: 6) ~lovember 2012
* If the fire is still burning after the specified duration has elapsed, the EAL is exceeded, no further assessment is necessary, and the emergency declaration would be made promptly.
As used here, "promptly" means at the first available opportunity (e.g., if the Shift Manager is receiving an update from the fire brigade at the 15-minute mark, it is expected that the declaration will occur as the next action after the call ends).
* If, for example, the fire brigade notifies the shift supervision 5 minutes after detection that the brigade itself cannot extinguish the fire such that the EAL will be met imminently and cannot be avoided, the NRC would not consider it a violation of the licensee's emergency plan to declare the event before the EAL is met (e.g., the 15-minute duration has elapsed).
While a prompt declaration would be beneficial to public health and safety and is encouraged, it is not required by regulation.
* In all of the above, the fire duration is measured from the time the alarm, indication, or report was first received by the plant operators. Validation or confirmation establishes that the fire started as early as the time of the alaim, indication, or report. A planned work activity that results in an expected event or condition which meets or exceeds an EAL doe s 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-&sect;-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-rninute 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 necessar y 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 NEI 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 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. 31 5.2 CLASSIFICATION METHODOLOGY NB! 99 01 (Revision 
: 6) Jlolovember 20 I 2 To mak e 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 EA L that spec i fies a time duration for the off-normal condition , the "cl ock" for the EAL time duration run s concurrently with the emergency classification process "cloc k." For a full discussion of this timing requirement, refer to NSIR/DPRISG-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 example: If an Alert EAL and a Site Area Emergency EAL are met , whether at one unit or at two different units, a Site Area Emergency should be declared.
There is no "a dditive" effect from multiple EALs meeting the same ECL. For examp l e: If two Alert EALs are met , whether at one unit or at two different units, 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 of NRC Guidance for Emergency Notifications During Quickly Changing Events. 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 l evel 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 appl ic able 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 32  
" NE! 99 01 (ReYision 
: 6) ~Joyember 2012 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 oftime (i.e., a change in the E CL 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 cl a ssification 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 E AL 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.
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 Te rminate 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 RlS 2007-02. 33 5.7 CLASSIFICATION OF SHORT-LIVED EVENTS NEI 99 01 (Revision 
: 6) November 2012 As discussed in Section 3.2 , event-based I Cs and E ALs 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 d e claration.
Examples of such events include a failure of the reactor protection system to automatically scram/trip the reactor follow e d by a successful manual scram/trip or an earthquake.
5.8 CLASSIFICATION OF TRANSIENT CONDITIONS Many of the I Cs 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 b e met for a brief period of tim e (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.
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 A TWS 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. 34 l>Jel 99 01 (Revis i eR e) l>Jevember 2012 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 -&sect;-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 DECLARATlON Guidance on the retraction of an emergency declaration reported to the NRC is discussed in NUREG-1022.
35  
:t>IEI 99 0 I (Revision 
: 6) Deeember 20 I 0 6 ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT ICS/EALS Table AR 1: Ree0geiti0e Categen* "1A .. R" Ieitiatieg Ceeditiee Matrix UNUSUAL EVENT AUlRUl Release of gaseous or liquid radioactivity greater than 2 times the (site specific effluent release controlling document)ODCM limits for 60 minutes or longer. Op. }.fades: All AU2RU2 illJ"PLAN1'rnD loss of water level above irradiated fuel. Op. }./odes: All ALERT Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE. Op. }.fades: All Significant lov,ering of water level above, or damage to, irradiated fuel. Op. }.fades: All AA3RA.3 Radiation levels that impede access to equipment necessary for normal plant operations, cooldovm or shutdovm.
Op. },fades: All 36 8ITEAREA El\iERGENCY A81R81 Release of gaseous radioactivity resulting in offsite dose greater than 1 00 mrem TEDE or 500 mrem thyroid CDE. Op. A/odes: All Spent fuel pool level at (site specific Level 3 description).1.Q ft 8 in (Level 3}. --GENERAL EMERGENCY AGlRGl Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE. Op. 1\1odes: AG2RG2 Spent fuel pool level cannot be restored to at least (site specific Level 3 description)4 0 ft 8 in (Level 3) for 60 minutes or longer. All -
NEI 99 0 I (Revision 
: 6) :t>JoYeA'lber 20 I 2 AU 1 RU1 ECL: Notification of Unu s ual Event Initiating Condition:
R e l ease of gaseous or liquid radioactivity greater than 2 times the specific effluent release controlling doc..1ment)ODAM limits for 60 minutes or longer. Operating Mode Applicability:
All Emergency Action Levels: Exam13le EmeFgeaey Aetioa Le'7els: (1 or 2 or 3) Notes:
* The Emergency Director should declare the Unusual Event promptly upon determining that 60 minutes h as b een excee ded , or will likely be exceeded.
* If an ongoin g release i s detected and the release start time is unknown, assume that the release duration has exceeded 60 minutes.
* If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path , then the effluent mon i tor reading is no longer valid for classification purposes.
Reading on ANY of the fo ll owing effluent radiation monitor.&sect; greater than the reading shown for 2 times the (site specific effluent release controlling document) limits for 60 minutes or longer: Effluent Monitor Classification Thresholds Monitor NOUE Reactor Building ventilation rad monitor (Kaman 3/4, 5/6, 7/8} 1.0E-03 uCi/cc "' I .OE-03 uCi/cc = Turbine Building ventilation rad monitor (Kaman 1/2} 0 <U "' co: Offgas Stack rad monitor (Kaman 9/1 O} 2.0E-01 uC i/cc c., LLRPSF rad monitor (Kaman 12} I .OE-03 uCi/cc GSW rad monitor (RIS-4767}
2.0E+03 CPS ; RHRSW & ESW rad monitor (RM-1997}
8.0E+02 CPS RHRSW & ESW Rugture Disc rad monitor (RM-4268}
I.OE+03 CPS (site specific monitor list and threshold values corresponding to 2 times the controlling document limits) R ading on AN&#xa5;ANY effluent radiation monitor greater than 2 times the alarm setpoint established by a current radioactivity discharge permit for 60 minutes or l onger. ~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. 37 Definitions:
None Basis: " NEI 99 0 I (Revision 
: 6) }lo 1 rember 2012 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 ohime (e.g., an uncontrolled rel ease). It includes any gaseous or liquid radiological release, monitored or monitor e d , including those for which a radioactivity discharge permit is normally prepared. Nuclear power plantsDAEC incorporates design features intended to control the release of radioactive effluents to the env ironment.
Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases.
The occurrence of an extended, uncontrolled radioactive re l ease to the environment is indicative of degradation in these features and/or controls.
Radiolo g ical effl uent EALs are a l so included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions a lon e. The inclusion of both plant cond ition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
C la ssification 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 lon ger valid for classification purposes. Releases shou ld 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 radioac t ivity 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 d etected by samp l e analyses 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+RA l. Develeper Netes: The "site specific effluent release controlling document" is the R-0:diological Effluent Technical Specifications (RETS) or, for plants that have implemented Generic Letter 89 01 \-the + implementation ef Progranimatie Controls fer Radiolegieal Ejfh,1ent Teehnieal Speeifieations in the Administrnti11e Controls Section o:f'the Tcehnirnl Speeifieations and the Relorntion of Procedural Details of RE TS to the Ojfsite Dose Calculation M-flm1al or to the Process Control Program 38 NEI 99 01 (ReYision 
: 6) }Jovember 2012 Offsite Dose Calculation Manual (ODCM). These documents implement regulations related to effluent controls (e.g., 10 CFR Part 20 and 10 CFR Part 50 , AppendiJc I). As appropriate , the RETS or ODCM methodology should be used for establishing the monitor thresholds for this IC. Listed monitors should include the effluent monitors described in the RETS or ODCM. Developers may also consider including installed monitors associated v1ith other potential effluent pathv,rays that are not described in the RETS or ODCM+;?,.
If included , EA.L values for these monitors should be determined using the most applicable dose/release limits presented in the RETS or ODCM. It is recognized that a calculat e d EAL value may be belmv what the monitor can read; in that case, the monitor does not need to be included in the list. Also, some monitors may not be governed by Technical Specifications or other license related related requirements; therefore , it is important that the associated EAL and basis section clearly identify any limitations on the use or availability of these monitors.
Some sit e s may find it advantageous to address gaseous and liquid releases with separate Ei\Ls. Radiation monitor readings should reflect values that correspond to a radiological release exceeding 2 times a release control limit. The controlling document typically describes methodologies for determining effluent radiation monitor setpoints; these methodologies should be used to determine EAL values. In cases *where a methodology is not adequately defined , developers should determine values consistent with effluent control regulations (e.g., 10 CFR Part 20 and 10 CFR Part 50 i\.ppendix I) and related guidance.
For EAL #2 Values in this El .. L should be 2 times the setpoint established by the radioactivity discharge permit to \Varn of a releas e that is not in compliance with the specified limits. Inde1cing the value in this manner ensures consistency betw=een the EAL and the setpoint established by a specific discharge permit. Developers should research radiation monitor design documents or other information sources to ensure that 1) the EAL value being con s idered is v,rithin the usable response and display range of the instrument, and 2) there ar e no automatic features that may render the monitor reading invalid (e.g., an auto purge feature triggered at a particular indication level). It is recognized that the condition describ e d by this IC may result in a radiological effluent value beyond the operating or display range of the installed effluent monitor. In those cases , EAL values should be determined 
\Yith a margin sufficient to ensure that an accurate monitor reading is available. For example, an EAL monitor reading might be set at 90% to 95% of the highest accurate monitor reading. This provision nohvithstanding , if the estimated/calculated monitor reading is greater than approximately 110% of the highest accurate monitor reading , then developers may choose not to include the monitor as an indication and identify an alternate EAL threshold.
Indications from a real time dose praj ection system are not included in the generic EALs. Many licensees do not have this capability. For those that do , the capability may not be within the + This includes consideration of the effluent monitors de s cribed in the site emergency plan section(s) which address the requirements of 10 CFR 50.47(b)(8) and (9). ;i Developers s hould keep in mind the requirement s of IO C F R 50.54(q) and the guidance provided by INPO related to e mergency response equipment when con s idering the addition of other effluent monitors. 39 NE! 99 o J (Revision 
: 6) }lovember 2012
* A licensee may request to include an EA~ using real scope of the plant Tech.'lical Spec1~cat10n:;al, ~r ill be considered on a case by case basis. time dose projection system results, awr<h . . ll
* l,s . . , m are not included m the genenc , L
* Indications from a perimeter m?~1tonn~ 
:~:!: that do these monitors may not be co?trolled Many licensees do not have this capalllhty.
F."' ent or withi~ the SCOjle of the plant Teelm1:*I and maintained to the same level_ as plant , equ_1p:en~ed by environmental or other factors_. , L Specifications.
In addit~on, readmgs i: perimeter monitoring system; approval will be licensee may request to mclude E, L g considered on a case by case basis. EGL Assignment Attributes:
3.1.1.B 40 Nel 99 01 (RevisioR 
: 6) 1-loYemeer 2012 AU2RU2 ECL: Notification of Unusual Event Initiating Condition:
UNPLANNED loss of water level above irradiated fuel. Operating Mode Applicability:
All Enmple Emergency Action Levels: -l-a. UNPLANNED water l eve l 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 THANless than 36 feet and lowering
* WR GEMAC Floodup indication (LI-4541) coming on scale(site specific level indications).
AND b. UNPLANNED rise in area radiation levels as indicated by ANY of the following radiation monitors. 
* (site specific list of area radiation monitors)
Spent Fuel Pool Area, RI-9178
* North Refuel Floor, RI-9163
* 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. REFU E LING PATHWAY: The reactor refueling cavity, spent fuel pool and fuel transfer canal. 41 NEI 99 OJ (Re11ision 
: 6) No11ember 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 l eve l decrease will b e primarily determined by indications from available level instrum enta tion. _ Other sources of l eve l 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 ma y also cause an incr ease in th e radiation levels of a djacent areas that can be detected by monitors in those locations.
42  
}/El 99 0 I (Re\1 ision 6) }I OYember 2012 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) of reactor cavity level coming on span for this instrument is used at DAEC as an indicator of uncontrolled reactor cavity level decrease. DAEC Technical Specifications require a minimum of 36 feet of water in the spent fuel pool 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 1 inch. Symptoms of inventory loss at DAEC include visual observation of decreasing water levels in reactor cavity or spent fuel storage pool, Reactor Building (RB) fuel storage pool radiation monitor or refueling area radiation monitor alarms, observation of a decreasing trend on the spent fuel pool water level indicator, and actuation of the spent fuel pool low water level alarm. To eliminate minor level perturbations from concern, DAEC uses LI-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 AA+/-RA2. Denleper Notes: The "site specific level indications" are those indications that may be used to monitor \vater level in the various portions of the REFUEUNG PATH\VAY.
Specify the mode applicability of a particular indication if it is not available in all modes. The "site specific list of area radiation monitors" should contain those area radiation monitors that 1.vould be expected to have increased readings following a decrease in water level in the site specific REFUELING PATHWAY. In cases where a radiation monitor(s) is not available or would not provide a useful indication, consideration should be given to including alternate indications such as DNPLAN1'mD changes in tank and/or sump levels. Development of the EALs should consider the availability and limitations of mode dependent, or other controlled but temporary, radiation monitors. Specify the mode applicability of a particular monitor if it is not available in all modes. EGL Assignment Attributes:
3.1.1.A and 3.1.1.B 43 NBI 99 0 I (RevisioA 
: 6) ~10 1 ,ember 2012 AA1RA1 ECL: Alert Initiating Condition:
Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 rnrem T E DE or 50 rnrem thyroid CD E. Operating Mode Applicability:
All Emergency Action Levels: Example EmeFgeney Aetion LeYels: (1 or 2 or 3 or 4 ) Notes: * *
* I
* Bal l The E mergency Director should declare th e Alert 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 minutes. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolat e the release path , then the effluent monitor reading is no longer valid for classification purpose s. The pre-calculated effluent monitor values presented in EAL RA 1.1 should only be used for emergency classification assessments until the results from a dos e assessment using actual meteorology are available.
Reading on ANY of the following radiation monitors greater than the reading shown for 15 minutes or longer: Effluent Monitor Classification Thresholds Monitor Alert Reactor Building ventilation rad monitor (Kaman 3/4, 5/6, 7/8) I .OE-02 uCi/cc "' = Turbine Building ventilation rad monitor (Kaman 1/2) I .OE-02 uCi/cc 0 "' C: Offgas Stack rad monitor (Kaman 9/ I 0) 4.5E+Ol uCi/cc c.., LLRPSF rad monitor (Kaman 12) I .OE-02 uCi/cc GSW rad monitor (RlS-4767) 2.0E+04 CPS j RHRSW & ESW rad monitor (RM-1997)
I.OE+04 CPS RHRSW & ESW Ruriture Disc rad monitor (RM-4268) 2.0E+04 CPS 44 (site specific monitor list and threshold values) }I.el 99 0 I (Re11ision 
: 6) }Joyember 2012 se assessment using actual meteoro l ogy indicates doses greater than 10 mrem TEDE ___ or 50 mrem thyroid CDE at or beyond (site specific dose receptor point)SITE B UNDARY. [Preferred]
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 specific dose receptor point)the SITE BOUNDARY for one hour of exposure. Field survey results indicate EITHER of the following at or beyond (site specific dose receptor point)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.
45 Definitions:
NEl 99 0 I (Rev i sion 6) }loYember 2012 SITE BOUNDARY:
That line beyond which the land is neither owned, nor leased, nor otherwise controlled by th e 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.
Re le ases of this magnitude represent an actua l 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 RA 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. Radiological effluent EALs are a lso included to provide a basis for classifying events and conditions that cannot be r eadily or appropr iatel y classified on the basis of plant conditions alone. _ T he 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 P AG of 1 , 000 mrem while the 50 mrem thyroid CDE was estab li shed in consideration of the 1 :5 ratio of the EPA P AG 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 lon ger va lid for classification purposes. ---Esca lation of the emergency classification level would be via IC AS+RS 1. De~1 eloper Notes: :\Vhile this IC may not be met absent challenges to one or more fission product barriers, it provides classification diversity and may be used to classify events that 1.vould not reach the same EGL based on plant status or the fission product matrix alone. For many of the DBA.s analyzed in the Updated Final Safety A.nalysis Report, the discriminator will not be the number of fission product barriers challenged , but rather the amount of radioactivity released to the environment.
The EPA PAGs are e)cpressed in terms of the sum of the effective dose equivalent (EDE) and the committed effective dose equivalent (CEDE), or as the thyroid committed dose equivalent (CDE). For the purpose of these IC/EALs, the dose quantity total effective dose equivalent (TEDE), as defined in 10 CFR &sect; 20, is used in lieu of" ... sum of EDE and CEDE .... ". The EPA PAG guidance provides for the use of adult thyroid dose conversion factors; however , some states have decided to base protective actions on child thyroid CDE. Nuclear power 46 Jl,IEI 99 0 I (ReYisioR 
: 6) No,'ember 2012 plant ICs/EALs need to be consistent with the protective action methodologies employed by the States within their EPZs. The thyroid CDE dose used in the IC and EA.Ls should be adjusted as necessary to align 1 n1ith State protective action decision making criteria.
The "site specific monitor list and threshold values" should be determined with consideration of the following:
* Selection of the appropriate installed gaseous and liquid effluent monitors.
* The effluent monitor readings should con-espond to a dose of 10 mrem TEDE or 50 mrem thyroid CDE at the "site specific dose receptor point" (consistent with the calculation methodology employed) for one hour of e)cposure.
* Monitor readings will be calculated using a set of assumed meteorological data or atmospheric dispersion factors; the data or factors selected for use should be the same as those employed to calculate the monitor readings for ICs ASl and AGl. Acceptable sources of this information include, but are not limited to, the RETS/ODCM and values used in the site's emergency dose assessment methodology.
* The calculation of monitor readings will also require use of an assumed release isotopic mi)c; the selected mix should be the same as that employed to calculate monitor readings for ICs AS 1 and AGL Acceptable sources of this information include, but are not limited to, the RETS/ODCM and values used in the site's emergency dose assessment methodology.
* Depending upon the methodology used to calculate the EAL values, there may be overlap of some values behveen different ICs. Developers will need to address this overlap by adjusting these values in a manner that ensures a logical escalation in the EGL. The " site specific dose receptor point" is the distance(s) and/or locations used by the licensee to distinguish between on site and offsite doses. The selected distance(s) and/or locations should reflect the content of the emergency plan , and the procedural methodology used to determine offsite doses and Protective Action Recommendations.
The variation in selected dose receptor points means there may be some differences in the distance from the release point to the calculated dose point from site to site. Developers should research radiation monitor design documents or other information sources to ensure that 1) the EAL value being considered is within the usable response and display range of the instrument, and 2) there are no automatic featlffes that may render the monitor reading invalid (e.g., an auto purge feature triggered at a particular indication level). It is recognized that the condition described by this IC may result in a radiological effluent value beyond the operating or display range of the installed effluent monitor. In those cases , EAL values should be determined with a margin sufficient to ensure that an accurate monitor reading is available. For mmmple, an EAL monitor reading might be set at 90% to 95% of the highest accurate monitor reading. This provision notwithstanding, if the estimated/calculated monitor reading is greater than approximately 110% of the highest accurate monitor reading, then developers may choose not to include the monitor as an indication and identify an alternate EAL threshold. Although the IC references TEDE, field survey results are generally available only as a " 1 n1hole body" dose rate. For this reason , the field survey EAL specifies a "closed windo1,1,*" survey reading. Indications from a real time dose projection system are not included in the generic EALs. 47 NET 99 OJ (Revision 
: 6) ~lo*1ember 2012 . . e that do the capability may not be w_ithin the Many licensees do not have this c~pabi_hty. FAo~i:::ssee may ;eqaest to include an EA~ usmg real scope of the plant Tech."lical 8peci~catw~
: a{ ~vill be considered on a case by case basis. time dose proj eetiea system results, appro ' . I, A b . . 'Stem are not included in the genenc i i s. Indications from a perimeter m?~itor~ng 
:~ose that do these monitors may not be co_ntrolled Many licensees do not have this capability.
_or nt or "'ithi~ the scope of the plant Techmcal .:.a ffiQilllaiaed te the sa1Be level_ as plaftl' ~quljl;;::'...;,ed i,'y eiwiromnental er ether faeter\1 d'
* dmgs ma) em 0 , , al "'I e Specifications.
In ad itl_on , rea E AI, . ng a perimeter monitoring system; appr " licensee may request to mclude a? i i us1 . d d on a case by case basis. cons1 ere EGL A.ssignment Attributes:
3.1.2.C 48 ECL: Alert "NEI 99 01 (ReYision 
: 6) ~loYeniber 2012 AA2RA2 Initiating Condition:
Significant lowering of water level above , or damage to , irradiated fuel. Operating Mode Applicability:
All Emergency Action Levels: Example Emergency Action Levels: (1 or 2 or 3) Uncovery of irradiated fuel in the REFUELING PATHWAY. Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by AN:Y of the following radiation monitors::Hi Rad alarm for 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 Reading greater than 5 R/hr on AN-&#xa5;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-9184B RA2.3 (site specific listing of radiation monitors , and the associated readings, setpoints and/or alarms) L wering of spent fuel pool level to (site specific Level 2 value). [See Developer IVotes]25.
l 7 ft. 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 ,Votes). 49 1'JEl 99 0 I (Re*,ision 
: 6) 1'Joyember 2012 These events present radiological safety challenges to plant personnel and are precursors to a re l ease 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 con s idered valid alarms for the purpose of comparison to these EALs. 50 NEI 99 0 I (Revision 
: 6) NoYember 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. E scalation of the emergency would be based on either Recognition Category A-R or C I Cs. EAL RA2.1 This EAL escalates from Am-RU2 in that the loss of level , 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.
51 NE! 99 0 I (Re*,ision 
: 6) NoYemser 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 abov e irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes. EAL RA2.2 This EAL addresses a release of radioactive material caused by mechanical damage to inadiated fuel. Damaging events m a y include the dropping , bwnping or binding of an assembly, or dropping a heavy load onto an assembly.
A 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. EAL RA2.3 Spent fuel pool water level at this value is within the low er 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 wou ld be via I Cs AS+--RS 1 or ~RS2(see AS2 De'llcloper Notes). DevelepeF Notes: For EAL #1 Depending upon the availability and range of instrumentation , this El'.,L may include specific readings indicative of fuel uncovery; consider water and radiation level readings.
Specify the mode applicability of a particular indication if it is not available in all modes. For EAL #2 The "site specific listing of radiation monitors, and the associated readings, setpoints and/or alarms" should contain those radiation monitors that could be used to identify damage to an in-adiated fuel assembly (e.g., confirmatory of a release of fission product gases from irradiated For EALs # 1 and #2 Developers should research radiation monitor design documents or other information sources to ensure that l) the EAL value being considered is v,ithin the usable response and display 52 NEI 99 0 l (Revision 
: 6) No*,cefflber 2012 range of the instrument , and 2) there are no automatic features that may render the monitor reading invalid (e.g., an auto purge feature triggered at a particalar indication level). It is recognized that th e condition described by this IC may result in a radiation value beyond the operating or display range of the installed radiation monitor. In those cases , EAL values should be d e termined with a margin sufficient to e nsure that an accurate monitor reading is available. For example , an EAL monitor reading might be set at 90% to 95% of the highest accurate monitor reading. This provision notwith s tanding , if the estimated/calculated monitor reading is greater than approximately 110% of th e highest accurate monitor reading , then developers may choos e not to includ e the monitor as an indication and identify an alternate EAL threshold.
To further promote accurate classification , developers should consider if some combination of monitors could be specified in the EAL to build in m1 appropriate level of corroboration between monitor readings into th e cla ss ification ass e ssm e nt. Development of the EALs should also consider the availability and limitations of mode dependent , or other controll e d but temporary , radiation monitors.
Specify the mode applicability of a particular monitor if it is not available in all modes. For EAL #3 In accordance with the discussion in Section 1.4 , NRG Order EA 12 051 , it is recommended that this EAL be implemented 1.vhen the enhanced spent fuel pool level instrumentation is available for use. The " site specific Level 2 value" is usually the spent fuel pool level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck. This site specific level is determin e d in accordance with NRG Order EA 12 051 and l'ffil 12 02 , and applicable owner's group guidance.
D e v e lop e rs should modify the EAL and/or Basis s ec tion to reflect any site specific constraints or limitations associated vt'ith the desi g n or operation of instrumentation used to dete1mine the Level 2 value. EGL Assignment Attributes:
3.1.2.B and 3.1.2.C 53 ECL: Alert l'Jel 99 0 I (Revision 
: 6) l'Jovember 2012 AA3RA3 Initiating Condition:
Radiation levels that impede access to equipment necessary for normal plant operations, cooldown or shutdown.
Operating Mode Applicability:
All Emergency Action Levels: Example Emergeney Aetian Levels: (1 or 2) Note: If the equipment in the listed room or area was already inop era ble or out-of-service befo re the eve nt occurred , then no emergency classification is warranted.
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)
An UNPLANNED eve nt results in radiation levels that prohibit or impede access to aey ANY of the following plant rooms or areas: BUILDING ROOM MODE Reactor Building HPCIRoom 1, 2, or 3 Reactor Building RCIC Room l, 2, or 3 Reactor Building SE or NW Corner Rooms 3 or 4 Reactor Buildin2:
Pumo House ESW / RHRSW Pumo Room 3 or 4 _(site specific list of plant rooms or areas with entry related mode applicability identified)
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 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.
54 NEl 99 0 I (Re11ision 
: 6) "f!>lovember 2012 For EAL RA3.2, an Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect at the time of the elevated radiation levels. The emergency classification is not contingent upon whether entry is actua ll y necessary at the time of the increased radiation levels. Access should be considered as impeded if extraordinary measures 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 extension in dose limits beyond normal administrative limits). An emergency declaration is not warranted if any of the following 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 during the operating mode in effect at the time of the elevated radiation levels). For example , the plant is in Mode 1 when the radiation increase occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4.
* The increased radiation levels are a result of a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area ( e.g., radiography, spent filter or resin transfer, etc.).
* The action for which room/area entry is required is of an administrative or record keeping nature (e.g., normal rounds or routine inspections).
* The access control measures are of a conservative or precautionary nature, and would not actually prevent or impede a required action. The list of plant rooms or areas in EAL RA3.2 was generated from a step-by-step review ofIPOI-3, Power Operations (35% -100% Rated Power) and IPOI-4, Shutdown.
Escalation of the emergency classification l evel would be via Recognition Category AR , C or F ICs. DeYeloper Notes: EAL#l The value of l 5mR/hr is derived from the GDC 19 value of 5 rem in 3 0 days with adjustment for e)(pected occupancy times. The "other site specific areas/rooms" should include any areas or rooms requiring continuous occupancy to maintain normal plant operation, or to perform a normal cooldovm and shutdown.
EAL#2 The "site specific list of plant rooms or areas v,ith entry related mode applicability identified" should specify thos e 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.
Do not include rooms or areas in which actions of a contingent or emergency nature would be performed. (e.g., an action to address an off normal or emergency condition such as emergency repairs, corrective measures or emergency operations).
In addition , the list should specify the plant mode(s) during i.vhich entry would be required for each room or a-rea. 55 J>lel 99 0 I (Revision 
: 6) November 2012 The list should not include rooms or areas for which entry is required solely to perform actions of an administrative or record keeping nature (e.g., normal rounds or routine inspections). If the equipment in the listed room or area was already inoperable, or out of service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event. Rooms and areas listed in EAL # 1 do not need to be included in EAL #2, including the Control Room. EGL Assignment Attributes:
3.1.2.C 56 ECL: Site Area Emergency NEI 99 0 J (Revision 
: 6) ~loYember 20 J 2 AS1RS1 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: Example Emergeney AetieB Levels: (1 or 2 or 3) Notes:
* The Emergency Director should declare the Site Area Emergency 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 minutes.
* If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path , then the effl uent monitor reading is no longer valid for classification purposes.
I
* The pre-calculated effluent monitor values presented in EAL RS 1.1 should only be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
R&sect;l.1 Reading on ANY of the following radiation monito rs greater than the reading shown for 15 minutes or longer: Effluent Monitor Classification Thresholds Monitor SAE Reactor Bui ldin g venti lation rad monitor (Kaman 3/4, 5/6, 7/8) l .OE-0 I uCi/cc "' = Turbine Building venti l ation rad monitor (Kaman 1/2) 1.0 E-01 uCi/cc 0 "' C: Offgas Stack rad monitor (Kaman 9/10) 4.5E+02 uCi/cc LLRPSF rad monitor (Kaman 12) I .OE-01 uCi/cc 57 I NEl 99 0 I (RevisioA 
: 6) }loYember 2012 J---+("'-'Sil-fite-spR<eF>fc:;+ifi=c-+1mAio~n~itu:o~r+li 
.... s+-t A-iaB:Aid:i--tuh~l:f..,es.;.i:hu:o..i.ld=i-&#xa5;v~a1~ul4",e=s) 2 R 1.3 Dose assessment using actual meteorology indicates doses greater than 100 mrem TEDE or 500 mrem thyroid CDE at or beyond (site specific dose receptor point)the SITE BOUNDARY. 
[Preferred]
Field survey results indicate EITHER of the following at or beyond (site specific dose receptor point)the SITE BOUNDARY:
* Closed window dose rates greater than 100 mR/hr expected to continue for 60_-minutes or longer.
* Analyses of field survey san1ples indicate thyroid CDE greater than 500 mrem for one hour of inhalation.
58 59 NEI 99 0 I (Revision 
: 6) }loYemeer 2012 Definitions: 
}JE[ 99 0 I (ReYisioR 
: 6) NoYel'l'lber 2012 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 rad i oactivity 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 proi ection 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 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 estab lish ed._ If the effluent flow past an effl uent 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 RSl.3. Esca lation of the emergency classification l evel wo uld be via IC AGl-RG 1. DevelepeF Netes: While this IC may not be met absent chall e ng e s 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 plant status or the fission product matri>( alone. For many of the DBAs analyzed in the Updated Final Safety Analysis Report , the discriminator will not be the number of fission product barriers challenged , but rather the amount of radioactivity released to the environment.
The EPA Pl.cGs are e)(pressed in terms of the sum of the effective dose equivalent (EDE) and the committed effective dose equivalent (CEDE), or as the thyroid committed dose equivalent (CDE). For the prn=pose of these IC/EALs , the dose quantity total effective dose equivalent (TEDE), as defined in 10 CFR &sect; 20, is used in lieu of" ... sum of EDE and CEDE .... ". The EPA PAG guidance provides for the use of adult thyroid dose conversion factors; hmvever , some stat es hav e d e cided to ba s e protectiv e actions on child thyroid CDE. Nuclear power 60 NEI 99 0 I (Revision 
: 6) :t-lovember 2012 plant ICs/EALs need to be consistent with the protective action methodologies employed by the States within their EPZs. The thyroid CDE dose used in the IC and EALs should be adjusted as necessary to align with State protective action decision making criteria. The "site specific monitor list and threshold values" should be determined with consideration of the following:
* Selection of the appropriate installed gaseous effluent monitors.
* The effluent monitor readings should con-espond to a dose of 100 mrem TEDE or 500 mrem thyroid CDE at the "site specific dose receptor point" (consistent 1.vith the calculation methodology employed) for one hour of e)cposure.
* Monitor readings \Vill be calculated using a set of assumed meteorological data or atmospheric dispersion factors; the data or factors selected for use should be the same as those employed to calculate the monitor readings for ICs AA.I and AGI. Acceptable sources of this information include, but are not limited to, the RETS/ODCM and values used in the site's emergency dose assessment methodology.
* The calculation of monitor readings will also require use of an assumed release isotopic mix; the selected mhc should be the same as that employed to calculate monitor readings for ICs AAl and AGl. Acceptable sources of this information include , but are not limited to , the RETS/ODCM and values used in the site's emergency dose assessment methodology.
* Depending upon the methodology used to calculate the EAL values, there may be overlap of some values behveen different ICs. Developers 1.vill need to address this overlap by adjusting these values in a manner that ensures a logical escalation in the EGL. The "site specific dose receptor point" is the distance(s) and/or locations used by the licensee to distinguish betw een on site and offsite doses. The selected distance(s) and/or locations should reflect the content of the emergency plan, and the procedural methodology used to determine offsite doses and Protective Action Recommendations.
The variation in selected dose receptor points means there may be some differences in the distance from the release point to the calculated dose point from site to site. Developers should research radiation monitor design documents or other information sources to ensure that 1) the EAL value being considered is within the usable response and display range of the instrument, and 2) there are no automatic features that may render the monitor reading invalid (e.g., an auto purge feature triggered at a particular indication level). It is recognized that the condition described by this IC may result in a radiological effluent value beyond the operating or display range of the installed effluent monitor. In those cases, EAL values should be determined 1.vith a margin sufficient to ensure that an accurate monitor reading is available.
For example, an EAL monitor reading might be set at 90% to 95% of the highest accurate monitor reading. This provision notwithstanding , if the estimated/calculated monitor reading is greater than apprmcimately 110% of the highest accurate monitor reading , then developers may choose not to include the monitor as an indication and identify an alternate EAL threshold.
Although the IC references TEDE, field survey results are generally available only as a "who le body" dose rate. For this reason, the field survey EAL specifies a "c losed window" survey reading. 61 NE! 99 0 l (Revision 
: 6) November 2012 Indications from a real time dose projection system are not included in the generic EALs. Many licensees do not have this capability. For those that do , the capability may not be v,ithin the scope of the plant Technical Specifications.
A licensee may request to include an EAL using real time dose projection system results; approval 1.vill be considered on a ease by ease basis. Indications from a perimeter monitoring system are not included in the generic EALs. Many licensees do not have this capability. For those that do, these monitors may not be controlled and maintained to the same level as plant equipment, or \Vithin the scope of the plant Technical Specifications.
In addition , readings may be influenced by environmental or other factors. A licensee may request to include an EAL using a perimeter monitoring system; approval 1.vill be considered on a case by case basis. EGL Assignment Attributes
: 3.1.3.C 62 NEI 99 0 I (Revision 
: 6) }loYember 2012 AS2RS2 [See Developer Notes] ECL: Site Area Emergency Initiating Condition:
Spent fuel pool level at (site specific Level 3 description) 16.36 fee t. Operating Mode Applicability:
All Example Emergency Action Levels: R 2.1 Lowering of spent fuel pool level to 16.36 feet.(site specific Level 3 value). Definitions:
None Basis: This IC addresses a significant loss of s pent 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 1 or -AG+/-RG2. Develeper Netes: In accordance with the discussion in Section 1.4 , NRG Order EA 12 051 , it is recommended that this IC and EAL be implemented when the enhanced sp e nt fuel pool level instrumentation is available for *.ise. The "site specific Level 3 value" is usually that spent fuel pool level 1.vhere fuel r e mains covered and actions to implem e nt make up 1.vat e r addition should no longer be deferred.
This site sp e cific lev e l is d e t e nnin e d in accordanc e with NRG Order EA 12 051 and NEI 12 02 , and applicable ovmer's group guidance. Developer s should modify the EAL and/or Basis s e ction to reflect any site specific constraints or limitations associated with the design or operation of instrum e ntation used to determine the Level 3 &#xa5;alue-: EGL Assignment Attributes
: 3.1.3.B 63 NEI 99 0 l (Re:visioB e) ~loYember 2012 AR G1 ECL: General Emergency Initiating Condition:
Release of gaseo u s 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 Emergeeey A .. etien Levels: (1 or 2 or 3) Notes:
* The Emergency Director shou ld declare the General Emerge nc y promptly upon determining that the applica ble time has been exceede d , or will likel y be exceeded.
* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
* If the effluent flow past an effl uent monitor i s known to have stopped due to actions to isolate the release path , then the effluent monitor reading is no longer valid for clas sifica tion purposes.
I
* The pre-calculated effluent monitor values presented in EAL RG 1.1 should only be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
Rfil l Reading on ANY of the following radiation monitors greater than the reading shown for 15 minutes or longer: Effluent Monitor Classification Thresholds Monitor GE "' Reactor Building ventilation rad monitor (Kaman 3/4, 5/6, 7/8) I .OE+OO uCi/cc :::, 0 Turbine Building ventilation rad monitor (Kaman 1/2) l .OE+OO uCi/cc Q) "' Cl! C!) Offgas Stack rad monitor (Kaman 9/10) 4.5E+03 uCi/cc _(site specific monitor list and threshold values) Dose assessment using actual meteorology indicates doses greater than 1 , 000 mrem TEDE or 5 , 000 mrem thyroid CDE at or beyond (site specific dose receptor point)the SITE BOUNDARY. [Preferred]
Field survey results indicate EITHER of the following at or beyond (site specific dose receptor point)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.
64 65 ~lei 99 01 (ReYision 
: 6) ~loYemeer 2012 D e fin it ion s: NE! 99 0 I (RevisioA 
: 6) }fo , ,cemeer 2012 SITE BOUN D ARY: That line beyond wh i ch the l and is neither owned, nor l ease d , nor ot h erwise co nt ro ll e d b y th e l icensee. 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 relea ses. Rel eases of this magnitude will require implementation of protective actions for the public. T h is IC is modified by a note that EAL RG 1.1 is only assessed for emergency class i ficat i on u n t i l a qua li fied dose assessor i s performing assessments using dose pro j ection software in corporat in g actual meteoro l ogical data and current radiologica l conditions. However, if Kama n monitor readings are s u stained for 15 minutes or longer and the required MIDAS dose assessments cannot be completed within this p e riod , then the declaration can be made using Kaman rea di ngs PROVIDED the readings are not from an i so l ated 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 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, fie l d survey results may be uti l ized to assess this IC using EAL RG l.3. Deve l epeF Netes: The effluent ICs/EALs are included to provide a basis for classifying events that cannot be readi l y classified on the basis of plant conditions alone. The inclusion of both types ofICs/EALs more fully addresses the spectrum of possible events and accidents.
Whi l e t his IC may not be met absent challenges to multip l e fission product b arr i ers, it provides class i fication diversity and may be used to classify events that 1.vould not reach the same EGL based on p l ant status or the fission product matrix alone. For many of the DBAs analyzed in the Updated Final Safety Analysis Report, t he discriminator 1.v ill not be the nu m b er of fiss i on product barr i ers challenged , but rather the amount of radioactiv i ty released to the e n v i ronment. The EPA PAGs are expressed in terms of the sum of the effective dose equiva l ent (EDE) and the committed effective dose equivalent (CEDE), or as the thyroid committed d ose equiva l ent (CDE). For the plupose of these IC/EALs, the dose quantity total effective dose equivalent (TEDE), as defined in 10 CFR &sect; 20, is used in lieu of" ... sum of EDE and CEDE .... ". 66 NEI 99 0 I (RevisiOfl e) 1'1oYember 2012 The EPA PAG guidance provides for the use of adult thyroid dose conversion factors; however, some states have decided to base protective actions on child thyroid CDE. Nuclear power plant ICs/EALs need to be consistent with the protective action methodologies employed by the States within their EPZs. The thyroid CDE dose used in the IC and EALs should be adjusted as necessary to align with State protective action decision making criteria.
The "site specific monitor list and threshold values" should be determined with consideration of the following:
* Selection of the appropriate installed gaseous effluent monitors.
* The effluent monitor readings should correspond to a dose of 1 , 000 mrem TEDE or 5,000 mrem thyroid CDE at the "site specific dose receptor point" (consistent with the calculation methodology employed) for one hour of exposure.
* Monitor readings will be calculated using a set of assumed meteorological data or atmospheric dispersion factors; the data or factors selected for us e should be the same as those employed to calculate the monitor readings for ICs ,.'\.Al and AS 1. Acceptable sources of this information include, but are not limited to, the RETS/ODCM and values used in the site's emergency dose assessment methodology.
* The calculation of monitor readings will also require use of an assumed release isotopic mix; the selected miJ( should be the same as that employed to calculate monitor readings for ICs i'nt\l and i\.Sl. Acceptable sources of this information include , but are not limited to, the RETS/ODCM and values used in the site's emergency dose assessment methodology.
* Depending upon the methodology used to calculate the EAL values, there may be overlap of some values between different ICs. Developers will need to address this overlap by adjusting these values in a manner that ensures a logical escalation in the EGL. The "site specific dose receptor point" is the distance(s) and/or locations used by the licensee to distinguish between on site and offsite doses. The selected distance(s) and/or locations should reflect the content of the emergency plan, and procedural methodology used to determine offsite doses and Protective Action Recommendations.
The variation in selected dose receptor points means there may be some differences in the distance from the release point to the calculated dose point from site to site. Developers should research radiation monitor design documents or other information sources to ensure that 1) the EAL value being considered is 1.vithin the usable response and display range of the instrument, and 2) there are no automatic features that may render the monitor reading invalid (e.g., an auto purge feature triggered at a particular indication level). It is recognized that the condition described by this IC may result in a radiological effluent value beyond the operating or display range of the installed effluent monitor. In those cases, EAL values should be detem1ined with a margin sufficient to ensure that an accurate monitor reading is available.
For example, an EAL monitor reading might be set at 90% to 95% of the highest accurate monitor reading. This provision notwithstanding , if the estimated/calculated monitor reading is greater than apprm(imately 110% of the highest accurate monitor reading, then developers may choose not to include the monitor as an indication and identify an alternate EAL threshold.
Although the IC references TEDE , field survey results are generally available only as a "whole body" dose rate. For this reason , the field survey EAL specifies a "closed window" survey reading. 67 NEI 99 0 l (Revi s ion 6) _IR<lieatioes from* real tiR>e dose . . ~''""""" WIJ Many licensees ao not ha, *e this c . . prOJ ect10n system are not incluaea in the ene . " seope of the plaet Teehni;al Sp e ei-:::i1;z~
F .. ~_those -do, the eapability may i!1 !Je";!tH-,:,...._tshe time aose projection system results; appro\;al, ~vtl~e:see m~x request to incluae an EAL using real e consitterea on a case hv case h . l d" * ' v!l5!S . . n ications from a p e rim e ter monito . , . ~:'y he..,_sees do oot have this capability ~=:'9 are eot melud e d ie the geeerie EAL,. Sp e:i:::::d :: '.;;d~:: , :::i::::. ~~eel , o~~!';ft i:::::~~';:~;:.:~:::.~:i"lled hcen_see may request to inclua e an Et, LY . mflu e~cea by environmental or other factors t, eoes d d '' usmg a p e nmet e * * * '* l ere on a case by case basis.r momtonng system; approval will be EGL Assignment Attributes
: 3.1.4.C 68 NEI 99 Q 1 (Re&#xa5;ision 
: 6) ~loYember 
?Ql2 AG2RG2 [See Developer Notes] 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 Example Emergency Action Levels: Note: The Emergency Director should declare the General Emergency promptly upon determining that 60 minutes has been exceeded , or will likely be exceeded.
R Q 2.1 Spent fuel pool level cannot be restored to at least 16.36 feet-: (site specific Level 3 value) for 60 minutes or longer. Definitions:
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 included to provide classification diversity.
De,relaper Notes: In accordance with the discussion in Section 1.4, NRG Order EA 12 051, it is recommended that this IC and E,.\L be implemented 1.v-hen the enhanced spent fuel pool level instrumentation is available fer use. The "site specific Level 3 value" is usually that spent fuel pool level where fuel remains covered and actions to implement make up water addition should no longer be deferred.
This site specific level is determined in accordance with NRG Order EA 12 051 and NEI 12 02, and applicable ovmer's group guidance.
Developers should modify the EAL and/or Basis section to reflect any site specific constraints or limitations associated with the design or operation of instrumentation used to determine the Level 3 &#xa5;affie-: EGL Assignment Attributes:
3.1.4.C 69 NEI 99 OJ (Revi s ion 6) ~lo*1ember 2012 7 COLD SHUTDOWN/
REFUELING SYSTEM MALFUNCTION ICS/EALS Table C 1: Reeognition Category "C" Initiating Condition 1\e'latrix UNUSUAL EVENT CUl ill>l"PLANNED loss of (reactor vessel/RCS 
[PWR] or RPV [BWR]) inventory for 15 minutes or longer. Op. }Jades: Geld Shutde, 1 m , Refueling2,.
fr CU2 Loss of all but one AC power source to emergency buses for 15 minutes or longer. Op. },1odes: LJCeld Shi1ffl.own , Refaeling, Defueled CU3 ill>JPLAN1'tED increase in RCS temperature. Op. A/odes: LJCold Shi1ffl.own , Refuelin g CU4 Loss of Vital DC power for 15 minutes or longer. Op. A/odes: LJCeld Shutdewn , Refuelin g CUS Loss of all onsite or offsite communications capabilities.
Op. }Jades: LJCold Shutdown, Refueling , Defueled ALERT CAl Loss of (r e actor v es s el/RCS [PWR] orRPV [BWR]) inventory. Op. },1ode s: LJCeld Shuffl.own , R(}fueling CA.2 Loss of all offsite and all onsite AC povrer to emergency buses for 15 minutes or longer. Op. ,\/odes: LJCold Shutdewn , Refae!i,qg , Defaeled C,A .. J Inability to maintain the plant in cold shutdovm. Op. }Jades: LJCold Shutdewn , R(}fueling 70 SITE AREA GENERAL EMERGENCY EMERGENCY C81 Loss of (reactor vessel/RCS 
[PWR] or RPV [BWR]) CC 1 Loss of (reactor vessel/RCS 
[PWR] or RPV [BWR]) inventory affectin g fuel clad integrity with containment challenged.
inventory affecting core decay heat removal capability.
Op. A/odes: LJCeld Shutdmvn , Ref ue ling Op. A/odes: LJCold Shutdev,m , R(}faelin g ,-------------------, 1 Table in tended for u s e b y
* I I , EAL de\'elopers. , : Inclusion in licensee I ,.J * * ,.J , uocuments 1s not requtr e u. ~------------------*
UNUSUAL EVENT A.LERT CA6 Hazardous event affecting a SAFETY SYSTEM needed for the cmTent operating mode. Op. ,\lodes: LJCeld Shutdo,1*r1 , Rcfi1cbn g 71 SITE AREA EMERGENCY I NEI 99 0 I (Revi s ion 6) }lovember 2012 GENERA.L EMERGENCY Table inteneee for use by
* EAL ee~*elopers. : InolusioH in licensee I ..J ' * ..J 1 uocuments 1s not requireu.
L------------------J NEI 99 01 (Re11ision 
: 6) No11ember 2012 CU1 ECL: Notification of Unusual Event Initiating Condition:
UNPLANNED loss of (reactor vessel/RC8 
[PWR] or RPV [BWR]) inventory for 15 minutes or longer. Operating Mode Applicability:
Cold 8hutdovm, Refuelin~
Emergency Action Levels: Example Emergeney Aetion Levels: (1 or 2) Note: The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded , or will likely be exceeded.
C 1.1 UNPLANNED loss of reactor coolant results fin (reactor vessel/RC8 
[PWR] or RPV [BW~]) level less than ANY of the following for 15 minutes or longer: a. In Mode 4 RPV water level less than 170" OR b. In Mode 5 if RPV level band is established above the RPV flan e and RPV water level drops below the RPV flangea required lov,zer limit for 15 minutes or longer. OR c. In Mode 5 if RPV level band is established below the RPV flan e and RPV water level drops below RPV level band. a. (Reactor vessel/RC8 
[PWR] or RPV_ [BWR]) level 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 Pool or Drywell and Reactor Building floor and equipment drain sump levels. 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 (reactor vessel/RC8 
[PWR] er-RPV [BWR]) 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. 72 NEI 99 01 (Revision 
: 6) November 2012 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. 73 NE! 99 01 (RevisioR 
: 6) ~lovember 2012 EAL CUI .1 recognizes that the minimum required (reactor vessel/RC&#xa3; 
[PWR] or RPV [BWR]) 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. +lle 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 crit e rion excludes transient conditions causing a brief lowering of water level. 74 1'lEI 99 01 (Revision 
: 6) 1'Jovember 2012 ---EAL CUI .2 addresses a condition where all means to determine (reactor vessel/RCS 
[PWR] or RPV [BWR]) level have been lost. In this condition , 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 flov, to ensure they are indicative ofleakage from th e (reactor vessel/RCS 
[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 l e akage. Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CAI or CA3. Develef)eF Netes: EAL #1 It is recognized that the minimum allowable reactor vessel/RCS/RPV level may have many values over the course of a refueling outage. Developers should solicit input from licensed op e rators concerning the optimum wording for this EAL statement.
In particular , determine if the generic wording is adequate to ensure accurate and timely classification, or if specific setpoints can be included 1 Nithout making the EAL statement unwieldy or potentially inconsistent with actions that may be tak e n during an outage. If specific setpoints are included, these should be dravm from applicable operating procedures or other controlling documents.
E AL #2.b E nter any " sit e sp e cific sump and/or tank" levels that could be expected to increase if there v;ere a loss of invento r y (i.e., the lost inventory 
\Vould enter the listed sump or tank). EGL Assignment ,i\ttributes:
3.1.1.A 75 ECL: Notification of Unu s u a l Event NEI 99 0 l (Revision 
: 6) }loYember 2012 CU2 Initiating Condition:
Loss of all but one AC power source to emergency essential buses for 15 minutes or longer. Operating Mode Applicability:
Cold 8hutdovm, Refueling4, 5 , D e fueled Example Emergency Action Levels: Note: The Emergency Dir ector should declar e the Unusual Eve nt promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
c&#xb5;2.1 a. AC power capability to (site specific emergency buses)IA3 and 1A4 buses is reduced to a sing le power s ource for 15 minutes or lon ge r. 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 p l ant and/or placing it in the cold shutdown condition, including the ECCS. Systems classified as 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 los s of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power so urce may be powering one , or more than one, train of safetyrelated 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 sys tems. Thus, when in these modes , this condition is considered to be a potential degradation of the l evel of safety of the plant. An "A C 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 b e lo w.
* 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 all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of emergency essential buses being back fed from the unit main generator.
* A loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency essential bu ses being back-fed from an offsite power source. 76 77 NEI 99 01 (RevisieR 
: 6) ~leYember 2012 NEI 99 01 (ReYision 
: 6) 1'foYemeer 2012 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. 78 Developer Notes: l>JEI 99 01 (Re&#xa5;isioA 
: 6) l>Jo&#xa5;ember 2012 For a power source that has multiple generators, the EAL and/or Basis section should reflect the minimum number of operating generators necessary for that source to provide required power to an AC emergency bus. For e1cample , if a backup power source is comprised of two generators (i.e., two 50% capacity generators sized to feed 1 i\.C emergency bus), the EAL and Basis section must specify that both generators for that source are operating.
The " site specific emergency buses" are the buses fed by offsite or emergency AC power sources that supply power to the electrical distribution system that powers SAFETY SYSTEMS. There is typically 1 emergency bus per train of SAFETY SYSTEMS. D e v e lop e rs should modify th e bull e t e d e1 campl e s provid e d in the basi s section , above, as needed to reflect their site specific plant designs and capabilities.
The EALs and Basis should reflect that each independent offsite power circuit constitutes a single power source. For e>mmple, three independent 345kV offsite power circuits (i.e., incoming pmver lines) comprise three separate power sources. Independence may be detem1ined from a review of the site specific UFSAR, SBO analysis or related loss of electrical power studies. The EAL and/or Basis section may specify use of a non safety related power source provided that operation of this source is recognized in AOPs and EOPS , or beyond design basis accident response guidelines (e.g., FLEX support guidelines). Such pov,er sources should generally meet the " Alternate ac source" definition provided in 10 CFR 50.2. J'A multi unit stations , the EALs may credit compensatory measures that are proceduralized and can be implemented within 15 minutes. Consider capabilities such as power source cross ties , " swing" generators , other po 1.ver sources described in abnormal or emergency operating procedures , etc. Plants that have a proceduralized capability to supply offsite A.C power to an affected unit via a cross tie to a companion unit may credit this pov,'er source in the EAL provided that the planned cross tie strategy meets the requirements of 10 CFR 50.63. EGL Assignment Attributes:
3 .1.1.A 79 ECL: Notification of Unusual Event Initiating Condition:
UNPLANNED increase in RCS temperature.
Operating Mode Applicability:
Cold Shutdown, Refueling4, 5 Emergency Action Levels: *NE! 99 0 I (Revision 
: 6) N 01rember 2012 CU3 Example Emergeeey Aetioe Leyels: (1 or 2) Note: The Emergency Dir ector should declare the Unusual Event promptly upon determinin g that 15 minutes ha s been exceeded, or will likely be exceeded. 
~1 ~2 UNPLANNED incr ease in RCS temperature to greater than (site specific Technical Sp e cification cold s hutdown temperature limit)212&deg;F. Loss of ALL RCS temperature and (reactor vessel/RCS 
[PWR] or RPV [BWR]) le ve l 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 tran s ient. 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.
CONTAIN.MENT CLOSURE: The procedurally defined conditions or actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.
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 avai l able does not warrant a classification.
EAL CU3.1 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 80 N E I 99 01 (Re11ision 
: 6) l>l oyember 20 I 2 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.
81  
:t>JEI 99 01 (RevisieR e) "t>levember 2012 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 l oad has been reduced since the cessation of power operation.
Fifteen minutes was se l ected 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.
Devel013eF Netes: For E,i\L #1, enter the "site specific Technical 
&#xa3;pecification cold shutdown temperature limit" where indicated.
EGL Assignment Attributes:
3 .1. l .i\ 82 ECL: Notification of Unusual Event Initiating Condition:
Loss of Vital DC power for 15 minutes or lon ger. Operating Mode Applicability:
Cold &#xa3;hutdown , Refueling1-_,_j_
Example EmeFgeeeyEmergency Action Levels: NEI 99 0 l (Revision 
: 6) *November 2012 CU4 Note: The Emergency Dir ector should declare the Unusual Eve nt promptly upon determining that 15 minutes ha s been exceeded , or will likely be exceeded.
Indicated voltage is less than (site specific bus voltage value) 105 VDC on BOTH Div 1 and Div 2 125 VDC busesrequired Vital DC buses for 15 minutes or lon ger. Definitions:
SAFETY SYSTEM: A system required for safe plant operation, cooling down the p l ant and/or placing it in the cold shutdown condition, including the ECCS. 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 deca y 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 (inop era ble) for scheduled outage maintenance work and Train Bis 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 pow er to Train A would not warrant an emergency classification. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Minimum DC bus voltage selected due to automatic trip of the inverters at 105 VDC decreasing.
Depending upon the event, escalation of the emergency classification level would be via IC CAl or CA3, or an IC in Recognition Category AR. DevelopeF Notes: The "site specific bus voltage value" should be based on the minimum bus voltage necessary for adequate operation of &#xa3;A.FETY &#xa3;Y&#xa3;TEM equipment.
This voltage value should incorporate a margin of at least 15 minutes of operation before the onset of inability to operate 83  
:t>JEI 99 01 (RevisioA 
: 6) November 2012 those loads. This voltage is usually near the minimum voltage selected when battery sizing is performed.
The typical value for an entire battery set is approximately 105 VDC. For a 60 cell string of batteries, the cell voltage is approximately 
: 1. 75 Volts per cell. For a 5 8 string battery set, the minimum voltage is apprmdmately 1.81 Volts per cell. EGL Assignment Attributes:
3.1.1.A 84 NEI 99 01 (Re,cision 
: 6) ~J oyemeer 2012 ECL: Notification of Unusual Event Initiating Condition:
Lo ss of all on s it e or off s ite communication s capabilities.
Operating Mode Applicability:
Cold Shutdown, Refueling5, 64, 5 , Defueled Emergency Action Levels: Aetien Le\'els: (1 or 2 or 3) Lo ss of ALL of th e followin g onsit e communication methods: CU5 !._(site specific list of communications methods)Plant Op e rations Radio System
* In-Plant Phone System
* Plant Paging System (Gaitronics) 
~2 Loss of ALL of the following GR.Qoffsite response organization communications methods:
* D AEC 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 communications methods) Los s 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 GRGoffsite response organization s and the NRC. 85 NEI 99 01 (Re,,ision 
: 6) ~Joyemser 2012 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 tran s mission points , individuals being sent to offsite locations , etc.). 86 NEI 99 01 (Rev i sion 6) }Jovember 2012 EAL CU5.l addresses a total Joss 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 GRGoffsite response organ i zation s of an emergency declaration.
The offsite response organizat i ons referred to here are the State of Iowa, Linn County, and Benton CountyThe OROs referred to here are (see Developer Notes). ---E AL CU5.3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration. De,,eloper Notes: 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 not items owned and maintained by individuals.
EAL #2 The "site specific list of communications methods" should include all communicat i ons methods used to perform initial emergency notifications to OROs as described in the site Emergency Plan. The listing should include installed plant equipment and components, and not items owned and maintained by individuals.
fammple methods are ring downldedicated telephone lines, commercial telephone lines, radios, satellite telephones and internet based communications technology. In the Basis section, insert the site specific listing of the OROs requiring notification of an emergency declaration from the Control Room in accordance with the site Emergency Plan, and typically within 15 minutes. EAL #3 The "site specific list of communications methods" should include all communications methods used to perfom1 initial emergency notifications to the NRG as described in the site Emergency Plan. The listing should include installed plant equipment and components, and not items owned and maintained by individuals.
These methods are typically the dedicated Emergency Notification System (ENS) telephone line and commercial telephone lines. EGL /\.ssignment Attributes:
3.1.1.C 87 NEI 99 01 (Revision 
: 6) l'lovember 2012 CA1 ECL: Alert Initiating Condition:
Loss of (reactor vessel/RC&#xa3; 
[PWR] or RPV [BWR]) inventory.
Operating Mode Applicability:
Cold &#xa3;hutdovm , Refueling4, 5 Emergency Action Levels: Example EmeFgeeeyEmeFgeeey Aetioe L~1 els: (1 or 2) Note: The Emergency Dir ect or should declare the Alert promptly upon determining that 15_ minutes h as be en exceeded , or will likel y be exceeded. 
~1 ~2 Loss of (reactor vessel/RC&#xa3; 
[PWR] or RPV [BWR]) inventory as indicated by level les s than (site specific level)l 19.5 inches. a. (Reactor vessel/RO, [PWR] or RPV [BWR]) 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 UNPL,"..1't"NED increase in (site specific sump and/or tank) levels due to a loss of (reactor vessel/RC&#xa3; 
[P\VR] or RPV [B'.VR]) 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 CAl.1 , a lowering of water level below (site specific level)l 19.5 inches indicates that operator actions have not been successful in restoring and maintaining (reactor vessel/RC&#xa3; 
[PWR] or RPV [BWR]) 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. 88 NE! 99 01 (RevisioR 
: 6) November 2012 For EAL CAl.2 , the inability to monitor (reactor vessel/RCS 
[PWR] or RPV [BWR]) level may be caus e d 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 1.vater flow to ensure they a.re indicative of leakage from the (reactor vessel/RCS 
[PWR] or RPV [BWR]).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. 89 NE! 99 01 (ReYisioA 
: 6) Jl-lovemeer 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 (reactor vessel/RCS 
[PWR] or RPV [BWR]) inventory l evel continues to lower, then escalation to Site Area Emergency would be via IC CS 1. DevelopeF Notes: For EAL #1 the "site specific level" should be based on either: * [BWR] Low Low EGGS actuation setpoinULevel 
: 2. This setpoint was chosen because it is a standard operationally significant setpoint at v,rhich some (typically high pressure EGGS) injection systems would automatically start and is a value significantly below the low RPV water level RPS actuation setpoint specified in IC CUI. * [PWR] The minimum allowable level that supports operation of normally used decay heat removal systems (e.g., Residual Heat Removal or Shutdovm Cooling).
If multiple levels e>cist, specify each along 1,vith the appropriate mode or configuration dependency criteria.
For EAL #2 The type and range of RCS level instrumentation may vary during an outage as the plant moves thrnugh various operating modes and refueling evolutions, particularly for a PWR. As appropriate to the plant design, alternate means of determining RCS level are installed to assure that the ability to monitor level within the range required by operating procedures ,vill not be interrupted.
The instrumentation range necessary to support implementation of operating procedures in the Cold Shutdown and Refueling modes may be different (e.g., narrower) than that required during modes higher than Cold Shutdown.
Enter any "site specific sump and/or tank" levels that could be e>&#xa3;pected to increase if there v,ere a loss of inventory (i.e., the lost inventory would enter the listed sump or tank). EGL Assignment Attributes:
3 .1.2.B 90 ECL: Alert l>JEI 99 0 I (Revision 
: 6) l>JoYember 2012 CA2 Initiating Condition:
Loss of all offsite and all onsite AC power to emergency essential buses for 15 minutes or longer. Operating Mode Applicability:
Cold &#xa3;hutdovm, Refueling4, 5 , Defueled Emergency Action Levels: Example EmeFgeneyEmergeney Aetiee Le*vels: Note: The Emergency Dir e ctor should d e clar e th e Alert promptly upon determining that 15_ minutes has been exceeded, or will likely be exceeded.
C 2.1 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. Systems classified as 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 los ses. ---Escalation of the emergency classification level would be via IC CS 1 or A&#xa3;.l-RS 1. DevelepeF Notes: For a power source that has multiple generators , the El.1.L and/or Basis section should reflect the minimum number of operating generators necessary for that source to provide adequate pov1er to an AC emergency bus. For eJrnmple , if a backup power source is comprised of two generators (i.e., two 50% capacity generators sized to feed l AC emergency bus), the EAL and Basis section must specify that both generators for that source are operating.
91 NE! 99 01 (Revision 
: 6) No,*ember 2012 The " site specific emergency buses" are the buses fed by offsite or emergency AC power sources that supply power to the electrical distribution system that powers SAFETY SYSTEMS. There is typically 1 emergency bus per train of SAFETY SYSTEMS. The EAL and/or Basis section may specify us e of a non safety related power source provided that operation of this source is controlled in accordance 1.vith abnormal or emergency operating procedures , or beyond design basis accid e nt r e sponse guidelines (e.g., FLEX support guidelines).
Such pov,cer sources should g enerally meet th e " Alternate ac source" d e finition provided in 10 CFR50.2. A.t multi unit stations , the EA.Ls may credit compensatory measures that are proceduralized and can be implement e d within 15 minutes. Consider capabilities such as power sourc e cross ties , " swing" generators , oth e r powe r sources described in abnormal or emergency operating proc e dur es, etc. Plant s that hav e a p roc e duralized capability to supply offsite AC power to an affected unit via a cros s tie to a companion unit may credit this power source in the EAL provided that the planned cross tie strategy meets the requirements of 10 CFR 50.63. EGL Assignment i\.ttributes:
3 .1.2.B 92  
:t>lel 99 O I (Re 1 1isioR 6) "t>loYember 2012 CA3 ECL: Alert Initiating Condition:
Inability to maintain the plant in cold shutdown.
Operating Mode Applicability:
Cold Shutdown , Refuelin~
Emergencv Action Levels: Example Emergeney Aetion LeYels: (1 or 2) Note: The Emergency Director should declare the A l ert promptly upon determining that the applicable time has been exceeded, or will likely be exceeded. ~1 C 3.2 UNPLANNED increase in RCS temperature to greater than (s ite specific Technical Specification cold shutdown temp e rature limit)212&deg;F for greater than the duration specified in the following table. Table: RCS Heat-up Duration Thresholds RCS 8tatusintegritv CONTAINMENT CLOSURE Heat-up Duration Status Intact (eut not at reduced Not applicable 60 minutes* in,,,entory f PWR]) Not intac t (or at reduced Esta blished 20 minutes* in'fentOFJ
' fi2 WR]) Not Established 0 minutes
* Jf an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL i s not applicable.
UNPLANNED RCS pressure increase greater than (site specific pressure reading) 10 psig due to a loss of RCS cooling.. (Thi s E AL does not apply during water solid plant conditions. fPWR]) Definitions:
UNPLANNED:
A parruneter 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. CONTAINMENT CLOSURE: The procedurally defined conditions or actions taken to secure containment and its associated structures , systems, and components as a functional barrier to fission product release under shutdo~'fl conditions. 93 Basis: l>IEI 99 0 I (Revision 
: 6) l>Jo 1 1ember 2012 This IC addresses conditions involving a lo ss 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 substantia l degradation of the l eve l 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). 94 NEl 99 01 (R!wision 
: 6) }Jo11e1~0er 20 I 2 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 RC8 inventory is reduced (e.g., mid loop operation in P'.l/Rs) . .,_ 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 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 AS-l-RS 1. Develeper Netes: For El.rL #1 Enter the " site specific Technical 8pecification cold shutdown temperature limit" where indicated.
The RC8 should be considered intact or not intact in accordance with site specific criteria.
For EAL #2 The " site specific pressure reading" should be the lowest change in pressure that can be accurately determin e d using installed in s trumentation , but not less than 10 psig. For P\VRs , this IC and its associated EA.Ls address the concerns raised by Generic Letter gg 17, Los s ofD ec ay Heat Removal. A number of ph e nomena such as pressurization , vorte>Eing , steam generator U tube draining , RC8 level differences when operating at a mid loop condition , decay heat removal system design , and level instrum e ntation problems can lead to conditions where decay heat r e moval 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 1.vithin an hour after decay heat removal i s lo s t. The allow e d time frames are consistent with the guidance provided by Generic Lett e r gg 17 and b e liev e d to be conservative given that a low pressure Containment barrier to fission product release is established.
EGL Assignment Attributes:
3.1.2.B 95 NEI 99 0 I (Revision 
: 6) l'J ovember 2012 CA6 ECL: Alert Initiating Condition:
Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode. Operating Mode Applicability:
Cold Shutdovm, RefuelingU Emergency Action Levels: Exam(lle Emergeney Aetien LeYels: _;_ Notes: C li\6.1
* If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occtmed, 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 wan-anted. 
: 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)River level above 757 feet
* River Water Supply (RWS) pit low level alarm
* Other events with similar hazard characteristics as determined by the Shift Manager or Emergency Director AND b. EITHER of the following: 
-----1. Event damage has caused indications of degraded performance in-at !east one train of a SAFETY SYSTEM needed for the cunent 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 cunent operating mode,BF. 96 NEI 99 01 (RevisioR 
: 6) J>fo 1 ,ce1f!ber 2012 .!__ The event has caused resulted in VISIBLE DAMAGE to the second train of a SAFETY SYSTEM component or structure needed for the current operating mode~
* Loss of the safety function of a sing l e train SAFETY SYSTEM. 97 Definitions:
1'JEI 99 01 (Revision 
: 6) l>l oyemeer 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. 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. Systems classified as 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 CA6.1. 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 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. 98 l>JEl 99 0 l (Revision 
: 6) November 2012 This IC addres s es a hazardous event that causes damage to a SA.CETY SYSTEM , or a structure 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 f)lant; EAL 1.b.1 addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available.
Th e indications of degraded performance should be significant e nough to cause concern regarding th e op e rability or reliability of the SAFETY SYSTEM train. EAL l .b.2 addr e sses damage to a SAF E TY SYSTEM component that is not in service/operation or readily apparent thrnugh indications alone , or to a structure containing SAFETY SYSTEM components.
Operators 1.vill make this d e termination based on the totality of available event and damag e r e port information. This i s i n t en d e d t o b e a brief assessment not r e quiring lengthy analysis or quantification of the damage. Escalation of the emergency classification level would be via IC CS 1 or AS 1 RS 1. DeYelef)eF Netes: For (site specific hazards), developers should consider including other significant , site specific hazards to the bulleted list contained in E,i\.L I .a (e.g., a seiche). Nuclear power plant Sl,FETY SYSTEMS are comprised of two or more separate and redundant trains of equipment in accordance with site specific design criteria.
EGL Assignment Attributes:
3 .1.2.B 99  
~IEI 99 0 I (Revision 
: 6) November 2012 CS1 ECL: Site Area Emergency Initiating Condition:
Loss of (reactor vessel/RCS 
[PWR] or RPV [BWR]) inventory affecting core decay heat removal capability.
Operating Mode Applicability:
Cold Shutdown , Refueling4, 5 Emergency Action Levels: Example Emergeney A.etion Levels: (1 or 2 or 3) Note: The Emergency Dir ec tor should declare the Site Area Emergency promptly upon determining that 30 minutes ha s b ee n exceeded, or will likely be exceeded. 
~1 a. b. ~2 a. b. CONTAINMENT CLOSURE not established.
AND (Reactor vessel/RCS 
[PWR] or RPV [BWR]) level LESS THANless than specific level)+64 inches.:.:
CONTAINMENT CLOSURE established.
AND (Reactor vessel/RCS 
[PWR] or RPV [BWR]) level LESS THA1'Hess than specific level).+ 15:2 inches ~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 ANY of the following:
I Definitions:
* Drywell Monitor (9184A/B) reading greater ~~rBetc-H+e-&#xa5;.a-H+e-+
.5. 0 R/hr
* Erratic source range monitor indication 
[PWR]
* UNPLANN E D level rise in Drywell/Reactor Building Equipment or Floor Drain sump, or Suppression Pool UNPLANt-rnD increase in (site specific sump and/or tank)_levels of sufficient magnitude to indicate core uncovery * (Other site specific indications) 100 NEl 99 OJ (Revision 
: 6) }>lovember 20 I 2 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 b e 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. 101 Basis: l>JEI 99 01 (ReYision 
: 6) NoYember 2012 This IC addresses a significant and prolonged loss of (reactor vessel/RCS 
[PWR] or RPV [BWR]) inventory control and makeup capability l eading 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 wanant a Site Area Emergency declaration.
10 2  
------1 NEI 99 0 I (R.e,,ision 
: 6) , November 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 RCS/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 RCS/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 s ince it is calibrat e d 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 install e d (including the ability to monitor level visually) to assure that the ability to monitor level will not be interrupted.
In EAL CS 1.3 .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 bas 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 vessel/RCS 
[PWR] or RPV [BWR]) 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 (reactor vessel/RO; 
[PWR] or RPV [BWR]). These EALs address concerns raised by Generic Letter 88-17 , Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management. 
---Escalation of the emergency classification level would be via IC CG 1 or AG+RG 1. DeYelapeF Nates: Accident analyses suggest that fuel damage may occur within one hoar of uncovery depending upon the an1ount of time since shutdovm; refer to Generic Letter 88 17, SECY 91 283, 1'RJREG 1449 andNUMARC 91 06. The type and range of RCS level instrumentation may vary during an outage as the plant moves through various operating modes and refueling evolutions, particularly for a P'.VR. As 103  
}JEI 99 0 I (Revision 
: 6) }fo*rember 2012 appropriate to the plant design, alternate means of determining RCS level are installed to assure that the ability to monitor level within the range 1*equired by operating procedures will not be interrupted. The instrumentation range necessary to support implementation of operating procedures in the Cold Shutdovm and Refueling modes may be different (e.g., narrower) than that required during modes higher than Cold Shutdovm. PV/R For EI'. .. L #1.b the "site specific level" is 6" below the bottom ID of the RCS loop. This is the level at 6" below the bottom ID of the reactor vessel penetration and not the low point of the loop. If the availability of on scale level indication is such that this level value can be determined during some shutdovm modes or conditions, but not others, then specify the mode dependent and/or configuration states during which the level indication is applicable. If the design and operation of vvater level instrumentation is such that this level value cannot be determin e d at any time during Cold Shutdovm or Refueling modes , then do not include EAL #1 (classification will be accomplished in accordance with EAL #3). For EI'. .. L #2.b The "site specific level" should be apprmcimately the top of active fuel. If the availability of on scale level indication is such that this level value can be determined during some shutdown modes or conditions , but not others, then specify the mode dependent and/or configuration states during 1 which the level indication is applicable. If the design and operation of water level instrumentation is such that this level value car.not be determined at any time during Cold Shutdovm or Refueling modes, then do not include EAL #2 (classification will be accomplished in accordance with EAL #3). For EAL #3.b first bullet A.s water level in the reactor vessel lowers, the dose rate above the core 1.vill increase.
Enter a "site specific radiation monitor" that could be used to detect core uncovery and the associated "site specific value" indicative of core uncovery. It is recognized that the condition described by this IC may result in a radiation value beyond the operating or display range of the installed radiation monitor. In those cases, EAL values should be determined with a margin sufficient to ensure that an accurate monitor reading is available.
For example, an EAL monitor reading might be set at 90%, to 95% of the highest accurate monitor reading. This provision notwithstanding, if the estimated/calculated monitor reading is greater than appro)cimately 110% of the highest accurate monitor reading, then developers may choose not to include the monitor as an indication and identify an alternate EAL threshold.
To further promote accurate classification , developers should consider if some combination of monitors could be specified in the EAL to build in an appropriate level of corroboration between monitor readings into the classification assessment.
For EAL #3.b second bullet Post TMI accident studies indicated that the installed PWR nuclear instrumentation will operate erratically 
\Vhen the core is uncovered and that this should be used as a tool for making such determinations.
For EAL #3. b third bullet Enter any 'site specific sump and/or tank" levels that could be e)cpected to change if there were a loss of RCS/reactor vessel inventory of sufficient magnitude to indicate core uncovef)'. Specific level values may be included if desired. For Et".L #3. b fourth bullet Developers should determine if other reliable indicators e)(ist to identify fuel uncovery (e.g., r~mote viewing using cameras). The goal is to identify any unique or 104 NEI 99 0 I (Revision 
: 6) ~Joyernber 2012 site specific indications, not already used elsewhere, that will promote timely and accurate emergency classification.
For EAL #1.b "site specific level" is the Low Lov,r Low EGGS actuation setpoint / Level 1. The B\1/R Lov,r Low Low EGGS actuation setpoint / Level 1 was chosen because it is a standard operationally significant setpoint at which some (typically low pressure EGGS) injection systems would automatically start and attempt to restore RPV level. This is a RPV water level value that is observable below the Low Low/Level 2 value specified in IC CAI, but significantly above the Top of Active Fuel (TQ,A.c,.V) threshold specified in EAL #2. For EAL #2.b The "site specific level" should be for the top of active fuel. For EAL #3.b first bullet As water level in the reactor vessel lov;ers, the dose rate above the core will increase.
Enter a "site specific radiation monitor" that could be used to detect core uncovery and the associated "site specific value" indicative of core uncovery.
It is recognized that the condition described by this IC may result in a radiation value beyond the operating or display range ofthe installed radiation monitor. In those cases, EAL values should be determined with a margin sufficient to ensure that an accurate monitor reading is available.
For example, an EAL monitor reading might be set at 90% to 95% of the highest accurate monitor reading. This provision notwithstanding, if the estimated/calculated monitor reading is greater than approximately 110% of the highest accurate monitor reading, then developers may choose not to include the monitor as an indication and identify an alternate EAL threshold.
To further promote accurate classification, developers should consider if some combination of monitors could be specified in the EAL to build in an appropriate level of corroboration betv,reen monitor readings into the classification assessment.
For BWRs that do not have installed radiation monitors capable of indicating core uncovery, alternate site specific level indications of core uncovery should be used if available. For EAL #3. b second bullet Because BWR source range monitor (SRJ\.4) nuclear instrumentation detectors are typically located below core mid plane , this may not be a viable indicator of core uncovery for BWRs. For EAL #3.b third bullet Enter any "site specific sump and/or tank" levels that could be expected to change if there vt'ere a loss of RPV inventory of sufficient magnitude to indicate core uncovery.
Specific level values may be included if desired. For EAL #3. b fourth b .. 1llet Developers should determine if other reliable indicators exist to identify fuel uncovery (e.g., remote viewing using cameras). The goal is to identify any unique or site specific indications, not already used elsevmere, that will promote timely and accurate emergency classification.
EGL l\ssignment Attributes:
3.1.3.B 105 1-JEI 99 0 l (Revision 
: 6) Novem.eer 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, RefuelingU Emergency Action Levels: Example EmergeueyEmergeuey 
,,\ ... etien Levels: (1 or 2) Note: Th e Emergency Director should declare the General Emergency promptly upon determining that 30 minutes has been exceeded , or w ill lik ely be exceeded. C 1.1 a. (Reactor vessel/RCS 
[PWR] or RPV [BWR]) lev e l LESS THANless than fsite-specific level)+ 15 !:inches for 30 minutes or lon ger. AND b!z. ANY indication from the Secondary Containment Challenge Table (see below). (Reactor vessel/RCS 
[PWR] or RPV [BWR]) l evel cannot be monitored for 30 minutes or longer. -----AND b. Core uncovery is indicated by ANY of the following:
* Drywell Monitor (9184A/B) (Site specific radiation monitor) reading GREA,TER THA.Ngreater than (site specific value)5.0 R/hr
* Errat ic source range monitor indication 
[PWR]
* UNPLANNED level rise in Drywell/Reactor Building Equipment or Floor Drain sump, or Suppression Pool illl"PLA}il':rED increase in (site specific sump and/or tank) levels of sufficient magnitude to indicate core uncovery * (Other site specific indications)
AND c. ANY indication from the Secondary Containment Challenge Table (see below). Secondary Containment Challenge Table 106
* CONTAINMENT CLOSURE not established
* NE! 99 01 (Re 1.'ision 6) l>foyemeer 2012
* Drywell Hydrogen or Torus Hydrogen GREATER TH.1\Ngreater than 6% AND Drywell Oxygen or Torus Oxygen GREATER THANgreater than 5% (E>cplosive miJcture) eJcists inside containment
* UNPLANNED increase in containment pressure
* Secondary containment radiation monitors above max safe operating limits (MSOL) ofEOP 3, Table 6radiation monitor reading above (site specific value) [BWR]
* If CONTAINMENT CLOSURE is re-established prior to exceed i ng the 30-_minute time limit , th en declaration of a General Emergency is not required.
107 Definitions:
l>lEI 99 0 I (RevisioA 
: 6) l>loYeFAber 20 !'.l CONTAINMENT CLOSURE: Procedmally defined actions taken to secure containment and its associated structures, systems, and components as a functional ban-ier to fission product release under existing plant conditions.
For DAEC, this is considered to be Secondary Containment as required by Technical Specifications.CONTAINMENT CLOSURE: The procedurally defined conditions or actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdovm 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 ad dresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged.
This condition repr esen ts actual or IMMINENT substantial core degrad a tion or melting with potential for lo ss of containment integrity. Releases can be reasonably expected to exceed EPA P AG exposme 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 RCS/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 established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required. The existence of an explosive mixtme means , at a minimum , that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen bmn (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 integrit y. 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 dming 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 con-elate reactor and plant conditions to determine if core uncovery has actually occun-ed (i.e., to account for various accident progression and instrumentation uncertainties).
It also allows sufficient time 108 1-lEI 99 01 (Re1rision 
: 6) No\'ember 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. 109 NEI 99 0 I (Revision 
: 6) l>love1:ni:Jer 20 I 2 The inability to monitor (reactor vessel/RCS 
[PWR] or RPV [BWR]) 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 (reactor 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: (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. +I hese EALs address concerns raised by Generic Letter 88-17, Los s of D ecay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issu es; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Indu stry Actions to Assess Shutdown Management.
DevelepeF Netes: Accident analyses suggest that fuel damage may occur within one hour of ,mcovery depending upon the amount of time since shutdovm; refer to Generic Letter 88 17 , SECY 91 283, NUREG 1449 andNUMARC 91 06. The type and range of RCS level instrumentation may vary during an outage as the plant moves through various operating modes and refueling evolutions , particularly for a P\l/R. As appropriate to the plant design , alternate means of determining RCS le 1 1el are installed to assure that the ability to monitor level \Vithin the range required by operating procedures 
\Yill not be interrupted. The instrumentation range necessary to support implem e ntation of operating procedures in the Cold Shutdown and Refueling modes may be different (e.g., narrower) than that required during modes higher than Cold Shutdovm.
For EAL #I.a The "site specific level" should be apprmcimately the top of active fuel. If the availability of on scale level indication is such that this level value can be determined during some shutdovm modes or conditions , but not oth e rs , th e n specify the mode dependent and/or configuration states during which the level indication is applicable.
If the design and operation of 1.vater level instrumentation i s s uch that this level value cannot be det e rmined at any time during Cold Shutdown or Refueling modes , th e n do not include EAL #1 (classification will be accomplished in accordance v1ith EAL #2). For EAL #2.b first bullet As water level in the reactor vessel lowers , the dose rate above the core will increase. Enter a "site specific radiation monitor" that could be used to detect core uncovef)' and the associated 
" site specific value" indicative of core uncovery. It is recognized that th e condition described by this IC may result in a radiation valu e beyond the operating or display range of the installed radiation monitor. In those cases , EAL values should be determined with a margin sufficient to ensure that an accurate monitor reading is available.
For example , an E AL monitor reading might be set at 90% to 95% of the highest accurate monitor reading. This provi s ion notwithstanding , if the estimated/calculated monitor reading i s greater than 110 NEI 99 01 (RevisioA 
: 6) *November 2012 apprmcimately 110% of the highest accurate monitor reading, then developers may choose not to include the monitor as an indication and identify an alternate EAL threshold.
To further promote accurate classification , developers should consider if some combination of monitors could be specified in the EAL to build in an appropriate level of corroboration between monitor readings into the classification assessment.
For BWRs that do not have installed radiation monitors capable of indicating core uncovery , alternate site specific level indication s of core uncovery should be used if available.
For EAL #2.b second bullet Post TMI accident studies indicated that the installed PWR nuclear instrumentation will operate erratically when the core is uncovered and that this should be used as a tool for making such determinations.
Because B\VR Source Range Monitor (SRM) nuclear instrumentation detectors are typically located below core mid plane , this may not be a v iable indicator of core uncovery for BWRs. For EAL #2.b third bullet Enter any " site specific sump and/or tank" levels that could be e)cpected to change if there w e re a loss of inventory of sufficient magnitude to indicate core uncovef)'. Specific level values may be included if desired. F or EAL #2.b fourth bullet Developers should determine if other reliable indicators e)dst to identify fuel uncovery (e.g., remote viewing using cameras).
The goal is to identify any unique or site specific indications, not already used elsevmere , that 1.vill promote timely and accurate emergency classification.
For the Containment Challenge Table: Site shutdown contingency plans typically provide for re establishing CONTA]}lMENT CLOSURE follO\Ying a loss of RCS heat removal or inventory control functions.
For " E)(plosive mh(ture" , developers may enter the minimum containment atmospheric hydrogen concentration necessary to support a hydrogen burn (i.e., the lower deflagration limit). A concurr e nt containment oxygen conc e ntration may be included if the plant has this indication available in the Control Room. For BWRs , the use of secondary containment radiation monitors should provide indication of increas e d release that may be indicative of a challenge to secondary containment.
The " site specific value" should be ba s ed on the EOP maximum safe values because these values are easily recognizable and have a defined basis. EGL Assignment Attributes:
3 .1. 4 .B 111 NEI 99 0 l (RevisieR e) }>Je,,ceffiber 2012 8 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS Table E 1: Reeegeitien Categen' "E" Initiating Condition Matrix UNUSUAL EVENT E HUl Damage to a loaded cask CONFINEMENT BOUNDARY.
Op. Afodes: All 11 2 I Table intended for use by 1 EAL developers. : Inclus i on in licensee I ,J , * ,.i 1 uocuments 1s not reqmreu. L------------------1 ISFSI "MALFUNCTION ECL: Notification of Unusual Event Initiating Condition:
Damage to a loaded cask CONFINEMENT BOUNDARY.
Operating Mode Applicability:
All Example Emergency Action Levels: E-HU1 E HUI.I Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading greater than the values shown below(2 times the site specific cask specific technical specification allowable radiation level) on the surface of the spent fuel cask. 61BT DSC HSM FfOnt BiFd ScFeen3 feet from HSM 800 mrem/hr Surface Outside HSM Door -Centerline of DSC 200 mrem/hr End Shield Wall Exterior 40 mrem/hr Definition:
CONFINEMENT 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 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 env ironm ent, 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-R IC RA UI, 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 !Cs HUI and HAI. 113 18F81 MALFUNCTION Developer Notes: The results of the ISFSI Safety Analysis Report (SAR) [per 1'illREG 1536], or a Si\R referenced in the cask Certificate of Compliance and the related NRG Safety Evaluation Report, identify the natural phenomena events and accident conditions that could potentially affect the CONFINEMENT BOU1'JDARY. This EAL addresses damage that could result from the range of identified natural or man made events (e.g., a dropped or tipped over cask, EXPLOSION, FIRE , EARTHQUKE, etc.). The allowable radiation level for a spent fuel cask can be found in the cask's technical specification located in the Certificate of Compliance. EGL Assignment Attributes:
3.1.1.B 114 9 FISSION PRODUCT BARRIER ICS/EALS Table 9 F 1 : Recognition Category " F" Initiating Condition MatriJc ALERT Any Loss or any Potential Loss of either the Fuel Clad or RCS barrier. Loss or Potential Loss of any two barriers.
Loss of any tv ,z o barriers and Loss or Potential Loss of th e third barrier. See Table 9 F 2 far ll\\'R EALs See Table 9 F 3 far P~'R EALs }leI 99 0 I (Re*,isioR e) }1ovember 2012 Devel0per Nate: The adjacent logic flow diagram is for use by developers and is not required for site specific implementation
; however , a site specific scheme must include some type of user aid to facilitate timely and accurate classification of fission product barrier losses and/or potential losses. Such aids are typically comprised of logic flov,z diagrams , " scoring" criteria or checkbmc type matrices. The user aid logic must be consistent v,ith that of the adjacent diagram. 115 m m L OSS POTENTIAL LOSS FUEL CLA D LOSS POTENTIAL LOSS F UEL CLAD LOSS POTENTI A L LOSS FU EL CLA D LOSS L OSS L OSS POTENTIAL LOSS RCS POTENTIAL LOSS RCS 2/3 POTENTIAL LOSS RCS LOSS POTENTIAL LOSS CONTAINMENT L O SS Ila -L oss of ANY T w o B arriers A.fil2 Loss or Po ten t ial Loss o f Th i rd Barrier POTENTIAL L OSS CONTA INMENT ES.l -Loss or Potential Loss of ANY Two Ba rri ers '---------------
-----------1 fAl -ANY Loss or ANY Pote nt i al L oss of .EIIHE.R F uc l Clad .QR RCS 116 NEI 99 0 I (R1wisioR 
: 6) November 2012  

}>lEJ 99 0 I (RevisioR a) November 2012 Table 9-F 4: B,VR DAEC EAL Fissi on Product Barrier Table Thresholds for LOSS or POTENTIAL LOSS of Barriers FAlALERT Any-ANY Loss or ANY aflY Potential Loss of either the Fuel Clad erOR RCS barrier. Fuel Clad Barrier LOSS POTENTIAL LOSS 1. Priman: Containment Conditions!.
R.(;8 AetP<*ity Not Am~lica blek Not Applicable E8ite s13eeifie iH:aieations that reaetor eoolant aetivity is greater than J gg 1-+/-Gi,lgm ease equivalent I FSl SITE AREA EMERGENCY RCS Barrier LOSS POTENTIAL LOSS 1. Primary Containment PFessuFeConditions A. Primary Not Applicable containment pressure greater than Esite s13eeifie 
&#xa5;a-l-aet~
due to RCS leakage. 118 FGlGENERALEMERGENCY Loss of fmY-ANY two barriers and Loss erOR Potential Loss of the third barrier. Containment Barrier LOSS POTENTIAL LOSS 1. Primary Containment Conditions A. UNPLANNED A. Primary rapid drop in eontainmentTorus 13nmary pressure greater eontainmentDrywe than Esite s13eeifie ti pressure &#xa5;a-l-aet53 psig following 13rimary OR eontainmentDrywe B. Drywell or Torus ti pressure rise H2 cannot be OR detennined to be B. Primary be88 +HAf.Hess eontainmentDrywe than 6% and ti pressure Drywell &OR response not Torus 02 cannot consistent with be determined to LOCA conditions.
be less than 5% OR Esite speeifie C. UNISOLABLE explesiYe direct downstream mixtuFe) exists pathway to the inside pFimary environment exists eeetaiemeet after primary OR containment C. HC+L (Graph 4 of isolation si@al EOP 2) exceeded.
Fuel Clad Barrier RCS Barrier LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS D. 119 }ffiI 99 01 (Revision 
: 6) NoveFAber 2012 Containment Barrier LOSS POTENTIAL LOSS OR Intentional Qrimary containment venting Qer EOPs I Fuel Clad Barrier RCS Barrier I LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS 2. RP V Wa t e r L eve l 2. RPV Water Leve l A. SAG entry is A. RPV water level A. RPV water level Not Applicable requiredPrimary cannot be restored cannot be restored eentaiA:ment and maintained and maintained fleeding required. above Esite Sf)eeifie above Esite-R,P\l :,,1,iater le*rel Sf)eeifie RP&#xa5; eerresJ3ending te water le>rel the tef) ef aeti>re eerresJ3ending te fueB+ 15 inches OR the tef) ef aetive cannot be fuelj+ 15 inches determined. OR cannot be determined. 3. RCS Leak Rate3.Net Applieable 
: 3. R CS L ea k Rate Not Applica b le Not Applicable A. UNISOLA B LE A. UNISOLABLE break in t ...... ~Y ef primary s y stem the fullewing:
leaka g e that (site Sf)eeifie results in systems with exceeding the f)Stential for high Max Normal energy line 0Qerating Limit breaks)Main (MNOL) ofEOP Steam, HPCI, 3, Table 6 for Feed water, EITHER of the RWCU, or RCIC following: as indicated by the
* 1. Max failure of both :J'.1.formal isolation valves in Operating anvANY one line Temperature to close AND OR -EITHER:
* 2. Max
* HighMSL +/-",formal 120 2. NEr 99 01 (RevisioR e) November 2012 Containment Barrier LOSS POTENTIAL LOSS RPV Water Level Not Applicable A. SAG entry is requiredPrimary eeataififfient fleediag required. 
: 3. RCS Leak Rate3. Primary Containment lselatiee Failure A. UNISOLABLE Not Applicable Qrimary system leakage that results in exceeding the Max Safe 0Qerating Limit (MSOL) of EOP 3, Table 6 for EITHER of the following:
* TemQerature OR
* Radiation Level --A: UNISO bA8bE direet I Fuel Clad Barrier RCS Barrier I LOSS I POTENTIAL LOSS LOSS POTENTIAL LOSS flow or steam Operating Area tunnel Radiat i on tem2erature 
+/-::itWelLevel:-
annunciators OR
* Direct re2ort of steam release OR -B. Emergency RPV Depressurization required. 121 }lei 99 01 (Revision 
: 6) l>Jovember 2012 Containment Barrier LOSS POTENTIAL LOSS downstream patlwray to the envirornnent exists after pnmary contairnnent isolation signal OR --Pr. Intentio nal primary containment venting per EGP-5 OR --G: YNI80 LA,BLE primary system leakage that rest1lts in exceeding El'.fllER o:f the foU01.ving: 
: l. Mm( 8afe Operating Temperature.
OR 2. MmE 8afe Gperating Area Radiation I Fuel Clad Barrier RCS Barrier I LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS 4. Primary Containment Radiation 
: 4. Primary Containment Radiation A. Drywell Monitor Not Applicable A. Dr~ell Monitor Not Applicable (9184A/B} 
(9184A/B}
reading greater reading greater than 200 R/hr. than 5 R/hr after OR reactor B. Torus Monitor shutdownA-(9185AIB}
Primary reading greater containment than 200 R/hr radiation monitor reading greater than (site specific value). 5. Other Indications 
: 5. Other Indications A. Fuel damage Not A1mlicableA.-
Not A1mlicableA-Not Ai;mlicableA.-
assessment (site specific as (site specific (site specific as indicates at least applicable) as applicable) applicable) 5% fuel clad damage. fsite-specific as applicable) 
: 6. Emergency Director Judgment 6. Emergency Director Judgment 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 Emergency Emergency Emergency Emergency Director that Director that Director that Director that indicates Loss of indicates Potential indicates Loss of indicates Potential 122 4. l>IBI 99 01 (ReYision 
: 6) l>fo*,ember 201? Containment Barrier LOSS POTENTIAL LOSS Primary Containment Radiation Not Applicable A. Drywell Monitor (9184A/B}
reading greater than 5000 R/hr. OR B. Torus Monitor (9185A/B}
reading greater than 500 R/hrA.-Primary containment radiation monitor reading greater than (site specific ,,,alue). 
: 5. Other Indications Not A1mlicableA.-
A. Fuel damage (site specific as assessment applicable)
EPA8AP +.'.t~ indicates at least 20% fuel clad damage.fsite-specific as applicable) 
: 6. Emergency Director Judgment A. ANY condition in A. ANY condition in the opinion of the the opinion of the Emergency Emergency Director that Director that indicates Loss of indicates Potential I Fuel Clad Barrier RCS Barrier I LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS the Fuel Clad Loss of the Fuel the RCS Barrier. Loss of the RCS Barrier. Clad Barrier. Barrier. 123 Jl.IBI 99 0 I (RevisioR e) November 2012 Containment Barrier LOSS POTENTIAL LOSS the Containment Loss of the Barrier. Containment Barrier.  

Basis Information For B'\\'R DAEC EAL Fission Product Barrier Table 9-F-2-B"'R 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 AetivityPrimary Containment Conditions Loss 1.A }Tm 99 0 I (Revision 
: 6) *November 2012 This threshold indicates that RCS radioactivity concentration is greater than 3 00 &#xb5;Ci/gm dose equivalent I 131. Reactor coolant activity above this level is greater than that e>cpected for iodine spikes and corresponds to an approximate 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. There is no Loss or Potential Loss threshold associated with RCS ActivityPrimary Containment Condition. De, 1 el0per Netes: Threshold values should be determined assuming RCS radioactivity concentration equals 300 &#xb5;Ci/gm dose equivalent I 131. Other site specific units may be used (e.g., &#xb5;Ci/cc). Depending upon site specific capabilities, this threshold may have a sample analysis component and/or a radiation monitor reading component.
Add this paragraph (or similar v,rording) to the Basis if the threshold includes a sample analysis component , "It is recogni2:ed 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 inch1ded as a backup to other indications." 2. RPV Water Level 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 represents the EOP requirement for primary containment flooding.
This is identified in the B'NROG EPGs/SAGs when the phrase , "Primary Containment Flooding Is Required," appears. Since a site specific RPV water level is not specified here , the Loss threshold phrase, "Primary containment 125  
}ffil 99 0 I (Revision 
: 6) No,;ember 2012 flooding required ," also accommodates the EOP need to flood the primary containment when RPV water level car,not be determined and core damage due to inadequate core cooling is believed to be occurring. 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. 126 B'\&#xa5;R FUEL CLAD BARRIER THRESHOLDS:
l>lBI 99 0 I (Revision 
: 6) NO'tember 2012 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 Emergenc y. 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 , automaticall y 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 pre v ent 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 emergenc y RPV depressurization has been attempted , giving the operator an opportunity to assess the capability oflow-pressure injection sources to restore RPV water le v el or 2) no low pressure RPV injection systems are available , precluding RPV depressurization in an attempt to minimize loss of RPV inventory. 127 DAEC FUEL CLAD BARRIER THRESHOLDS 
{cont.):~ 
}-tEJ 99 0 I (Revision 
: 6) November 2012 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 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 acti v e fuel in order to reduce reactor power. RPV water level is then controlled between the top of acti v e 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 SSS 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.
128 "t-l"EII 99 0 I (RevisioR a) NO\'ember 
?O 12 BWR FUEL CLAD BARRIER THRESHOLDS: 
: 3. DeYel0J3er Netes: Loss 2.A The phrase , "Primary containment flooding required," should be modified to agree \Vith the site specific EOP phrase indicating exit from all EOPs and entry to the SA.Gs (e.g., drywell flooding required , etc.). Potential Loss 2.A The decision that "RPV water level car.not be determined" is directed by guidance given in the RPV water level control sections of the EOPs. Net Af)f)lieable (ineluded fer numbering eensisteney between barrier tables)RCS Leak Rate ---There is no Loss or Potential Loss threshold associated with RCS Leak Rate. M......:_Primary Containment Radiation Loss 4.A This threshold indicates that RCS radioactivity concentration is greater than 300 &#xb5;Ci/gm dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate 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. Loss 4.B The Torus radiation monitor rea d ing corresponds to an instantaneous release of all reactor coolant mass into the primary Toruscontainment , assuming that reactor coolant activity equals 300 &#xb5;Ci/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate 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 reading in this threshold is 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.
129 0 (") ......
BWR DAEC FUEL CLAD BARRIER THRESH O LDS (cont.): 1'ffil 99 0 I (Revision 
: 6) 1'Jo 1 1ember 2012 her lndicationst 
: 5. 1. Other Iedieatiees Loss and/or Potential Loss 5.A Results obtained from procedure PASAP 7.2. Fuel Damage Assessment indicate at least 5% fuel clad damage. ffhis subcategory addresses other site specific thresholds that may be included to indicate loss or potential loss of the Fuel Clad barrier based on plant specific design &#xa3;!:l aracteristics not considered in the generic guidance.
There is no Potential Loss threshold associated with Other Indications.
D eve l o p er No t es: Loss and/or Potential Loss 5.A Developers should determine if other reliable indicators exist to evaluate the status of this fission product barrier ( e.g., re v iew accident analyses described in the site Final Safety Analysis Report , as updated).
The goal is to identify an y unique or site-specific indications that will promote timely and accurate assessment of barrier status. Any added thresholds should represent approximately the same relative threat to the barrier as the other thresholds in this column. Basis information for the other thresholds may be used to gauge the relative barrier threat level. ~E m erge nc y Dir e c tor J ud gment 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. Po t entia l 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 potent i al l y lost. The Emergency Director should also consider whether or not to declare the barrier potentiall y lost in the e v ent that barrier status cannot be monitored. 131}}