Text

River Bend Station, Unit 1, License Amendment Request Adoption of Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 6 - Evaluation of Proposed Changes
	ML18128A052
Person / Time
Site:	River Bend
Issue date:	04/30/2018
From:	; Entergy Operations
To:	; Office of Nuclear Reactor Regulation
References
	RBG-47847
	Download: ML18128A052 (715)
	v • d • e

ENCLOSURE 1 RBG-47847 EVALUATION OF PROPOSED CHANGES I Enclosure 1 RBG-47847 Page 1 of 8 EVALUATION OF THE PROPOSED CHANGES 1.0

SUMMARY

DESCRIPTION Entergy Operations, Inc. (Entergy) proposes to revise the River Bend Station, Unit 1 (RBS) currently approved Emergency Plan (EP) Emergency Action Level (EAL) scheme, which is based on the Nuclear Energy lnstitute's (NEl's) guidance established in NEI 99-01, Revision 5, "Methodology for Development of Emergency Action Levels" (Reference 1 ). Entergy is proposing to adopt the EAL schemes based on the guidance provided in NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors," which has been endorsed by the NRC (Reference

2) 2.0 DETAILED DESCRIPTION The proposed changes involve revising RBS's EAL scheme, which is currently based on NEI 99-01, Revision 5, to a scheme based on NEI 99-01, Revision 6. Enhancements provided by Revision 6 of the guidance include:

Clarification of numerous EALs that have been typically misinterpreted by the industry in the development of their site-specific EAL scheme, -* Clarification of the intent of EALs that have been historically misclassified, 2.1

Providing additional guidance for the development of EALs for current non-passive reactor designs as well as possible future reactor designs that are non-passive,

Incorporating lessons learned from industry events (i.e., Fukushima and others) and NUREG/CR-7154, "Risk Informing Emergency Preparedness Oversight:

Evaluation of Emergency Action Levels -A Pilot Study of Peach Bottom, Surry and Sequoyah," and

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.

Proposed Initiating Conditions (ICs) and EALs Enclosure 2, "Proposed EAL Technical Basis Document (Markup)," provides a markup of the current RBS EAL scheme basis illustrating changes incorporating the guidance of NEI 99-01, Revision 6. Enclosure 3 provides a clean copy of RBS EAL Technical Basis document.

2.2 Deviations and Differences Enclosure 4 contains a matrix that provides a comparison of the Initiating Conditions (ICs) and EALs in NEI 99-01 to the ICs and EALs proposed for RBS. The comparison evaluates differences and deviations consistent with the similar exercise performed during the RBS EP upgrade to NEI 99-01, Revision 5 (Reference 3). The NRC approved RBS's EP upgrade to NEI 99-01, Revision 5, on July 31, 2012 (Reference 4).

Enclosure 1 RBG-47847 Page 2 of 8 A difference is an EAL change where the basis scheme guidance 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 re-formatting 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.). A number of differences are the result of adding plant-specific information to the EALs. In these cases, Enclosure 4 may refer the reader to an associated document in Enclosure 5, "Supporting Referenced Document Pages, "which provides the technical basis for plant-specific information.

2.3 Generic Differences The differences below apply throughout the set of EALs and are not specifically identified in each instance in the comparison matrix as a difference.

NEI 99-01, Rev 6 EALs RBS Station EALs References PWRs Deleted PWR references as appropriate Uses E-HU for ISFSI ICs Uses EU for ISFSI ICs Designates ICs and EALs as, Designates ICs and EALs as Example: (IC)HU1 EAL 2 Example: HU1 .2 Emergency Classification ICs are presented Emergency Classification ICs are presented together by Emergency Classification level by Emergency Classification "family" (A. .. 1, (all NOUEs grouped together, then all A ... 2, A ... 3, etc.) for each category, in Alerts, etc.) for each category (A, C, H, ascending order (UE -GE) etc.), in ascending order (UE -GE) 2.4 Operational Mades Mode applicability of the proposed !Cs and EALs is consistent with the NEI 99-01, Revision 6. The following tables provide the operating modes for RBS as defined by the respective Technical Specifications (TSs).

L Enclosure 1 RBG-47847 Page 3 of 8 '.MODE TITLE 1 Power Operation 2 Startup 3 Hot Shu.tdown (a) 4 Cold Shutdown (a) 5 Refueling (bl RBS AVERAGE REACTOR COOLANT TEMPERATURE

(°F) NA NA > 200 :::; 200 NA (a) All reactor vessel head closure bolts fully tensioned. (b) One or more reactor vessel head closure bolts less than fully tensioned.[CA1]

In addition to the TS defined operational modes, NEI 99-01, Revision 6, defines the following additional mode: Defueled:

All reactor fuel removed from the reactor vessel (i.e., full core off load during refueling or extended outage). RBS procedures recognize and are consistent with the definition of a defueled condition.

2.5 Instrumentation Used for EALs RBS has verified that the specified values used as EAL setpoints are within the* calibrated range of the referenced instrumentation.

2.6 Background Technical Information Enclosure 2 provides the existing EAL Technical Basis document marked to illustrate the proposed changes. Enclo 9 ure 3 provides the revised (clean) EAL Technical Basis document.

Enclosure 4 provides a deviation-difference document comparing NEI 99-01, Revision 6, with the proposed changes to the RBS EAL schemes. Enclosure 5 provides specific reference documents or excerpts which support related RBS Emergency Plan proposed changes. 3.0 TECHNICAL EVALUATION

,

Enclosure 1 RBG-47847 Page 4 of 8 Entergy has evaluated the proposed changes to determine whether applicable regulations and requirements have been met. NEI 99-01 guidance methodology includes many years of development, along with use and implementation.

The guidance has been subject to NRC reviews and approval.

The RBS EAL scheme currently in place is based on the methodology in NEI 99-01, Revision 5. NEI 99-01, Revision 6, is the latest version endorsed by the NRC and provides guidance to nuclear power plant operators for the development of a site-specific emergency action level scheme.' 10 CFR 50.47(b )(4) requires that emergency plans include a standard emergency classification and action level scheme. This scheme is a fundann['

ental component of an EP in that it provides the defined thresholds that will allow site personne to rapidly implement a range of pre-planned emergency response measures.

'An emergency cl ssification scheme also facilitates timely

decision-making by an offsite response organization concerning the implementation of precautionary or protective actions for the public. NEI 99-01, Revision 6, contains a generic set of ICs, EALs, and fission product barrier status thresholds.

The guidance also includes supporting technical basis information, developer notes, and recommended classification instructions for users. The methodology described ir, this document is consistent with NRC requirements and guidance.

In particular, this methodology was specifically endorsed by the NRC in a March 28, 2013, letter from NRC to NEI (Reference

2) and determined to provide an acceptable approach in meeting requirements of 1 O CFR 50.47(b)(4), applicable requirements of 10 CFR 50, Appendix E, and the associated planning standard evaluation elements contained in NUREG-0654/FEMA-REP-1, Revision 1, "Criteria for Preparation and Evaluation Df Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants," November 1980. 10 CFR 50, Appendix E, Section IV.R2, requires that a licensee desiring to change its entire EAL scheme shall submit an application for an amendment to its license and receive NRC approval before implementing the change. The proposed change to the RBS EAL scheme from NEI 99-01, Revision 5, to NEI 99-01, Revision 6, guidance does not reduce the capability to meet the applicable emergency planning standards and requirements in 10 CFR 50.47(b) and 10 CFR 50, Appendix E. Accordingly, pursuant to the requirements of 10 CFR 50, .Appendix E,Section IV.B.2, Entergy requests NRC review and approval of the proposed changes to the EAL scheme as a license amendment request in accordance with 10 CFR 50.90: 4.0 REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria The regulations in 10 CFR 50.54(q) provide direction to licensees seeking to revise emergency plans. The requirements related to nuclear power plant emergency plans are contained in the standards in 10 CFR 50.47, "Emergency Plans," and the requirements of Appendix E, "Emergency Planning ~md Preparedness for Production and Utilization Facilities." Paragraph 10 CFR 50.47(a)(1) states that no operating license for a nuclear power reactor will be issued unless a finding is made by the NRC that there is reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency.

Section 50.47(b) contains standards that onsite and offsite emergency response plans must meet for the NRC staff to make a positive finding that there is reasonable assurance that Enclosure 1 1 RBG-47847 Page 5 of 8 adequate protective measures can and will be taken in the event of a radiological emergency.

One of these standards, 10 CFR 50.47(b )( 4 ), requires that emergency plans include a standard emergency classification and action level scheme. 10 CFR 50, Appendix E, Section IV.B, "Assessment Actions," requires that emergency plans include emergency action levels (EALs) that are to be used as criteria for determining the need for notification and participation of local and state agencies, and for determining when and what type of protective measures should be considered to protect the health and safety of individuals both onsite and offsite. EALs are to be based on plant conditions and instrumentation, as well as onsite and offsite radiological monitoring.

Section IV.B provides that initial EALs shall be . discussed and agreed on by the applicant and state and local authorities, be approved by the NRC, and reviewed annually thereafter with state and local authorities.

Therefore, a revision to EALs will require NRC approval prior to implementation, if it involves (1) changing from one EAL scheme to another (e.g., NEI 99-01, Revision 4 to NEI 99-01, Revision 6), (2) proposing an alternate method to comply with the regulations, or (3) the EAL revision proposed by the licensee decreases the effectiveness of the emergency plan. NRC Regulatory Issue Summary (RIS) 2005-02, Revision 1, "Clarifying the Process for Making Emergency Plan Changes", issued April 19, 2011, says that a change in an EAL scheme to incorporate the improvements provided in NUMARC/NESP-007 or NEI 99-01 would not decrease the overall effectiveness of the emergency plan, but due to the potential safety significance of the change, the change needs prior NRC review and approval.

The proposed changes meet the above regulatory requirements.

4.2 Precedent The following commercial nuclear power plants have received license amendments that approved EALs based on NEI 99-01, Revision 6: ' Callaway -Amendment 212 (Reference

5) Fermi 2 -Amendment 202 (Reference

6) South Texas -Amendment 206 and 194 (Reference

7) 4.3 No Significant Hazards Consideration Analysis Entergy Operations, Inc. (Entergy) proposes to revise the currently approved Emergency Plan (EP) Emergency Action Level (EAL) scheme for River Bend Station, Unit 1 (RBS), which is based on the Nuclear Energy lnstitute's (NEl's) guidance established in NEI 99-01, Revision 5, "Methodology for Development of Emergency Action Levels." Entergy is proposing to adopt the EAL schemes based on the guidance provided in NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors," which has been endorsed by the NRC. Entergy has ev~luated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment." As required by 1 O CFR 50.91(a), Entergy analysis of the issue of no significant hazards consideration is presented below.

Enclosure 1 RBG-47847 Page 6 of 8 1. Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated?

Response:

No The proposed changes to the RBS EALs db not involve any physical changes to plant equipment or systems and do not alter the assumptions of any accident analyses.

The proposed changes do not adversely affect accident initiators or precursors and do not alter design assumptions, plant configuration, or the manner in which the plant is operated and i maintained.

The proposed changes do not adversely affect the ability of structures, systems or components (SSCs) to perform intended safety functions in mitigating the consequences of an initiating event within the assumed acceptance limits. Therefore, the changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated?

Response:

No No new accident scenarios, failure mechanisms, or limiting single failures are introduced

~s a result of the proposed changes. The changes do not challenge the integrity or performance of any safety-related systems. No plant equipment is installed or removed, and the changes do not alter the design, physical configuration, or method of operation of any plant SSC. Because EALs are not accident initiators and no physical changes are made to the plant, no new causal mechanisms are introduced.

Therefore, the changes do not create the possibility of a new or' different kind of accident from an accident previously evaluated.

3. Do the proposed changes involve a significant reduction in a margin of safety? Response:

No Margin of safety is associated with the ability of the fission product barriers (i.e., fuel cladding, reactor coolant system pressure boundary, and containment structure) to limit the level of radiation dose to the public. ,The proposed changes do not impact operation of the plant and no accident analyses are affected by the proposed changes. The changes do not , affect the Technical Specifications or the method of operating the plant. Additionally, the proposed changes will not relax any criteria used to establish safety limits and will not relax any safety system settings.

The safety analysis acceptance criteria are not affected by these changes. The proposed changes will not result in plant operation in a configuration outside the design basis. The proposed changes do not adversely affect systems that respond to safely shut down the plant and to maintain the plant in a safe shutdown condition.

Therefore, the changes do not involve a significant reduction in a margin of safety.

Enclosure 1 RBG-47847 Page 7 of 8 Based upon the reasoning presented above, Entergy concludes that the requested change .involves no significant hazards consideration, as set forth in 1-0. CFR 50.92(c), "Issuance of Amendment." 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. 5.0 ENVIRONMENTAL CONSIDERATION The proposed changes are applicable to emergency planning requirements involving the proposed adoption of the NRG-endorsed EAL guidance as described in NEI 99.-01, Revision 6, and do 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 changes do not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed changes meet the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Therefore, pursuant to j O CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed changes ..

6.0 REFERENCES

1. NEI 99-01, Revision 5, "Methodology for Development of Emergency Action Levels" February 2008 (ML080450149)

2. NRC letter "U.S. Nuclear Regulatory Commission Review and Endorsement of NE/ 99-01, Revision 6, Dated November, 2012 (TAC No. 092368)," March 28, 2013 (ML 12346A463)

3. Entergy letter dated August 1, 2011, "Proposed Emergency Action Levels Using NE/ 99-01 Revision 5 Scheme," (ML 11216A055) lRBG-47165)

4. NRC Safety Evaluation dated July 31, 2012, "River Bend Station, Unit 1 -Emergency Action Level Scheme Upgrade Based on Nuclear Energy Institute (NE/) 99-01, Revision 5, "Methodology for Development of Emergency Action Levels" (TAC Nos. ME6846) (ML 12178A567) (RBC-51043)

5. NRC letter "Callaway Plant, Unit 1 -Issuance of Amendment Re: Upgrade to Emergency Action Level Scheme (GAG No. MF4945),"0ctober 7, 2015 (ML 15251A493)

6. NRC letter "Fermi 2 -Issuance of Amendment to Revise the Emergency Action Level Scheme for the FERMI 2 Emergency Plan (TAC No. MF5048)," September 29, 2015 (ML 15233A084) )

Enclosure 1 RBG-47847 Page 8 of 8 7. NRC letter "South Texas Project, Units 1 and 2 -Re: Upgrade to Emergency Action Level Scheme (TAC Nos. MF4195 and MF4196)," August 20, 2015 (ML 15201A195)

ENCLOSURE 2 RBG-47847 PROPOSED EAL TECHNICAL BASIS DOCUMENT (NEI REVISION 6 MARKUP)

-===-Entergy River Bend Station EAL Basis Document Revision XXX River Bend Station EAL Technical Basis Page 1 of 296

==-Entergy River Bend Station EAL Basis Document Revision XXX Table of Contents

1.0 INTRODUCTION

........................................................

Error! Bookmark not defined. 2.0 DISCUSSION

.......................

.' .....................................................................................

3 2.1 Background

.....................................................................................................

3 2.2 Fission Product Barriers ..................................................................................

4 2.3 Fission Product Barrier Classification Criteria .................................................

4 2.4 EAL Organization

............................................................................................

4 2.5 Technical Bases Information

...........................................................................

7 2.6 Operating Mode Applicability

...........................................................................

8 3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS

..................................

9 3.1

General Considerations

...................................................................................

9 3.2 Classification Methodology

............................................................................

1 O

4.0 REFERENCES

........................................................................................................

14 4.1 Developmental

..............................................................................................

14 4.2 Implementing

.................................................................................................

14 5.0 DEFINITIONS, ACRONYMS &ABBREVIATIONS

..................................................

15 5.1 Definitions (ref. 4.1.1 except as noted) ..........................................................

15 5.2 Abbreviations/Acronyms

................................................................................

20 6.0 RBS-TO-NEI 99-01 Rev. 6 EAL CROSS-REFERENCE

..........................................

23 -7.0 ATTAGHMENTS

......................................................................................................

27 7.1 Attachment 1, Emergency Action Level Technical Bases .............................

27 . Category A -Abnormal Rad Levels/ Rad Effluent .......................................

28 Category C -Cold Shutdown/

Refueling System Malf~nction

......................

69 Category E -Independent Spent Fuel Storage Installation (ISFSI) ............

114 Category F -.Fission Product Barrier Degradation

......................................

117 Table F-1 Fission Product Barrier Threshold Matrix & Bases 122 Category H -Hazards and Other Conditions Affecting Plant Safety ...........

176 Category S -System Malfunction

...............................................................

221 7.2 Attachment 2, Safe Operation

& Shutdown Areas TablesA-3

& H-2 Bases262 Page 2 of 296

---Entergy River Bend Station EAL Basis Document Revision XXX

1.0 INTRODUCTION

This document provides an explanation and rationale for each Emergency Action Level (EAL) included in the EAL Upgrade Project for River Bend Station (RBS). It should be used to facilitate review of the RBS EALs and provide historical documentation for future reference.

Decision-makers responsible for implementation of EIP-2-001, Classification of Emergencies, may use this document as a technical reference in support of EAL interpretation.

This information may assist the Emergency Director in making classifications, particularly those involving judgment or multiple events. The basis information may also be useful in training and for explaining event classifications*to off-site officials.

The expectation is that emergency classifications are to be made as soon as conditions are present and recognizable for the classification, but within 15 minutes or less in all cases of conditions present. Use of this document for assistance is not intended to delay the emergency classification.

Because the information in a basis document can affect emergency classification making (e.g., the Emergency Director refers to it during'an event), the NRC staff expects that changes to the basis document will be evaluated in accordance with the provisions of 10 CFR 50.54(q).

2.0 DISCUSSION 2.1 Background EALs are the plant-specific indications, conditions or instrument readings that are utilized to classify emergency conditions defined in the RBS Emergency Plan. In 1992, the NRC endorsed NUMARC/NESP-007 "Methodology for Development of Emergency Action Levels" as an alternative to NUREG-0654 EAL guidance.

NEI 99-01 (NUMARC/NESP-007)

Revisions 4 and 5 were subsequently issued for industry implementation.

Enhancements over earlier revisions included:

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

Initiating conditions and example emergency action levels that fully address conditions that may be postulated to occur at permanently Defueled Stations and Independent Spent Fuel Storage Installations (ISFSls).

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

Subsequently, Revision 6 of NEI 99-01 has been issued which incorporates resolutions to numerous implementation issues including the NRC EAL Frequently Asked Questions (FAQs). Using NEI 99-01 Revision 6, "Methodology for the Development of Emergency Action Levels for Non-Passive Reactors," November 2012 (ref. 4.1.1), RBS conducted an EAL implementation upgrade project that produced the EALs discussed herein. Page 3 of 296 River Bend Station EAL Basis Document Revision XXX 2.2 Fission Product Barriers Fission product barrier thresholds represent threats to the defense in depth d~sign concept that precludes the release of radioactive fission products to the environment.

This concept relies on multiple physical barriers, any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment.

Many of the EALs derived from the NEI methodology are fission product barrier threshold based. That is, the conditions that define the EALs are based upon thresholds that represent the loss or potential loss of one or more of the three fission product barriers. "Loss" and "Potential Loss" signify the relative damage and threat of damage to the barrier. A "Loss" threshold means the barrier no longer assures containment of radioactive materials.

A "Potential Loss" threshold implies a greater probability of barrier loss and reduced certainty of maintaining the barrier. The primary fission product barriers are: A. Fuel Clad Barrier (FCB): the Fuel Clad Barrier consists of the-cladding material that contains the fuel pellets. B. Reactor Coolant System Barrier (RCB): The RCS Barrier is the reactor coolant system pressure boundary and includes the RPV and all reactor coolant system piping out to and including the isolation valves. C. Containment Barrier (CNB): The Primary Containment Barrier includes the drywell, the containment, their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves. Containment Barrier thresholds are used as criteria for escal_ation of the Emergency Classification Level (EGL) from Alert to a Site Area Emergency or a General Emergency.

2.3 Fission Product Barrier Classification Criteria The following criteria are the bases for event classification related to fission product barrier loss or potential loss: Alert: Any loss or any potential loss of either Fuel Clad or RCS Barrier Site Area Emergency:

Loss or potential loss of any two barriers General Emergency:

Loss of any two barriers and loss or potential loss of the third barrier 2.4 . EAL Organization The .RBS EAL scheme includes the following features:

Division of the EAL set into three broad groups: Page 4 of 296 River Bend Station EAL Basis Document Revision XXX o EALs applicable under any plant operating modes -This group wol'.lld be reviewed by the EAL-user any time emergency classification is considered.

o EALs applicable only under hot operating modes -This group would only be reviewed by the EAL-user when the plant is in Hot Shutdown, Startup, or Power Operation mode .. o EALs applicable only under cold operating modes -This group would only be reviewed by the EAL-user when the plant is in Cold Shutdown, Refueling or Defueled mode. The purpose of the groups is to avoid review of hot condition EALs when the plant is in a cold condition and avoid review of cold condition EALs when the plant is in a hot condition.

This approach significantly minimizes the total number of EALs that must' be reviewed by the EAL-user for a given plant condition, reduces EAL-user reading burden and, thereby, speeds identification of the EAL that applies to the emergency.

Within each group, assignment of EALs to categories and subcategories:

Category and subcategory titles are selected to represent conditions that are operationally significant to the EAL-user.

The RBS EAL categories are aligned to and represent the NEI 99-01" Recognition Categories." Subcategories are used in the RBS scheme as necessary to further divide the EALs of a category into logical sets of possible emergency classification thresholds.

The RBS EAL categories and subcategories are listed below. The primary tool for determining the emergency classification level is the EAL Classification Matrix. The user of the EAL Classification Matrix may (but is not required to) consult the EAL technical bases in order to obtain additional information concerning the EALs under classification consideration.

The user should consult Section 3.0 and Attachments 1 & 2 of this document for such information.

Page 5 of 296 River Bend Station EAL Basis Document Revision XXX EAL Groups, Categories and Subcategories EAL Group/Category I EAL Subcategory Any Operating Mode: A-Abnormal Rad Levels/ Rad Effluent 1 -Radiological Effluent 2 -Irradiated Fuel Event 3 -Area Radiation Levels H -Hazards and Other Conditions 1 -Security Affecting Plant Safety 2 -Seismic Event \ ' 3 -Natural or Technological Hazard 4-Fire 5 -Hazardous Gas 6 -Control Room Evacuation 7 -Emergency Director Judgment E -Independent Spent Fuel Storage 1 -Confinement Boundary Installation (ISFSI) Hot Conditions:

S -System Malfunction 1 -Loss of Emergency AC Power 2 -Loss of Vital DC Power 3 -Loss of Control Room Indications 4 -RCS Activity 5 -RCS Leakage 6 -RPS Failure 7 -Loss of Communications 8 -Hazardous Event Affecting Safety Systems F -Fissio~ Product Barrier Degradation None Cold Conditions:

C -Cold Shutdown / Refueling System 1 -RPV Level Malfunction 2 -Loss of Emergency AC Power 3 -RCS Temperature 4 -Loss of Vital DC Power 5 -Loss of Communications I 6 -Hazardous Event Affecting Safety Systems Page 6 of 296

-===-Entergy River Bend Station EAL Basis Document Revision XXX 2.5 Technical Bases Information EAL technical bases are provided in AUachm~nt 1 for each EAL according to EAL group (Any, Hot, Cold), EAL categol)' (A, C, E, F, Hand S) and EAL subcategory.

A summary explanation of each category and subcategory is given at the b_eginning oUhe technical bases discussions of the EALs included in the category.

For each EAL, the following information is provided:

Category Letter & Title Subcategory Number & Title Initiating Condition (IC) Site-specific description of the generic IC given in NEI 99-01 Rev. 6. EAL Identifier (enclosed in rectangle)

Each EAL is assigned a unique identifier to support accurate communication of the emergency classification to onsite and offsite personnel.

Four characters define each EAL identifier:

1. First character (letter):

Corresponds to the EAL category as described above (A, C, E, F, H orS) 2. Second character (letter):

The emergency classification (G, S, A or U) G = General Emergency S = Site Area Emergency A= Alert U = Unusual Event 3. Third character (number):

Subcategory number within the given category.

Subcategories are sequentially numbered beginning with the number one (1 ). If a category does not have a subcategory, this character is assigned the number one ( 1 ). 4. Fourth character (number):

The numerical sequence of the EAL within the EAL subcategory.

If the subcategory has only one EAL, it is given the number one (1 ). Classification (enclosed in rectangle):

Unusual Event (U), Alert (A), Site Area Emergency (S) or General Emergency (G) EAL (enclosed in rectangle) . Exact wording of the EAL as it appears in the EAL Classification Matrix. Page 7 of 296

-:;:::::-Entergy River Bend Station EAL Basis Document Revision XXX Mode Applicability One or more of the following plant operating conditions comprise the mode to which each EAL is applicable:

1 -Power Operation, 2 -Startup, 3 -Hot Shutdown, 4 -Cold Shutdown, 5 -Refueling, DEF -Defueled, or All. (See Section 2.6 for operating mode definitions)

Definitions:

If the EAL wording contains a defined term, the defin'ition of the term is included in this section. These definitions can also be found in Section 5.1. Basis: An-EAL basis section that provides RBS-relevant information concerning the EAL as well . as a description of the rationale for the EAL as provided in NEI 99-01 Rev. 6. Reference(s):

Source documentation from which the EAL is derived 2.6 Operating Mode Applicability 1 Power Operation Reactor is critical and the mode switch is in RUN 2 Startup The mode switch is in REFUEL (with all reactor vessel head closure bolts fully tensioned) or STARTUP/HOT STANDBY 3 Hot Shutdown The mode switch is in SHUTDOWN and average reactor coolant temperature is >2QQ°F 4 Cold Shutdown The mode switch is in SHUTDOWN and average reactor coolant temperature is 200°F . 5 Refueling The mode switch is in REFUEL or SHUTDOWN with one or more reactor vel?sel head closure bolts are less than fully tensioned DEF Defueled RPV contains no irradiated fuel The plant operating mode that exists at the time that the event occurs (prior to any protective system or operator action being initiated in response to the condition) should be compared to the mode applicability of the EALs. If a lower or higher plant operating mode is reached before the emergency classification is made, the declaration shall be based on the mode that existed at the time the event occurred.

Page 8 of 296

--::::=-Entergy River Bend Station EAL Basis Document Revision XXX 3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS 3.1 General Considerations When making an emergency classification, the Emergency Director must consider all information having a gearing 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 based on loss or potential loss of Fission Product Barrier Thresholds.

EAL matrices should be read from left to right, from General Emergency to Unusual Event, and top to bottom. Declaration decisions should be independently verified before declaration is made except when gaining this verification would exceed the 15 minute declaration requirement.

Place keeping should be used on all EAL matrices.

3.1.1 Classification Timeliness NRC regulations require the licensee to establish and maintain the capability to assess, classify, and declare an emergency condition within 15 minutes after the availability of indications to plant operators that an emergency action level has been exceeded and to promptly declare the emergency condition as soon as possible following identification of the appropriate emergency classification level. The NRC staff has provided guidance on implementing this requirement in NSIR/DPR-ISG-01, "Interim Staff Guidance, Emergency Planning for Nuclear PowerPlants" (ref. 4.1.8). 3.1.2 Valid Indications All emergency classification assessments shall 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, verification could be accomplished through an instrument channel check, response on related or redundant J indicators, or direct observation by plant personnel.

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

3.1.3 Imminent Conditions For ICs and EALs that have a stipulated time duration (e.g., 15 minutes, 30 minutes, etc.), the Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time. If an ongoing radiological release is detected and the release start time is unknown, it should be assumed that the release duration specified in the IC/EAL has been exceeded, absent data to the contrary.

Page 9 of 296 River Bend Station EAL Basis Document Revision XXX 3.1.4 Planned vs. Unplann~d Events A planne~ work activity that results in an expected event or coddition which meets or exceeds an EAL does not warrant an emergency declaration provided that: 1) the activity proceeds as planned, and 2) the plant remains within the limits imposed by the operating license. Such activities include planned work to test, manipulate, repair, maintain or modify a system or component.

In these cases, the controls associated with the planning, preparation and execution of the work will ensure that compliance is maintained with all aspects of the operating license provided that the activity proceeds and concludes as expected.

Events or conditions of this type may be subject to the reporting requirements of 10 § CFR 50. 72 (ref. 4.1.4). 3.1.5 Classification Based on Analysis 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.). For these EALs, the EAL wording or the associated basis discussion will identify the necessary analysis.

In these cases, the 15-minute declaration period starts with the availability of the analysis results that show the threshold to be exceeded (i.e., this is the time that the EAL,i,nformation is first available).'

The NRG expects licensees to establish the capability to initiate and complete EAL-related analyses within a reasonable period of time (e.g., maintain the necessary expertise on-shift).

3.1.6 Emergency Director Judgment 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 EAL 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 (EGL) 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 EGL definition.

A similar provision is incorporated in the Fission Product Barrier Tables; judgment may be used to determine the status of a fission product barrier.

3.2 Classification Methodology To make an emergency classification, the user will compare an event or condition (i.e., the relevant plant indications and reports) to an EAL(s) and determine if the EAL has been met or exceeded.

The evaluation of an EAL must be consistent with the related Operating Mode Applicability and Notes. If an EAL has been met or exceeded, the associated IC is likewise met, the emergency classification process "clock" starts, and the EGL must be declared in accordance with plant procedures no later than fifteen minutes after the process "clock" started. When assessing an EAL that specifies a time duration for the off-normal condition, the "clock" for the EAL time duration runs concurrently with the emergency classification process "clock." For a full discussion of this timing requirement, refer to NSIR/DPR-ISG-01 (ref. 4.1.8). Page 10 of 296 Q "'===* Entergy River Bend Station EAL Basis Document Revision XXX 3.2.1 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 EGL 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 units, a Site Area Emergency should be declared.

There is no "additive" effect from multiple EALs meeting the same EGL. For example:

If two Alert EALs are met, whether at one unit or at two units, an Alert should be declared.

Related guidance conoorning classification of rapidly escalating events or conditions is provided in Regulatory Issue Summary (RIS) 2007 02, Clarification of lVRC Guidance for Emergency Notifications During Quickly-.-G#aAging-Events (ref. 4.1.2). 3.2.2 Consideration of Mode Changes During Classification The mode in effect at the time that an event or condition occurred, and prior to any plant or operator response, is the mode that determines whether or not an IC is applicable.

If an event or condition occurs, and results in a mode change before the emergency is declared, the emergency classification level is still based on the mode that existed at the time that the event or condition was initiated (and not when it was declared).

Once a different mode is reached,

any new event or condition, not related to the original event or condition, requiring emergency classification should be evaluated against the ICs and EALs applicable to the operating mode at the time of the new event or condition.

For events that occur in Cold Shutdown or Refueling, escalation is via EALs that are applicable in the Cold Shutdown or Refueling modes, even if Hot Shutdown (or a higher mode) is entered during the subsequent plant response.

In particular, the fission product barrier EALs are applicable only to events that initiate in the Hot Shutdown mode or higher. 3.2.3 Classification of Imminent Conditions Although EALs provide specific thresholds, the Emergency Director must remain alert to events or conditions that could lead to meeting or exceeding an EAL within a relatively short period of time (i.e., a change in the EGL is IMMINENT).

If, in the judgment of the Emergency Director, meeting an EAL is IMMINENT, the emergency classification should be made as if the EAL has been met While applicable to all emergency classification levels, this approach is particularly important at the higher emergency classification levels since it provides additional time for implementation of protective measures.

3.2.4 Emergency Classification Level Upgrading and Termination An EGL may be terminated when the event or condition that meets the classified IC and EAL no longer exists, and other site-specific termination requirements are met. As noted above, guidance concerning classification of rapidly escalating events o~ provided in RIS 2007 02 (ref. 4.1.2). Page 11 of 296

-::::::-Entergy River Bend Station EAL Basis Document Revision XXX 3.2.5 Classification of Short-Lived Events Event-based ICs and EALs define a variety of specific occurrences that have potential or actual safety significance.

By their nature, some of these events may be short-lived and, thus, over before the emergency classification assessment can be completed.

If an event occurs* that meets or exceeds an EAL, the associated ECL must be declared regardless of its continued presence at the time of declaration.

Examples of such events include an earthquake or a failure oJ the reactor protection system to automatically scram the reactor followed by a successful manual scram. 3.2.6 Classification of Transient Conditions Many of the I Cs and/or EALs employ time-based criteria.

These criteria will require that the IC/EAL conditions be present for a defined period of time before an emergency declaration is warranted.

In cases where no time-based criterion is specified, it is recognized that some transient conditions may cause an EAL to be met for a brief period of time (e.g., a few seconds to a few minutes).

The following guidance should be applied to the classification of these conditions.

EAL momentarily met during expected plant response -In instances in which 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 ATWS occurs and the high pressure ECCS systems fail to automatically start. The plant enters an inadequate core cooling condition (a potential loss of both the Fuel Clad and RCS Barriers).

If an operator manually starts a high pressure ECCS system in accordance with an EOP step and clears the inadequate core cooling 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 (process clock) 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 when an operator is able to take a successful corrective action prior to the Emergency Director completing the review and steps necessary to make the emergency declaration.

This provision is included to ensure that any public protective actions resulting from the emergency classification are truly warranted by the plant conditions.

Page 12 of 296

~==-Entergy River Bend Station EAL Basis Document Revision XXX 3.2. 7 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 (ref. 4.1.3) is applicable.

Specifically, the event should be reported to the NRC in accordance with 10 CFR § 50. 72 (ref. 4.1.4) 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.

3.2.8 Retraction of an Emergency Declaration Guidance on the retraction of an emergency declaration reported to the NRC is discussed in NUREG-1022 (ref. 4.1.3). \ Page 13 of 296 f

-::::::-Entergy River Bend Station EAL Basis Document Revision XXX

4.0 REFERENCES

4.1 , Developmental 4.1.1 NEI 99-01 Revision 6, Methodology for the Development of Emergency Action Levels for Non-Passive Reactors, ADAMS Accession Number ML 12326A805 4.1.2 4.1.3 4.1.4 4.1.5 4.1.6 4.1.7 4.1.8 RIS 2007-02 Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events, February 2, 2007. *' NUREG-1022 Event Reporting Guidelines:

10CFR50.72 and 50.73 10 § CFR 50.72 Immediate Notification Requirements for Operating Nuclear Power Reactors 10 § CFR 50. 73 License Event Report System RBS Technical Specifications Table 1.1-1, Modes RBS USAR Section 2.1 Geography and Demography , NSIR/DPR-ISG-01 Interim Staff Guidance, Emergency Planning for Nuclear Power Plants 4.1.9 RBS Emergency Plan 4.1.10 RBS USAR 9.1.2.5 Holtec HI-STORM Dry Spent Fuel Storage System ' -4.1.11 RBS USAR 9.1.4.2.3.11 Fuel Transfer System 4.1.12 OSP-0037 Shutdown Operations Protection Plan (SOPP) 4.1.13 RBS Security Plan 4.2 Implementing 4.2.1 EIP-2-001 Classification of Emergencies 4.2.2 NEI 99-01 Rev. 6 to RBS EAL Comparison Matrix . 4.2.3 RBS EAL Matrix Page 14 of 296

-::::=-Entergy River Bend Station EAL Basis Document Revision XXX 5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS 5.1 . Definitions (ref. 4.1.1 except as noted) Selected terms used in Initiating Condition and Emergency Action Level statements are set in all capital letters (e.g., ALL CAPS). These words are defined terms that have specific meanings as used in this document.

The definitions of these terms are provided below. 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 small fractions of the EPA Protective Action Guideline exposure levels. Confinement Boundary The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the RBS ISFSl, Confinement Boundary is defined as the Holtec System Multi-Purpose Canister (MPC) (ref. 4.1.10). Containment Closure The f>FeGB~ReEl-eeRe#rEms-or-actions taken to secure primary containment (pr+rAaR,f-Bf seooAEi-aFy-fer:-~J.P.)-and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.

Containment Closure is established when the containment requirements of OSP-0037 (ref. 4.1.12) are met ancl at least one integral barriw to the release of radioactive material is provided.

within the spedfied limits usina STP-057-3804.

Emergency Action Level (EAL) A pre-determined, site-specific, observable threshold for an Initiating Condition that, when met or exceeded, places the plant in a given emergency classification level. Emergency Classification Level (ECL) One of a set of names or titles established by the US Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions according to (1) potential or actual effects or consequences, and (2) resulting onsite and offsite response actions. The emergency classification levels, in ascending order of severity, are:

Noti#ea#e-R-ef-Unusual Event (NGUE)

Alert

Site Area Emergency (SAE)

General Emergency (GE) Page 15 of 296

@ -::::::-Entergy River Bend Station EAL Basis Document Revision XXX 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 require a post-event inspection to determine if the attributes of an explosion are present. f.aulteG :r-he-term applied te a steam-9 enerator tl=l-at-has a steam leak en-tJ:le--seGOOEIBFy-Siae-e-f sl::!#H:rient size to cabi-se-an-1:mcentrolleEH:lfop in $team-geAerafe.r-pressure or the steam ge-A-6fatef-ffi-9*00me-eeffi)3-!-etety depress u ri 7 ed. 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.

Fission Product Barrier Threshold A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier. Flooding A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area. General Emergency Events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility.

Releases can be reasonably I expected to exceed EPA PAG exposure levels offsite for more than the immediate site area. Hostage A person(s) held as leverage against the station to ensure that demands will be met by t~e station. Hostile Action An act toward ~l-PP-RBS or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes . attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.

Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the-N,l';2j2R8S.

Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA). Page 16 of 296

~~----~ -===* Entergy River Bend Station EAL Basis Document Revision XXX Hostile Force One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.

Imminent The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. lmoede(d}

Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely emploved).

Independent Spent Fuel Storage Installation (ISFSI} A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. Initiating Condition (IC} An event or condition that aligns with the definition of one of the four emergency classification levels by virtue of the potential or actual effects or consequences.

---f'i-Jo

...... rmal-be-vcls

--~1,.....J\s,.._,..,a+tp~~i-eal I C/EAb&,--tl:!e-rng-Aest reading ifi-ffie-f>a-&t--P,veRty-fooF-Ae-tlfS 6*B+l:lffiA§-t~peak-val-He-:

Owner Controlled Area .(OCA} For the purposes of classification, the Security Owner Controlled Area (SOCA) or the area between the SOCA Fence and the PROTECTED AREA Boundary (ref 4.1.13). Projectile An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety. Protected Area The area within the perimeter of the RBS security fence. (ref. 4.1.9). Refueling Pathway Reactor cavity (well), containment spent fuel pool, fuel transfer canal, and fuel building spent fuel pools, but not including the reactor vessel. comprise the refueling pathway (ref. 4.1.11 ). Restore Take the appropriate action required to return the value of an identified parameter to the applicable limits. Page 17 of 296 ---

-==-Entergy River Bend Station EAL Basis Document Revision XXX The-eooEl#ieR-of a stearn-generator-i-A-which prirnary-te-se~ge is of sufficieRt fflqJAitude to require a safety i-R-jection.

Safety System A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as related (as defined in 1 OCFR50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

Security Condition Any security event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A security condition does not involve a HOSTILE ACTION. Site Area Emergency Events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; (1) toward site personnel or equipment that could lead to the likely failure of or; (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA PAG exposure levels beyond the SITE BOUNDARY.

Site Boundarv For classification and dose projection purposes, the Site Boundary is the area defined as the exclusion area or exclusion zone in 1OCFR100.3 (a) which is a boundary of approximately 3.000 feet (or 0.5748 mile) from the RBS reactor centerline. (ref. 4.1. 7) Unisolable An open or breached system line that cannot be isolated, remotely or locally. Unplanned A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.

The cause of the parameter change or event may be known or unknown. Unusual Event Events are in progress or have occurred which indicate a potential degradation in the level of safety of the plant or indicate a security threat to facility protection has been initiated.

No Page 18 of 296

-*===-Entergy River Bend Station EAL Basis Document Revision XXX releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of SAFETY SYSTEMS occurs. Valid An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

Visible Damage Damage to a SAFETY SYSTEM train that is readily observable without measurements.

testing. or analysis.

The visual impact of the~ is sufficient to cause concern reqardlng the operability or reliability of the affected SAFETY SYSTEM train.Qa+1*¥'g&-te-a--~t--e-F s-trnEm:tre--#-1-ahs-reaEJ.~~ooservaB+e-wi-toout--FAeasttfements, testi R§,eF-aRa+y-s-i&-+J:ie-vi-sl:1-a~

+mpae-t--0-f-t-he-Ei-afflag e i s-s~-&a8-S8-GBHeern reg a rd ~A§}-t-he-epe-i:a-e-f+i-t-y-er--re!+ahll+ty-e-f-the a-ff-eB-t-eEl--c--0-mpe-A-effi-0-r---&ffi:tBttlfe-;-

Page 19 of 296 I

"'::::=-Entergy River Bend Station EAL Basis Document Revision XXX 5 .2 Abbreviations/Acronyms

~F .......................................................................................................

Degrees Fahrenheit 0 .....................................................................................................

*********************

Degrees AB .................................

.' .........................................................................

Auxiliary Building AC ........................................................................................................

Alternating Current AOP .................................................................................

Abnormal Operating Procedure APRM ..................................................................................

Average Power Range Meter ATWS ......................................................................

Anticipated Transient Without Scram BWR ...............................................................................................

Boiling Water Reactor BWROG ..................................................................

Boiling Water Reactor Owners Group CDE .......................................................................................

Committed Dose Equivalent CFR .....................................................................................

Code of Federal Regulations CNS ..............................................................................................................

Containment Barrier CS ..............................................................................................................................

Core Spray CTMT .............................................................................................................

Containment DEF ......................................................................................................................

Defueled OBA ...............................................................................................

Design Basis Accident DC ...............................................................................................................

Direct Current DIG .........................................................................................................

Diesel Generator ORMS .......................................................................

Digital Radiation Monitoring System EAL .............................................................................................

Emergency Action Level ECCS ............................................................................

Emergency Core Cooling System ECL. .................................................................................

Emergency Classification Level EOF .......................................

..........................................

Emergency Operations Facility ~, EOP ...............................................................................

Emergency Operating Procedure EPA ..............................................................................

Environmental Protection Agency I -EPG .....................................................

.........................

Emergency Procedure Guideline EPP .......................................................................................

Emergency Plan Procedure ERO ...........................................................................

Emergency Response Organization ESF.: .......................................................................................

Engineered Safety Feature FAA .........................................................

........................

Federal Aviation Administration FBI ...................................................................................

Federal Bureau of Investigation FEMA. ..............................................................

Federal Emergency Management Agency FSAR ....................................................................................

Final Safety Analysis Report GE .....................................................................................................

General Emergency HCTL ............................................................................

Heat Capacity Temperature Limit Page 20 of 296

~Entergy River Bend Station EAL Basis Document Revision XXX HPCS .......................................................................................

High Pressure Core Spray IC ..........................................................................................................

Initiating Condition IPEEE .................

Individual Plant Examination of External Events (Generic Letter 88-20) ISFSI. ...........................................................

Independent Spent Fuel Storage Installation Kett .........................................................................

Effective Neutron Multiplication Factor LCO ..................................................................................

Limiting Condition of Operation LER ............

........................*....................

.-......................................

Licensee Event Report LOCA .........................................................................................

Loss of Coolant Accident LPCS ........................................................................................

Low Pressure Core Spray LRW ........................................................................................................

Liquid Radwaste LWR ....................................................................................................

Light Water Re.actor MPG ...................................

Maximum Permissible Concentration/M,ulti-Purpose Canister MPH .......................................................................................

...................

Miles Per Hour mR, mRem, mrem, mREM ..............................................

milli-R9entgen Equivalent Man MSCRWL. ........................................................

Minimum Steam Colling RPV Water Level MSIV .......................................................................................

Main Steam Isolation Valve MSL ........................................................................................................

Main Steam Line MW ....................................................................................................................

Megawatt NEI. .........................................................

....................................

Nuclear Energy Institute NEIC ...................................................................

National Earthquake Information Center NESP ...................................................................

National Environmental Studies Project . NORAD .....................

.............................

North American Aerosp~ce Defense Command (NO)UE ................................................................................

Notification of Unusual Event NPP ..................................................................................................

Nuclear Power Plant NRC ................................................................................

Nuclear Regulatory Commission NSSS ................................................................................

Nuclear Steam Supply System OBE ......................................................................................

Operating Basis Earthquake ODCM .............................................................................

Offsite Dose Calculation Manual ORO .................................................................................

Offsite Response Organization PA ...............................................................................................................

Protected Area PAG .............. , ..........................................................................

Protective Action Guideline PB ....................................................................................................................

Pushbutton PCIS .....................................................................

Primary Containment Isolation System PRA/PSA .....................

Probabilistic Risk Assessment

/ Probabilistic Safety Assessment PSID .; .......................................................................

Pounds Per Square Inch Differential PSIG .. ' ..............................................................................

Pounds per Square Inch Gauge Page 21 of 296 River Bend Station EAL Basis Document Revision XXX R ........................................................................................................................

Roentgen RCS .......................................

.'. .......................................................................

RCS Barrier RCIC ..............

' ...................................................................

Reactor Core Isolation Cooling RCS ....................................

...................................

...................

Reactor Coolant System Rem, rem, REM ............................

..........................................

Roentgen Equivalent Man RETS .........................................................

Radiological Effluent Technical Specifications RHR .............................................................................................

Fesidual Heat Removal RPS ........................................................................................

Reactor Protection System RPV ...........................................................................................

Reactor Pressure Vessel RWCU ..........................................................................................

Reactor Water Cleanup

SAG .........................................................................................

Severe Accident Guideline SAP .............................................................

.........................

Severe Accident Procedure SAR ...................................

.' ...........................................................

Safety Analysis Report SBO ..........................................................................................................

Station Blackout SCBA .......................................................................

Self-Contained Breathing Apparatus SOCA ..............................................................................

Security Owner Controlled Area SPDS ............................................................................

Safety Parameter Display System SRO ............................................................................................

Senior Reactor Operator SRV ..........................

..........................................................................

Safety Relief Valve . SSE ........................................................................................

Safe Shutdown Earthquake TEDE ...............................................................................

Total Effective Dose Equivalent TAF .............................................................................

.... : ....................

Top of Active Fue.1 TSC ..........................................................................................

Technical Support Center UFSAR ...................................................................

Updated Final Safety Analysis Report USGS ............................................................................

United States Geological Survex Page 22 of 296

=-::::: Entergy River Bend Station EAL Basis Document Revision XXX 6.0 RBS-TO-NEI 99-01 Rev. 6 EAL CROSS-REFERENCE This cross-reference is provided to facilitate association and location of an RBS EAL within the NEI 99-01 IC/EAL identification scheme. Further information regarding the development of the RBS EALs based on the NEI guidance can be found in the EAL Comparison Matrix. "RBS NEI 99-01 Rev. 6 EAL IC Example EAL AU1.1 AU1 1, 2 AU1.2 AU1 3 AU2.1 AU2 1 AA1.1 AA1 1 AA1.2 AA1 2 AA1.3 AA1 3 AA1.4 AA1 4 AA2.1 AA2 1 AA2.2 AA2 2 AA2.3 AA2 3 AA3.1 AA3 1 , AA3.2 AA3 2 AS1.1 AS1 1 AS1.2 AS1 2 AS1.3 AS1 3 AS2.1 AS2 1 AG1.1 AG1 1 AG1.2 AG1 2 AG1.3 AG1 3 Page 23 of 296

~-Entergy River Bend Station EAL Basis Document Revision XXX RBS NEI 99-01 Rev. 6 EAL IC Example EAL AG2.1 AG2 1 CU1.1 CU1 1 CU1.2 CU1 2 CU2.1 CU2 1 CU3.1 CU3 1 CU3.2 CU3 2 CU4.1 CU4 1 ' CU5.1 CU5 1, 2, 3 CA1.1 CA1 1 CA1.2 CA1 2 CA2.1 CA2 1 CA3.1 CA3 1, 2 CA6.1 CA6 1 CS1.1 CS1 1 CS1.2 CS1 2 CS1.3 CS1 3 CG1.1 CG1 1 CG1.2 CG1 2 EU1.1 EU1 1 FA1.1 FA1 1 FS1 .1 ,FS1 1 FG1.1 FG1 1 HU1.1 HU1 1, 2, 3 Page 24 of 296

~::::=-Entergy River Bend Station EAL Basis Document Revision XXX RBS NEI 99-01 Rev. 6 EAL IC Example EAL HU2.1 HU2 1 HU3.1 HU3 1 HU3.2 HU3 2 HU3.3 HU3 3 HU3.4 HU3 4 HU4.1 HU4 1 HU4.2 HU4 2 HU4.3 HU4 3 HU4.4 HU4 4 HU7.1 HU? 1 HA1.1 HA1 1, 2 HA5.1 HAS 1 HA6.1 HA6 1 HA7.1 HA? 1 HS1.1 HS1 1 HS6.1 HS6 1 HS7.1 HS7 1 HG7.1 HG? 1 SU1.1 SU1 1 SU3.1 SU2 1 SU4.1 SU3 1 SU4.2 SU3 2 SU5.1 SU4 1, 2, 3 Page 25 of 296

-~ Entergy River Bend Station EAL Basis Document Revision XXX RBS NEI 99-01 Rev. 6 EAL IC Example EAL SU6.1 SU5 1 SU6.2 1 SU5 2 SU7.1 SU6 1, 2, 3 N/A SU? 1, 2 SA1.1 SA1 1 SA3.1 SA2 1 SA6.1 SA5 1 SA8.1 SA9 1 SS1.1 SS1 1 SS2.1 SS8 1 SS6.1 SS5 1 SG1.1 SG1 1 \ SG1.2 SGS 1 Page 26 of 296 Q -:;::=-Entergy River Bend 'Station EAL Basis Document Revision XXX 7.0 ATTACHMENTS 7.1 Attachment 1, Emergency Action Level Technical Bases 7.2 Attachment 2, Safe Operation

& Shutdown Areas Tables A-3 & H-2 Bases Page 27 of 296

~-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category A -Abnormal Rad Levels / Rad Effluent EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.) Many EALs are based on actual or potential degradation of fission product barriers because of the elevated potential for offsite radioactivity release. Degradation of fission product barriers though is not always apparent via non-radiological symptoms.

Therefore, direct indication of elevated radiological effluents or area radiation levels are appropriate symptoms for emergency classification.

At lower levels, abnormal radioactivity releases may be indicative of a failure of containment systems or precursors to more significant releases.

At higher release rates, offsite radiological conditions may result which require offsite protective actions. Elevated area radiation levels in plant may also be indicative of the failure of containment systems _or preclude _access to plan\ vital equipment necessary to ensure plant safety. Events of this category pertain to the following subcategories:

1. Radiological Effluent Direct indication of effluent radiation monitoring systems provides a rapid assessment mechanism to determine releases in excess of classifiable limits. Projected offsite doses, actual offsite field measurements or measured release rates via sampling indicate doses or dose rates above classifiable limits. ; 2. Irradiated Fuel Event Conditions indicative of a loss of adequate shielding or damage to irradiated fuel may preclude access to vital plant areas or result in radiological releases that warrant emergency classification.

3. Area Radiation Levels Sustained general area radiation levels which may preclude access to areas requiring continuous occupancy also warrant emergency classification.

Page 28 of 296 Q -::::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

A -Abnormal Rad Levels / Rad Effluent Subcategory:

1 -Radiological Effluent Initiating Condition:

Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer EAL: AU1.1 Unusual Event Reading on any Table A-1 effluent radiation monitor> column "UE" for ;;:: 60 min. (Notes 1, 2, 3) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.

The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Note 3: 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.

Table A-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE Main Plant Vent -Primary RE125 9.56E+08 µCi/sec 9.56E+07 µCi/sec 9.63E+06 µCi/sec 1.01 E+05 µCi/sec Main Plant Vent -RE126 1.66E-01 µCi/ml 1.74E-03 µCi/ml Secondary

II) Fuel Bldg Vent -Primary RE5A 7.75E+08 µCi/sec 7.75E+07 µCi/sec 7.75E+06 µCi/sec 6.50E+03 µCi/sec :s 0 Cl) II) Fuel Bldg Vent -Secondary RE5B 1.72E-01 µCi/ml 1.38E-03 µCi/ml ns --------(!) Radwaste Bldg Vent -RE6A 8.03E+08 µCi/sec 8.03E+07 µCi/sec 8.03E+06 µCi/sec 6.96E+04 µCi/sec Primary Radwaste Bldg Vent -RE6B --------2.12E-01 µCi/ml 1 .71 E-04 µCi/ml Secondary

E :s Liquid Radwaste RE107 ------------2 x Alarm Setpoint C" :::i Page 29 of 296

,~*Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Mode Applicability:

All Definition(s):

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

Basis: This IC addresses a potential Ef.e-&rea&e-reduction in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release).

It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.

Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment.

Further, there are administrative controls established to prevent unintentional releases, ahd to control and monitor intentional releases.

The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features arid/or controls.

Radiological effluent E 1 ALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis* of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

Classification based on effluent monitor readings assumes that a release path to the environment is established.

If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.

Releases should not be prorated or averaged.

For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL. * ( EAL #1 This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways 7 *AL #2 This EAL addressesas well as radioactivity releases that cause effluent radiation monitor readings to exceed 2 times the limit established by a radioactivity discharge permit. This EAL '.viii Such releases are typically be-associated with planned batch releases from continuous release pathways (e.g., radwaste, waste gas). EAL #3 This EAL addresses uncontrolled gaseoup or liquid releases that are detectee-by sample analyses or environmental surveys, particularly on unmonitored pathway&fe-cg., spills of radioactive liquids into storm drains, heat exchanger lea~age in river 'Nater systems, etc.). Page 30 of 296

7:.::::.-

Entergy , River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Escalation of the emergency classification level would be via IC AA 1. Reference(s):

1. SOP-0086 Digital Radiation Monitoring System 2. RSP-0008 Offsite Dose Calculation Manual 3. EP-CALC-RBS-1801 Radiological Effluent EAL Threshold Values 4. NEI 99-01 AU1 Page 31 of 296 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

A -Abnormal Rad Levels / Rad Effluent Subcategory:

1 -Radiological Efflue,rt Initiating Condition:

Release of gaseous or liquid radioactivity greater than 2 times the . ODCM limits for 60 minutes or longer. EAL: AU1.2 Unusual Event Sample analysi~ for a gaseous or liquid release indicates a concentration or release rate > 2 x ODCM limits for~ 60 min. (Notes 1, 2) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.

The Emergency Director is not allowed an additional 15 minutes fo declare after the time limit is exceeded.

Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability:

All Definition(s):

None Basis: This IC addresses a potential decrease-reduction in the level of safety of the plant as indicated by a low-level radiological release tha( exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release).

It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.

Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment.

Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases.

The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation*in these features and/or controls.

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

Classification based on effluent monitor readings assumes that a release path to the e-rwironment is esta.blished.

If the effluent flow past-an

~ed due to actions te-isolate the release path, then the effluent monitor~ longer valid for classification p-1::tFf)eSe&.-

Page 32 of 296 I

-===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Releases should not be prorated or averaged.

For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL. EAL #1 This EAb--aeElresses normally occurring continuoos-raffi0-a&t~vity releases from moofteree-ga~ffltteR.t pathVv'ay&.-

E/\L #2 This El\b--addresses raEJ.fo-acUv-i-ty-re~ea-se-s-rnat cause-e-f#Hent-r-ae+atkm-moni-tef reaEHR~ee-2--1:+mes the lim~t-e&t-ahl+sRe4-by--a-radioacti-vity discharge permit.

wi+l-t-ypi-ca!-1-y be assoc~taAAe4-bat*fl-feleases from-non cofftiRt10t1-s-release-pat-hvvays fe.g., radwaste, waste-gas}.

EAL #a-This EAL addresses uncontrolled gaseous or liquid releases that are detected by sample 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 AA1A. Reference(s):

1. RSP-0008 Offsite Dose Calculation Manual 2. NEI 99-01 AU1 ,. Page 33 of 296

-~-----------------------

River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

A -Abnormal Rad Levels / Rad Effluent 1 -Radiological Effluent Subcategory:

Initiating Condition: . Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid COE EAL: AA1.1 Alert Reading on any Table A-1 effluent radiation monitor> column "ALERT" for ;?: 15 min. (Notes 1, 2, 3, 4) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.

The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Note 2: Note 3: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.

Note 4 The pre-calculated effluent monitor values presented in EALs AA 1.1, AS 1.1 and AG1 .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Table A-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE Main Plant Vent -Primary RE125 9.56E+08 /µCi/sec 9.56E+07 µCi/sec 9.63E+06 µCi/sec 1.01 E+OS µCi/sec ' Main Plant Vent -RE126 1.66E-01 µCi/ml 1.74E-03 µCi/ml Secondary

en Fuel Bldg Vent -Primary RESA 7.75E+08 µCi/sec 7.75E+Q7 µCi/sec 7.75E+06 µCi/sec 6.50E+03 µCi/sec :I 0 Q) en Fuel Bldg Vent -Secondary RESB 1. 72E-01 µCi/ml 1.38E-03 µCi/ml c,:J --------(.!) Radwaste Bldg Vent -RE6A 8.03E+08 µCi/sec 8.03E+07 µCi/sec 8.03E+06 µCi/sec 6.96E+04 µCi/sec Primary .. Radwaste Bldg Vent -RE6B --------2.12(01 µCi/ml 1. 71 E-04 µCi/ml Secondary

E :I Liquid Radwaste RE107 ------------2 x Alarm Setpoint Cl" r Page 34 of 296 \

1::::=-Entergy River Bend Station EAL Basis Document Revision XXX

Attachment 1 -Emergency Action Level Technical Bases Mode Applicability:

All Definition(s):

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

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 (PAG§i). It includes both monitored and un-monitored releases.

Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).

Radiologica'I effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

The TEDE dose is set at 1 % of the EPA PAG of 1,000 mrem while the 50 mrem thyroid COE was established in consideration of the 1 :5 ratio of the EPA PAG for TEDE and thyroid COE. Classification based on effluent monitor readings assumes that a release path to the environment is established.

If the effluent flow past an' effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes.

Escalation of the emergency classification level would be via IC AS1. Reference(s):

1. SOP-0086 Digital Radiation Monitoring System 2. RSP-0008 Offsite Dose Calculation Manual 3. EP-CALC-RBS-1801 Radiological Effluent EAL Threshold Values 4. NEI 99-01 AA 1 Page 35 of 296

===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

A -Abnormal Rad Levels / Rad Effluent Subcategory:

1 -Radiological Effluent Initiating Condition:

Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid COE EAL: AA1.2 Alert Dose assessment using actual meteorology indicates doses > 10 mrem TEDE or 50 mrem , thyroid COE pt or beyond the SITE BOUNDARY (Note 4) Note 4: The pre-calculated effluent monitor values presented in EALs AA 1.1, AS 1.1 and AG 1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

'Mode Applicability:

All Definition(s):

SITE BOUNDARY -For classification and dose projection purposes, the Site Boundary is the area defined as the exclusion area or exclusion zone in 1OCFR100.3 (a) which is a boundary of approximately 3,000 feet (or 0.5748 mile) from the RBS reactor centerline.

~asis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1 % of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases.

Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

The TEDE dose is set at 1 % of the EPA PAG of 1,000 mrem while the 50 mrem thyroid COE was established in consideration of the 1 :5 ratio of the EPA PAG for TEDE and thyroid COE. Classification baseEl--On--e-ffiHe-A-t monito t a release path to the eA-v-ironment-~s-estabiisheEl.

If the effluent fle-w-past-aR-efflH-e-A-t-ffiefl-i-tor-~

stopped due to aet+ons to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.

Escalation of the emergency classification level would be via IC AS4AS1. Page 36 of 296

-::::=-Enterrav bJ I Reference(s):

River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases 1. EIP-2-024 Offsite Dose Calculations

2. NEI 99-01 AA1 Page 37 of 296

~Entergy River Bend Station EAL Basis Document Revision XXX /' Attachment 1 -Emergency Action Level Technical Bases .Category:

A -Abnormal Rad Levels / Rad Effluent Subcategory:

1 -Radiological Effluent Initiating Condition:

Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid COE EAL: AA1.3 Alert Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses> 10 mrem TEDE or 50 mrem thyroid COE at or beyond the SITE BOUNDARY for 60 min. of exposure (Notes 1, 2) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded:

The Emergency Dir~ctor is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability:

All Definition(s):

SITE BOUNDARY -For classification and dose projection purposes, the Site Boundary is the area defined as the exclusion area or exclusion zone in 1OCFR100.3 (a) which is a boundary of approximately 3,000 feet (or 0.5748 mile) from the RBS reactor centerline.

Basis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1 % of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases.

Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly e?{ceeds regulatory limits (e.g., a significant uncontrolled.

release).

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

-The TEDE dose is set at 1 % of the EPA PAG of 1,000 mrem while the 50 mrem thyroid COE was established in consideration of the 1 :5 ratio of the EPA PAG for TEDE and thyroid COE. Classification based on efflt.+en-t monitor readings assumes that a release path to the environment is established.

If the effluent flmv past an effluent monitor is knovm to have Page 38 of 296 e -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases stop!}Ba-ooe-te-aeti on s to i so~e-t:A-e-relea-se--f*HA,fRBR-tRe-e#l-tl e nt mcmitof-Few+Ag-i-s-oo ieft§&F-Va~ia-fo r cl a ssifisatiBn-pHFf}OSe&.-

Th is EAL is assessed per the ODCM (ref. 2) Escalation of the emergency classification level would be via IC AS4-AS1. Reference(s):

1. EIP-2-024 Offsite Dose Calculations

2. RSP-0008 Offsite Dose Calculation Manual 3. NEI 99-01 AA1 Page 39 of 296

-===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases -Category:

A -Abnormal Rad Levels / Rad Effluent Subcategory:

1 -Radiological Effluent Initiating Condition:

Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid COE EAL: AA1.4 Alert Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:

Closed window dose rates > 10 mR/hr expected to continue for~ 60 min.

Analyses of field survey samples indicate thyroid COE > 50 mrem for 60 min. of -inhalation. ,r (Notes 1, 2) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.

The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

' Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability:

All Definition(s):

SITE BOUNDARY -For classification and dose projection purposes, the Site Boundary is the area defined as the exclusion area or exclusion zone in 1OCFR100.3 (a) which is a boundary of approximately 3,000 feet (or 0.5748 mile) from the RBS reactor centerline.

Basis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1 % of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases.

Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled rel~ase).

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

The TEDE dose is set at 1 % of the EPA PAG of 1,000 mrem while the 50 mrem thyroid COE was established in consideration of the 1 :5 ratio of the EPA PAG for TEDE and thyroid COE. Page 40 of 296

--===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases ---C-la-.s-s~fkr-Jfie.'1-1:3-aseEl-efl-e#!Hent moRi-t-&Headings assumes-that a release path to the e-A-ViffiAn1ent-i-s-estael+she&.-l-f-#:le effl ueRt-#0W-J3ast-an-efnl:l&Rt-mRitef-is-Kt10WR-ffi-AaVe

~EH:H::1~0late the re-1-eas * * ~e longer va l+El-fer-e~f#satioo-13Hf13GSe&.-

Escalation of the emergency classification level would be via IC Ag4Afil. Reference(s):

1. EIP-2-014 Offsite Radiological Monitoring

2. NEI 99-01 AA 1 Page 41 of 296 I I River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

A -Abnormal Rad Levels / Rad Effluent Subcategory:

1 -Radiological Effluent Initiating Condition:

Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid COE EAL: AS1.1 Site Area Emergency Reading on any Table A-1 effluent radiation monitor> column "SAE" for :::: 15 min. (Notes 1, 2, 3, 4) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.

The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Note 3: 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.

Note 4: The pre-calculated effluent monitor values presented in EALs AA 1.1, AS 1.1 and AG1 .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Table A-1 Effluent Monitor Classification Thresholds Release Point I Monitor I GE I SAE I Alert I UE Main Plant Vent -Primary RE125 9.56E+08 µCi/sec 9.56E+07 µCi/sec 9.63E+06 µCi/sec 1.01 E+05 µCi/sec Main Plant Vent -RE126 1.66E-01 µCi/ml 1.74E-03 µCi/ml Secondary

UJ Fuel Bldg Vent -Primary RE5A 7.75E+08 µCi/sec 7.75E+07 µCi/sec 7.75E+06 µCi/sec 6.50E+03 µCi/sec ::J 0 Q) UJ Fuel Bldg Vent -Secondary RE5B 1.72E-01 µCi/ml 1.38E-03 µCi/ml nl --------(!) Radwaste Bldg Vent -RE6A 8.03E+08 µCi/sec 8.03E+07 µCi/sec 8.03E+06 µCi/sec 6.96E+04 µCi/sec Primary Radwaste Bldg Vent -RE6B --------2.12E-01 µCi/ml 1. 71 E-04 µCi/ml Secondary

E ::J Liquid Radwaste RE107 ------------2 x Alarm Setpoint C" ::::i Page 42 of 296 I

( ~Entergy \ River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Mode Applicability:

All Definition(s):

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

Basis: This IC addresses a release of gaseous radioactivit/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. Radiological effluent EALs are also included to provide a basis for classifyipg events and conditions that cannot be readily or appropriately classified on the basis of plant conditions I 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 COE was established in consideration of the 1 :5 ratio of the EPA PAG for TEDE and thyroid COE. 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 mon.itor reading is no longer VALID for classification purposes.

Escalation of the emergency classification level would be via IC AG1. Reference(s):

1. SOP-0086 Digital Radiation Monitoring System 2. RSP-0008 Offsite Dose Calculation Manual 3. EP-CALC-RBS-1801 Radiological Effluent EAL Threshold Values 4. NEI 99-01 AS1 Page 43 of 296

-==::-, Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

A -Abnormal Rad Levels / Rad Effluent Subcategory:

1 -Radiological Effluent Initiating Condition:

Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid COE EAL: AS1.2 Site Area Emergency Dose assessment using actual meteorology indicates doses> 100 mrem TEDE or 500 mrem thyroid COE at or beyond the SITE BOUNDARY (Note 4) Note 4: The pre-calculated effluent monitor values presented in EALs AA 1.1, AS 1.1 and AG1 .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Mode Applicability:

All Definition(s):

SITE BOUNDARY -For classification and dose projection purposes, the Site Boundary is the area defined as the exclusion area or exclusion zone in 1OCFR100.3 (a) which is a boundary of approximately 3,000 feet (or 0.5748 mile) from the RBS reactor centerline.

Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases.

Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public. 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 do~e is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid COE was established in consideration of the 1 :5 ratio of the EPA PAG for TEDE and thyroid COE. ---'Clc-,f-;-lat!-is"""smsa#on based on e#H:!ent momtOHea4ngs ass1:1-me-s-that a release path-te-#te environment is established.

lf'the effl-Hem-#o-wf,ast an effluent monitor is knovvn to have s-tof)ped due to actiens to isolate ti:=le releasB-fJati:=l, then ti:=le effluent moniteF-reading is no longer valid for classification purpe-se&.-

Escalation of the emergency classification level would be via IC AG1. Page 44 of 296

~--===* Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Reference(s):

1. EIP-2-024 Offsite Dose Calculations

2. NEI 99-01 AS1 Page 45 of 296

-~-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases , Category:

Subcategory:

Initiating Condition:

EAL: A -Abnormal Rad Levels / Rad Effluent 1 -Radiological Effluent Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid COE r AS1 .3 Site Area Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:

Closed window dose rates> 100 mR/hr expected to continue for~ 60 min.

Analyses of field survey samples indicate thyroid COE > 500 mrem for 60 min. of inhalation. (Notes 1, 2) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.

The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability:

All Definition(s):

SITE BOUNDARY -For classification and dose projection purposes, the Site Boundary is the area defined as the exclusion area or exclusion zone in 1OCFR100.3 (a) which is a boundary of approximately 3,000 feet (or 0.5748 mile) from the RBS reactor centerline.

Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases.

Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public. 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 COE was established in consideration of the 1 :5 ratio of the EPA PAG for TEDE and thyroid COE. Classification based on effluent monitor readings assumes that a release path to the Page 46 of 296

-===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases efWi.r-OAment is establiSfl6*1.,-Jf the effk!--eAHffiW--f}a&t--ar+e#ltieRt-+nonitor is knowR-4e-A-ave s-t9ffl':)e

* -te--the-release-pa#+,#1-eR-the efflooFlt--FFl&RHer reading is no le-Rgef-Val+EHo r c! a ssifiBatisr+~rpose&.-

Esca la tio n of the emergency classification level would be via IC AG1. Reference(s):

1. El P-2-014 Off site Radiological Monitoring

2. NEI 99-01 AS1 Page 47 of 296 /'

-===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

A -Abnormal Rad Levels / Rad Effluent Subcategory:

1 -Radiological Effluent Initiating Condition:

Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid COE EAL: AG1.1 General Emergency Reading on any Table A-1 effluent radiation monitor> column "GE" for 15 min. (Notes 1, 2, 3, 4) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.

The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Note 3: 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.

Note 4: The pre-calculated effluent monitor values presented in EALs AA 1.1, AS 1.1 and AG1 .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Table A-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE Main Plant Vent -Primary RE125 9.56E+08 µCi/sec 9.56E+07 µCi/sec 9.63E+06 µCi/sec 1.01 E+05 µCi/sec Main Plant Vent -RE126 1.66E-01 µCi/ml 1.74E-03 µCi/ml Secondary

U) Fuel Bldg Vent -Primary RE5A 7.75E+08 µCi/sec 7.75E+07 µCi/sec 7.75E+06 µCi/sec 6.50E+03 µCi/sec ::I 0 Q) U) Fuel Bldg Vent -Secondary RE5B 1. 72E-01 µCi/ml 1.38E-03 µCi/ml ra --------(!) Radwaste Bldg Vent -RE6A 8.03E+08 µCi/sec 8.03E+07 µCi/sec 8.03E+06 µCi/sec 6.96E+04 µCi/sec Primary Radwaste Bldg Vent -RE6B --------2.12E-01 µCi/ml 1.71 E-04 µCi/ml Secondary

5! ::I Liquid Radwaste RE107 ------------2 x Alarm Setpoint C" :J Page 48 of 296

===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Mode Applicability:

All Definition(s):

VALID -An indication, report'. or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. , Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases.

Releases of this magnitude will require implementation of protective actions for the public. 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 COE was established in consideration of the 1 :5 ratio of the EPA PAG for TEDE and thyroid COE. 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.

Reference(s):

1. SOP-0086 Digital Radiation Monitoring System 2. RSP-0008 Offsite Dose Calculation Manual 3. EP-CALC-RBS-1801 Radiological Effluent EAL Threshold Values 4. NEI 99-01 AG1 Page 49 of 296 Q -::=::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

A -Abnormal Rad Levels / Rad Effluent Subcategory:'

1 -Radiological Effluent Initiating Condition:

Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid COE EAL: AG1.2 General Emergency Dose assessment using actual meteorology indicates doses> 1,000 mrem TEDE or 5,000 mrem thyroid COE at or beyond the SITE BOUNDARY (Note 4) Note 4: The pre-calculated effluent monitor values presented in EALs AA1 .1, AS1 .1 and AG1 .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Mode Applicability:

All Definition(s):

SITE BOUNDARY -For classification and dose projection purposes, the Site Boundary is the area defined as the exclusion area or exclusion zone in 1OCFR100.3 (a) which is a boundary of approximately 3,000 feet (or 0.5748 mile) from the RBS reactor centerline.

Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases.

Releases.

of this magnitude will require implementation of protective actions for the public. 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 COE was established in consideration of the 1 :5 ratio of the EPA PAG for TEDE and thyroid COE. ---+-C.,....la....-.s..-sifisa#on based on effi.Hem-FROR-i-tef-rea4ngs assUffles that a release path to the e-A-V+rooment is established.

If the effl.He11-t-f!&11v past an effluent monitor is known to have ~ue to actions to isolate the release path, then the effiueRt-rnoniror reading is no lORfJ8-f-VaH-E!-1'ofclassification purposes.

Reference(s):

1. EIP-2-024 Offsite Dose Calculations Page 50 of 296 I -===-Entergy 2. NEI 99-01 AG1 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Page 51 of 296

-:::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

A -Abnormal Rad Levels / Rad Effluent Subcategory:

1 -Radiological Effluent ' Initiating Condition:

Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid COE EAL: AG1 .3 General Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:

Closed win~fow dose rates> 1,000 mR/hr expected to continue for;::: 60 min.

Analyses of field survey samples indicate thyroid COE> 5,000 mrem for 60 min. of inhalation. (Notes 1, 2) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be. exceeded.

The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability:

All Definition(s):

SITE BOUNDARY -For classification and dose projection purposes, the Site Boundary is the area defined as the exclusion area or exclusion zone in 1OCFR100.3 (a) which is a boundary of approximately 3,000 feet (or 0.5748 mile) from the RBS reactor centerline Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases.

Releases of this magnitude will require implementation of protective actions for the public. 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 COE was established in consideration of the 1 :5 ratio of the EPA PAG for TEDE and thyroid COE. Page 52 of 296

-~=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases -G~a&S-i-fiB-affon based eR-effl-ueHt-menitor readiAgs asstlffles that a Fe~e-too eAvi-reF1me-F1HS-&stablisheEl,-l-f-ttte-Bffl-Hera-f.lew past an effl-uent-mo:n1tor is kRBWA-4e-Rave s-te-we*1--Gu-e-to actions to iso+at-e the release path, then the-e-ffiHe-nt-moHfter readin-g-f-s-oo kmge-F-valid for classification pHfP&se&.

Reference(s):

1. EIP-2-014 Offsite Radiological Monitoring

2. NEI 99-01 AG1 Page 53 of 296 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

A -Abnormal Rad Levels / Rad Effluent Subcategory:

2 -Irradiated Fuel Event Initiating Condition:

UNPLANNED loss of water level above irradiated fuel EAL: AU2.1 Unusual Event UNPLANNED water level drop in the REFUELING PATHWAY as indicated by level instrumentation, low water level alarm or visual observation.

AND UNPLANNED rise in corresponding area radiation levels as indicated by any of the following radiation monitors:

RMS-RE140 Refueling Floor Near North Entrance

RMS-RE141 Refueling Floor Near South Entrance

RMS-RE192 Fuel Building Operating Floor -South

RMS .. RE193 Fuel Building Operating Floor -North Mode Applicability:

All Definition(s):

UNPLANNED-.

A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.

The cause of the parameter change or event may be known or unknown. REFUELING PATHWAY-Reactor cavity (well), containment spent fuel pool, fuel transfer canal, and fuel building spent fuel pools, but not including the reactor vessel, comprise the refueling pathway. Basis: This IC addresses a decrease-drop 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 contrpl radiation levels within the plant. It is therefore a potential degradation in the level of safety of the plant. *, A water level decrease-drop will be primarily determined by indications from available level instrumentation.

Other sources of level indications may include reports from plant personnel (e.g., from a refueling crew) or video camera observations (if available).

A significant drop in J the water level may also cause aA tAGre-ase-rise in the radiation levels of adjacent areas that 1 can be detected by monitors in those locations.

Page 54 of 296

-====-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases The effects of planned evolutions should be considered.

For example, a refueling bridge area radiation monitor reading may iBB-Fease-rise 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. The following low level alarms on P870 are related to low level in the REFUELING PATHWAY (ref. 4): o SFP low alarm #0111 (H13-P870

/ 56,q / E02) o Uooer Transfer Pool Low alarm #0336 (H13-P870

/ 56A / E03l o Cask Pool Low alarm #0337 (H'l3-P870 I 56A / 003) o Lower Transfer Pao! low alarm #0335 (H13-P870

/ 56A / F03) o Rx Bldo Storaae Pool low alarm #0'112 (H13-P870

/ 56A / H03) 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 AA2. Reference(s):

1. AOP-0027 Fuel Handling Mishaps 2. USAR 12.3 Table 12.3-1 Area Direct Radiation Monitor Locations

3. USAR 9.1.4.2.3.11 Fuel Transfer System 4. ARP-870-0034 P870-56 Alarm Response 5. NEI 99-01 AU2 Page 55 of 296

.~ -;::==-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

A -Abnormal Rad Levels / Rad Effluent Subcategory:

2 -Irradiated Fuel Event Initiating Condition:

Significant lowering of water level above, or damage to, irradiated fuel EAL: AA2.1 Alert IMMINENT uncovery of irradiated fuel in the REFUELING PATHWAY Mode Applicability:

All Definition(s): , CONFINEMENT BOUNDARY-The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the RBS ISFSI, Confinement Boundary is defined as the Holtec System Multi-Purpose Canister (MPC). IMMINENT -The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. REFUELING PATHWAY-Reactor cavity (well), containment spent fuel pool, fuel transfer canal, and fuel building spent fuel pools, but not including the reactor vessel, comprise the refueling pathway. 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 Sf)BRHOOt-f*)e+REFUELI NG PATHWAY (see Developer l\lotes).

These events present radiological safety challenges to -' plant personnel and are precursors to a release of radioactivity to the environment.

As such, they represent an actual or potential substantial degradation of the level of safety of the plant. 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--t4U1. ----;Essalation of the emergency would be baseE!-oA-etther Reeo§Ri-t+en Categ91y-A-ef-G EAL#1 This EAL escalates from AU2J_ 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-Page 56 of 296 \

e -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases off curve). Classification of an event using this EAL should be based on the totality of available indications, reports and observations.

While an area radiation monitor could detect aPt tRerease rise in a dose rate due to a lowering of water level in some portion of the REFUELING PATHWAY, the reading may not be a reliable indication of whether or not the fuel is actually uncovered.

To the degree possible, readings should be considered in combination with other available indications of inventory loss. A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance

'Nith Recognition Category C during the Cold Shutdown and Refueling modes.-c*A~

This E,L\L addresses a release of rae+EHIBtWe material caused by trradi-a-ted-ftteh-Q.a~RtS-may-1RBJ.uEl-e-tfte-El-ro-~F*ft§.OOffif*A~e+A-§-Bf-a*R assem&l.y,e.r-e-ro-pping a heavy load onto an assembly;-A-rise-~R-Feadings on radiation meRito-r&BJ:i~e-ooRSiEl ered in conjHRetioo-w#~o-rts-ef-G&ser-vatien&-ef-a porential-fuel-aarna§tR§-eveA-f-t&.§r-.a-ruet-hBFIGJ.i~dem}.

---Ercl\b-#'J

---S-p*em-ruel-poe+-\,vatef--!Bvel at t-ru&-\;afHB--is-w+tAtR--ffie-+EPNef-6RB-e-f-ffi*--+evel-FaR§B f1-0*e&&afy'-te-p.revent si g nifi cant dose con seEtt1BRCe&4rem-ei rect g f)eff&rmmg-e-~y-e-f--me-s-pen-HBel--po&h-Tui&-ee-nElilio n reflects a sig nifi-cant l-oss of spent fuel pee+-water inventory anEl--tttt!s it is also a 13f0CtlfSor to a loss of the ability-to 006EJt!atB~y-Bool-m&tffa*l+ateEl-ruel-assem&i.e&--stored-+A-t-Re-f}OOl-:-

Escalation of the emergency classification level would be via ICs AS1~see-A-S2 9evelepef-Nete&J.

Reference(s):

1. NEI 99-01 AA2 Page 57 of 296

,:::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

A -Abnormal Rad Levels / Rad Effluent

Subcategory:

2 -Irradiated Fuel Event Initiating Condition:

Significant lowering of water level above, or damage to, irradiated fuel .EAL: AA2.2 Alert Damage to irradiated fuel resulting in a release of radioactivity AND High alarm on any Table A-2 radiation monitor Table A-2 Fuel Damage Radiation Monitors

RMS-RE140 Refueling Floor Near North Entrance

RMS-RE141 Refueling Floor Near South Entrance

RMS-RE192 Fuel Building Operating Floor -South

RMS-RE193 Fuel Building Operating Floor -North

RMS-RE5A(B)

Fuel Building Ventilation Exhaust Mode Applicability:

All Definition(s

): CONFINEMENT BOUNDARY-The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the RBS ISFSI, the Confinement Boundary is comprised the Holtec System Multi-Purpose Canister (MPG). Basis: This G-EAL addresses events that have caused i-mm~nent or actual damag_e to an irradiated fuel assembly, or a significant lowering of vvater level within the spent fuel peel (see Developer Nete-s). These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment.

As such, they represent an actual or potential substantial degradation of the level of safety of the plant. This lG-EAL 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-FtU1. Escalation of the -e-A-e+th er Re cognition Gateger-y-A--e-r-G Page 58 of 296

Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases EAL #Thi-s-*Ah-esca:ates freffi-A~at-me-loss of level,in the affeete&fffiRffi-A-OHhe REFU Eb+NG-PA-i:FI-VVAY, ~s of suff1BteA-t-magrntl:!Ele-t~~--+Acevery-e-f-kradiatod flleh-1-ndicatioA-S---&f-tFr-aEi+ated-fHel uncoveiy-may-iRGl-ude direct oF-i-Roirect viSl::1-ai-eB-servatieR (e.g., reports freFR-f}eF-SOfH1e-l-or-eamera-+mages), as well as significant changes in v,ateF-aoo ffif#ation levels, or other pl-am-par-ameter-&.-GemfcH+rational aias may al-s~~e.g., a eff-&1:1-FV&).

Glassificaoo-H-e-f-an event llSiAg-m+s-EAL should be-eased on me-toral+ty-of-avaHaele indicati-e-Rs,re~ooser-vatiefls-:-

---\J\_,,_/, h-ite--an-a-Fea-radiatten--FOOflttof-oot1ta-detect an iflGFea-Se in a dose --i:ate due te-a +eweri-R~r level ~A-e-f-tqe-REFUEll-NG-PA-1=~/\Y, the a reliable indication of-vvhether or not the fuel is actually uncov~e-degree possible, readings SAetH-El-be considered in come+na-HHA-v.4#1-ether availa&le-i-AEHO-atiefl-s-of inventery-4e-s&.-

---1fa'-l-,-t+dffii:H~ove irraEi+ated fuel wi~rnactor v~ssi-fi.e*1-iR oocefE!anee-R-eee§flffi9fl-GatO§ery-G-*1-uritt§-tl:te-Get*i--&oot~l+ng-meEk.,ns:-

This EAL addresses a release of radioactive material caused by mechanical damage to irradiated fuel. Damaging events may include the dropping, bumping or binding of an assembly, or dropping a heavy load onto an assembly.

A rise in readings on 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).

Ei\L #3Spent fuel--f*}E>>-wateF

+evel at this value is withifl--the-tewer end-of the leve1--Fa-A§e necessary to prevent signmeam oose-BORs~ff-f>fFI-El+reet-gai;nma--Faa-iation to perseA-Ae+-pBffe-rmi-Rg-eperations iR--tfie viBi-Rit-y of the spent fue+-pe~oRElition reft-eet-s-a-~eant-~l-J3eel--wateF

+nventory an*1--thus-+t-i-~ef40-a-+e-ss-~~lity to fuel assemWes stored in the--~ Escalation of the emergency classification level would be via I Cs AS+-AfiloF

/l.S2 (see AS2 f)e-~. Reference(s):

1. AOP-0027 Fuel Handling Mishaps 2. USAR 12.3 Table 12.3-1 Area Direct Radiation Monitor Locations

3. USAR 12.3 Table 12.3-2 Airborne Process and Effluent Radiation Monitors 4. NEI 99-01 AA2 Page 59 of 296 I/ e *::::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

A -Abnormal Rad Levels / Rad Effluent Subcategory:

2 -Irradiated Fuel Event Initiating Condition:

Significant lowering of water level above, or damage to, irradiated fuel EAL: AA2.3 Alert Lowering of spent fuel pool level to 108.0 ft. (Level 2) on SFC-Ll29A/B Mode Applicability:

All Definition(s

): None Basis: This lG-EAL addresses events that have caused fu-e~-asseffib+y,e-F-a significant lowering of water level within the spent fuel pool-fsee-Qeveloper Netes). These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment.

As such, they represent an actual or potential substantial degradation of the level of safety of the plant. This IC applies-to irradiated-rue!

that ls-l-i*ensed for~~-e~t-the

~eaded stora§e cask is seale4-GR-*e-Seale4,Elamage te-a--~ed cask causing loss of the CONFINEMENT BG~QARY is e+assified in accer-*1-artse-w+th IC E Hl:J4:---E-s-ealaooA-e+-tAB-er'11e-F§BRey-weuk:l-l:Je-b-asee-e-A-eithef-R:eeeg-A-i#EtA-GategGPj-,l\:er-G lG-~Ab-#+-h-is EAL esc-alates-fmm-Af.J&in that the loss of level, in the affected portieR-GHhe REFUELING-PA-Ž-WAY, is of stJfflBteA-t magniruee to have resulted iA-l:fneevery of irmetatea fuel. Indications of irraeiated--f1::1BI uncovery--may inclu-d-e-El+ree:t-or indirect vist1a-l-el3-ser:vati-oo fe.g., report&-froFFl-f}erse-nnel-0r-GarRera-i-mages), as well as signifi-eant-changes in-water and faGiatiB-A-levels, or otheF-f*&At-f:IBramffiefS~t+ooal-a+El-s-may also be o#-&u-rve).

Classi#ca#eA-Bf-aR-evem-using this EAL should be-eased on tRe-tet-aUty-ef-avatlable tnElications, reports and-observation&:

While an area radiation memteF-GetHd-El-etect an increase in a dose rate-due to a +eWer~ng of water level in some portioo-ef-t-he-R~i;;:.LJ-~J-G-PA+HWA-Y,--#:le readiA-§-may--f!e-t-ee a-rel+ahle-tAEH*a-H~et--U-1e-ruel is actuaUy-unoo-vered.

To the-eeg-ree-pesstble, -reaeings should be consi-deffid-i-A-OOffiBtflatie-A-wHA--ether available indications of inven-ter-y4o&&

/'-. drop in water level above irrad+ated-fuel-w+thin the reactor vessel may be classifted-tn accordance Recognition Cate§ory C during the Cold Shutdown and Refueling modes. Th~es a release-ef-FaEHe-active material caused by mechaniea-1-damage--te irradiated fuel. Damag+Rg events may in~pping, bumping or bindirig of an Page 60 of 296

=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases ~~tteavy-+e-a*1--onto an mBflOOfS-Sho uld be-B-ens+eeree-m-c-eAjuoc-tioo-w#A-+fl-f

)la nt ref*}~ser-vat+oo-s-e:f-a petootial-:f'l::let-*1-arnagi ng event (e.g., a fae~FtE!-l+Rg-aw~

EAL #3Spent fuel pool water level at this value is within the lower end of the level range necessary to prevent significant dose consequences from direct gamma radiation to personnel performing operations in the vicinity of the spent fuel pool. This condition reflects a significant loss of spent fuel pool water inventory and thus it is also a precursor to a loss of the ability to adequately cool the irradiated fuel assembles stored in the pool. Escalation of the emergency classification level would be via I Cs AS1 or AS2-fsee-A&2 neveloper l\lofc, C' \ t:7 1 I nr r....,,----;;::ry.

9 ny fuel rack providing added marqin . . Buildinq on the lnterior of the West exterior wall (ref. ?). Page 61 of 296

-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Reference(s):

1. RBG-47570 Completion of Required Action by.NRG Order EA-12-051.Reliable SFP Instrumentation

2. RBS-FSG-011 Alternate Spent Fuel Pool Makeup and Cooling 3. NEI 99-01 AA2 ) Page 62 of 296 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

Subcategory:

A -Abnormal Rad Levels / Rad Effluent 2 -Irradiated Fuel Event Initiating Condition:

Spent fuel pool level at the top of the fuel racks EAL: AS2.1 Site Area Emergency Lowering of spent fuel pool level to 86.0 ft. (Level 3) on SFC-Ll29A/B Mode Applicability:

All Definition(s):

IMMINENT -The trajectory of events or conditions is suc.h that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. Basis: This .i.G-EAL addresses a significant loss of spent fuel pool inventory control and makeup capability leading to IMMINENT fuel damage. This condition entails major failures of plant functions needed for protection of;the public and thus warrant a Site Area Emergency declaration.

It is recognized that this IC would likely not be met until well after another Site Area Emergency IC was met; however, it is included to provide classification diversity.

Escalation of the emergency classification level would be via IC AG1 or AG2:AG2. (SFC-Ll29/.VB) capable of identifyina normal level (Level i ). SFP level approximately 23 ft. above the top of the fuel racks. (Level 2) 107 ft. 10 5/16 in. (rounded to 108.0 ft.) which is that level adequate to provide substantial radiation shielding for a person standinq on the SFP operating deck, and SFP level at the too of the fuel racks (Level 3) 85 ft. 10 5/'l 6 in. (rounded to 86.0 ft.) {ref. 1 ). RBS uses a Level 3 of approximately one foot above the highest point of any fuel rac!, providing added marqin. Soent Fuel Pool Level indicators SFC-U29A and B are read on the 98 ft. elevation Control Bui!dinq on the interior of the West exterior wall (ref. 2). Reference(s):

1. RBG-47570 Completion of Required Action by NRG Order EA-12-051 Reliable SFP Instrumentation

2. RBS-FSG-011 Alternate Spent Fuel Pool Makeup and Cooling 3. NEI 99-01 AS2 Page 63 of 296 e -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

A -Abnormal Rad Levels / Rad Effluent Subcategory:

2 -Irradiated Fuel Event Initiating Condition:

Spent fuel pool level cannot be restored to at least the top of the fuel racks for 60 minutes or longer EAL: AG2.1 General Emergency Spent fuel pool level cannot be restored to at least 86.0 ft. (Level 3) on SFC-Ll29NB for 60 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.

The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Mode Applicability:

All Definition(s):

None Basis: This lG-EAL 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.

Post-Fukushima order EA-12-05'1 required the installation of reliable SFP level indication (SFC-Ll29A/B) capable of identifying normal level (Level *1 ), SFP level approximately

?3 ft. above the top of the fuel racks. (Level 2) *107 ft. 10 5/16 in. (rounded to 108.0 ft.) whicr1 is that level adequate to provide substantial radiation shielding for a person standina on the SFP ooeratinq deck, and SFP level at the top of the fuel racks (Level 3) 85 ft. -10 5/16 in. (rounded to 86.0 ft.) (ref. 1 ). RBS uses a Level 3 of approxirnatelv one foot above the hiqhest point of any fuel rack providing added margin. Spent Fuel Pool Level indicators SFC-Ll29A and B are read on the 98 ft. elevation Control Buildina on the interior of the West exterior wall (ref. 2). Reference(s):

1. RBG-47570 Completion of Required Action by NRC Order EA-12-051 Reliable SFP Instrumentation

2. RBS-FSG-011 Alternate Spent Fuel Pool Makeup and Cooling 3. NEI 99-01 AG2 Page 64 of 296

---::::.=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

A -Abnormal Rad Levels / Rad Effluent Subcategory:

3 -Area Radiation Levels Initiating Condition:

Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown EAL: AA3.1 Alert Dose rate > 15 mR/hr in EITHER of the following areas:

Control Room (RMS-RE170)

Central Alarm Station (by survey) Mode Applicability:

All Definition(s):

IMPEDE(D)

-Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).

Basis: Areas that meet this threshold include the Control Room (CR) and the Central Alarm Station (CAS). The Control Room is monitored for excessive radiation by one detector.

RMS-RE170 (ref. i ). The CAS is included in this EAL because of its importance to permitting access to areas required to assure safe plant operations.

There are no permanently installed area radiation monitors in CAS that may be used to assess this EAL tr,reshold.

Therefore, this threshold is evaluated usina local radiation survey for this area. 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 i-R&re-as-ee-rise in radiation levels and determine if another IC may be applicable._For EAL #2, an Aleft-El-ec-laratiefH&warranted if emry-+me-#te-affestoo room/area is, or may be, f}FB-CeE:itl-rally reql:ltfeEi-d-Hfi-ng the p~t operating mode in effect at the time of the elevated rad-i-a-He-n levels. The emergeney-classification is whe-tAB-r entry is actl:lBUy-necessary at the time of the incre-as-ed-Fadiation levels. /\ccess should be considered as ~oded if extraordinary measures are necessary to facilitate entry-of f:}efSOnnel iAtO-t-Ae affecteEl-roefB/area (e.g., install~ng-temporary shielding, requiring use of Page 65 of 296

-~=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases OOR-Fet:Hine protoetiv-e-ef!*Mpment,reqHestiH§-&fl-vXtension in dose lim~ts beyond Aefma+ administrative limi-ts}.-

AA-emeFgency declaration is oot-warraflte&-~evvifi.g--ooAffitions apply. 111-+he-pl-ant is in an operatin§-mode different-than the mode specified for-#te-a#ee-ted room/area (i.e., entry is not ffi*1l:H-FBEI--G1:.1r~eE!*-+R-effeet at the time of the elevated radiation le~he plant is in Mode 1 when the raeiation increase occurs, and-ffie-f)ffi~d for nOfrnal-ef}erafioo,roetfiewA-arn:l-SP.ti:fEl-ewA-ae-ABt require entry into the affected room uRtH--Mede-4.-

11-+he increased-radia~are a resul~J:iat includes compensat9fY -measures vvhich address the-tempora-ry-i-naecessibilfty of a room or area (e.*r-,fadiograp-hy, spent fil.ter or resin-transfer, etc.). ia-+he-action for 1 Nhich room/area erni:y-i-s-FeqH~r~trative-e-F-f600fe-k-eefRg Aature (e.g., normal rounds-er-rootine inspecttOftS}.- llf-+he access co.ntrol measures are of a cOfts-ervati-ve or precautioi,ary nature,--aflB--WOOki-Ret acruaH-y-prevent or-impede a reqHtred-actioA-c Escalation of the emergency classification level would be via Recognition Category A, C or F ICs. Refe re nee( s): 1. USAR 12.3 Table 12.3-1 Area Direct Radiation Monitor Locations

2. NEI 9~-01 AA3 Page 66 of 296

-::==-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: A -Abnormal Rad Levels / Rad Effluent Subcategory: 3 -Area Radiation Levels Initiating Condition: Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown EAL: AA3.2 Alert I An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to any Table A-3 room or area (Note 5) Note 5: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted. Table A-3 Safe Operation & Shutdown Rooms/Areas Room/Area Auxiliary Building 70' RHR B Pump Room Auxiliary Building 80' RHR A Pump Room Auxiliary Building 114' West Control Building 95' Div 1 RSS Room Mode Applicability: 3 -Hot Shutdown Definition(s): Mode 3. 3 3 3 IMPEDE(D) -Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed). 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 radi~tion 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 increaseEl-rise in radiation levels and determine if another IC may be applicable. Page 67 of 296 ~==-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases For EAL #2 AA3.2, 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 actually necessary at the time of the inc:-easee-higher 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 emer~ency 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

+Rerea&e-rise occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4~.

The +Berea-see-higher 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.). o--Toe-astlB-fl-fe.f-wl:ti-sh roe mf'.ar-:ea-entr-y-~&-ffitru+Fe*H&*Of-a n ad mi
OOf*ng nalHFe-(&.-fr;,Re-r-FAal--F8H*R*t&Bf-fOOti-Ae~tiBHSj-;*
The access control measures are of a conservative or precautionary nature, and would not actually prevent or IMPEDE a required action. Escalation of the emergency classification level would be via Recognition Category A, C or F ICs. lf 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 alreadv allowed by Technical Specifications at the time of the event. The list of plant rooms or areas with entry-related mode applicabiiitv identified specify those rooms or areas that contain equipment vvhich require a manual/iocal action as SReci'fied in operating procedures used for normal plant operation, cooldown and shutdown. Rooms or areas in which actions of a continqent or emergency nature would be performed (e.o., an action to address an off-normal or emeraency condition such as emergency repairs, corrective measures or emergency operations) are not included. In addition, the list specifies the plant mode(s) during which entry would be required for each room or area (ref. 1 ). FAL AA3.2 mode applicability has been limited to the mode limitations of Table A-3 (Mode 3 only). Reference(s):

1. Attachment 2 Safe Operation

& Shutdown Areas Tables A-3 & H-2 Bases 2. NEI 99-01 AA3 Page 68 of 296

-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category C -Cold Shutdown / Refueling System Malfunction EAL Group: Cold Conditions (RCS temperature

200°F); EALs in this category are applicable only in one or more cold operating modes. Category C EALs are directly associated with cold shutdown or refueling system safety functions.

Given the variability of plant configurations (e.g., systems out-of-service for maintenance, containment operi, reduced AC power redundancy, time since shutdown) during these periods, the consequences of any given initiating event can vary greatly. For example, a loss of decay heat removal capability that occurs at the end of an extended outage has less significance than a similar loss occurring during the first week after shutdown. Compounding these events is the likelihood that instrumentation necessary for assessment may also be inoperable. The cold shutdown and refueling system malfunction EALs are based on performance capability to the extent possible with consideration given to RCS integrity, CONTAINMENT CLOSURE, and fuel clad integrity for the applicable operating modes (4 -Cold Shutdown, 5 -Refueling, DEF -Defueled).

The events of this category pertain to the following subcategories:

1. RPV Level RPV water level is directly related to the status of adequate core cooling and, therefore, fuel clad integrity . . 2. Loss of Emergency AC Power Loss of vital plant electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity.

This category includes loss of onsite and offsite power sources for 4.16 KV ENS buses. 3. RCS Temperature Uncontrolled or inadvertenttemperature,or pressure rises are indicative of a potential loss of safety functions.

4. Loss of Vital DC Power Loss of emergency plant electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity.

This category includes loss of power to or degraded voltage on the 125V DC vital buses. 5. Loss of Communications Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification. Page 69 of 296 -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases 6. Hazardous Event Affecting Safety Systems Certain hazardous natural and technological events may result in VISIBLE DAMAGE to or degraded performance of safety systems warranting classification. Page 70 of 296 A -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: C -Cold Shutdown / Refueling System Malfunction Subcategory: 1 -RPV Level Initiating Condition: UNPLANNED loss of RPV inventory EAL: CU1.1 Unusual Event UNPLANNED loss of reactor coolant results in RPV water level less than a required lower limit for~ 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 4 -Cold Shutdown, 5 -Refueling Definition(s): 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: River Bend is equipped with multiple RP\/ water level instruments includinq: Wide Range, Fuel Zone, Shutdown Ranqe, Upset Range, and Narrow *Range (ref. 1 ). Multiple instruments on different reference and variable legs should be monitored. The Upset Range and Shutdown Range instruments share a common reference leg; therefore, Narrow Range instruments should be routinely monitored when relying on Shutdown or Upset Range instrument as the primarv indication. ,, With the plant in Cold Shutdown, RPV water level is norma!lv maintained above the RPV low level scram setpoint of 9.7 in. (ref. 2). However, if RPV level is being controlled below the RPV low level scram setpoint, or if level is being maintained in a desianated band in the reactor vessel it is the inability to maintain level above the low end of the designated control band due to a loss of inventory resultino from a leak in the RCS that is the concern. With the plant in Refueling mode, RPV water level is normally maintained at or above the reactor vessel flange. Technical Specifications require at least 23 fl of water above the tog_ of the reactor vessel flange in the refueling cavity during refuelino operations (ref. 3). The RPV flange is at approximately 200 in. on the Shutdown Range. {ref. !11. This G-EAL addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor (reactor vessel/RCS Page 71 of 296 -:::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases fP.WR}-oF-RPV -[-B-WRj}level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant. Refueling evolutions that El-eern-aw-lower RCS water inventory are carefully planned and controlled. An UNPLANNED event that results in water level Ge-GFeas+R§--lowerinq 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. This EAL-#4 recognizes that the minimum required (feaB:t-e-F-Ve-&Sel/RGS-[-PVI/R] Of---RPV f-BWRH-level can change several times during the course of a refueling outage as different plant configurations and system lineups are implemented. This EAL is met if the minimum level, specified for the current plant conditions, cannot be maintained for 15 minutes or longer. The minimum level is typically specified in the applicable operating procedure but may be -specified in another controlling document. The 15-minute threshold duration allows sufficient time for prompt operator actions to restore and maintain the expected water level. This criterion excludes transient conditions causing a brief lowering of water level. ---E-,-'\b-#2-ad d resse-s-a-oo-ooi#e-R-WheH~-all-meaFl-&4e-aet-emii n e (reactor vesse~IRG& f e been lost. In U=i-is-oone-itie-n,eperators-ffia-y-*letormine that an m-vente-r-y-loss is occurring by obse~n sump and/or tank levels. Sump and/8-f tank-level changes must be eva I l::lated ag a i tRey-are-im:l4Battve-e-f--tea kag e fi:e-m the (reac-te-F-vessel!RGS-f~~ Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA 1 or CA3. Reference(s):

1. EOP-1 RPV Control 2. Technical Requirements Table 3.3.1.1-1 RPS Instrumentation

3. Technical Specification 3.9.6 Reactor Pressure Vessel (RPV) Water Level -Irradiated Fuel 4. GMP-0102 Reactor Vessel Disassembly

5. NEI 99-01 CU1 Page 72 of 296 JfOA *~ -====-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

C -Cold Shutdown / Refueling System Malfunction -* Subcategory: 1 -RPV Level Initiating Condition: UNPLANNED loss of RPV inventory EAL: CU1.2 ) Unusual Event ~PV water level cannot be monitored AND EITHER

UNPLANNED rise in any Table C-1 sump or pool level due to a loss of RPV inventory
Visual observation of UNI.SOLABLE RCS leakage Mode Applicability:

Table C-1 Sumps/Pool

Drywell equipment drain sump
Drywell floor drain sump
Pedestal floor drain sump
CTMT equipment drain sump
CTMT floor drain sump
Suppression Pool
RHR A, B, C, HPCS, LPCS, RCIC room sumps
Auxiliary Building floor drain sump 4 -Cold Shutdown, 5 -Refueling Definition(s):

UN/SOLABLE -An open or breached system line that cannot be isolated, remotely or locally. UNPLANNED-. A parameter changes 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. r Page 73 of 296 -~-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Basis: Shutdown Ranqe instrument. 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-lRCS [PVV-R-} er-RPV-[BWRft-level concurrent with indications of coolant leakage .. Either of these conditions is considered to be a potential degradation of the level of safety of the plant. Refueling evolutions that El-eGFease-lower RCS water inventory are carefully planned and controlled. An UNPLANNED event that results in water level El-esrea&iR§-!owerinq below a procedurally required limit warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the core covered. EAL #1 recOfj-AfZes that the-miftiR-'lum requ+red (react-er vessel/RCS [PVI/R] er-RP-V [BWR]) level can change severaW~mes-El-1::1-r~ng the course of-a-retooling outage as diffefe.A.t p+am-oon#gurations and system lin~ented. This E/\L is met ~f-#te-mi-nifnHm level, speBi-fieG-fef-the-eu-frent-p+a-Rt-Be-rtEl+tions, cannot be maintaineEl-for 15 The mmimum-l~speeified in-too-awJ.i.e.ablB-efIBFatifig procedure but may-be spe~ntffitl+ng-EieCHmenh The-1-5-minute threshold dl::l-FatieR-al+e-ws-st1-ffisient time for r:irompt operatoF-aBtieR-s-te-restere and maintaif'I the expected v,1ater level. This erneFien excludes transient conditions cauffing-a Bf~-lewering of water level. This EAL #2-addresses a condition where all means to determine ff-easter vessel/RCS [PVVR} er-RPV-fS\A/R]) 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 (Table C-1 ). 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 [PVVR] oi:-RPV ~-Page 7 4 of 296 ------------- Q -~*Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Continued loss of RCS inventory may resulfin escalation to the Alert emergency classification level via either IC CA 1 or CA3. Reference(s):

1. SOP-0104 Floor and Equipment Drains System 2. SOP-0033 Drywell and Containment Leak Detection System 3. EOP-3 Secondary Containment and Radioactivity Release Control 4. NEI 99-01 CU1 Page 75 of 296

-::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: C -Cold Shutdown / Refueling System Malfunction Subcategory: 1 -RPV Level Initiating Condition: Significant Loss of RPV inventory EAL: CA1.1 Alert Loss of RPV inventory as indicated by RPV water level < -43 in. (Level 2) Mode Applicability: 4 -Cold Shutdown, 5 -Refueling Definition(s): None Basis: The threshold RPV water level of -43 in. is the Level 2 actuation setpoint for HPCS and RCIC. Although RClC cannot restore RPV inventory in the cold condition, the Levei 2 actuation setpoint is ooerationallv significant and is indicative of a loss of RPV inventorv significantl:L below the low RPV water level scram setpoint soecified in CU1 .1 (ref. 1. 2). 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 this EAL-#4, a lowering of RPV water level below-fs~ecific level-)-# the specified level indicates that operator actions have not been successful in restoring and maintaining R-G-S (reactor vessel/RCS [PVVR] oF-RPV {WAJ.R})-water level. The heat-up rate of the coolant will i-Fterease--rise as the available water inventory is reduced. A continuing e-eernase-drop in water level will lead to core uncovery. Although related, this EAL#-1-is concerned with the loss of RPV inventory and not the potential concurrent effects on systems needed for decay heat removal (e.g., loss of a Residual-Decay Heat Removal suction point). AFJ i-Ftc-rease-rise in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3. For E/\L #2, the inab+tity to me-n-i-tor (reactor vessel/RCS fPWR}-er RPV may-ee caused by-+A-st-Fillflefttatiefl-a.ooler-13ewef-fail u res, or watefiev&l-er-epp~~e FaH§-e-e-f-av-a+la-ete-i-n-stru-me-nt-at+&R-:-lf 'Nater level cannot be-mG-fl-HOfed,eperators may Eletermine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level-eftanges mtl-S-t--f:}e evaluated agains~~s-&t-water

evJ to ensure they are iHEl-+Batwe-ef-.eakage from the ( f B'"'R]) rv . Page 76 of 296 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases =P.=to 15 mtAute duratie-A-fof--me-+e-ss-e-f-~vel-ifldication was chos&A-besat1-&e-i-t-+s-Ra#-ef-tl=te E/\L EllifatiOH-speetHBEl-tR-I-G-GS4 If RGS-the-freaBtor vessel/RCS-fP-WR}-ef-RPV fBVVR]) iAveRte-ry-water level continues to lower, then escalation to Site Area Emergency would be via IC CS1. Reference(s):

1. Technical Requirements Table 3.3.5.1-1, Emergency Core Cooling System Instrumentation

2. Technical Requirements Table 3.3.5.2-1, RCIC Instrumentation

3. NEI 99-01 CA1 Page 77 of 296

-===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: C -Cold Shutdown / Refueling System Malfunction Subcategory: 1 -RPV Level Initiating Condition: Significant Loss of RPV inventory EAL: CA1.2 Alert RPV water level cannot be monitored for~ 15 min. (Note 1) AND EITHER

UNPLANNED rise in any Table C-1 sump or pool levels due to a loss of RPV inventory
Visual observation of UNISOLABLE RCS leakage Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.

The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: Table C-1 Sumps/Pool

Drywell equipment drain sump
Drywell ~oor drain sump
Pedestal floor drain sump
CTMT equipment drain sump
CTMT floor drain sump
Suppression Pool
RHR A, B, C, HPCS, LPCS, RCIC room sumps
Auxiliary Building floor drain sump 4 -Cold Shutdown, 5 -Refueling Definition(s):

UNISOLABLE -An open or breached system line that cannot be isolated, remotely or locally. J 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. Page 78 of 296 --===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases ,, Basis: In Cold Sl1utdown mode, the RCS will normally be intact and standard RPV level monitorina

means are available.

RPV level in the Refueling mode is normaliy monitored using the Shutdown Range instrument. In this EAL. all water level indication is unavailable and the RPV inventory loss must be detected b\L the leakage indications listed in Table C-1. Level rises must be evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure thev are indicative of RPV leakage. Rise in clrywell equipment drain sump level and drywell floor sumo level is the normal method of monitoring and calculating leakage from the RPV (ref. 1. 2). An Auxiliary Building sump level rise may also be indicative of RCS inventory losses external to the Containment from systems connected to the RPV. With RHR System operating in the Shutdown Coolinci mode, an unexplained rise in suporession pool water level could be indicative of RHR valve misalignment or !eakage. If the make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS that cannot be isolated could be indicative of a loss of RPV inventor~ 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 #1, a lowering of ',Nater level below (site specific level) indicates tha-t-eperator aetie~een-st1ccessful in restOfi.ng and main-ta+n+ng (reactor vessel/RCS [PVl/R] or RPV-f-Bl/1/R]) \vater level. The ooakfp-rate--ef-th-e-eeeta-Ftt-will increase as t~Me-watef -i-rwentory is reduce4-A

* * 'Nill lead to core t:mGovery.

---+fa*,lt!-Hhff-l-ough-rel-ai:ed, EAL #1 is concern~he loss of RCS inventory-a~ f*}te-nti-al-Boocurrent effects on systems needed for decay heat removal (e.g., loss of a -Residua! Heat Removal suction point). An-i-n-crease in RCS temperarure caused by a loss of ~t-removal-capability is evaluated under IC C/\'J-c For this EAL-#2, the inability to monitor RCS (reactor vesse+/RCS [PWR]-e-r-RPV [8\NR]) 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 freactor vessel/RCS [P'NR] or-RPV f P\l\f PJ\ E) VI/~]* The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CS1.:. If the (reactor vessel/RCS [PVVR] or RPV-fBVVR]) inventory level continues to lower, then escalation to Site Area Emergency would be via IC CS1. Page 79 of 296 Q -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Reference(s):

1. SOP-0104 Floor and Equipment Drains System 2. SOP-0033 Drywell and Containment Leak Detection System 3. EOP-3 Secondary Containment and Radioactivity Release Control 4. NEI 99-01 CA1 Page 80 of 296

~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: C -Cold Shutdown/ Refueling System Malfunction Subcategory: 1 -RPV Level Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability EAL: CS1.1 Site Area Emergency CONTAINMENT CLOSURE not established AND RPV water level < -143 in. (Level 1) Mode Applicability: 4 -Cold Shutdown, 5 -Refueling Definition(s): CONTAINMENT CLOSURE -The actions taken to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. Containment Closure is established when the containment requirements of OSP-0037 are met and at least one integral barrier to the release of radioactive material is provided, within the specified limits using STP-057-3804. IMMINENT -The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. Basis: The threshold RPV water level of -143 in. is the low-low-low ECCS actuation setpoint (1 eve! 1 ). The magnitude of this loss of water indicates that makeup systems have not been effective and mav not be capable of preventinq further lowerino of RPV water level and potential core uncoverv. The inabilitv to restore and maintain level after reaching this setooint infers a failure of the RCS barrier and Potential Loss of the Fuel Clad barrier (ref. 1, 2). This IC addresses a significant and prolonged loss of (reaetor vessel/RCS [PWR}-Or-RPV fB-WR})-inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration. Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/roae-teF vesse!RPV level cannot be restored, fuel damage is probable. Page 81 of 296 e -::::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases 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 vesselRPV levels of EALs CS1 .1 and CS1 .2 4--A3-a-AG--be-reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment. -I-A-EAL 3.a, the 3G-mtnute criterion is tied to a rea~voot-sffift-#me-(i.e., the total-loss of abil#y-te--r:oon-i-tOf-level), aAG-al!ews--5llffie.fent time-1:e-monitor, assess and correlate -FeaBtor and pl-ant condiHBfl-S-te4ete-rmi-Re-i-f-cere-Hn-oovery has actually ac-ooUHHEtr-vafi-e-1::1-&-aec~El-e-A-t-progressioo--afte--i-AstrumontatioA----1::1-Ree-rtai-Rt~}c--tt-a~se-alle-vvs SH#+e.feR-t-t+me-fe-F--pBfformance of actie-A-s-te--te-r:miflate leakage, recover invente-r:y sentrol/ma~ipment-aREl+e-F-fOS-te-Fe-~1--moo#ormg-c +t-IB-~Ra-9H+ty-ffi-moM0f-fre-actof--vessel/-R-GS-fPWRte-r-RP-V-fB'A' R]) Io vol may be caHSee--ey i-A-st-r:u me ntat_io n an*1-!'0f-pe-wef-4aHtlfe&;-Of-Wffi*f-le-vel--dm~~~ng-ookaw#te-raRge-ef-ava~ i-A-s-tfHmeA-tation. If 1.vateF-~bva....moo#ored,o!c}Orators-may-determiR~ tfl-veA-tefy-loss is ocwfriH§-b-y-oosewi-R§--BA-afl§BS-in sump and/or tan-k-~~Ei/or tan-k-~evel-chang e~~uatee--aga+ABf-ot-Ref-Petenti-al--sol:ffCe&ef-wateF-#ew-te-e-n-s-uro mey-are-iREl-icat~ve-e-f-leaka§e--frem-#to-freaBt-0-F-vess-et!RGS-fP-WR}-oi:--RP--V-fSWRt):- Thi sese EALs addresses 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 CG1 or AG1. Reference(s):

1. Technical Requirements Table 3.3.5.1-1, Emergency Core Cooling System Instrumentation

2. NEI 99-01 CS1 Page 82 of 296

-~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: C -Cold Shutdown / Refueling System Malfunction Subcategory: 1 -RPV Level Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability , EAL: CS1 .2 Site Area Emergency CONTAINMENT CLOSURE established AND RPVwater level< -162 in. (TAF) Mode Applicability: 4 -Cold Shutdown, 5 -Refueling Definition{s): CONTAINMENT CLOSURE -The actions taken to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. Containment Closure is established when'the containment requirements of OSP-0037 are met and at least one integral barrier to the release of radioactive material is provided, within1the specified limits using STP-057-3804. IMMINENT:.. The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. Basis: When RPV level drops to the top of active fuel (TAF) (an indicated RPV level of-162 in.), core uncovery starts to occur ( ref. 1 ). This IC addresses a significant and prolonged loss of RPV inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration. Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level.* If RPV 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 vesselRPV levels of EALs 1.b and 2.b CS1 .1 and CS1 .2 reflect the fact that with' CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment. Page 83 of 296 ~-Entergy River Bend Station 'EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases This EAL addresses 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 CG1 or AG1. Reference(s):

1. EPSTG*0002 Appendix B EOP and SAP Bases 2. NEI 99-01 CS1 Page 84 of 296 C!\ ~=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

C -Cold Shutdown / Refueling System Malfunction Subcategory: 1 -RPV Level Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability EAL: CS1 .3 Site Area Emergency RPV level cannot be monitored for~ 30 min. (Note 1) AND Core uncovery is indicated by any of the following:

UNPLANNED rise in any Table C-1 sump or pool levels of sufficient magnitude to indicate core uncovery
Visual observation of UNISOLABLE RCS leakage of sufficient magnitude to indicate core uncovery
Note 1: \ RMS-RE140 Refueling Floor Near North Entrance, RMS-RE141 Refueling Floor Near South Entrance.

or RMS-RE16 A/B Primary containment -PAM A/B reading > 9 R/hr The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Table C-1 Sumps/Pool

Drywell equipment drain sump
Drywell floor drain sump
Pedestal floor drain sump
CTMT equipment drain sump
CTMT floor drain sump *
Suppression Pool
RHR A, B, C, HPCS, LPCS, RCIC room sumps
Auxiliary Building floor drain sump Mode Applicability:

4 -Cold Shutdown, 5 -Refueling Page 85 of 296 e *z=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Definition(s): IMMINENT -The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigat,ion or corrective actions. UNISOLABLE -An open or breached system line that cannot be isolated, remotely or locally. UNPLANNED -A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The. cause of the parameter change or event may be known or unknown. Basis: In Cold Shutdown mode, the RCS will normally be intact and standard RPV level monitoring means are available. RPV level in the Refueling mode is normal!v monitored usina the, Shutdown Range instrument. ln this EAL, al! water level indication is unavailable and the RPV inventory loss must be detected by the leakaae indications listed in Table c--1. Level rises must be evaluated against other potential sources of leakage such as cooling water sources inside the drywe!i to ensure they are indicative Df RPV leakage. Rise in drywell equipment drain sump level and drywe!I floor sump level is'the normal method of monitoring and calculating leakaae from the RPV (ref. 1, 2). An Auxiliary Building sump level rise may also be indicative of RCS inventorv losses external to the Containment from systems connected to the RPV. With RHR System operating in the Shutdovvn Cooling mode, an unexplained rise in suppression pool water level could be indicative of RHR valve misalignment or leakage. If the make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from* systems connected to the RCS that cannot be isolated could be indicative of a loss of RPV inventory. ln the Refueling Mode, as water level in the RPV lowers, the dose rate above the core will rise. The dose rate due to this core shine should result in indications on area radiation monitors. The 9 R/hr value is selected for this EAL because it is 90% of the scale for RMS-RE140 and 141 (lower range monitors) and on scale for the hiqher range monitors. This value represents a reading that is higher than that likely to be attributable to normal refuel floor operations. These monitors are located in the Containment on the refuel floor. This IC addresses a significant and prolonged loss of (reactor vessel/RCS RGS-.:{PVVR] or RPV {£WR]) inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration. Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel_RCSRPV level cannot be restored, fuel damage is probable. Outage/shutdown contingency plans typically provide for re establishing or verifying Page 86 of 296 ~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases GGN=r-AI-NME~lewif-1§-3-tess-sf--heat remevaJ.:-OF-RCS invente~~ fu-ABOORs-,-1ltB-4i#erence in the S13BBi-ffeEI-RCS/reaet-sr-ve&Se{-!eve~s-ef E/\Ls 1.b-aREl-2~f~Bt

le-facHhat-w~E-N+-GbGSURE establish~is a lower proeaehlty-e-f--a
sstefl-pfeeuct release te-t:~eR-t~

In this EAL~, the 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring. The inability to monitor_RCS (reactqr vessellRCS-f,-oWR}eF---RPV fl3WRB-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 RGS-freaete-F-Vesse4/-RGS-f,~RPV-fBWRB. +A-ese-This EAL-s addresses 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 CG1 or AG1 Reference(s):

1. SOP-0104 Floor and Equipment Drains System 2. SOP-0033 Drywell and Containment Leak Detection System 3. EOP-3 Secondary Containment and Radioactivity Release Control 4. NEI 99-01 CS1 Page 8 7 of 296

":;::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: C -Cold Shutdown / Refueling System Malfunction Subcategory: 1 -RPV Level Initiating Condition: Loss of RPV inventory affecting fuel clad integrity with Containment challenged EAL: CG1 .1 General Emergency RPV level< -162 in. (TAF) for~ 30 min. (Note 1) AND Any Containment Challenge indication, Table C-2 Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required. Table C-2 Containment Challenge Indications

CONTAINMENT CLOSURE not established (Note 6)
Drywell or containment hydrogen concentration>

4%

UNPLANNED rise in containment pressure
Exceeding one or more Secondary Containment Control MAX SAFE area radiation levels: Area ORMS Grid 2 Max. Safe Operating Value RHR Equip Rm A 1213 9.5E+03 mR/hr RHR Equip Rm B 1214 9.5E+03 mR/hr RHR Equip Rm C 1215 9.5E+03 rhR/hr Mode Applicability:

4 -Cold Shutdown, 5 -Refueling Page 88 of 296 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Definition(s): CONTAINMENT CLOSURE -The actions taken to secure* primary containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. Containment Closure is established when the containment requirements of OSP-0037 are met and at least one integral barrier to the release of radioactive material is provided, within the specified limits using STP-057-3804. IMMINENT -The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. ) 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: When RPV level drops below -16? in .. core uncoverv starts to occur (ref. 1 ). Four conditions are associated with a challenge to Containment integrit:t: CONTAINMENT CLOSURE is not established. I') In the early stages of a core uncovery event. it is unlikely that hydrogen buildup due to a core uncoverv could result in an explosive mixture of dissolved Qases in the containment. However. containment monitoring and/or sampling should be performed to verifv this assumotion and a General Emergency declared if it is determined that hvdroaen concentration has exceeded U1e minimum necessary to support a hydrogen burn {4%). The Igniter Svstem is designed to prevent hydrogen accumulation by locally burning hydrogen in a controlled manner as soon as the hydrogen enters the containment atmosphere and reaches the igniters. For hioh rates of hydrogen production. ignition occurs at the lowest concentration that"can support ignition. Following ignition, hvdrogen Is consumed through formation of diffusion flames where the gas enters the containment. thus controlling hydrogen concentration at approximately 4% (ref. 2). 1/) Any UNPLAI\JNFD rise in containment pressure in the Cold Shutdown or Refueling mode indicates a ootentia! loss of CONTAINMENT CLOSURE capability. UNPLANNED containment pressure rise indicates CONTAINMENT CLOSURE cannot be assured and the containment cannot be relied upon as a barrier to fission product release. Secondary Containment radiation monitors should provide indication of a larger release that may be indicative of a challenqe to CONTAINMENT CLOSURE. The MAX SAFE radiation levels are indicative of problems in the secondary containment that are spreading. The locations into which the orimary system discharge is of concern correspond to the areas addressed in Table SC-2 of EOP-3, Secondary Containment ' Page 89 of 296 -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases and Radioactivity Release Control that are in service under Cold Shutdown conditions (ref. 3). This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area. Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RPV 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 mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a challenge to Containment integrity. In the early stages of a core uncovery event, it is*unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment. If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged. Toe-3-G-rn~m,;1-1:8-Bfi-ter-ie-R-i-s-t+eEi-te-a-reaEl-Hy--reeegRiBb~effi-start-t+me ( i . e., the--rota1-l-e-&s-Bf alc}~~i+y-te--mo n ito r I e vel), a nEl-a+!Bws-sumereffi-4~me-te-moo-i-te-F,a&Ses-s-aoo co rrerat-e-Fea&t:ef-aR*1 plant-eoooole-RS-te-d ete rm~Fte-i-f-Bore-l:lHOO-VeF-y-FIBS-aB-tHa+!-y-o ccl:if~;;&.,-t.e-a-eeotJRt-fef-vaH01:t& ac-si*1eA-t-p-~SA-aHEHA-SiruffiOAfati&R-Hf1B-GHaifme-st,--lt-a!-se-al-JBws-su#+e+ent-ti-me-f8-f peff o rman co e-t-a&l:loo-&4e-ter-mtfl-ate-leak-afre,reee-vef-i-RVBA-t-BfY-Ge-R-tfo+/makel:lp-eEJ-bl-tfafRe-At a-Rdlef...rnstore leve+-me-R-i-te-F~R-§b This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-2a3, 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. Reference(s):

1. EOP-1 RPV Control 2. EPSTG*0002 Appendix B EOP and SAP Bases Page 90 of 296

-:::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases 3. EOP-3 Secondary Containment and Radioactivity Release Control 4. NEI 99-01 CG1 Page 91 of 296 -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: C -Cold Shutdown / Refueling System Malfunction Subcategory: 1 -RPV Level Initiating Condition: Loss of RPV inventory*affecting fuel clad i.ntegrity with Containment challenged EAL: CG1 .2 General Emergency RPV water level cannot be monitored for~ 30 min. (Note 1) AND Core uncovery is indicated by any of the following:

UNPLANNED rise in any Table C-1 sump or pool levels of sufficient magnitude to indicate core uncovery
Visual observation of UNISOLABLE RCS leakage of sufficient magnitude to indicate core uncovery
RMS-RE140 Refueling Floor Near North Entrance, RMS-RE141 Refueling Floor Near South Entrance or RMS-RE16 NB Prirhary containment

-PAM NB reading > 9 R/hr AND Any Containment Challenge indication, Table C-2 Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency D'irector is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required. Page 92 of 296 Q -::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Table C-1 Sumps/Pool

Drywell equipment drain sump
Drywell floor drain sump
Pedestal floor drain sump
CTMT equipment drain sump
CTMT floor drain sump
Suppression Pool
RHR A, B, C, HPCS, LPCS, RCIC room sumps
Auxiliary Building floor drain sump Table C-2 Containment Challenge Indications
CONTAINMENT CLOSURE not established (Note 6)
Drywell or containment hydrogen concentration>

4%

UNPLANNED rise in containment pressure
Exceeding one or more Secondary Containment Control MAX SAFE area radiation levels: Area DRMS Grid 2 Max. Safe Operating Value RHR Equip Rm A 1213 9.5E+03 mR/hr RHR Equip Rm B 1214 9.5E+03 mR/hr RHR Equip Rm C 1215 9.5E+03 mR/hr Mode Applicability:

4 -Cold Shutdown, 5 -Refueling Definition(s): CONTAINMENT CLOSURE -The actions taken to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. Page 93 of 296 A -:;::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Containment Closure is established when the Containment requirements of OSP-0037 (ref. 4.1.12) are met with the following exception: a functional barrier must exist at the time of the event (i.e., cannot rely on contingency methods to establish a functional barrier). IMMINENT -The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. UNISOLABLE -An open or breached system line that cannot be isolated, remotely or locally. UNPLANNED -A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: In Cold Shutdown mode, the RCS will normally be intact and standard RPV level monitoring means are available. RPV level in the Refueling mode is normallv monitored using the Shutdown Range instrument. In this EAL all water level indication is unavailable and the RPV inventory loss must be detected by the leakaae indications listed in Table C-1. Level rises must be evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure they are indicative of RPV leakage. Rise in drywell equipment drain sump level and drywell floor sump level is the normal method of monitoring and calculating leakage from the RPV (ref. *1, 2). An Auxiliary Building sump level rise may also be indicative of RCS inventory losses external to the Containment from systems connected to the RPV. With RHR System operating in the Shutdown Cooling mode. an unexplained rise in suppression pool water level could be indicative of RHR va!ve misalignment or leakage. If the make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from ~ystems connected to the RCS that cannot be isolated could be indicative of a loss of RPV inventory. In the Refueling Mode, as water level in the RPV lowers, the dose rate above the core will rise. The dose rate due to this core shine should result in indications on area radiation monitors. The 9 R/hr value is selected for this EAL because it is 90% of the scale for RMS-RE140 and 141 (lower range monitors) and on scale for the higher range monitors. This value represents a reading that is hiqher than that likely to be attributable to normal refuel floor operations. These monitors are located in the Containment on the refuel floor. Four conditions are associated with a challenge to Containment integrity:

CONTAINMENT CLOSURE is not established.

0 In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive mixture of dissolved gases in the containment. However. containment monitoring and/or sampling should be performed to verify this assumption and a General Emergency declared if it is determined that hydrogen concentration has exceeded the minimum necessary to support a hydrogen Page 94 of 296 e ,~-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases burn (4%). The lanlter System is designed to prevent hydrogen accumulation by locally burning hydrogen in a controlled manner as soon as the hvdrogen enters the containment atmosghere and reaches the ia'liters. For high rates of hydrogen production, ignition occurs at the lowest concentration that can support ignition. Following ignition, hydrogen is consumed throuah formation of diffusion flames where the gas enters the containment, thus controlling hydrogen concentration at aoproximatelv 4% (ref. 3). o Any UNPLANNED rise in containment pressure in the Cold Shutdown or Refueling mode indicates a ootential loss of CONTAJN!V1ENT Ci OSURE caoabiiity. U1\JPLANNED containment pressure rise indicates CONTAINMENT CLOSURF cannot be assured and the containment cannot be relied uoon as a barrier to fission product release. o Secondary Containment radiation monitors should provide indication of a larger release that mav be indicative of a challenae to CONTAINMFNT Cl OSURE. The MAX SAFE radiation levels are indicative of problems in the secondary containment that are spreadina. The locations into which the primary system discharge is of concern gorr§§.R.ond to the areas addressed in Table SC-2 of EOP-3. Secondary Containment and Radioactivity Release Control that are in service under Cold Shutdown conditions (ref. 4). This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area. Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RGS ~el-RPV 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 mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a challenge to Containment integrity. In .the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a-core uncovery could result in an explosive gas mixture in containment. If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed Page 95 of 296

~-~----

-===-Entergy I River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases 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 2.b, tihe 30-minute criterion is tied to a readily recognizable event start time (i.e., the ~otal loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core 'uncovery has actually occurred (i.e., to account for various accident progrespion 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/RGS-fP-WR}-ef-RPV-RGS::fB-WRB-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 1 they are indicative of leakage from the_-frea-eteF--vess-e~RPV-f&WRB. Thisese EALs addresses 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 Manageff}ent.

Reference(s):

1. SOP-0104 Floor and Equipment Drains System

2. SOP-0033 Drywell and Containment Leak Detection System 3. EPSTG*0002 Appendix B EOP and SAP Bases 4. EOP-3 Secondary Containment and Radioactivity Release Control 5. NEI 99-01 CG1 Page 96 of 296

-::::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -. Emerge,ncy Action Level Technical Bases Category: C -Cold Shutdown / Refueling System Malfunction Subcategory: 2 -Loss of Emergency AC Power Initiating Condition: Loss of all but one AC power source to ENS buses for 15 minutes or longer EAL: CU2.1 Unusual Event AC power capability, Table C-3, to DIV I and DIV II 4.16 KV ENS buses reduced to a single power source for~ 15 min. (Note 1) AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS , Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Table C-3 AC Power Sources Offsite

1RTX-XSR1C
1 RTX-XSR 1 D Onsite
EGS-EG1A
EGS-EG1B Mode Applicability:

4 -Cold Shutdown, 5 -Refueling, DEF -Defueled Definition{s): SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. \These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; Page 97 of 296 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis: The DIV ill bus (1 E22*S004) is not credited because it onlv supplies power to the HPCS oump and associated loads. not any long term decay heat removal systems. ln particular, suppression pool coolina mechanisms would be essential subsequent to a station blackout. This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment. When in the cold shutdown, refueling, or defueled mode, this condition is not classified as an Alert because of the ~nsrea-see-areater time available to restore another power source to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition is considered to be a potential degradation of the level of safety of the plant. An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an ENS bus. Some examples of this condition are presented below.

A loss of all offsite power with a concurrent failure of all but one emergency ENS power source (e.g., an onsite diesel generator).
A loss of all offsite power and loss of all emergency ENS power sources (e.g., onsite
diesel generators) with a single train of emergency ENS buses being back-fed from the unit main generator.
A loss of emergency ENS power sources (e.g., onsite diesel generators) with a single train of emergency ENS buses being 0-aek-fed from an offsite power source. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power. The subsequent loss of the remaining single power source would escalate the event to an Alert in accordance with IC CA2. This EAL is the cold condition equivalent of the hot condition EAL SA 1.1. Reference(s):

1. USAR Section 8.1 Electric Power Introduction

2. EE-001AC Startup Electrical Distribution Chart 3. SOP-0046 4.16 KV System 4. AOP-0004 Loss of Offsite Power 5. NEI 99-01 CU2 Page 98 of 296 --------

Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: C -Cold Shutdown / Refueling System Malfunction Subcategory: 2 -Loss of Emergency AC Power Initiating Condition: Loss of all offsite and all onsite AC power to ENS buses for 15 minutes or longer EAL: CA2.1 Alert Loss of all offsite and all onsite AC power capability to DIV I and DIV II 4.16 KV ENS buses for~ 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 4 -Cold Shutdown, 5 -Refueling, DEF -Defueled Definition(s): SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 1 OCFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. 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. Mitigative strategies using other power sources (Division Ill diesel generator. FLEX generators. etc.) may be effective in supplying power to these buses. These power sources must be controlled in accordance with abnormal or emergency operating procedures. or beyond design basis accident response guidelines (e.g., FLEX support guidelines) and must be capable (alone or in combination) of supplying power for long term decay heat removal systems. In particular. suppression pool cooling systems would be essential subsequent to a station blackout. Page 99 of 296 -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases When in the cold shutdown, refueling, or defueled mode, this condition is not classified as a Site Area Emergency because of the +R-ereasee-greater time available to restore an *-SF-ENS bus to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition represents an actual or potential substantial degradation of the level of safety of the plant. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via IC CS1 or AS1. This EAL' is the cold condition equivalent of the hot condition EAL SS1 .1. Reference(s):

1. USAR Section 8.1 Electric Power Introduction

2. EE-001AC Startup Electrical Distribution Chart 3. SOP-0046 4.16 KV System 4. AOP-0004 Loss of Offsite Power 5. NEI 99-01 CA2 Page 100 of 296

) River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: C -Cold Shutdown / Refueling System Malfunction Subcategory: 3 -RCS Temperature Initiating Condition: UNPLANNED rise in RCS temperature EAL: CU3.1 Unusual Event UNPLANNED rise in RCS temperatur~ to> 200°F due to loss of decay heat removal capability Mode Applicability: 4 -Cold Shutdown, 5 -Refueling Definition(s): .CONTAINMENT CLOSURE -The actions taken to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. Containment Closure is established when the containment requirements of OSP-0037 are met and at least one integral barrier to the release of radioactive material is provided, within the specified limits using STP-057-3804. 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 an UNPLANNED iRB-Fease rise in RCS_temperature above the Technical Specification cold shutdown temperature limit, GHRe-+Aaetl~em~ and level,and represents a potential degradation of the level of safety of the plant. If the RGS RCS_is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Director should also refer to +G-EAL CA3j_. A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification. EAL #1 This EAL This EAL involves a loss of decay heat removal capability, or an addition of heat to the RCS in excess of that which can currently be removed, such that reactor coolant temperature cannot be maintained below the cold shutdown temperature limit specified in Technical Specifications. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation. During an outage, the level in the reactor'vessel will normally be maintained at-efat or 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 Page 101 of 296 ~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases inventory may result in a rapid ti:ter-eas-e-rise in reactor coolant temperature depending on the time after shutdown. EAL #2 reflec~oo-wRBre there has beon-a-&f§flificant loss of instrumentati-GR ea!*ffitltt-y-Aecessary-to mo niter RCS OOAGi-ti9fl-S-aftEl-9f)Brate-~el-e-tEHflefl-HOf-key parameters-Reeessary to assure core decay heat reFRe-Va-h-~ndition, there is no +mmoo+ate-threat-e-f-fuel dama§B-beeause the core decay heat load 1:"t-as-1:IBen reduceE!-si-Aee the cessation of povvef-ef}efa-Ue-A-:- Fifteen minUfO-S--',Nas selected as a threshok+-f&-ffi{e+t1El-e-tra-n-si*m-ef-me-meRtar-y-k}sses ef-~Affi&atien-: Escalation to Alert would be via IC CA 1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria. Reference(s):

1. Technical Specification Table 1.1f1 Modes 2. STP-050-0700 RCS Pressure and Temperature Limits Verification

3. AOP-0051 Loss of Decay Heat Removal 4. NEI 99-01 CU3 Page 102 of 296

,22, ~~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: C -Cold Shutdown / Refueling System Malfunction Subcategory: 3 -RCS Temperature Initiating Condition: UNPLANNED rise in RCS temperature EAL: CU3.2 Unusual Event Loss of all RCS temperature and RPV water level indication for~ 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 4 -Cold Shutdown, 5-Refueling Definition(s): CONTAINMENT CLOSURE -The actions taken to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. Containment Closure is established when the containment requirements of OSP-0037 are met and at least one integral barrier to the release of radioactive material is provided, within the specified limits using STP-057-3804. 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 .J.C EALEAL addresses a-A-lJN.PlA~mease iA-RCS tempera~oo Technical Specification cel-El--sfitl-teewA--temf}-er-ffitJfe-l+mit,ef-the inability to determine RCS temperature and level, aooand 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 Emerge~cy Director should al~o refer to IC CA3. ---1-/\-\-,* Ž~~6*BHF&i-e-n-aoove-tR-e-Technical Specificat-i~ewn ~A-tt:ie heat removal fune#e-R-is availae-le does oot vvarrant a classification. I EAL #1 involves a loss of decay heat-removal capaei+i-ty,or an addition of heat to the RGS-i-n-excess of tfl.at-wA.i-*1:l can curFeruly-ee-removed, such tflffi-fe-aC-toF-BOelaA-t-temf}-erak:J.re cannot be maintained belevv the cold shilld-e-wA-temperarure-l+mit specified in Techni-caJ. Sf)ecifications. During this condition, there is no immediate threat of fuel damage because the cere-El-~been reduced sinee-t-he cessation of pmver operation. Page 103 of 296 -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases --WFffig-BA-B-tltage, the level-i-A--ffie-reaeto~+J-y-l:}e-maintained aee-ve-too rea-Gt~#aR{r&.-Refuel+ng-eve-ltl-tieR-S-#!a-t-+ewef-Watef-l-e-vel--0elevv-t-A&reactof-VeSS*+

ange are caref.u~!y planned a~e4-A-le-s-~-remev-al-at-reduced inventory may rest1-~n a rapid-iftcrease in reaetor coolant temperawre a#er shutdmvn-:-

EAL #2This EAL reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor RCS conditions and operators would be unable to monitor key parameters necessary to assure core decay heat removal. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication. Escalation to Alert would be via IC CA1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria. Reference(s):

1. EOP-1 RPV Control 2. Technical Specifications Table 1.1-1 Modes 3. STP-050-0700 RCS Pressure and Temperature Limits Verification

4. AOP-0051 Loss of Decay Heat Removal 5. NEI 99-01 CU3 Page 104 of 296 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

C -Cold Shutdown / Refueling System Malfunction Subcategory: 3 -RCS Temperature Initiating Condition: Inability to maintain plant in cold shutdown EAL: CA3.1 Alert UNPLANNED rise in RCS temperature to> 200°F for> Table C-4 duration (Note 1) OR UNPLANNED RPV pressure rise> 10 psig Note 1: The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Table C-4 RCS Heat-up Duration Thresholds RCS Status CONTAINMENT Heat-up Duration CLOSURE Status Intact N/A 60 min.* Not intact Established 20 min.* Not established 0 min.

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

Mode Applicability: 4 -Cold Shutdown, 5 -Refueling Definition(s): CONTAINMENT CLOSURE -The actions taken to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. Containment Closure is established when the containment requirements of OSP-0037 are met and at least one integral barrier to the release of radioactive material is provided, within the specified limits using STP-057-3804. 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.

Page 105 of 296

-===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Basis: In the absence of reliable RCS temperature indication caused by the loss of decay heat removal capability, classification should be based on the RCS pressure rise criteria when the RCS is intact in Mode 4 or based on time to boil data when in Mode 5 or the RCS is not intact in Mode 4. This +C-EAL addresses conditions involving a loss of decay heat removal capability or an addition of heat to the RCS in excess of that which can currently be removed. Either condition represents an actual or potential substantial degradation of the level of safety of the plant. A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification. The RCS Heat-up Duration Thresholds table addresses aR mcrease* rise in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact, or RCS iflventory is fe*H:tGe~mi*1-leep operation in PW-Rs). The 20-minute criterion was included to allow time for operator action to address the temperature inc-reaserise. The RCS Heat-up Duration Thresholds table also addresses aFt tRGfease--rise 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 iAerease-rise without a substantial degradation in plant safety. Finally, in the case where there is an +nsrease-rise in RCS temperature, the RCS is not intact or is at reduced ITT¥eR-~and CONTAINMENT CLOSURE is not established, no 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. The RCS pressure rise threshold EAL #2 provides a pressure-based indication of RCS_heat-up in the absence of RCS temperature monitoring capability. Escalation of the emergency classification level would be via IC CS1 or AS1. Reference(s): 1 Technical Specifications Table 1.1-1 Mods 2. STP-050-0700 RCS Pressure and Temperature Limits Verification

3. AOP-0051 Loss of Decay Heat Removal 4. NEI 99-01 CA3 Page 106 of 296

-::::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: C -Cold Shutdown / Refueling System Malfunction Subcategory: 4 -Loss of Vital DC Power Initiating Condition: Loss of Vital DC power for 15 minutes or longer EAL: CU4.1 Unusual Event Indicated voltage is < 105 VDC on required Safety Related DIV I and DIV II 125 voe buses for~ 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time .limit is exceeded. Mode Applicability: 4 -Cold Shutdown, 5 -Refueling Definition(s): SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 1 OCFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition;

(3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

Basis Safety Related DC buses ENB-SWG01A (DlV l) and ENB-SWGOi B (DIV II) feed the Division I and Division II loads respectively. The Division I and Division II batteries each have 60 cells with a specific minimum voltage of 1.75 volts/cell. These cell voltages 'Lield minimum design 'bus voltaaes of 105 VDC ( ref. 1 ). This IC addresses a loss of vital DC power which compromises the ability to monitor and control operable SAFETY SYSTEMS when the plant is in the cold shutdown or refueling mode. In these modes, the core decay heat load has been significantly reduced, and coolant system temperatures and pressures are lower; these conditions i-AGr:ease-raise 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 Page 107 of 296 fr\ ' -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Train A is out-of-service (inoperable) for scheduled outage maintenance work and Train B is in-service (operable), then a loss of Vital DC power affecting Train B would require the declaration of an Unusual Event. A loss of Vital DC power to Train A would not warrant an emergency classification. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Depending upon the event, escalation of the emergency classification level would be via IC CA1 or CA3, or an IC in Recognition Category A. This EAL is the cold condition equivalent of the hot condition EAL SS2.1. Reference(s):

1. Safety Related Battery Specification 244.521 2. USAR 8.3.2 DC Power Systems 3. NEI 99-01 CU4 Page 108 of 296

. -~=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: C -Cold Shutdown / Refueling System Malfunction Subcategory: 5 -Loss of Communications Initiating Condition: Loss of all onsite or offsite communications capabilities EAL: CUS.1 Unusual Event Loss of all Table C-5 onsite communication methods OR Loss of all Table C-5 State and local agency communication methods OR Loss of all Table C-5 NRC communication methods Table C-5 Communication Methods System Onsite Plant radio system X Plant Paging System X Sound powered phones X In-plant telephones X Emergency Notification System (ENS) Commercial Telephone System Satellite Phones State of Louisiana Radio State and Local Hotline radio INFORM Notification System Mode Applicability: 4 -Cold Shutdown, 5 -Refueling, DEF -Defueled Page 109 of 296 State/ Local X X X X X NRC X X X Definition(s): None Basis: River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases 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 GR-Gs-State and local agencies and the NRC. This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.). The first EAL condition ~b-#4-addresses a total loss of the communications methods used in support of routine plant operations. The second EAL condition EAL #2 addresses a total loss of the communications methods used to notify all GRGs State and local aqencies of an emergency declaration. The ORGs State and local aciencies referred to here are fsee---Qevel-e-i::,eF-Netes-)the Louisiana Department of Environmental Quality, Governor's Office of Homeland Security and Emergency Preparedness, five Local Parishes Office of Homeland Security and Emergency Preparedness and 24 hour notification points, Mississippi Emergency Management Agencv and the Mississippi Highway Patrol. The third EAL condition EAL #3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration. This EAL is the cold condition equivalent of the hot condition EAL SU?. 'I. Reference(s):

1. RBS Emerge.ncy Plan Section 13.3.6.1.5.4 Communications

2. RBS Emergency Plan Section 13.3.6.2.1 Site Communications

3. NEI 99-01 CU5 Page 110 of 296

-===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: C -Cold Shutdown / Refueling System Malfunction ( Subcategory:

5 -Hazardous

§vent Affecting Safety Systems Initiating Condition: Hazardous event affecting SAFETY SYSTEMS needed for the current operating mode EAL: CA6.1 Alert The occurrence of any Table C-6 hazardous event AND Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode *AND EITHER:

Event damage has caused indications of degraded performance to the second train of the SAFETY SYSTEM needed for the current operating mode
Event damage has resulted in VISIBLE DAMAGE to the second train of the SAFETY SYSTEM needed for the current operating mode. (Notes 9, 10) Note 9: If the affected SAFETY *sYSTEM train was already inoperable or out of service before the hazardous event occurred, then emergency classification is not warranted.

Note 1 O: If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted. ' Table C-6 Hazardous Events

Seismic event (earthquake)
Internal or external FLOODING event
High winds or tornado strike
FIRE
EXPLOSION
Other events with similar hazard characteristics as determined by the Shift Manager Page 111 of 296 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Mode Applicability:

4 -Cold Shutdown, 5 -Refueling Definition(s): 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 require a event inspection to determine if the attributes of an explosion are present. FIRE -Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed. FLOODING -A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area. SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 1 OCFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. , VISIBLE DAMAGE -Damage to a SAFETY SYSTEM. train that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected SAFETY SYSTEM train. 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 SYSTFM 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 VISlBLE DAMAGE such that the potential exists for performance issues. Noie that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has indications of degraded performance: Page 112 of 296 e *::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases commercial nuclear power plants are designed to be able to supp6rt single svstem issues without compromising public health and safety from radiological events. Indications of degraded performance addresses damage to a SAFFTY SYSTEM train that is in service/operation since indications for it will be readily avaiiable. The indications of degraded performance should be significant enough to cause concern reaarding the operability or reliability of the SAFETY SYSTEM train. V1Sl8LE DAMAGE addresses damage to a SAFETY SYSTEM train that is not in service/ooeration and that potentially could cause performance issues. Operators will make this determination based on the totalitv of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damaqe. This VISIBLE DAMAGE should be significant enouqh to cause concern regardinq the operability or reliability of the SAFETY SYSTEM train. +J:l.i-s-+G-aee-resses a ha~&HS--eveA-t-rnat-&a:Hses-eamage-te-a-&AF-*TY. S~+e-M,eF-a str1:mtu-re-eema~A+R§-SA.-F-*-~-M-e&ffip0-neA-ts,-ReerleEl-fe.F-#te-sb!-f,FeA-H3-f)eratl n g +ru&-OORGittoo-sfg-r-'ti:HBa-AtJ-..J-redu-Bes the maF§-fn-te a loss oF-peteRtial-l o ss of a fi s&f&A-pfWH&t OOfF~-anEl-there-fefe-Fepre-&00t&-at1-a ctual o r-ic1eteRti-al-st:1B&taA#al-eew.-aEJati&Fi-0f-rne-1ev-el-e:f safe-ty-ef the pl-a-A-h *Ah-4-J3+aoores-ses-Ei-amage-to a SA F ETY-S¥-S+E-l\A-tr-a+ft-t.A.at-tS-i-A-seF-Vi*el-e-peratisA-SfABB i-fl.El~raaei+y---a-va+J.able. The ifl-*l+&atioo-s-ef deg radee-peoorma-Ase-sheH~ sigA-i-#eaA-t-&AGl:t§-Me-ca use ooRee-FR-Fe§-aFEl-H-1§-me-el:}eFaei+i-t-y-eF-reli a bi I ity of-#te-SA.f:-§.1:¥ SY-ST~ E-Ab-1.b.2 aoorosses-eama-§e-to a SAFETY S¥ST-EM compeRe-A-t-#1-at-i-s-AE:H-m S*fVfG-eff}f)effiH&fi-eHBaffi~L3i3i38 rent th ffi-1:1§-A-+AGi cations ale ne,-eF-~ture contain fR§--&AF-ET¥ SYSTEM-Bompooe-nts. OpefffiefS-will make this-El-etefffi+Ration basee-oo-the-teta+ity e-f-avai+ae-le-event-aAe-El-am-agEl--fepB-FHffieFmatie&.-Tus--i-s-+AteA-Gee-1B-f)e-a-ooef-assessmeHt oo-keq u iring I e ng-tRy-a-Rawsi-s-e-f-.ft1:1a n tifi catie-A-ffi-ffie-e-amage-:- Escalation of the emergency classification level would be via IC CS1 or AS1. This EAL is the cold condition equivalent of hot condition EAL SA8.1. Reference(s):

1. EP FAQ 2016-002 2. NEI 99-01 CA6 Page 113 of 296 Q -::-:==-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category E -Independent Spent Fuel Storage Installation (ISFSI) EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)

An independent spent fuel storage installation (ISFSI) is a complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. A significant amount of the radioactive material contained within a canister must escape its packaging and enter the biosphere for there to be a significant environmental effect resulting from an accident involving the dry storage of spent nuclear fuel. An Unusual Event is declared on the basis of the occurrence of an event of sufficient magnitude that a loaded cask CONFINEMENT BOUNDARY is damaged or violated.

The RBS ISFSI is located wholly within the plant PROTECTED AREA. Therefore any security event related to the ISFSI are classified under Category H1 security event related EALs. Page 114 of 296 ~=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Catego'ry: . ISFSI Subcategory: Confinement Boundary Initiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARY EAL:* EU1.1 Unusual Event Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading on the surface of a loaded spent fuel cask (HI-STORM overpack) > EITHER of the following:

60 mRem/hr (Y + 11) on the top of the overpack
600 mRem/hr (Y + 11) on side of the overpack (excluding inlet and outlet ducts) Mode Applicability:

All I Definition(s): CONFINEMENT BOUNDARY-The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the RBS ISFSI, Confinement Boundary is defined as the Holtec System Multi-Purpose Canister (MPC). INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI): A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. Basis: This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of a storage cask containing spent fuel. It applies to irradiated fuel that is licensed for dry storage beginning at the point that the loaded storage cask is sealed. The issues of concern are the creation of a potential or actual release path to the environment, degradation of one or more fuel assemblies due to environmental factors, and configuration changes which could cause challenges in removing the cask or fuel from storage. \_ The existence of "damage" is determined by radiological survey. The specified EAL threshold values* correspond to 2 times the cask technical specification values. The technical specification (licensing bases document) multiple of "2 times", which is also used in Recognition Category A IC AU1, is used here to distinguish between non-emergency and emergency conditions (ref. 2). 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 Page 11 5 of 296 -~-Entergy

River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases "on-contact" dose rate limit is exceeded may be determined based on measurement of a dose rate at some distance from the cask.
Security-related events for ISFSls are covered under ICs HU1 and HA1. Reference(s):

1. USAR 9.1.2.5 Holtec HI-STORM Dry Spent Fuel Storage System 2. RBS HI-STORM 100 SYSTEM Certificate of Compliance for Spent Fuel Storage Casks 1 Amendment 5, Appendix A Technical Specifications for the HI-STORM 100 Cask System Section 5.7.4 3. NEI 99-01 E-HU1 Page 116 of 296 Q -::::::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category F -Fission Product Barrier Degradation EAL Group: Hot Conditions (RCS temperature

> 200°F); EALs in this category are applicable only in one or more hot operating modes. EALs in this category represent threats to the defense in depth design concept that precludes the release of highly radioactive fission products to the environment. This concept relies on multiple physical barriers any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment. The primary fission product barriers are: A. Fuel Clad Barrier (FCB): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets. B. Reactor Coolant System Barrier (RCB): The RCS Barrier is the reactor coolant system pressure boundary and includes the RPV and all reactor coolant system piping out to and including the isolation valves. C. Containment Barrier (CNB): The Containment Barrier includes the drywell, the containment, their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves. Containment Barrier thresholds are used as criteria for escalation of the Emergency Classification Level (ECL) from an Alert to a Site Area Emergency or a General Emergency. The EALs in this category require evaluation of the loss and potential loss thresholds listed in the fission product barrier matrix of Table F-1. "Loss" and "Potential Loss" signify the relative damage and threat of damage to the barrier. "Loss" means the barrier no longer assures containment of radioactive materials. "Potential Loss" means integrity of the barrier is threatened and could be lost if conditions continue to degrade. The number of barriers that are lost or potentially lost and the following criteria determine the appropriate emergency classification level: Alert: Any loss or any potential loss of either Fuel Clad or RCS Barrier Site Area Emergency: Loss or potential loss of any two barriers General Emergency: Loss of any two barriers and loss or potential loss of third barrier The logic used for emergency classification based on fission product barrier monitoring should reflect the following considerations:

The Fuel Clad Barrier and the RCS Barrier are weighted more heavily than the Containment Barrier. Page 11 7 of 296

-::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases

Unusual Event ICs associated with RCS and Fuel Clad Barriers are addressed under System Malfunction ICs.
For accident conditions involving a radiological release, evaluation of the fission product barrier thresholds will need to be performed in conjunction with dose assessments to ensure correct and timely escalation of the emergency classification.

For example, an evaluation of the fission product barrier thresholds may result in a Site Area Emergency classification while a dose assessment may indicate that an EAL for General Emergency IC AG1 has been exceeded.

The fission product barrier thresholds specified within a scheme reflect plant-specific RBS design and operating characteristic's.
As used in this category, the term RCS leakage encompasses not just those types defined in Technical Specifications but also includes the loss of RCS mass to any location-inside the primary containment, an interfacing system, or outside of the primary containment.

The release of liquid or steam mass from the RCS due to the designed/expected operation of a relief valve is not considered to be RCS leakage.

At the Site Area Emergency level, EAL users should maintain cognizance of how far present conditions are from meeting a threshold that would require a General Emergency declaration.

For example, if the Fuel Clad and RCS fission product barriers were both lost, then there should be frequent assessments of containment radioactive inventory and integrity. Alternatively, if both the Fuel Clad and RCS fission product barriers were potentially lost, the Emergency Director would have more assurance that there was no immediate need to escalate to a General Emergency. \ i. Page 118 of 296' e -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Any loss or any potential loss of either Fuel Clpd or RCS EAL: FA1.1 Alert Any loss or any potential loss of either Fuel Clad or RCS barrier (Table F-1) Mode Applicability: 1 -Power Operation, 2 -Startup, 3 -Hot Shutdown Definition(s): None Basis: Fuel Clad, RCS and Containment comorise the fission product barriers. Table F-1 lists the fission product barrier thresholds, bases and references. At the Alert classification level. Fuel Clad and RCS barriers are weiqhted more heavily than the Containment barrier. Unlike the Containment barrier, loss or potential loss of either the Fuel Clad or RCS barrier mav result in the relocation of radioactive materials or degradation of core coolina capability. Note that the loss or potential loss of Containment barrier in combination with loss or ootential loss of either Fuel Clad or RCS barrier results in declaration of a Site Area Emergency under EAL FS1 .1 Reference(s):

1. NEI 99-01 FA1 Page 119 of 296 e -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss or potential loss of any two barriers EAL: FS1.1 Site Area Emergency Loss or potential loss of any two barriers (Table F-1) Mode Applicability: 1 -Power Operation, 2 -Startup, 3 -Hot Shutdown Definition(s): IMMINENT -The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. Basis: Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 lists the fission product barrier thresholds, bases and references. At the Site Area Emeraency classification level, each barrier is weighted equa!lv. A Site Area Emergency is therefore appropriate for any combination of the following conditions: e One barrier loss and a second barrier loss (i.e., loss -loss) o One barrier loss and a second barrier ootenjial loss (i.e .. loss -potential loss) @ One barrier potential loss and a second barrier potential loss (i.e., potential loss -potential loss) At the Site Area Emergency classification level. the abilitv to dynamical Iv assess the proximity of present conditions with respect to the threshold for a General Emergency is important. For examole. the existence of Fuel Clad and RCS Barrier loss thresholds in addition to offsite dose assessments would require continual assessments of radioactive inventory and Containment inteqrity in anticipation of reaching a General Emergency classification. Alternatively, if both Fuel Clad and RCS potential loss thresholds existed, the Emeraency Director wouid havtl greater assurance that escalation to a General Emergency is less IMMINENT. Reference(s):

1. NEI 99-01 FS1 Page 120 of 296 e -::==-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss of any two barriers and loss or potential loss of third barrier EAL: FG1.1 General Emergency Loss of any two barriers AND Loss or potential loss of the third barrier (Table F-1) Mode Applicability: 1 -Power Operation, 2 -Startup, 3 -Hot Shutdown Definition(s): None Basis: Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 lists the fission product barrier thresholds, bases and references. At the General Emergency classification leve! each barrier is weiqhted equally. A General Emergency is therefore aQpropriate for an~ombination of the foliowing conditions: e Loss of Fuel C!ad, RCS and Containment Barriers 0 Loss of Fuel Clad and RCS Barriers with potential loss of Containrnent Barrier Loss of RCS and Containment Barriers with potential loss of Fuel Clad Barrier 0 Loss of Fuel Clad and Containment Barriers with potential loss of RCS Barrier Reference(s):

1. NEI 99-01 FG1 Page 121 of 296

. -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Table F-1 Fission Product Barrier Threshold Matrix & Bases Table F-1 lists the threshold conditions that define the Loss and Potential Loss of the three fission product barriers (Fuel Clad, Reactor Coolant System, and Containment). The table is structured so that each of the three barriers occupies adjacent columns. Each fission product barrier column is further divided into two columns; one for Loss thresholds and one for Potential Loss thresholds. The first column of the table (to the left of the Fuel Clad Barrier Loss column) lists the categories (types) of fission product barrier thresholds. the fission product barrier categories are: A. RPV Water Level B. RCS Leak Rate C. Containment Conditions D. Containment Radiation / RCS Activity E. Containment Integrity or Bypass F. Emergency Director Judgment Each category occupies a row in Table F-1 thus forming a matrix defined by the categories. The intersection of each row with each Loss/Potential Loss column forms a cell in which one or more fission product barrier thresholds appear. If NEI 99-01 does not define a threshold for a barrier Loss/Potential Loss, the word "None" is entered in the cell. Thresholds are assigned sequential numbers within each barrier column beginning with number one (ex., FCB1, FCB2 ... FCB6). C If a cell in Table F-1 contains more than one numbered threshold, each of the numbered thresholds, if exceeded, signifies a Loss or Potential Loss of the barrier. It is not necessary t9 exceed all of the thresholds in a category l;)efore declaring a barrier Loss/Potential Loss. Subdivision of Table F-1 by category facilitates association of plant conditions to the applicable fission product barrier Loss and Potential Loss thresholds. This structure promotes a systematic approach to assessing the classification status of the fission product barriers. When equipped with knowledge of plant conditions related to the fission product barriers, the EAL-user first scans down the category column of Table F-1, locates the likely category and then reads across the fission product barrier Loss and Potential Loss thresholds in that category to determine if a threshold has been exceeded. If a threshold has not been exceeded, the EAL-user proceeds to the next likely category and continues review of the thresholds in the new category If the EAL-user determines that any threshold has been exceeded, by definition, the barrier is lost or potentially lost -even if multiple thresholds in the same barrier column are exceeded, only that one barrier is lost or potentially lost. The EAL-user must examine each of the three fission product barriers to determine if other barrier thresholds in the category are lost or Page 122 of 296 Q -=-==-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases potentially lost. For example, if containment radiation is sufficiently high, a Loss of the Fuel Clad and RCS Barriers and a Potential Loss of the Containment Barrier can occur. Barrier Losses and Potential Losses are then applied to the algorithms given in EALs FG1 .1, FS1 .1, and FA 1 .1 to determine the appropriate emergency classification. In the remainder of this Attachment, the Fuel Clad Barrier threshold bases appear first,

followed by the RCS Barrier and finally the Containment Barrier threshold bases. In each barrier, the bases are given according to category Loss followed by category Potential Loss beginning with Category A, then 8, ... ,F. Page 123 of 296 Category A RPVWater FCB1 Level B RCS Leak Rate C CTMT Conditions D FCB3 CTMTRad / FCB4 RCS Activity E CTMT Integrity or Bypass F FCB5 Emergency Director Judgment -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Table F-1 Fission Product Barrier Threshold Matrix Fuel Clad Barrier (FCB) Reactor Coolant System Barrier (RCB) Containment Barrier (CNB) Loss Potential Loss Loss Potential Loss Loss Potential Loss FCB2 RPV water level cannot be RCB1 RPV water level cannot be restored and maintained>

restored and maintained> SAP entry is required -162 in. (TAF) -162 in. (TAF) None None CNB1 SAP entry is required or cannot be determined or cannot be determined CNB2 UNISOLABLE primary RCB2 UNISOLABLE break in any of RCB4 UNISOLABLE primary system system leakage that results the following: leakage that results in in exceeding EITHER: . Main steam lines exceeding EITHER: , One or more EOP-3 Max RCIC steam Line , One or more EOP-3 Max Safe area radiation None None . Normal area radiation operating value that can be None . RWCU operating value (Table F-2) read in the Control Room . Feedwater

One or more Isolation (Table F-2) RCB3 Emergency Depressurization Temperature alarms , One or more EOP-J Max is required (Table F.-2) Safe area temperature operatinq value (Table F-2) CNB5 Containment pressure>

15 psig CNB3 UNPLANNED rapid drop in containment pressure following CNB6 Drywell or containment RCB5 Drywell pressure > 1.68 psid containment pressure rise hydrogen concentration> 4% None None None due to RCS leakage CNB4 Containment pressure CNB7 Parameters cannot be restored response not consistent with and maintained within the safe LOCA conditions zone of the HCTL curve (EOP Figure 2) Containment radiation (RMS-RE16) > 3,000 R/hr RCB6 Drywell radiation (RMS-RE20) None > 30 R/hr None None CNB8 Containment radiation Coolant activity (RMS-RE16) > 12,000 R/hr > 300 µCi/gm dose equivalent 1-131 CNB9 UNISOLABLE direct downstream pathway to the environment exists after None None None None Containment isolation signal None ' CNB101ntentional Containment venting per EOPs Any condition in the opinion FCB6 Any condition in the opinion of RCB7 Any condition in the opinion of RCB8 Any condition in the opinion of CNB11 Any condition in the opinion of CNB12Any condition in the opinion of of the Emergency Director the Em erg ency Director that the Emergency Director that the Emergency Director that the Emergency Director that the Emergency Director that that indicates loss of the fuel indicates potential loss of the indicates loss of the RCS indicates potential loss of the indicates loss of the indicates potential loss of the clad barrier fuel clad barrier barrier RCS barrier Containment barrier Containment barrier Page 124 of 296 e -~*Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Fuel Clad Category: A. RPV Water Level Degradation Threat: Loss Threshold: I FCB1 SAP entry is required Definition(s): None Basis: Emergency Operating Procedure (EOPsl.§.Qecifv entry to ihe Severe Accident Procedures (SAPs) when core cooling is severely challenged. These EOPs orovide instructions to ensure adequate core coolina by maintaining RPV water level above prescribed limits or operatina sufficient RPV injection sources when level cannot be determined (ref. 1, 2). The EOP conditions requiring SAP entry represent a challenae to core cooling and are the minimum va!u_es to assure core coolinq without further deqradation of the clad. This threshold is also a Potentiai Loss of the Containment bar[jer (Ci\!B1 ). Since SAP entry occurs after core uncoverv has occurred a Loss of the RCS barrier exists (RCB1 ). SAP entry_,_ therefore. reoresents a Loss of two barriers and a Potential Loss of a third, which requires a General Emergency classification. The Loss threshold represents the EOP requirement for entry into the SAPs. This is identified in the BVVROG EPGs/SAGs when adequate core coolinq cannot be assured.+fle Loss threshold repFeSe-Rt-s-th~1irement-fef-jmmary cofttai-RmeRt-#eoEl+Rg;-T-his is feefttifiea tft-#l~~Pftras~ff!Bfy-Geft~oatR§-1-s-Req-Hiffid~ appears. Since a s~Gi-fic RP\/ watei=-+e-vet-i-s-net speci:fieEI-Aere, the Los-s thresAOl-~hfa.se, ~ai-n-mefrt:--#eeffifl-§-f~lso accommodates the EOP need to flood tR& p#FRary coAt-atAffl~wate-r level canoot be determinee--aA-Ei core damage due to inadequa~-ere cooling is 13-e-Hevee-te-e-e-eeGtlffi-R&- Reference(s):

1. EOP-1 RPV Control 2. EOP-4 RPV Flooding 3. EP FAQ 2015-004 4. NEI 99-01, RPV Water Level Fuel Clad Loss 2.A Page 125 of 296 e *-===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Fuel Clad I Category:

A. RPV Water Level Degradation Threat: Potential Loss Threshold: FCB2 RPV water level cannot be restored and maintained> -162 in. (TAF) or cannot be rdetermined Definition(s): None Basis: An RPV water level instrument readinq of -162 in. indicates RPV level is at the top of active fuel (TAF) (ref. 1 ). When RPV level is at or above the TAF. the core is completely submerged. Core submergence is the most desirable means of core cooling. When RPV level is below TAF, the uncovered portion of the core must be cooled by less reliable means (i.e., steam cooling or spray coolinq). If core uncovery is threatened, the EOPs specify alternate, more extreme, RP\/ water level control measures in order to restore and maintain adequate core cooling. Since core uncoverv begins if RPV level drops to TAF. the level is indicative of a challenge to core cooling and the Fuel Clad barrier. When RPV water level cannot be determined, EOPs require entry to EOP-4. RPV Floodino. RPV water level indication provides the primary means of l<nowinq if adequate core cooling is being maintained (ref. 2). When al! means of determining RPV water level are unavailable, the fuel clad barrier is threatened and reliance on alternate means of assuring adequate core cooling must be attempted. The instructions in EOP-4 specify these means, which include emergency depressurization of the RP\/ and injection into the RPV at a rate needed to flood to the elevation of the main steam lines or hold RPV pressure above the Minimum Steam Cooling Pressure (in scram-failure events). If RP\/ water level cannot be determined with respect to the top of active fueL a potential loss of the fuel clad barrier exists. Note that EOP-1A, RPV Control. ATWS. may require intentionally lowering RPV water level to -162 in. and control level between the Minimum Steam Cooling RPV Water Level (MSCRWL) and the top of active fuel (ref. 3). Under these conditions, a high-power ATWS event exists and requires at least a Site Area Emerqency classification in accordance with the System Malfunction -RPS Failure EALs. In high-power ATWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level to the top of active fuel in order to reduce reactor power. RPV water Page 126 of 296

River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases level is then controlled between the top of active fuel and the Minimum Steam Cooling RPV Water Level (MSCRWL). Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority. For such events, G-s EALs SA-e-SA6.1 or ~SS6.1 will dictate the need for emergency classification. This water level corresponds to the top of the active fuel and is used in the EOPs to indicate a challenge to core cooling. The RPV water level threshold is the sarne as RCS barrier Loss threshold 2-c-ARCB1. Thus, this threshold indicates a Potential Loss of the Fuel Clad barrier and a Loss of the RCS barrier that appropriately escalates the emergency classification level to a Site Area Emergency. This threshold is considered to be exceeded when, as specified in the site-specific EOPs, RPV water cannot be restored and maintained above the specified level following depressurization of the RPV (either manually, automatically or by failure of the RCS barrier) or when procedural guidance or a lack of low pressure RPV injection sources preclude Emergency RPV depressurization. EOPs allow the operator a wide choice of RPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurization of the RPV in order to facilitate RPV water level control with low-pressure injection sources. In some events, elevated RPV pressure may prevent restoration of RPV water level until pressure drops below the shutoff heads of available injection sources. Therefore, this Fuel Clad barrier Potential Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV

depressurization in an attempt to minimize loss of RPV inventory.

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

1. EOP-1 RPV Control 2. EOP-4 RPV Flooding 3. EOP-1A RPV Control, ATWS 4 NEI 99-01 RPV Water Level Potential Loss 2.A Page 127 of 296 /

River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases -B,arrier: Fuel Clad Category: B. RCS Leak Rate Degradation Threat: Loss Threshold: Page 128 of 296 "=:::;::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases / Barrier: Fuel Clad Category: B. RCS Leak Rate Degradation Threat: Potential Loss Threshold: I None Page 129 of 296 -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier:. Fuel Clad Category: C. CTMT Conditions Degradation Threat: Loss Threshold: None Page 130 of 296 @ -:::::::-, Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Fuel Clad Category: C. CTMT Conditions Degradation Threat: Potential Loss Threshold: None Page 131 of 296 Q *===-Entergy River Bend Station EAL Basis D,ocument Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Fuel Clad Category: D. CTMT Radiation/ RCS Activity Degradation Threat: Loss Threshold: FCB3 Containment radiation (RMS-RE16) > 3,000 R/hr Definition(s): None Basis: The containment radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals 300 µCi/gm dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to a-R-a-J3-Pfffi<+FAate-range el-2% to §..%.significant fuel clad damageJref. 1 ). 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 -Gc-+RCB6 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 EGL to a Site Area Emergency. There is no Fuel Clad barrier Potential Loss threshold associated with RGS--Aeti-v+t-y-/ Gootai-Affiem-CTMT Radiation/RCS Activitv. Reference(s):

1. Calculation G13.18.9.4-045 Containment Doses for E;mergency Action Levels (EALs) 2. NEI 99~01 Primary Containment Radiation Fuel Clad Loss 4.A Page 132 of 296 I, River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Fuel Clad Category:

D. CTMT Radiation/ RCS Activity Degradation Threat: Loss Threshold: FCB4 Coolant activity> 300 µCi/gm dose equivalent 1-131 Definition(s): None Basis: This threshold indicates that RCS radioactivity concentration is greater than 300 µCi/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. It is recognized that sampre coilection and anaiysis of reactor coolant with highly elevated activity levels could require several hours to complete. Nonetheless, a sample-related threshold is included as a backup to other indications. There is no Fuel Clad barrier Potential Loss threshold associated with RGS-A6tivrty-/ Ge-Rta~A-mentCTfVlT Radiation/RCS Activity. Reference(s):

1. NEI 99-01 RCS Activity Fuel Clad Loss 1.A L_ Page 133 of 296

~. *===* Entergy Barrier: Category:

River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Fuel Clad D. CTMT Radiation/ RCS Activity Degradation Threat: Potential Loss Threshold: None Page 134 of 296 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Fuel Clad Category: E. CTMT Integrity or Bypass Degradation Threat: Loss Threshold: I None Page 135 of 296 --Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Fuel Clad Category: E. CTMT Integrity or Bypass Degradation Threat: Potential Loss Threshold: None Page 136 of 296 Q ~Enteigy Barrier: Category: River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Fuel Clad F. Emergency Director Judgment Degradation Threat: Loss Threshold: FCBS Any condition in the opinion of the Emergency Director that indicates loss of the Fuel Clad barrier Definition(s): None Basis: This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad barrier is lost. Reference(s):

1. NEI 99-01 Emergency Director Judgment Fuel Clad Loss 6.A Page 137 of 296

~* -:"-::::=* Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Fuel Clad Category: F. Emergency Director Judgment Degradation Threat: Potential Loss Threshold: FCB6 Any condition in the opinion of the Emergency Director that indicates potential loss of the Fuel Clad barrier Definition(s): None Basis: This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored. Reference(s):

1. NEI 99-01 Emergency Director Jud,gment Fuel Clad Potential Loss 6.A Page 138 of 296 e -::==-Entergy Barrier: Category:

River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Reactor Coolant System A. RPV Water Level Degradation Threat: Loss Threshold: RCB1 RPV water level cannot be restored and maintained> -162 in. (TAF) or cannot be determined Definition(s): None Basis: An RPV *water level instrument reading of -162 in. indicates level is at the top of active fuel (TAF) __ {.@f. 1 'L TAF is siqnificantly lower than Hie normal operating RPV level control band. To reacbJhis level. RPV inventory loss would have greviously required isolation of the RCS and Containment barriers, and initiation of all ECCS. If RPV water level cannot be maintained above TAF, FCCS and other sources of RPV injection have been ineffective or incapable of reversing the lowerinq level trend. Tr1e cause of the loss of RPV inventoryj_s therefore assumed to be a LOC,~. By definition. a LOCA event is a Loss of the RCS barrier. When RPV water level cannot be determined, EOPs require entry to EOP-4, RPV Floodina. RPV water ievel indication provides the primarv means of knowina if adequate core coolinq is being maintained (ref. 2). The instructions in EOP-4 specify emergencv depressurization of the RPV. which is defined to be a Loss of the RCS barrier (RCS Loss RCB3). Note that EOP-1A. RPV Control, ATWS, may reguire intentionally lowering RPV water level to -'162 in. and control level betvveen the Minimum Steam Coo!inq RPV Water Level (MSCRWL) and the top of active fuel (ref. 3). Under these conditions. a high-power ATWS event exists and reauires at least a Site Area Emergency classification in accordance with the System Malfunction -RPS Failure EALs. In high-power A TWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level to the top of active fuel in order to reduce reactor power. RPV water level is then controlled between the top of active fuel and the Minimum Steam Cooling RPV Water Level (MSCRWL). Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority. For such events, ICs SA-e-SA6 or S&e-SS6 will dictate the need for emergency classification. Page 139 of 296 -~ Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases This water level corresponds to the top of active fuel and is used in the EOPs to indicate challenge to core cooling. The RPV water level threshold is the same as Fuel Clad barrier Potential Loss threshold &.AFCB2. Thus, this threshold indicates a Loss of the RCS barrier and Potential Loss of the Fuel Clad barrier and that appropriately escalates the emergency classification level to a Site Area Emergency. This threshold is considered to be exceeded when, as specified in the site-specific EOPs, RPV water cannot be restored and maintained above the specified level following depressurization of the RPV (either manually, automatically or by failure of the RCS barrier) or when procedural guidance or a lack of low pressure RPV injection sources preclude Emergency RPV depressurization , EOPs allow the operator a wide choice of RPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurizatio.n of the RPV in order to facilitate RPV water level control with low-pressure injection sources. In some events, elevated RPV pressure may prevent restoration of RPV water level until pressure drops below the shutoff heads of available injection sources. Therefore, this RC$ barrier Loss is met only after either: 1) the RPV has been depressurized, or requir,ed emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory. The term, "cannot be restored and maintained above," means the value of RPV water level is not able to be brought above the specified limit (top of active fuel). The determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the current value is below the top of active fuel, but does not permit extended operation beyond the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained. There is no RCS barrier Potential Loss threshold associated with RPV Water Level. Reference(s):

1. EOP-1 RPV Control 2. EOP-'4 RPV Flooding 1 3. EOP-1A RPV Control, ATWS 4. NEI 99-01 RPV Water Level RCS Loss 2.A Page 140 of 296 n .. -:::::::-Entergy Barrier: Category:

River Bend Station EAL-Basis Document Revision XXX I Attachment 1 -Emergency Action Level Technical Bases Reactor Coolant System A. RPV Water Level Degradation Threat: Potential Loss Threshold: None Page 141 of 296 \ e -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Category: Reactor Coolant System B. RCS Leak Rate Degradation Threat: Loss Threshold: RCB2 UNISOLABLE break in any of the following:

Main steam line
RCIC steam line
RWCU
Feedwater Definition(s):

UNISOLABLE -An open or breached system line that cannot be isolated, remotely or locally. Basis: Failure to isolate the leak (from the Control Room or locally), within 15 minutes or if knovvn that the leak cannot be isolated within 15 minutes. from the start of the leak requires immediate ciassiFlcation. The conditions of this threshold include required containment isolation failures allowing a flow path to the environment. A release pathway outside primarv containment exists when flow is not prevented by downstream isolations. ln the case of a failure of both isolation valves to close but in which no downstream flowoath exists. emergency declaration under this threshold would not be required. Similarly. if the emergency response requires the normal process* flow of a system outside containment (e.o .. EOP requirement to bypass MSIV low RPV water level interlocks and maintain the main condenser as a heat sink using main turbine bypass valves), the threshold is not met. The combination of these threshold conditions represent the loss of both the RCS and Containment (see Loss CNB9) barriers and justifies declaration of a Site Area Emergencv (i.e .. Loss or Potential Loss of anv two barriers). Even thouah RVVCU and Feedwater systems do not contain steam. they are included in the list because an UNISOLABLE break could result in the high-pressure discharge of fluid that is flashed to steam from relatively large volume svstems directly connected to the RCS. Page 142 of 296 -~=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Large high-energy lines that rupture outside primary containment can discharge significant amounts of inventory and jeopardize the pressure-retaining capability of the. RCS until they are isolated. If it is determined that the ruptured line cannot be promptly isolated. remotely or locally-fffif"R-the Comffil-Room, the RCS barrier Loss threshold is met. Reference(s):

1. NEI 99-01 RCS Leak Rate RCS Loss 3.A Page 143 of 296

-::::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: B. RCS Leak Rate Degradation Threat: Loss Threshold: RCB3 Emergency Depressurization is required Definition(s): None Basis: Plant symptoms reguiring Eme..rgencv RPV Depressurization per the EOPs are indicative of a loss of the RCS barrier (ref. 1. 2). Emergency RPV Depressurization in accordance with the EOPs is indicative of a loss of the RCS barrier. If Emergency RPV Depressurization is performed, the plant operators are directed to open safety relief valves (SRVs)-aREl-keep-them-BfIBR. Even though the RCS is being vented into the suppression pool, a Loss of the RCS barrier exists due to the diminished effectiveness of the RCS to retain fission products within its boundary. EOP-*l Emergency Depressurization allows terminating the depressurization if necessary to maintain RCIC as an injection source. This would require closing the SRVs. Even though the SRVs may be reclos.ed, this threshold is still met due to the requirement for an Emeraency Depressurization having been met (ref. 2). Reference(s):

1. EOP-1 RPV Control Emergency Depressurization

2. EP FAQ 2015-003 3. NEI 99-01 RCS Leak Rate RCS Loss 3.8 '\ Page 144 of 296

-::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category:

8. RCS Leak Rate Degradation Threat: Potential Loss Threshold:

RCB4 UNISOLABLE primary system leakage that results in exceeding EITHER:

One or more EOP-3 Max Normal area radiation operating value (Table F-2)
One or more Isolation Temperature alarms (Table F-2) Table F-2 Secondary Containment Operating Values Area Temperatures Parameter Isolation Temperature Max Safe Main Steam Line Tunnel 173°F (P601-19A-A 1/A3/B1/B3) 200°F RHR Equipment Area 1 (A) 117°F (P601-20A-B4) 200°F RHR Equipment Area 2 (B) 117°F (P601-20A-B4) 200°F RCIC Equipment Area 182°F (P601-21A-B6) 200°F RWCU Pump Room 1 (A)/ 2 (B) 165°F (P680-1 A-A2/B2) 200°F Area Radiation Levels Parameter Max Normal Max Safe HPCS Area (1212) Grid 2 8.20E+01 mR/hr-9.5E+03 mR/hr RHR Equipment Room A (1213) Grid 2 8.20E+01 mR/hr 9.5E+03 mR/hr RHR Equipment Room B (1214) Grid 2 8.20E+01 mR/hr 9.5E+03 mR/hr RHR Equipment Room C (1215) Grid 2 8.20E+01 mR/hr 9.5E+03 mR/hr LPCS Equipment Room (1216) Grid 2 8.20E+01 mR/hr 9.5E+03 mR/hr HPCS Penetration Area (1217) Grid 2 8.20E+01 mR/hr 9.5E+03 mR/hr LPCS Penetration Area (1218) Grid 2 8.20E+01 mR/hr 9.5E+03 mR/hr RCIC Equipment Room (1219) Grid 2 1.20E+02 mR/hr 9.5E+03 mR/hr Definition(s):

FLOODING -A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area. UN/SOLABLE -An open or brea~hed system line that cannot be isolated, remotely or locally. Page 145 of 296 e -:;=::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Basis: Failure to isolate the leak (from the Control Room or locally), within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification. The presence of elevated general area temperatures or radiation levels in the Secondary Containment may be indicative of UNlSOLABLE primary system leakage outside the containment. The EOP-3 Max Normal and Isolation Temperature alarm setooint values in Table F-2 define this RCS threshold because they are the maximum normal *operating/ Technical Specification Isolation values and signify the onset of abnormal system operation. When parameters reach this level, equipment failure or mis-operation may be occurrina. Elevated parameters may also adversely affect the ability to gain access to or operate equipme*nt within the affected area. The locations into which the primary system discharge is of concern correspond to the areas addressed in EOP-3. Secondary Containment and Radioactivity Release Control (ref. 1 ).

In general, multiple indications should be used to determine if a primary system is discharging outside Containment.

For example, a high area radiation condition does not necessarily indicate that a primary system is discharqing into the Secondary Containment since this may be caused by radiation shine from nearby steam lines or the movement of radioactive materials. Converselv, a hiah area radiation condition in conjunction with other indications (e.g. room FLOODING, high area temperatures, reports of steam in the secondary containment, an unexpected rise in feedwater flowrate, or unexpected main turbine control valve closure) may indicate that a primary system is discharging into the Secondary Containment. Potential loss of RCS based on primary system leakage outside the primary containment is determined from EOP temperature or radiation Max Normal Operating values in areas such as main steam line tunnel, RCIC, HPCI, etc., which indicate a direct path from the RCS to areas outside primary containment. A Max Normal Operating value is the highest value of the identified parameter expected to occur during normal plant operating conditions with all directly associated support and control systems functioning properly. An Isolation Temperature value is indicative of an UNISOLABLE leak when temperatures do not begin to recover as a result of the isolation actions following -the alarm and represents a Technical Specification limiting value. The indicators reaching the threshold barriers and confirmed to be caused by RCS leaka92. from a primary system warrant an Alert classification. A primary system is defined to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pr-essure will effect a reduction in the steam or water being discharged through an unisolated break in the system. Reference(s):

1. EOP-3 Secondary Containment and Radioactivity Release Control 2. NEI 99-01 RCS Leak Rate RCS Potential Loss 3.A Page 146 of 296 e ~Entergy Barrier: Category:

River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Reactor Coolant Systern C. CTMTConditions Degradation Threat: Loss Threshold: RCBS Drywell pressure > 1.68 psid due to RCS leakage Definition(s): None Basis: The drvwell high pressure scram setpoint is an entry condition to EOP-1, RPV Control. 'A high Containment pressure of greater than 0.3 psig is an entry condition to EOP-2, Primary Containment Control (ref. 1, ?). Normal containment pressure control functions (e.g., operation of drywell and containment cooling, vent using containment vessel purge. etc.) are specified in EOP-2 in advance of less desirable but more effective functions (e.g., Emergency Oepressurization. etc.). ln the design basis. containment pressures above the drywell high pressure scram setpoint are assumed to be the result of a high-energy release into the containment for which normal pressure control systems are inadequate or incapable of reversing the rising pressure trend. Pressures of this magnitude, however. can be caused by non-LOCA events such as a loss of drywell cooling or inability to control containment vent/purge (ref. 3). The threshold phrase" ... due to RCS leakaae" focuses the barrier failure on the RCS instead of the non-LOCA malfunctions that may adversely affect containment pressure. Drywell pressure

greater than i .68 psid*with corollary indications (e.g., drvwell temperature, indications of loss of RCS inventorv) should therefore be considered a Loss of the RCS barrier. Loss of drvwell cooling that results in pressure areater than -1.68 psid should not be considered an RCS barrier Loss. The (site specific valt!ej1 .68 psid p-FtffiafY-Bema+A-rnf-f:)rnssurevalue is the drywell high pressure setpoint which indicates a LOCA by automatically initiating the ECCS or equivalent makeup system. There is no RCS barrier Potential Loss threshold associated with Primary Containment PfessureCTMT Conditions.

Page 147 of 296 -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Reference(s):

1. EOP-1 RPV Control 2. EOP-2 Primary Containment Control 3. USAR Section 6.2.1 Containment Functional Design 4. NEI 99-01 Primary Containment Pressure RCS Loss 1.A Page 148 of 296

~Entergy Barrier: Category: River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Reactor Coolant System C. CTMT Conditions Degradation Threat: Potential Loss Threshold: Page 149 of 296 e -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: D. CTMT Radiation/ RCS Activity Degradation Threat: Loss Threshold: RCB6 Drywell radiation (RMS-RE20) > 30 R/hr Definition(s): N/A Basis: Under post-LOCA conditions coaxial cables used on the drvwell post-accident monitors RE20NB) are susceptible to Thermally Induced Currents (TIC). These currents may cause the drywell PAMs to read falsely high (-469 R/hr) on a rapid temperature rise and read falsely low on a rapid temperature drop. When accident temperature conditions stabilize indicated radiation dose rates would be more accurate. The duration of the spurious signal would last approximately 15 minutes. During the period of false readings operators should rely on other indications of RCS leakage including a rise in drywell temperature and pressure (RCB5). The drywell radiation monitor reading (38 R/hr rounded to 30 R/hr for readability) corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for Fuel Clad Barrier Loss threshold 4AFCB3 since it indicates a loss of the RCS Barrier only (ref. 1 ). There is no RCS barrier Potential Loss threshold associated with P-Rrna-Fy-Gofrt.ai-RmeAtCTMT Radiation/RCS Activity. Reference(s):

1. Calculation G13.18.9.4-045 Containment Doses for Emergency Action Levels (EALs) 2. NEI 99-01 Primary Containment Radiation RCS Loss 4.A Page 150 of 296

'$'Entergy Barrier:

Category:*

River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Reactor Coolant System D. CTMT Radiation/ RCS Activity Degradation Threat: Potential Loss Threshold: None Page 151 of 296

-:;::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: E. CTMT Integrity or Bypass Degradation Threat: Loss Threshold: None Page 152 of 296 ~Entergy Barrier: Category: River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Reactor Coolant System E. CTMT Integrity or Bypass Degradation Threat: Potential Loss Threshold: None Page 153 of 296 ! Q ~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: F. Emergency Director Judgment Degradation Threat: Loss Threshold: RCB7 Any condition in the opinion of the Emergency Director that indicates loss of the RCS barrier Definition(s): None Basis: This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is lost. Reference(s):

NEI 99-01 Emergency Director Judgment RCS Loss 6.A Page 154 of 296

-:;:::=-Entergy River Bend Station EAL Basis Document Revision XXX

Attachment 1 -Emergency Action Level Technical Bases ' Barrier: Reactor Coolant System Category:

F. Emergency Director Judgment Degradation Threat: Potential Loss Threshold: RCBS Any condition in the opinion of the Emergency Director that indicates potential loss of the RCS barrier Definition(s): None Basis: This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored. Refe re nee( s): 1. NEI 99-01 Emergency Director Judgment RCS Potential Loss 6.A Page 155 of 296 River Bend Station EAL Bc;isis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Containment Category: A. RPV Water Level Degradation Threat: Loss Threshold: None Page 156 of 296 e *===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Containment Category: A. RPV Water Level Degradation Threat: Potential Loss Threshold: CNB1 SAP entry is required Definition(s): None Basis: EOPs specify entry to the SAPs when core cooling is severelv challenged. These EOPs provide instructions to ensure adequate core cooling bv maintainina RPV water level above prescribed limits or operatinq sufficient RPV iniection sources when level cannot be determined (ref. 1. 2). The EOP conditions requiring SAP entrv represent a challenge to core coolinq and are the minimum values to assure core coolin.fU'.Yithout further c;!_ggradation of the clad. This threshold is also a Loss of the Fuel Clad barrier (Loss FCB1 ). Since SAP entry occurs after core uncovery has occurred a Loss of the RCS barrier exists (Loss RCBi ). SAP entr~ therefore, represents a Loss of two barriers and a Potentiai Loss of a third, which requires a General Emergencv classification. The Potential Loss threshold is identical to the Fuel Clad Loss RPV Water Level threshold A-1-FCB1. The Potential Loss requirement for entry into the S/\Gs SAPs indicates adequate core cooling cannot be assured and that core damage is possible. BWR EPGs/SAGs (RBS term SAPs) specify the conditions when the EPGs are exited and SAGs are entered. Entry into SAGs is a logical escalation in response to the inability to assure adequate core cooling. PRA studies indicate that the condition of this Potential Loss threshold could be a core melt sequence which, if not corrected, could lead to RPV failure and increa-see-greater potential for primary containment failure. In conjunction with the RPV water level Loss thresholds in the Fuel Clad and RCS barrier columns, this threshold results in the declaration of a General Emergency. There is no Containment barrier Loss threshold associated with RPV \Nater Level. Page 157 of 296

-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emerge~cy Action Level Technical Bases Reference(s):

1. EOP-1 RPV Control 2. EOP-4 RPV Flooding 3. EP FAQ 2015-004 4. NEI 99-01 RPV Water Level PC Potential Loss 2.A Page 158 of 296

@\ *::==-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Containment Category: B. RCS Leak Rate Degradation Threat: Loss Threshold: CNB2 UNISOLABLE primary system leakage that results in exceeding EITHER:

One or more EOP-3 Max Safe area radiation operating value (Table F-2)
One or more EOP-3 Max Safe area temperature operating value (Table F-2) Table F-2 Secondary Containment Operating Values Area Temperatures*

Parameter Isolation Temperature Max Safe Main Steam Line Tunnel 173°F (P601-19A-A1/A3/B1/B3) 200°F RHR Equipment Area 1 (A) 117°F (P601-20A-B4) 200°F RHR Equipment Area 2 (B) 117°F (P601-20A-B4) 200°F RCIC Equipment Area 182°F (P601-21A-B6) 200°F RWCU Pump Room 1 (A)/ 2 (B) 165°F (P680-1A-A2/B2) 200°F Area Radiation Levels Parameter Max Normal Max Safe HPCS Area (1212) Grid 2 8.20E+01 mR/hr 9.5E+03 mR/hr RHR Equipment Room A (1213) Grid 2 8.20E+01 mR/hr 9.5E+03 mR/hr RHR Equipment Room B (1214) Grid 2 8.20E+01 mR/hr 9.5E+03 mR/hr RHR Equipment Room C (1215) Grid 2 8.20E+01 mR/hr 9.5E+03 mR/hr LPCS Equipment Room (1216) Grid 2 8.20E+01 mR/hr 9.5E+03 mR/hr HPCS Penetration Area (1217) Grid 2 8.20E+01 mR/hr 9.5E+03 mR/hr LPCS Penetration Area (1218) Grid 2 8.20E+01 mR/hr 9.5E+03 mR/hr RCIC Equipment Room (1219) Grid 2 1.20E+02 mR/hr 9.5E+03 mR/hr Page 159 of 296 e -::==-Entergy River Bend Station EAL Basis Document Revision XXX \ ~-----~---------------------------~ Attachment 1 -Emergency Action Level Tec.hnical Bases Definition(s ): FLOODING -A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area. UNISOLABLE -An open or breached system line that cannot be isolated, remotely or locally. Basis: Failure to isolate the leak (from the Control Room or locallv), within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification. The presence of elevated qeneral area temperatures or radiation levels in the secondary containment may be indicative of UNISOLABLE primary svstem leakage outside the containment. The Max Safe conditions define this Containment barrier threshold because they are indicative of pr,oblems in the secondary containment that are spreading and pose a threat to achieving a safe plant shutdown. This threshold addresses problematic discharges outside containment that mav not originate from a hiah-eneray line break. The locations into which the primary system discharge is of concern correspond to the areas addressed in EOP-3, Secondary Containment and Radioactivity Release Control (ref . .Jl In general, multiple indications should be used to determine if a primary svstem is discharoing outside containment. For example. a high area radiation condition does not necessarily indicate that a primary system is discharging into the secondary containment since this may be caused bv radiation shine from nearby steam lines or the movement of radioactive materials. Conversely, a hiah area radiation condition in conjunction with other indications (e.a. room FLOODING, high area temperatures, reports of steam in the secondary containment, an unexpected rise in feedwater flowrate, or unexpected main turbine control valve closure) may indicate that a primary system is discharging into the secondary containment. I The Max Safe Operating Temperature and the Max Safe Operating Radiation Level are each the highest value of these parameters at which neither: (1) equipment necessary for the safe shutdown of the plant will fail, nor (2) personnel access necessary for the safe shutdown of the plant will be precluded. EOPs utilize these temperatures and radiation levels to establish conditions under which RPV depressurization is required. There is no Containment barrier Potential Loss threshold associated with RCS Leak Rate. Reference(s):.

1. EOP-3 Secondary Containment and Radioactivity Release Control 2. NEI 99-01 RCS Leak Rate PC Loss 3.C Page 160 of 296

-===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Containment Category: B. RCS Leak Rate Degradation Threat: Potential Loss Threshold: None Page 161 of 296 -::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Containment Category: C. CTMT Conditions Degradation Threat: Loss Threshold: CNB3 UNPLANNED rapid drop in containment pressure following containment pressure rise Definition(s): 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: Rapid UNPLANNED loss of f}Fif"Aar-y--containment pressure (i.e., riot attributable to El-r-ywen Sf:)faYcontainment coolina or condensation effects) following an initial pressure tABre-a-se-rise indicates a loss of primary containment integrity. Rfi-mary-eentaffiment pressure shst!l-*1-i-Rerease as a re&t--J!-t-e-Hfla&S--aflfl-efleF§Y-Fe-l-ease-fA-te-t~eRtai-nFABHHFOm a +h us, pri~ment-p-Fe-SSt1re-RBHAC-Fe-a&~er-these---esn4itkms--i-REH*ates a loss-o-f f}Ftma-Fy-GOHiatRfFle-At-fmegfi-t-y-c These-This thresholds rely-relies on operator recognition of an unexpected response for the condition and therefore a specific value is not assigned. The unexpected (UNPLANNED) response is important because it is the indicator for a containment bypass condition. Reference(s):

1. NEI 99-01 Primary Containment Conditions PC Loss 1.A Page 162 of 296 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Containment Category:

C. CTMT Conditions Degradation Threat: Loss Threshold: CNB4 Containment pressure response not consistent with LOCA conditions Definition(s): 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: Rapid UN-PL/\NN*-D loss of pr:j.mary con~~et-atml:)utable to Eifywe-1+ Sf*IY-8-F-OOMeR--SatiB-A-&ffee.t-&t-:fellB-Wffig-aR-tRi-1:ial-pFe&St:tre inc rease-inei-c-ate-s-a-loss of p ri oornai n me nt integrity. Primary containment pressure should +ABIBa-se-rise as a result of mass and energy release into the primary containment from a LOCA. Thus, primary containment pre*ssure not mGF&a-s+ng-risino under these conditions indicates a loss of primary containment integrity. These thresholds rely on operator recognition of an unexpected response for the condition and therefore a specific value is not assigned. The unexpected (UNPLANNED) response is important because it is the indicator for a containment bypass condition. Reference(s):

1. USAR Table 6.2-7, Results of Containment Response Analysis 2. USAR Table 6.2-1, Containment Design Parameters

3. NEI 99-01 Primary Containment Conditions PC Loss 1.8 Page 163 of 296

=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Containment Category:

C. CTMT -Conditions Degradation Threat: Potential Loss Threshold: CNBS Containment pressure > 15 psig Definition(s): VALID -An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. Basis: 1 When the containment pressure exceeds the maximum allowable value (15 psig) (ref. 1 ), containment venting may be required even if offsite radioactivity release rate limits will be exceeded (ref. 2). This pressure is based on the containment desiqn pressure as identified in the accident analysis. lf this threshold is exceedeci, a challenqe to the containment structure has occurred because assumotions used in the accident analysis are no longer VALID and an unanalyzed condition exists. ThJs constitutes a Potential Loss of the Containment barrier even if a containment breach has not occurred. The threshold pressure is the primary containment internal design pressure. Structural acceptance testing demonstrates the capability of the primary containment to resist pressures greater than the internal design pressure. A pressure of this magnitude is greater than those expected to result from any design basis accident and, thus, represent a Potential Loss of the Containment barrier. Reference(s):

1. USAR Table 6.2-1, Containment Design Parameters

2. EOP-2 Primary Containment Control 3. NEI 99-01, Primary Containment Conditions PC Potential Loss 1.A Page 164 of 296

e. -:::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Containment Category:

C. CTMT Conditions Degradation Threat: Potential Loss Threshold: CNB6 Drywell or containment hydrogen concentration > 4% Definition(s): None Basis: / In the early stages of'a core uncovery event, it is unlikely that hydroqen buildup due to a core uncovery could result in an explosive mixture of dissolved gases in the containment. However, containment monitoring and/or sampling should be performed to verify this assumption and a General Emergency declared if it is determined that hydrogen concentration has exceeded the minimum necessary to support a hydrogen burn (4%). The Igniter System is designed to prevent hydrogen accumulation by locally burning hydrogen in a controlled manner as soon as the hydrogen enters the containment atmosphere and reaches the igniters. For high rates of hydrogen production. ignition occurs at the lowest concentration that can support ignition. I Following ignition, hydrogen is consumed through formation of diffusion flames where the gas enters the containment, thus controlling hydrogen concentration at approximately 4% (ref. 1 }. If hydrogen concentration reaches or exceeds the lower flammability limit, as defined in plant EOPs, in an oxygen rich environment, a potentially explosive mixture exists. If the combustible mixture ignites inside the containment, loss of the Containment barrier could occur. Reference(s):

1. EPSTG*0002 Appendix B EOP and SAP Bases 2. NEI 99-01 Primary Containment Conditions PC Potential Loss 1.8 \ \ Page 165 of 296 I I !

~Entergy Barrier: Category: River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Containment C. CTMT Conditions Degradation Threat: Potential Loss Threshold: CNB7 Parameters cannot be restored and maintained within the safe zone of the HCTL curve (EOP Figure 2) Definition(s): None Basis: The Heat Capacity Temperature Limit (HCTL) is the highest suppression pool temperature from which Emergency RPV Depressurization will not raise:

Suppression chamber temperature above the maximum temperature capability of the suppression chamber and equipment within the suppression chamber which may be required to operate when the RPV is pressurized, OR
Suppression chamber pressure above Primary Containment Pressure Limit-A, while the rate of energy transfer from the RPV to the containment is greater than the capacity of the containment vent. The HCTL is a function of RPV pressure, suppression pool temperature and suppression pool water level. It is utilized to preclude failure of the containment and equipment in the containment necessary for the safe shutdown of the plant and therefore, the inability to maintain plant parameters below the limit constitutes a potential loss of containment.

The term "cannot be restored and maintained above" means the parameter value(s) is not able to be brought within the specified limit. The determination requires an evaluation of system performance and availability in relation to the parameter value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained within a specified limit does not require immediate action simply because the currer.it value is outside the limit, but does not permit extended operation outside the limit the threshold must be considered reached as soon as it is apparent that operation within the limit cannot be attained. Page 166 of 296 -:::::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Reference(s):

1. EOP-2 Primary Containment Control \ 2. NEI 99-01 Primary Containment Conditions PC Potential Loss 1.C Page 167 of 296

-::==-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Containment Category: D. CTMT Radiation/RCS Activity Degradation Threat: Loss Threshold: None Page 168 of 296 -=:::::-Entergy River Bend Station EAL Basis Document Revision XXX

Attachment 1 -Emergency Action Level Technical Bases Barrier: Containment Category:

C. CTMT Radiation/RCS Activity Degradation Threat: Potential Loss Threshold: CNB8 Containment radiation (RMS-RE16) > 12,000 R/hr Definition(s): None Basis: !n order to reach this Containment barrier Potential Loss threshold. a loss of the RCS barrier (Loss RCB6) and a loss of the Fuel Clad barrier (Loss FCB3) have alreadv occurred. This threshold, therefore. represents_a General Emergency classifLcatioll.:. The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that 20% of the fuel cladding has failed (ref. 1 ). This level of fuel clad failure is well above that used to determine the analogous Fuel Clad Barrier Loss and RCS Barrier Loss thresholds. NUREG-1228, Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents, indicates the fuel clad failure must be greater than approximately 20% in order for there to be a major release of radioactivity requiring offsite protective actions. For this condition to exist, there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential loss of containment which woyld then escalate the emergency classification level to a General Emergency. There is no Containment barrier Loss threshold associated with CTfvlT Radiation/RCS Activity. Reference(s):

1. Calculation G13.18.9.4-045 Containment Doses for Emergency Action Levels (EALs) 2. NEI 99-01 NEI 99-01 Primary Containment Radiation Potential Loss 4.A Page 169 of 296 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Containment Category:

E. CTMT Integrity or Bypass Degradation Threat: Loss Threshold: CNB9 UNISOLABLE direct downstream pathway to the environment exists after Containment isolation signal Definition(s): UNISOLABLE -An open or breached system line that cannot be isolated, remotely or locally. VALID -An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. Basis: Failure to isolate the leak (from the Control Room or locallv), within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification. This threshold addresses failure of open isolation devices, which should close upon receipt of a manual or automatic containment isolation signal resulting in a significant radiological release pathway directly to the environment. The concern is the UNISOLABLE open pathway to the environment. A failure of the ability to isolate anv one line indicates a breach of containment integritv. This threshold also applies to a containment bypass due to a HPCS or LPCS line break outside containment with iniection check valve failure allowing an UNlSOLABLE direct pathway for RCS release to the environment. The use of the modifier "direct" in defining the release path discriminates against release paths through interfacing liquid systems or minor release pathways, such as instrument lines, not protected by the Primary Containment Isolation System (PCIS). Leakage into a closed system is to be considered only if the closed system is breached and thereby creates a significant pathway to the environment. Examples include UNiSOLABLE main steam line or RClC steam line breaks. UNlSOLABLE RWCU system breaks, and UNISOLABLE containment atmosphere vent paths. If the main condenser is available with an UNlSOLABLE main steam line. there may be releases throuqh the steam jet air ejectors and gland seal exhausters. These pathways Page 170 of 296 I -1 River Bend Station.EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases are monitored, hovvever. and do not meet the intent of an Ui\llSOLABLE release path to the environment. These minor releases are assessed usinq the Category A, Abnormal Rad Levels / Rad Effiuent. EALs. The existence of a filter is not considered in the threshold assessment. Filters do not remove fission product noble gases. In addition, a filter could become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream. EOP-2 Primary Containment Control. mav specifv primary containment venting and intentional bypassina of the containment isolation valve logic, even if offsite radioactivitv release rate limits are exceeded (ref. 1 ). Under these conditions with a VALID containment isolation sianal, the Containment barrier should be considered lost. Refer to CNB10. Following the leakage of RCS mass into primary containment and a rise in primary containment pressure, there may be minor radiological releases associated with allowable primary containment leakage through various penetrations or system components. Minor releases may also occur if a primary containment isolation valve(s) fails to close but the primary containment atmosphere escapes to an enclosed system. These releases do not constitute a loss or potential loss of primary containment but should be evaluated using the Recognition Category A ICs. There is no Containment barrier Potential Loss threshold associated with CTMT Integrity or Bypass.

Reference(s):

1. EOP-2 Primary Containment Control 2. NEI 99-01 Primary Containment Isolation Failure PC Loss 3.A Page 171 of 296

-:::==-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Containment Category: E. CTMT Integrity or Bypass Degradation Threat: Loss Threshold: CNB10 Intentional Containment venting per EOPs Definition(s): VALID -An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. Basis: EOP-?. Primary Containment Control. mav specify containment ventina and intentional byoassing of the containment isolation valve !oqic. even if offsite radioactivity release rate limits are exceeded. Under these conditions, with a VALID primary containment isolation signal, the threshold is met when the operator be.£]ins venting the containment in accordance with Enclosure

21. not when actions are taken to bypass interlocks prior to opening the vent valves (ref. 11. ---EGP-s-rnay-Ei-i-reet-pfi+1tai-:y-eemaiflrneRt-~sB!atioo--val-vB-+~B(s+4e-ee--i-mentieRaU-y e-y~R-if offsite radfeaeti-v~ty-r:e+ease rate-l+FRH-S-Wi+!-ee-e~EJ.er-tRese eet1-d-iOORS-V11irn-a-vaiifl-f}rimary-oorn-atnmePJ-isel-a#oA-siffAal,--#1*-GBAt-ai-AfRoot-&hEH+~El-alse-0e BB-A&kl-&FeE:Hos-Hf-!3R-FAaF-y-B&ffiaiRment-vePrhR§-+S--aettraH-y-fIBFfGffHW-;-

l ntentio nal venting of primary containment for primary containment pressure or combustible gas control to the secondary containment and/or the environment is a Loss of the Containment. Venting for primary containment pressure control when not in an accident situation (e.g., to control pressure below the drywell high pressure scram setpoint) does not meet the threshold condition. There is no Containment barrier Potential Loss threshold associated with CTMT !nteqrity or Bvpass. Reference(s):

1. EOP-2 Primary Containment Control 2. NEI 99-01 Primary Containment Isolation Failure PC Loss 3.B Page 172 of 296

-:;::=-Entergy Barrier: Category: River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Containment E. CTMT Integrity or Bypass Degradation Threat: Potential Loss Threshold: None Page 173 of 296 tr'\ -:::::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Containment Category: F. Emergency Director Judgment Degradation Threat: Loss Threshold: CNB11 Any condition in the opinion of the Emergency Director that indicates loss of the Containment barrier Definition(s): None Basis: This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Containment Barrier is lost. Reference(s):

1. NEI 99-01 Emergency Director Judgment PC Loss 6.A Page 174 of 296

~Entergy ) River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Containment Category: E. Emergency Director Judgment Degradation Threat: Potential Loss Threshold: CNB12 Any condition in the opinion of the Emergency Director that indicates potential loss of the Containment barrier Definition(s): None Basis: This threshold addresses any other factors that may be used by the Emergency Director in determining Whether the Containment Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored. Reference(s):

1. NEI 99-01 Emergency Director Judgment PC Potential Loss 6.A Page 175 of 296

-::==-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category H -Hazards and Other Conditions Affecting Plant Safety EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.) Hazards are non-plant, system-related events that can directly or indirectly affect plant operation, reactor plant safety or personnel safety. 1. Security Unauthorized entry attempts into the PROTECTED AREA, bomb threats, sabotage attempts, and actual security compromises threatening loss of physical control of the plant. 2. Seismic Event Natural events such as earthquakes have potential to cause plant structure or equipment damage of sufficient magnitude to threaten personnel or plant safety. 3. Natural or Technological Hazard Other natural and non-naturally occurring events that can cause damage to plant facilities include tornados, FLOODING, hazardous material releases and events restricting site access warranting classification.

4. Fire FIRES can pose significant hazards to personnel and reactor safety. Appropriate for classification are FIRES within the plant PROTECTED AREA or which may affect operability of equipment needed for safe shutdown 5. Hazardous Gas Toxic, corrosive, asphyxiant or flammable gas leaks can affect normal plant operations or preclude access to plant areas required to safely shutdown the plant. 6. Control Room Evacuation Events that are indicative of loss of Control Room habitability.

l'f the Control Room must be evacuated, additional support for monitoring and controlling plant functions is necessary through the emergency response facilities. -7. Emergency Director Judgment I The EALs defined in other categories specify the predetermined symptoms or events that are indicative of emergency or potential emergency conditions and thus warrant classification. While these EALs have been developed to address the full spectrum of possible emergency conditions which may warrant classification and subsequent implementation of the Emergency Plan, a provision for classification of emergencies based on operator/management experience and judgment is still necessary. The EALs of this category provide the Emergency Director the latitude to classify emergency conditions consistent with the established classification criteria based upon Emergency Director judgment.

Page 176 of 296

-===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: H -Hazards Subcategory: 1 -Security Initiating Condition: Confirmed SECURITY CONDITION or threat EAL: HU1.1 Unusual Event A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by RBS Security Shift Supervision OR Notification of a credible security threat directed at the site OR A validated notification from the NRC providing information of an aircraft threat Mode Applicability: All Definition(s): HOSTAGE -A person(s) held as leverage against the station to ensure that demands will be met by the station. -HOSTILE ACTION -An act toward RBS or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile Action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on RBS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA). OWNER CONTROLLED AREA (OCA) -For the purposes of classification,.the Security Owner Controlled Area (SOCA)"or the area between the SOCA Fence and the PROTECTED AREA Boundary. PROJECTILE -An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety. PROTECTED AREA -1The area within the perimeter of the RBS security fence. SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 1 OCFR50i2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: Page 177 of 296 @ -::::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases ( 1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. SECURITY CONDITION -Any security event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A security condition does not involve a HOSTILE ACTION. Basis: This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR § 73.71 or 10 CFR § 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HA1, and HS1-af!G-~G4. Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event. Classification of these events will initiate appropriate threat-related notifications to plant personnel and Offsite Response Organizations. Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program]. The first threshold EAL #1 references the Securitv Shift Suoervision (-site specific securf-ty-&A-i# -&LJ-J3*-FVi-&ie-A-)because these are the individuals trained to confirm that a security event is occurring or has occurred. Training on security event confirmation and classification is controlled due to the nature of Safeguards and 10 CFR § 2.39 information. The second threshold EAL #2 addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with the Security Plan for (-s-i-1:e speci-f.i-e proced~RBS. The third threshold EAL #3 addresses the threat from the impact of an aircraft on the plant. The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may also be provided by NORAD through the NRC. Validation of the threat is performed in accordance with AOP-0063 Outside Threats (ref. 2)(-s-i-:f:~eeffi&-p-rocedure). Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Security Plan for RBS (ref. 1 ). Page 178 of 296 -~-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Escalation of the emergency classification level would be via IC HA1. Reference(s): 1 . RBS Security Plan 2. AOP-0063 Outside Threats 3. AOP-0054 Security Events 4. NEI 99-01 HU1 Page 179 of 296 -=::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: H -Hazards Subcategory: 1 -Security Initiating Condition: HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes EAL: HA1.1 Alert A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by RBS Security Shift Supervision OR A validated notification from NRC of an aircraft attack threat within 30 min. of the site Mode Applicability: All Definition(s}: HOSTAGE -A person(s) held as leverage against the station to ensure that demands will be met by the station. HOSTILE ACTION -An act toward RBS or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile Action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on RBS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA). HOSTILE FORCE -One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with *suitable weapons capable of killing, maiming, or causing destruction. OWNER CONTROLLED AREA (OCA) "'. For the purposes of classification, the Security Owner Controlled Area (SOGA) or the area between the SOGA Fence and the PROTECTED AREA Boundary. PROJECTILE -An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety. PROTECTED AREA -The area within the perimeter of the RBS security fence. Page 180 of 296 -~Entergy . River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Basis: This IC addresses the occurrence of a HOSTILE ACTION within the SECURITY OWNER CONTROLLED AREA or notification of an aircraft attack threat. This event will require rapid response and assistance due to the possibility of the attack progressing to the PROTECTED AREA, or the need to prepare the plant and staff for a potential aircraft impact. Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event. Security plans*and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program]. As time and conditions allow, these events require a heightened state bf readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Alert declaration will also heighten the awareness of Offsite Response Organizations (OROs), allowing them to be better prepared should it be necessary to consider further actions. This 1-G-EAL does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72. The first threshold ~l-#4-is applicable for any HOSTILE ACTION occurring, or that has. occurred, in the OWNER CONTROLLED AREA. =f-his ii:,ok::loo&-afl-y-aBt-ien-directed ag-amst-aR -hSffil-that-+s-looated oblt-sid-&4A*-i)+affi-P-R:G+EG+e-g ARE/\. The second threshold EAL-#2-addresses the threat from the impact of an aircraft on the plant, and the anticipated arrival time is within 30 minutes. The intent of this EAL is to ensure that . threat-related notifications are made in a timely manner so that plant personnel and OROs are in a heightened state of readiness. This EAL is met when the threat-related information has been validated in accordance with AOP-0063 Outside Threats (ref. 2)(site spe~. The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may be provided by NORAD

through the NRC. In some cases, it may not be readily apparent if an aircraft impact within the SECURITY OWNER CONTROLLED.AREA was intentional (i.e., a HOSTILE ACTION). It is expected, although not certain, that notification by an appropriate Federal agency to the site would clarify r this point. In this case, the appropriate-federal agency is intended to be NORAD, FBI, FAA or NRC. The emergency declaration, including one based on other ICs/EALs, should not be unduly delayed while awaiting notification by a Federal agency. Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information.

This includes information that may be Page 181 of 296 -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases \ advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Security Plan for RBS (ref. 1 ). Escalation of the emergency classification level would be via IC HS1. Reference(s):

1. RBS Security Plan 2. AOP-0063 Outside Threats 3. AOP-0054 Security Events 4. NEI 99-01 HA1 Page 182 of 296

-::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: H -Hazards Subcategory: 1 -Security Initiating Condition: HOSTILE ACTION within the PROTECTED AREA EAL: HS1.1 Site Area Emergency A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by RBS Security Shift Supervision Mode Applicability: All Definition(s): HOSTAGE -A person(s) held as leverage against the station to ensure that demands will be met by the station. HOSTILE ACTION -An act toward RBS or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile Action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on RBS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA). HOSTILE FORCE -One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction. OWNER CONTROLLED AREA (OCA) -For the purposes of classification, the Security Owner Controlled Area (SOGA) or the area between the SOGA Fence and the PROTECTED AREA Boundary. -PROJECTILE -An object directed toward a Nuclear Power Plant that could cause concern-for its continued operability, reliability, or personnel safety. PROTECTED AREA -The area within the perimeter of the RBS security fence. Basis: This IC addresses the occurrence of a HOSTILE ACTION within the PROTECTED AREA. This event will require rapid response and assistance due to the possibility for damage to plant equipment. Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event (ref. 1, 2. 3). Page 183 of 2~6 -::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program]. As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Site Area Emergency declaration will mobilize Offsite Response Organization (ORO) -resources and have them available to develop and implement public protective actions in the unlikely event that the attack is successful in impairing multiple safety functions. This .J-G-EAL does not apply to a-H'GS-1-l-hE--AG-T+ON-a~reetee-at-al-1-4S~PFeteetee-Afea +e-eal:ee-oots~~e-P--ROTEGTED AR~~sttBJ:1-an attack should be-as-sesseEl-1:Jsin§-I-G-H-A+.-l-t a-l-s-e-Ele-e-s---Fl-0t-aWl-y-te-incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73. 71 or 10 CFR § 50. 72. Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversa*ry, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Security Plan for RBS (ref. 1 ). E-SBalation of the-emergency classiflBati.e.~e via IC HG+ Reference(s):

1. RBS Security Plan 2. AOP-0063 Outside Threats 3. AOP-0054 Security Events 4. NEI 99-01 HS1 Page 184 of 296 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 2 -Seismic Event Initiating Condition: Seismic event greater than OBE levels EAL: HU2.1 Unusual Event Seismic event> OBE as indicated by EITHER of the following:

Annunciator P680-02A-C06, SEISMIC EVENT HIGH
Annunciator P680-02A-B06, SEISMIC EVENT HIGH/HIGH and amber lights illuminated on H13-P869 ERS-NBl101 Mode Applicability:

All Definition(s): None Basis: This IC addresses a seismic event that results in accelerations at the plant site greater than those specified for an Operating Basis Earthquake (OBE). An earthquake greater than an OBE but less than a Safe Shutdown Earthquake (SSE) should have no significant impact on safety-related systems, structures and components; however, some time may be required for the plant staff to ascertain the actual post-event condition of the plant (e.g., performs downs and post-event inspections). Given the time necessary to perform walk-downs and inspections, and fully understand any impacts, this event represents a potential degradation of the level of safety of the plant. Event verification with external sources should not be necessary during or following an OBE. Earthquakes of this magnitude should be readily felt by on-site personnel and recognized as a seismic event (e.g., lateral accelerations in exBe&S-ef-G-cG-8fil. The Shift Manager or Emergency Director may seek external verification if deemed appropriate (e.g., a call to the U. S. Geological Survey (USGS}, check internet news sources, etc.); however, the verification action must not preclude a timely emergency declaration. Depending upon the plant mode at the time of the event, escalation of the emergency

classification level would be via 1-G-EAL CA6~ or &A9SA8.1.

To avoid inapprooriate emerqency classification resulting from spurious actuation of the seismic instrumentation or felt motion not attributable to seismic activity, an offsite agency (USGS, National Earthquake Information Center (NEIC)) can confirm that an earthquake has occurred in the area of the plant. Such confirmation should not, however. preclude a timelv Page 185 of 296 e *===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases emergency declaration based on receipt of the OBE alarm. lf requested. provide the analyst with the following RBS coordinates: 30° 45' 26" north latitude, 9'!0 19' 54" west longitude (ref. 3). Alternatively, near real-time seismic activitv can be accessed via the NEIC website .. Reference(s): 1 . ARP-680-02 P680-02 Alarm Response 2. AOP-0028 Seismic Event 3. USAR section 2.1.1.1 Specification of Location 4. NEI 99-01 HU2 r Page 186 of 296 e -::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 -Natural or Technological Hazard Initiating Condition: Hazardous event EAL: HU3.1 Unusual Event A tornado strike within the PROTECTED AREA Mode Applicability: All Definition(s): PROTECTED AREA -The area within the perimeter of the RBS security fence. Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant. EAL #'1----This EAL addresses a tornado striking (touching down) within the PROTECTED AREA. *-AL #2 addresses fl.ooEling of a-ooile+R§-~FA-ef-afe-a-rn-at-results in oper-ate-r-s-i-selating pewe-r ro-a-S~~~e-nem-El-He-tB-wateF-level oF-etl=leF-w~erns. Clas-sifk:,-aoon t&-a~F-leve[-e-Helated-wetti-R§-Bauses an automatic ise-la~A~Y SYST~component--frefR-i-t&-pe-vver source (e.g., a breaker-er relay trip};-Te-warmt

-1-a-ssrneati-OR;--O-perab-H#y-e-Hhe-affect.eEJ-eempooent mu st be req u+reEl-f>y-+ec-AH+sa+

Spec+fieations for the current operatirig-mee.&.- EAL #3 addresses a hazarde-~eA-t-origi-Aat~ocation and-ef &l:fffisfe-nt--ma§ni-tooe to impede the-meveFAeR-t-8-f personne+-within the PROTECTED AR~ EAL #4 addresses a hazard-e-H-s-evOHHhat causes an on site impediment to vehic-le-mevemefrt aR4-s+w=Micant one-ugh to pre-h+eit the plant staff from--ac-ces-s+R§-tfl*-Si-te--using p8-fSOO-al. -vehicles. ExaFAp~vent incluee-s~te flooding caused by a humcane, Aea-vy-r-ai-ns, Hp-fiver water releases, dam failure, etc., e-F--aH--e-A-site roae,-Tffis--e-Ab--i&-Re-t-iRtend ed a pply-tG--fOilliA-e-tfflf}eci+FAG-ffiS-s-uel=l-as-feg,&Rew,+ce, or vefi+Cl-e breakdovms or accide-n-Is,eut rather to more sigA+fiBant conditions such as the Hurric-ane AR-drew strike on Turkey Point in 1992, the floodiAg around the Coop~A§4oo Miewest floods of 199-0, or the f100El-ffi§-Or0-ttAd Ft. Calhoun Station in 2011. EAL #&-addresses (site speei-#e-El-escription). Page 187 of 296 e -:::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Escalation of the emergency classification level would be based on I Cs in Recognition Categories A, F, Sor C. If damaqe is confirmed visually or by other in-plant indications, the event mav be escalated to an Alert under EAi CA6.1 or SA8.1. A tornado striking (touching down) within the PROTECTED AREA warrants declaration of an Unusuai Event regardless of the measured wind speed at the meteoroloaical tower. A tornado is defined as a violently rotatina column of air in contact with the ground and extendina from the base of a thunderstorm. Reference(s):

1. AOP-0029 Severe Weather Operation

2. NEI 99-01 HU3 Page 188 of 296

-:-:::=* Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 -Natural or Technological Hazard Initiating Condition: Hazardous event EAL: HU3.2 Unusual Event Internal room or area FLOODING of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component required by Technical Specifications for the current operating mode Mode Applicability: All Definition(s): FLOODING -A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area. SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 1 OCFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result* in potential offsite exposures. Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant. *:Ab-#4-aGEJresses a temade-stf~te-l:H3Ring down)-wi-tfim the PRG+eG+-ED AR~ This EAL addresses FLOODING of a building room or area that results in operators isolating power to a SAFETY SYSTEM component due to water level or other wetting concerns-. Classification is also required if the water level or related wetting causes an automatic isolation of a SAFETY SYSTEM component from its power source (e.g., a breaker or relay trip). To warrant classification, operability of the affected component must be required by Technical Specifications for the current operating mode. EAL #3 addresses a hazardeHs-ma-terials event origiffilting at an offsite location ane-ef sufficient magnitude to ~he-ffiovement of personnel within the PROTECTED ARE/\. Page 189 of 296 -::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases EAL #4 addces-se-s-a-ha-za-Fde-Hs event that causes an on site iFRf}eEl+FAent to vehicle moveme-At a-nEl--significant--ene-lcl§h to prohibit the plant-staff-from accessi-ng-#le-site using persOAa veAicles. Examples of such an event include--s}te flooding caused by a hurricane, heavy rains, up river water releases, dam-failure, etc., or an on site train derailment-blocking the access ma-4 ' This EAL is not intene--ed apply to routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the flooding-around the Cooper Station during the MiElwest floods of 1993, or the flooding around Ft. Calhoun Station in 2011. EAL #s-ae-dresses E-s+te-sf>ec~fic description). Escalation of the emergency classification level would be based on I Cs in Recognition Categories A, F, Sor C. Refer to EAL CA6.1 or SA8.1 for internal FLOODING affecting one or more SAFETY SYSTEM trains. Reference(s):

1. NEI 99-01 HU3 ( Page 190 of 296

-::::::-Entergy River Bend Station EAL Basis ,Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 -Natural or Technological Hazard Initiating Condition: Hazardous event EAL: HU3.3 Unusual Event Movement of personnel within the PROTECTED AREA is IMPEDED due to an event external to the PROTECTED AREA involving hazardous materials (e.g., an offsite chemical spill or toxic gas release) Mode Applicability: All Definition(s): IMPEDE(D) -Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed). PROTECTED AREA -The area within the perimeter of the RBS security fence. Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant. This EAL E/\L #1 addresses a tornado striking (teBSl:mlg down) within tAB-P.ROTECTE-Q AAEA-+tHs-EA-b-aooresse-s-fleoding of a bui+d+R§-FOOFA-or area that results in OJ)erators isolating 13eweF-ro-a-SAf-ETY SYSTEM component due-to-watef-1-evel-e-F-eth-e-F-We-Wng Classification is also reEjtl-ifeEH.:f-.tt:l.ffif--level or rel-ated-welliR§-Gauses an automatie-isolation of-a-SAFETY SYSTEM compon~ower-sot1-ffie-{e.g., a breaker or relay tri~-o warfafl:t classification,eperatm+t-y-ef-#te-affeet-ed-c-Gmf*3-nent must b&-f0*ru4red by Technicat Specifications for tAe-c-Bffe-n-k)perating mode. E/\L #3 addresses a hazardous materials event originating at an-e-f-t:si-te location outside the PROTECTED AREA and of sufficient magnitude to IMPEDE the movement of personnel within the PROTECTED AREA. E/\L #4-aooresses a hazardot1s event that causes aA-on site impediment to vehicle mevemem an-d-significant enou~~FOm-accessing the site using personal . vehicles. Examples of such an event include site flooding caused by a hurricane, heavy rains, up river water releases, dam failure, etc., or an on site train Gefai.1-ment----eteB-king the access roaEh-Page 191 of 296 e *::::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases +J:li-&-*AL is not intendeEl-apply to routine impeEl~rnent&-SHGR-a&-fog, snow, ice, or--veAtBle breakdovvns or accrdents, but rather to ~-a~ioos such as the Andrevv strike on Turkey Point in 1992-,---tRe-fle-oEli-Rg--arou-A*1-t:he-GoofIBF--Sta-ti-Em-dtlfi.Rg-the Mtewes-t---#eods of 1993, or the flooding around Ft. Gaihoun Sta~'l-h EAL #5 addresses (site specific descriptiont Escalation of the emergency classification level would be based on I Cs in Recognition Categories A, F, Sor C. Reference(s):

1. NEI 99.:01 HU3 / Page 192 of 296

-===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: Subcategory: Initiating Condition: EAL: H -Hazards and Other Conditions Affecting Plant Safety ( 3 -Natural or Technological Hazard Hazardous event HU3.4 Unusual Event A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles (Note 7) Note 7: This EAL does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents. Mode Applicability: All Definition(s): FLOODING -A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area. Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant._*-Ab-#-i--a*1affi5Ses a torA-a*lo strik~+R§--OOWAt-Wirnin t!cle-PF?G+E:G-+~A- +Ris-~b-aooresses-fle.eGi-A~ilffifi§-FB9ITT-or area tt:i-a-t-re-~perators isolat+ng power te-a-S~ETY SYS-~Emern-ooe-te--wat6f--l.e.v-e+-0F-&t!=tef-vvetting coRB-erA& Gl-a-ssmcation is alsEHeEJ-Hired if t+1B-watef-level or relateEl--wetting causes an ~i::EM-eempeneflt--fffirn-it:s-p~ree-(e.g., a ~ker-OHe+ay-tfi-13fr-i::& waFFant classificati&A-,ef:,era-e+l-ity of the-a#eeted-oomp&RB-At must b6-f6Etuired by +effi-Aieal. Specifieations for the-GH~~ EAL #J-..a.EI.Elresses a hazar*1e-Hs-materi-als--6-'IA...ant-&fi§+n-atiA§ at an offsite-l&eatie-n-aREl-&f sl::l#ieie-nt magnitude to impede-the-meve-meflt of personnel within the-PROTECTED /\R-EA-. This EAL E-Ab-#4-addresses a hazardous event that causes an on-site impediment to vehicle movement and significant enough to prohibit the plant staff from accessing the site using personal vehicles. Examples of such an event include site FLOODING caused by a hurricane, heavy rains, up-river water releases, dam failure, etc., or an on-site train derailment blocking the access road. This EAL is not intended apply to routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the FLOODING around the Cooper Station during the Midwest floods of 1993, or the FLOODING around Ft. Calhoun Station in 2011. Page 193 of 296

=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases *A-h-#5--aGGFesses (site specific des-ei:~13-tie-R-};--Escalation of the emergency classification level would be based on ICs in Recognition Categories A, F, Sor C. Reference(s):

1. NEI 99-01 HU3 Page 194 of 296

~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 -Fire Initiating Condition: FIRE potentially degrading the leyel of safety of the plant . EAL: HU4.1 Unusual Event _A FIRE is not extinguished within 15 min. of any of the following FIRE detection indications (Note 1 ):

Report from the field (i.e., visual observation)
Receipt of multiple (more than 1) fire alarms or indications
Field verifica,tion of a single fire alarm AND The FIRE is located within any Table H-1 area Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, pr will likely be exceeded.

The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: All Definition(s}: Table H-1 Fire Areas *

Reactor Building
Auxiliary Building
Fuel Building
Control Building * *standby Cooling Tower
Diesel Generator Building
Tunnels (B, D,E, F, G) 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.

VALID -An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the Page 195 of 296 e -::::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. Basis: This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant. * ~b-#-1-The intent of the 15-minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., smoldering waste paper basket). In additio.n to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation of a suppression system, etc. Upon receipt, operators will take .prompt actions to confirm the validity of an initial fire alarm, indication, or report. For EAL assessment purposes, the emergency declaration clock starts at the time that the initial alarm, indication, or report was received, and not the time that a subsequent verification action was performed. Similarly, the fire duration clock also starts at the time of receipt of the initial alarm, indication or report._e-Ab-#2 +h~s-EAb-aoofe&OOs-reB-e~t-e-f--a-s+A§le-:fire-alaFm,ane-me-e-Xt&te-Rce of a FIRE is-net-veFified (h&.,f}reveEl-er-El*t&proved-)--w~A-3G-miBHtes-of-tt:le-a~aRtt-b/.j3efHBee+f3t,Bf3eratGrs-witi-ta*ke pro mpt-aGHtffis--te-eemtFm the vaHE!+t-y-e-f--a-si-R§ I e fire al arm. Fer E-Ab-asse&&mef't~poses, the W-ffii-A-li~he-time-mat the initial-atarm--was rnceivea,and not Sl::IB-&eEftlef'tt-vefi-fic-atioo action-was peFfe.Fmefl-:. A single fire almm, absent otl:ter indicationfs~i-ne+Gati-ve-ef-efftfi-pment-famtre er-a-Sf*}rious amivation, anEl-Aet-aR-a&tual-R-RE. FOF this-reason, adE~ftional time-+s-aHewe*1-te verify the validity of the alarm. The-3{)-mi-A-lit-e-f)eReEl-i-s-a-reasonab!e amount of-t+me-te ooterm+Re-+f an actual-R-RE exists; ho 1 Ne-vef,-ffitef-t-J::iat-#me,anEl-aeseRt-~~ oontrary, it is assliffied that an actual F!R~i-R--f3rogress. l.f-an--ootual Fl Re-is verified by a report from-t-l=ie-fiele,-#lOR--eAL

1 is immediatel-y-awHeaeJ.e, and the emerge-n~e declared if the F+RE-i-&-Rot extin~~mint1tes of the report. If the-llia-rm-is--v-eft#ed to be-dtle-te-aR-eqttif3ment-faalire-er-a-spt100H-s-aBti-vatio n, aoo #=iIB-Vorification ocGHFS-~tes of the recet~&-alarm,41:ten this EAL is not B-!3f*fBaei-e-aREl-Ae-emerge.Aey-eeelaratieA-i&-wafraRteEh

AL #'J m--ade+tieA-te-a-R-~L-#4-or EAL #2, a FIRE within the-f3lant PROTECTED AR-E-A-Ftet-ex-t~~hee-wi-truH-@G minutes may also potentially-Ele§rade-me level of plant safety. This basis-extends to a FIRE occuF4A§-wi-thin the-PROTECTED AREA of an ISFS~ located outside the plant PRO+ECTED AREA. [Sentence for-pkffi-ts-with-aA-ISF-SI outside the ~~At-Protected Area] EAL#4 Page 196 of 296 e ~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases +fa FIRE withiFHhB--i)lant or ISFSI [for plants w#h an ISFSI outside tAe plant Protected Area] P-ROTECTED AREA-i-s-ef-st!#icient size to require a respo~te firefighting agency (e.g., a local town Fire Department), then the-~el of plant safety is potentially degraded.

The El-i-s-patGh-of an offsite fir-ef~ting agency to the site requires an emergency declaration only if it is needed to actively suppo~because the fire is F-i-re-Brigade to extinguish. Declaration is not necessary if the agency resources are placed on stand by, or sup-po-rting post extift§tH-shmern-rew-very or investigation actions. Basis Related Refremeri-ts from A:fmeftffi*-R Apj:IBRE!ix R to 10 GF-P~G,states in part-: Gri-tefie-A-3-ef-A~e nd lx A-te-tni-s~spoof+e-s-#tat-2.'Sffi:lctLI res, systems,aoo OOffipOOents important to safe-ty-&J:IB+J--ee-4esfg-Fl-Od and locatee-te-rni-n+mize, consisteA-t witA-e-ther-s-afety--Fe*j-uirements,-U-re-p-rooae~hly-aoo-e-ffeG-t..B-f-fiIB&-afle-e-xf}l-esie-~ -Wl:teA-ee-nstdefi-A§-the-e-ffee-1:&-ef-fi-re,those syste-ms-a-ssoei-a-t**!-wi-th-a~ievi-R§-a-Ad mai-Ata-}R-~ safe sh El-am-age-te-~ n lead to-oore-fi-amag e re-swtmg-ffB-m-le-ss-ef-Beo~A-HflFOO§A-eetk3-tf- ' . ge-c-ause-ftre-ma-y affect safe sh utd O\Nn-s-ysteFHS--a-R4--be-c-a-Ll-5*-4h-e-le-ss-e-f:--R:lfl*tie-R-ef s-y-ste-ms-u-seEl--te--rn~ti~-e-tAe-BOA-seEt-H&nees-e-f-e-es~n-baBi s a cdEi-Emts-t1A*1-ef-pe st fire ooA-di-tiens c!oes-ne-t-per so impact puel+c-sa-fety,-#le-R-eed to limit--fif~m-s ffi*!-\:;J-iree-t&-aehi eve and ma-i-R-t-a+n--safe-sh-u-tEIBwn-certe+t+B-A-S-is-§-re-ater-tA-af!--fh*-Ae0*l-to +i-mf-t-:fiFe-El-a-mage to these systems-reEf*;!-ifed-te-mitigate the b-as+s-aseffieflt&. +n--a*l*1-i-t-ie-~peRd~~G,Fefluires, arne-Ag-etl1-eF-Be.'1Si-EIBFa-t+e-n-s,t-h:e-use-of 1 hour fire baffi-e-r-s-:fe-r-~f-eaele--a-A~El-assooi-a-te*1--non safet-y-eircuits of one reooA*l-a+-1t-trai-A-(-G72 . c). /\s Hse*1-i-A--eAL

2, the-3 0 min utes-te-ve-r-i-fy-a-s-i-R§te-ala-r1-n-i-s-i,,vel+

~worst case 1 hetl-r-ti-rnB--i)efi-oEh-Depending upon the plant n;,ode at the time of the event, escalation of the emergency classification level would be via +C-EAL CA6.1 or8A9SA8.1. --' -The 15 minute requirement begins with a credible notification that a FIRE is occurring, or receipt of multiple VALID fire detection system alarms or field validation of a single fire alarm. The alarm is to be validated using available Control Room indications or alarms to prove that it is not spurious, or by reports from the field. Table H-1 Fire Areas are those areas that contain equipment necessary for safe operation and shutdown of the plant (ref. *1 ). Reference(s):

1. AOP-0052 Fire Outside the Main Control Room in Areas Containing Safety Related Equipment

(' Page 197 of 296 ~=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases 2. NEI 99-01 HU4 Page 198 of 296 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 -Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL: ,HU4.2 Unusual Event r Receipt of a single fire alarm (i.e., no other indications of a FIRE) AND The fire alarm is indicating a FIRE within any Table H-1 area AND The existence of a FIRE is not verified within 30 min. of alarm receipt (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Mode Applicability:

All Definition(s): Table H-1 Fire Areas

Reactor Building
Auxiliary Building
Fuel Building
Control Building
Standby CoolingTower
Diesel Generator Building
Tunnels (B, D,E, F, G) 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.

VALID -An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. Page 199 of 296 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Basis: This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant. EAL#1 +lqe intent of the 1 ~duration is to s0Hhe-i;:-IR*--aA-El to discriminate against small FIRES that are r-eae-i-1-y-e-xting*uished (e.g., smetdering vvaste papeF-l3asl<et). In addition to aJ.aFffiS,--O#IBF-tfl*1-i.e.at+e-ns of a FIRE c&!::1-ie--ee a drop i-A-fu:e main pressHre,aHtema-#c activation of a suppression sy&tefFl,6t&. Upon receip~ill tal<e prom~ie-n-s to conftfm-m-e-vaHeity of an initial fire al-a-Fm, ifld-ieati0A,Gf-Feport. For-e-AL assessment purposes, the emergency declaratkm-el-ocl< starts at me-time that the initial alarm, indication, or.report-was received, aRd not the time that a subsequent verification actio~Fformed. Simil-a4y,#1e fire duration cloek-alse-starts at t~FOBO~~~El-icatiOR-OH6-j3Bfh EAL #2 This EAL addresses receipt of a single fire alarm, and the existence of a FIRE is not verified (i.e., proved or disproved) within 30-minutes of the alarm. Upon receipt, operators will take prompt actions to confirm the validity of a single fire alarm. For EAL assessment purposes, the 30-minute clock starts at the time that the initial alarm was received, and not the time that a subsequent verification action was performed. A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failure or a spurious activation, and not an actual FIRE. For this reason, additional time is allowed to verify the validity of the alarm. The 30-minute period is a reasonable amount of time to determine if an actual FIRE exists; however, after that time, and absent information to the contrary, it is assumed that an actual FIRE is in progress. If an actual FIRE is verified by a report from the field, then HU4.1 EAL #1 is immediately applicable, and the emergency must be declared if the FIRE is not extinguished within 15-minutes of the report. If the alarm is verified to be due to an equipment failure or a spurious activation, and this verification occurs within 30-minutes of the receipt of the alarm, then this EAL is not applicable and no emergency declaration is warranted. EAL #3 ffi-a*1.dition to a FIRE addressed by EAL #1--ef-Ei-\L

2, a FIRE within the plant PROTEG=r-*9 AREA n~i-A§HisRBEl-wit~may also potentially degrade the level ef-f}lam safety. This basis-ffi<-tends t~!RE occuffiA§J-wjfhin the PROTECTED AREA-et an IS~ located outside the plant PROTECTED AREA. [Sentence-fer plants 11,lith an ISFSI otft.s-iEie-.t:he pkmt-Protected Area] EAi #4 If a FIRE within the plant or ISFSI [for plants with an ISFSI outside the plant Protected Area] PROTECTED AREA is of sufficient size to require a response by an offsite firefighting agency (e.g., a local tovm Fire Department), then the level of plant safety is potentially degraded.

The Page 200 of 296 ~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases dispatch of an offsite firefighting agency to the site requires an emergency declaraOOA--eRly-4-i-t is needed to actively support firefig~ti-Rg-efforts because the fire is beyond the capability of the Fire Brigade to extinguish. Declaration is not necessary if the agency resources are placed stand by, or supporting post extinguishment recovery or investigation actions. Basis Related Requirements from Appendix R Appendix R to 10 CFR 50, states in part: Criterion 3 of Appendix A to this part specifies that "Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions." When considering the effects of fire, those systems associated with achieving and maintaining safe shutdown conditions assume major importance to safety because damage to them can lead to core damage resulting from loss of coolant through boil-off. Because fire may affect safe shutdown systems and because the loss of function of systems used to mitigate the consequences of design basis accidents under post-fire conditions does not per se impact public safety, the need to limit fire damage to systems required to achieve and maintain safe shutdown conditions is greater than the need to limit fire damage to those systems required to mitigate the consequences of design basis accidents. In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of 1-hour fire barriers for the enclosure of cable and equipment and associated non-safety circuits of one redundant train (G.2.c). As used in HU4.2EAL #2, the 30-minutes to verify a single alarm is well within this worst-case 1-hour time period. Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via lG-EAL CA6J_ or ~9SA8.1. The 30 minute requirement begins upon receiot of a single VALID fire detection svstem alarm. The alarm is to be validated using available Control Room indications or a!arms to prove that it is not spurious, or by reports from the field. Actual field reports must be made within the 30 minute time limit or a classification must be made. lf a fire is verified to be occurring bv fielc;i report, classification shall be made based on EAL HU4.1. with the 15 minute requirement beqinning with the verification of the fire by: field report Table H-1 Fire Areas are those areas that contain equipment necessary for safe operation and shutdown of the plant (ref. 1 ). Reference(s):

1. AOP-0052 Fire Outside the Main Control Room in Areas Containing Safety Related Equipment

2. NEI 99-01 HU4 Page 201 of 296 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

Subcategory: H -Hazards and Other Conditions Affecting Plant Safety 4-Fire Initiating Condition:

FIRE potentially degrading the level of safety of the plant EAL: HU4.3 Unusual Event A FIRE within the PROTECTED AREA not extinguished within 60 min. of the initial report, alarm or indication (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.

The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: All Definition(s): FIRE -Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed. PROTECTED AREA -The area within the perimeter of the RBS security fence. Basis: This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant. EAL #1 +M intent e.f-#:le-%-m~nlite-Ei-1:tration is to size th~IRE and to dis&Fiffimate-aga~nst smal-1-~t are read+ly-e-xtiH§t!tshed (~mo-J.oooog vvaste paper basket). In addition to alarms, other inffiBffiieHs-of a FIRE cot1td-be a drop in fire main pressure, automatic activatiBH ~m,et&.-Ypon receipt, operators wi+l-tak&i)r9fflj3t--a*tions te--oonfirm the valte-ity-e-f-aA-iHi#al-fire alarm, tne-ication, or reporh---Fer EAL assessment purposes, the emergency dee~ the ti me that the in it~al-a!arffi,tA*l-~atio n, er-repeff-was received, and net-th*-tiffte-4hat-a subsequent verification action was-performed. Similarly, the fire du~Ft-s-at the time of receipt of the infti.al-alafm,iHd+eation or report. EAb-#2 This EAL addresses receipt of a single fire-alarm, and the existence of a FIRE is not-vef~~ (i.e., proved or disproved) within 30 minutes of the alarm. Upon receipt, operators will take f}FE)ffiftt actions to confirm the vaVidity of a single fire alarm. For EAL assessmen~ses, the Page 202 of 296 e -:-c:::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases 3-G-ffii.A-t.1-1-e-oocl< staFts-at the time-4~ial alarm-wa&-received, ane-ne-t-the time that a &Ue&eEJ-1:ffiffi-VOR-#GaHeR-affitefl-wa&f}B-OOFmeEh A-single fire-alafm,---a!:).seffi--etfler indicatfe-Rfs-)-ef a FIRE, may be indic-ative-of equipment faHtlffi 0r a spuFi-o-H&-acti-vtion, and net-an actua~r-4-h~d-#i-enal time is allov-1ed ts vefi-fy-tAe-v-aHEl-i-ty---e-f-fe-alarm. The 30 minute period is a reasonaele-ameHf1-t--e.f.--t~me-te-EletBfffii n e if an asrual-F I RE exi-st~ffiNe-ve-r,after-thaHme,ane-a-oseRHA-fefma#e-A-te-the contrary, it-is assumeEl-that an a&tua1-FIRE is in pfO§-res-&.--l-f--afl-aBtHa-!-FIRE is v-efifi**!-ey-a-report from the field, then EAL #1 is ifA-ffleG-i-atety-a!:}pUcame, anMhe emergency mt.1-s-t-be-eeooreITTf-t!IB-F--IRE is not-e-~sooEl-w~tA-~R-4-5-mifHJtes of tA-e ~f-the-a!-a-rm is veriHee-te-be due to an eEft!+f*fl6nt failure or a spurious actwation, and thi-s-ve-Fi-Hcatien-eccurs within 30 minutes--&f-.the-recei-pt-etthe-ala-rm,tAefl-t-R+s-E/\L is no-t app~~e-aREl--Re-emerg-ency-4-eclaratien-i-s-warrarnea E/\L #3 In addition to a FIRE addressed by EAL HU4.1 #-1-or HU4.2EAL #2, a FIRE within the plant PROTECTED AREA not extinguished within 60-minutes may also potentially degrade the level of plant safety. +A~El-s-to a FIR~eCBtlfrmg-w~thin the eutside the p~am-~G+E9-AREA. [Sentence for plam-~utside tR-e-/*1Rt -Proteet*-d Area]Mb-#4 tf a Fl RE-wi-m+A---the-pla-nt or IS~.fer-p!am-s with aR-f&F-&l-eutside the-plant Prote-Gted Area] P ROT EG+E-Q-AR*-A-i-s of suffiB+6flkt:2:e-te-feq-l::l-i-re a res po n-s-e-by-a-A-e#&ite-f~h#Rg--a§-e-A-GY (e.g., a loca!-tew~F-i-fe--Gef1artment},4t:l-en-tne-le-vel of plant safety is 13o-tenti-al-ly---aeg-raEled. The EH-s-patch-e-f-an-e#srte-firefightin§-8-§-8-RBY-to the site FOEJ-llire-s-a-A-emefQ-RCy declarat+9-fl-O-ffiy-tf-it +s-Aeeeed to ae#ve-~s-u-ppert firefig-htffig-~because-the-fi-re is beyeA-El-tA-e-*-a-paeility of tA-e -F-i-re-Bngaee-te-e-xti-A-g uish. Dec!-a-ratioFl-i-s-Aot-Hece-s-sary-i-f-th e ag stand by, or supf:}oftf-A§-f3ost extiA-g-uishmeA-t-recovery or-~vestigatioA-a-CtiOfl-&.- Basis Related Re@ffOments-froFl:tAfmeoo-B<-R Appendix R to 10 CFR 50, st-a-tes--i-A-pa G-Fiterion of /\ppene-i-x-A-te---#ti-s-part specifies tha~tr-uctures, systems, anEl oompene-A-ts-+mf)f)rtant to safety sh-all be designed and locate*! to minimi+/-e,w-Rsistent w~A-e-mer-s-ate-t-y-requirem~d effect oHire-s---a-A~ie-~ Whe-n-eenste-e-rin g the effects of fire,--#le-se systems a ssociai:eEl-with a chi evi ng-a-Ad-mai nta~n-~n§-safe-s-hutde.WA--OOflffiti o ns assume-maje-Hmp0ft-an-ee to safety because damage to the~mage resulting from loss of coolant througl:1 boil off. Because fire may affect safe shutdown systems and because tho loss of function of systems used to mitigate the consequen*e-S-ef-El-esign basis accidents under post fire eeHd-itions does not per so impact publie-s-afety, the neee-te-#mi-Hire damage to systems required to achieve and maintain safe shutdown conditions is greater than the need te-Pago 203 of 296

~~ ----~~--------------------------------

--::::::-Entergy River Bend Station EAL Basis Document Revision XXX 1 Attachment 1 -Emergency Action Level Technical Bases l-imit fire damage te-the-se-s-yste-ms require*l--te-fn#igate the consequences of design basi-s-aeci dents.

In addition,Appendix R to 10 CFR 50, requires, among other considerations, the use of 1 hoUf fire barriers-fer-t:fle-eflclosure of cable-and equipment and associated non safety circuits of one fedundant train (G.2.c). As used in EAL #2, the 30 m~At!IBS-te-veftfy a single alarm is well within this-worst case 1 hettr time pe#eEh Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IG-EAL CA6J.. or SA9SA8.1.

Reference(s):

1. NEI 99-01 HU4 Page 204 of 296 River Bend Station EAL Basis Document Revision XXX ' I Attachment 1 -Emergency Action Level Technical Bases Category:

Subcategory: H -Hazards and Other Conditions Affecting Plant Safety 4-Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL: HU4.4 Unusual Event A FIRE within the PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish Mode Applicability: All Definition(s):

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

PROTECTED AREA -The area within the perimeter of the RBS security fence. Basis: This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant. E-Ab---#'.'1-

r-he-iRffifl-t-e-Hh e 'l 5 mfAtlte-ffilffiOOR-i-s-te-&ize-th e FI RE-aA<l-te-*1-i-sBfim~Rate-a§-atRS-t-smaH FIRES Utat are readily-e-~i-she~dering vva-&te-f*lpor basket). In addition to alarms, other indications of a FIRE could be a drop in fire main pressure-,--at1-tomatic activatten

&f-a-sHppress+e-n-sy&tem,et&.- Ypoo-reeei-j3t, operators will take !c}FOmpt actisn~f~rm-4Jqe-vaooity of an in~tia!-fi-fe-al-arm-, ind i catkffi,oF-re po rt. Fer-EAL a ssessmem--ptl-F!c}OSe&;-the--em6f§ef}e-y-d e cl a raOOA-Blo ck starts-at the-Ume4ha-t4Re initial alarm, indIBation, -eF-ref}ert-was-i:esei-vw,a-A<l-not-the-#me that a St1BSettt1e nt verifie-a#en---ami&R-Was-f30ff'ermed. SimH-aFly,the-fire d uratie-n-G!oo-k-alse-sta rts at ~the initial alarm, indication or repOfh E/\L #2 +h+s-EAb-aooresses recetpt-ef-a singl-e-fi-re-al-arm,-a-R*Hhe-ex-i-stence of a FIRE is not-vermee fh&.,proved or disproved)-w+~s-ef-the-a-larm. UpsA-re-ceipt, 013erators wi.J.1-take f)fOffif*-actie-ns to confirm the validity of a single fire---alarm. For eAL assessment !*lfl30Ses, the aO minute-clock starts at the time that the initial alarm was rece*

SLI-B-SOEtttent-vef#icffif~rmeEh Page 205 of 296 e -::::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases A-s+n§+e-4ife-a!

a rm, assent ot-A-6f-fRB+BffiiE}Rfste-f-a-F-H~-l~,ma-y-ee-i-AGicative e:l'-6*/tf~pme-nt fa i I ure 0-F-a----Sp-l:lous activation,anEI-Rot an v-erify the valie#y-of the alarm-+R-e--3G--minute peooe--is a reasonable El-eterm-ifl..e-i-f-an actual-FlR~x~oowe-ver, after th-at time, and abse-At eontrary, it is assume*1-th-a-t-an-actHal4-J.RE is in pregress. lf-arl-actual FIRE is verified by a report from-the field, then EAL #1 is immediately appUea-bte, aAd the emergeRBY-mBst--ee-Eleel-a-red if the FlR-*:--+s--Ae-4*HAg-tH-SfleG-Within 1 e---m-inutes of the repe-1t---t4Ae-a-l-afm-t&--ver+nee-4o-be due ~m * * ~*

this veFi:f.ea#e-R---eswrs-v1,,4tfl.i-A-3G minutes of the rece~f the alarm, #teA-Ws EAL is-Rm ap-p-licable and no emergeney-Ei-e-Gl-aratiBn is 1.varranted.
-A~* In addition to a FIRE addressee-by EAL #1 or EAL #2, a F-IR*-witl::i-~~

A-REA-Rffi---B*~ifl-*G-m~nH-tes-fAay-al-s0--f)ffie-Rtiafl.y deg raee the !ev-el-ef-plaA-t safety. This basis e>ttends to a located-oot-side the plant PR-G-TECTED AR-EA-[Sentence for plants-w-i#J--aA-.l-&F-&l-01HSide the plam-Proteeted-A-r:ea] EAL#4 If a FIRE within the plant or ISFSI [fef-f)lams-with an !S.CSl--etJtside tho pfant-Pfetectod AFea} PROTECTED AREA is of sufficient size to require a response by an offsite firefighting agency (e.g., a local town Fire Department), then the level of plant safety is potentially degraded. The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish. Declaration is not necessary if tho agency resources are placed on stand-by, or supporting post-extinguishment recovery or investigation actions. Basi-s-Rel-a~tfeffients from Appendix R Criterion 3 of /\ppe-ndix A to this part specifies that "Structures, compo-nents important t0 safety shall-be-GOStftned and located to minimize,ee-nsistent wi-th-ffih6f-SBfety-feq-Hi-femeA-t&,--the-f>fe-e-asfJ.i-ty-a-ne-effec+ef-fi res and exp I osi o~ When consiOOfi.Rg the effects of fire, those systems-associated with ach~Ef. maintaining safe shutdown-ee-Rd-ffions assume major importance to safety because earn-age to them can leaEi-te-ee-re-Elamage-resulting from loss of coolant through boi+-o-#.- Because fire may affect safe shuteown systems and because the loss of function m systems usee to miHgate the conseqHOflces of eesign e-asis accidents under pest-fire coneitions does not per so-impact public safety, the need tCr!+mtt fire damage to systems required to achieve and maintain safe shutdown conditions is greater than the need to Page 206 of 296 --===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases #ffii.H~e-te-4hese systems-Fe<:ttlffee-t0--mi#gate-t-Re-OOASeEfBOOCe-&-0k:J.e.si§-n basis accidents. ~wendix R to 10 Cffi-BO, requires, amoog-e-ffief-Bonsiderations, the use of 1 hol::IF #re-ooi:r:ieFS-f.ef-the encles1::1Fe-ef-ea!:>~~d-non safety-etFCtfits of one Feffiffi*!.ant train (G.2.c). A-6-tIBBd in EAL #2, the 30-m~A-H-tes-te-verify a singl-e-ak3-rm is well within this 1.Norst case '.J-he-1.:,!F time peFie4 Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via !G-EAL CA6,...1 or ~9SA8.1. Reference(s): . 1. NEI 99-01 HU4 Page 207 of 296 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 5 -Hazardous Gas Initiating Condition: Gaseous release IMPEDING access to equipment necessary for normal plant operations, cooldown or shutdown EAL: HAS.1 Alert Release of a toxic, corrosive, asphyxiant or flammable gas into any Table H-2 room or area AND Entry into the room or area is prohibited or IMPEDED (Note 5) Note 5: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted. Table H-2 Safe Operation & Shutdown Rooms/Areas Room/Area Auxiliary Building 70' RHR B Pump Room Auxiliary Building 80' RHR A Pump Room Auxiliary Building 114' West Control ~uilding 95' Div 1 RSS Room Mode Applicability: 3 -Hot Shutdown Definition(s):

Mode 3 3 3 3 IMPEDE(D)

-Personnel

access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).
Basis: This IC addresses an event involving a release of a hazardous gas that precludes or IMPEDES access to equipment necessary to maintain normal plant operation, or required for a normal plant cooldown and shutdown.

This condition represents an actual or potential substantial degradation of the level of safety of the plant. 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 gaseous release. The Page 208 of 296

=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases emergency classification is not contingent upon whether entry is actually necessary at the time ' of the release. Evaluation of the IC and EAL does not require atmospheric sampling; it only requires the Emergency Director's judgment that the gas concentration in the affected room/area is sufficient to preclude or significantly IMPEDE procedurally required access. This judgment may be based on a variety of factors including an existing job hazard analysis, report of ill effects on personnel, advice from a subject matter expert or operating experience with the same or similar hazards. Access should be considered as IMPEDED if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).

An emergency declaration is not warranted if any of the following conditions apply: I

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 gaseous release).

For example, the plant is in Mode 1 when the gaseous release occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4~.

The gas release is a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., fire suppression system testing).

o--::f-hB-aBtie-A-fe.F-wrnGA-fBBFWa-Fe-&eRtr-y-~&-feEttl~Fed-i-s-0f an ad minis Rat+/-tFe-fe. g., A8-FfF!a-!-F8-t!HGS-Bf-ffit:H~He-+P.-&l:)BCT1SR-Sj-;

The access control measures are of a conseNative or precautionary nature, and would not actually prevent or IMPEDE a required action. o lf the equipment in the listed room or area was already inoperable, or out-of-service.

before the event occurred, then no emerqency should be declared since the event will have no adverse impact be;iond that already allowed by Technical Specifications at the time of the event. An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels. Most commonly, asphyxiants work by merely displacing air in an enclosed environment. This reduces the concentration of oxygen below the normal level of around 19%, which can lead to breathing difficulties, unconsciousness or even death. This EAL does not apply to firefighting activities that generate smoke and that automatically or manually activate a fire suppression system in an area , or to intentional inerting of ee-ntai-Ame-A4BV'/ R on I y). Escalation of the emergency classification level would be via Recognition Category A, C or F !Cs. The list of plant rooms or areas with entry-related mode applicability identified specify those rooms or areas that contain eauioment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an Page 209 of 296

--::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases action to address an off-normal or emergency condition such as emergency repairs. corrective measures or emergency operations) are not included.

In addition, the list specifies the plant mode(s) during which entry would be required for each room or area (ref. 1 ). EAL HA5.1 mode applicability has been limited to the mode limitations of Table H-2 (Mode 3 only). Reference(s):

1. Attachment 2 Safe Operation

& Shutdown Areas Tables A-3 & H-2 Bases 2. NEI 99-01 HAS Page 210 of 296 -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 6 -Control Room Evacuation Initiating Condition: Control Room evacuation resulting in transfer of plant control to alternate locations EAL: HA6.1 Alert An event has resulted in plant control being transferred from the Control Room to the Remote Shutdown Panels Mode Applicability: All Definition(s): None Basis: This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations outside the Control Room. The loss of the ability to control the plant from the Control Room is considered to be a potential substantial degradation in the level .of plant safety. Following a Control Room evacuation, control of the plant will be transferred to alternate shutdown locations. The necessity to control a plant shutdown from outside the Control Room, in addition to responding to the event that required the evacuation of the Control Room, will present challenges to plant operators and other on-shift personnel. Activation of the ERO and emergency response facilities will assist in responding to these challenges. Transfer of olant control begins when the last licensed operator leaves the Control Room. Escalation of the emergency classification level would be via IC HS6. Refe re nee( s): 1. AOP-0031 Shutdown from Outside the Main Control Room 2. NEI 99-01 HA6 Page 211 of 296 Enter.av . fjJ River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 6 -Control Room Evacuation Initiating Condition: Inability to control a key safety function from outside the Control Room EAL: HS6.1 Site Area Emergency An event has resulted in plant control being transferred from the Control Room to the Remote Shutdown Panels AND Control of any of the following key safety functions is not re-established within 15 min. (Note 1 ):

Reactivity (Modes 1 and 2 only)
RPV water level
RCS heat removal Note 1: The Emergency Director should declare the event promptly upon *determining that the time limit has been exceeded, or will likely be exceeded.

The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 1 -Power Operation, 2 -Startup, 3 -Hot Shutdown, 4 -Cold Shutdown, 5 -Refueling Definition{s): None Basis: This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations, and the control of a key safety function cannot be reestablished in a timely manner. The failure tb gain control of a key safety function following a transfer of plant control to alternate locations is a precursor to a challenge to one or more fission product barriers within a relatively short period of time. The determination of whether or not "control" is established at the remote safe shutdown location(s) is based on Emergency Director judgment. The Emergency Director is expected to make a reasonable, informed judgment within 15 fth~ site specific time for transfer) minutes whether or not the ope.rating staff has control of key safety functions from the remote safe shutdown location(s). Transfer of plant control and the time oeriod to establish control begins when the last licensed operator leaves the Control Room. Page 212 of 296 -~ -~=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Escalation of the emergency classification level would be via IC FG1 or CG1 Reference(s):

1. AOP-0031 Shutdown from Outside the Main Control Room 2. EP FAQ 2015-014 3. NEI 99-01 HS6 Page 213 of 296 Q -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 -Emergency Director Judgment Initiating Condition: Other conditions exist that in the judgment of the Emergency Director warrant declaration of a UE EAL: HU7.1 Unusual Event Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of SAFETY SYSTEMS occurs. Mode Applicability: All Definition(s): SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 1 OCFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: c.., (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis: 'This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for arr UNUSUAL EVENTNOUE. Reference(s):

1. NEI 99-01 HU? Page 214 of 296

-::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 -Emergency Director Judgment Initiating Condition: Other conditions exist that in the judgment of the Emergency Director warrant declaration of an ALERT EAL: HA7.1 Alert Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels. Mode Applicability: All Definition(s): HOSTAGE -A person(s) held as leverage against the station to ensure that demands will be met by the ~tation. HOSTILE ACTION -An act toward RBS or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile Action should not be construed to include acts of civil disob.edience or felonious acts that are not part of a concerted attack on RBS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA). OWNER CONTROLLED AREA (OCA) -For the purposes of classification, the Security Owner Controlled Area (SOGA) or the area between the SOGA Fence and the PROTECTED AREA Boundary. PROJECTILE -An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety. PROTECTED AREA -The area* within the perimeter of the RBS security fence. Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for an ALERT. Page 215 of 296 _J Q -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Reference(s):

1. NEI 99-01 HA? Page 216 of 296 Q -:::::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 -Emergency Director Judgment Initiating Condition: Other conditions exist that in the judgment of the Emergency Director warrant declaration of a SITE AREA EMERGENCY EAL: \ HS7.1 Site Area Emergency Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentionai damage or malicious acts, (1) toward site personnel or equipment that could lead to the likely failure of or, (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure .levels beyond the SITE BOUNDARY Mode Applicability: All Definition(s): HOSTILE ACTION -An act toward RBS or its personnel that includes the use of violent force to destroy equip'ment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile Action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on RBS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA). OWNER CONTROLLED AREA (OCA) -For the purposes of classification, the Security Owner Controlled Area (SOGA) or the area between the SOGA Fence and the PROTECTED AREA Boundary. PROJECTILE -An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety. PROTECTED AREA -The area within the perimeter of the RBS security fence. SITE BOUNDARY -For classification and dose projection purposes, the Site Boundary is the area defined as the exclusion area or exclusion zone in 1OCFR100.3 (a) which is a boundary of approximately 3,000 feet (or 0.5748 mile) from the RBS reactor centerline. Page 217 of 296 I~ ~-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a SITE AREA EMERGENCY. . Reference(s):

1. NEI 99-01 HS? Page 218 of 296 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 -Emergency Director Judgment Initiating Condition: Other conditions exist that in the judgment of the Emergency Director warrant declaration of a GENERAL EMERGENCY EAL: HG7.1 General Emergency Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA Protective Action Guidefine exposure levels offsite for more than the immediate site area

Mode Applicability:

All Definition(s): HOSTAGE -A person(s) held as leverage against the station to ensure that demand~ will be met by the station. HOSTILE ACTION -An act toward RBS or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be , included. Hostile Action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on RBS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA). IMMINENT -The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. OWNER CONTROLLED AREA (OCA) -For the purposes of classification, the Security Owner Controlled Area (SOGA) or the area. between the SOGA Fence and the PROTECTED AREA Boundary. PROJECTILE -An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety. PROTECTED AREA -The area within the perimeter of the RBS security fence. Page 219 of 296 '~ -~ Entergy Basis: River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant

declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a GENERAL EMERGENCY.

Reference(s):

1. NEI 99-01 HG? Page 220 of 296 Q ~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases . Category S -System Malfunction EAL Group: Hot Conditions (RCS temperature

> 200°F); EALs in

this category are applicable only in one or more hot operating modes. Numerous system-related equipment failure events that warrant emergency classification have been identified in this category.

They may pose actual or potential threats to plant safety. The events of this category pertain to the following subcategories:

1. Loss of Emergency AC Power ', Loss of emergency plant electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity.

This category includes loss of onsite and offsite power sources for 4.16 KV ENS buses. 2. Loss of Vital DC Power Loss of emergency plant electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of power to or degraded voltage on the 125V DC vital buses. 3. Loss of Control Room Indications Certain events that degrade plant operator ability to effectively assess plant conditions within the plant warrant emergency classification. Losses of indicators are in this subcategory.

4. RCS Activity During normal operation, reactor coolant fission product activity is very low. Small concentrations of fission products in the coolant are primarily from the fission of tramp uranium in the fuel clad or minor perforations in the clad itself. Any significant rise from these base-line levels (2% -5% clad failures) is indicative of fuel failures and is covered

under the Fission Product Barrier Degradation category.

However, lesser amounts of clad damage may result in coolant activity exceeding Technical Specification limits. These fission products will be circulated with the reactor coo'lant and can be detected by coolant sampling.

5. RCS Leakage The reactor pressure vessel provides a volume for the coolant that covers the reactor core. The reactor pressure vessel and associated pressure piping (reactor coolant system) together provide a barrier to limit the release of radioactive material should the reactor fuel clad integrity fail. Excessive RCS leakage greater than Technical Specification limits indicates potential pipe cracks that may propagate to an extent threatening fuel clad, RCS and containment integrity

.. Page 221 of 296 ----------------~ -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases 6. RPS Failure This subcategory includes events related to failure of the Reactor Protection System (RPS) to initiate and complete reactor scrams. In the plant licensing basis, postulated failures of the RPS to complete a reactor scram comprise a specific set of analyzed events referred to as Anticipated Transient Without Scram (ATWS) events. For EAL classification, however, t' ATWS is intended to mean any scram failure event that does not achieve reactor* shutdown. If RPS actuation fails to assure reactor shutdown, positive control of reactivity is at risk and could cause a threat to fuel clad, RCS and containment integrity.

7. Loss of Communications Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification.

8. Hazardous Event Affecting Safety Systems Various natural and technological events that result in degraded plant SAFETY SYSTEM performance or significant VISIBLE DAMAGE warrant emergency classification under this subcategory.

Page 222 of 296 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: S -System Malfunction Subcategory: 1 -Loss of Emergency AC Power Initiating Condition: Loss of all offsite AC power capability to ENS buses for 15 minutes or longer EAL: SU1.1 Unusual Event Loss of all offsite AC power capability, Table S-1, to DIV I and DIV II 4.16 KV ENS buses for~ 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Table S-1 AC Power Sources Offsite

1 RTX-XSR 1 C
1 RTX-XSR1 D Onsite
EGS-EG1A
EGS-EG1 B Mode Applicability:

1 -Power Operation, 2 -Startup, 3 -Hot Shutdown Definition(s): None Basis: The DlV 111 bus (1 E22*S004) is not credited because it only supplies power to the HPCS pump and associated loads, not any lonc1 term decay heat removal systems. In particular. suppression pool coolinq mechanisms would be essential subsequent to a station blackout. This IC addresses a prolonged loss of offsite power. The loss of offsite power sources renders the plant more vulnerable to a complete loss of power to AC ernergef1*Y-ENS buses. This condition represents a potential reduction in the level of safety of the plant. For emergency classification purposes, "capability" means that an offsite AC power source(s) is available to the &AIBrge-ABy-ENS buses, whether or not the buses are powered from it. Page 223 of 296 -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Fifteen minutes was selected as a threshold to exclude transient or momentary losses of offsite power. Escalation of the emergency classification level would be via IC SA1. Refe re nee{ s): 1. USAR Section 8.1 Electric Power Introduction

2. EE-001AC Startup Electrica1Distribution Chart 3. SOP-0046 4.16 KV System 4. AOP-0004 Loss of Offsite Power 5. NEI 99-01 SU1 Page 224 of 296

-~ Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: S -System Malfunction Subcategory: 1 -Loss of Emergency AC Power Initiating Condition: Loss of all but one AC power source to ENS buses for 15 minutes or longer EAL: SA1.1 Alert AC power capability, Table S-1, to DIV I and DIV II 4.16 KV ENS buses reduced to a single power source for~ 15 min. (Note 1) AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Table 5-1 AC Power Sources Offsite

1 RTX-XSR1 C
1 RTX-XSR1 D Onsite
EGS-EG1A
EGS-EG1 B Mode Applicability:

1 -Power Operation, 2 -Startup, 3 -Hot Shutdown

Definition(s):

SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 1 OCFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; Page 225 of 296

-~-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Leve.I Technical Bases (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

Basis: The DIV 111 bus ( 1 E22*S004) is not credited because it only supplies power to the H PCS pump and associated loads, not any long term decay heat removal systems. In particular. suooression pool cooling mechanisms wouid be essential subseauent to a station blackout. This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment. This IC provides an escalation path from IC SU1. An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency-ENS bus. Some examples of this condition are presented below.

A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).
A loss of all offsite power and loss of all ENS emergency power sources (e.g., onsite diesel generators) with a single train of emergeABy-ENS buses being back-fed from the unit main generator.
A loss of ENS emergency power sources (e.g., onsite diesel generators) with a single train of ENS emergency buses being e-aek-fed from an offsite power source. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power. Escalation of the emergency classification level would be via IC SS1. This EAL is the hot condition equivalent ofte the cold condition EAL CU2.1 . .:. Reference(s):

1. USAR Section 8.1 Electric Power Introduction

2. EE-001 AC Startup Electrical Distribution Chart 3. SOP-0046 4.16 KV System 4. AOP-0004 Loss of Offsite Power 5. NEI 99-01 SA1 Page 226 of 296

-===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: S -System Malfunction Subcategory: 1 -Loss of Emergency AC Power Initiating Condition: Loss of all offsite and all onsite AC power to ENS buses for 15 minutes or longer EAL: SS1 .1 Site Area Emergency Loss of all offsite and all onsite AC power capability to DIV I and DIV II 4.16 KV ENS buses for~ 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 1 -Power Operation, 2 -Startup, 3 -Hot Shutdown Definition(s): SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 1 OCFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. 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. Mitigative strategies using other power sources (Division Ill diesel generator, FLEX generators, etc.) may be effective in supplying power to these buses. These power sources must be controlled in accordance with abnormal or emergency operating procedures, or beyond design basis accident response guidelines (e.g., FLEX support guidelines) and must be capable (alone or in combination) of supplying power for long term decay heat removal systems. In particular, suppression pool cooling systems would be essential subsequent to a station blackout. In addition, fission *product barrier monitoring capabilities may be degraded under Page 227 of 296

~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases these conditions.

This IC represents a condition that involves actual or likely major failures of plant functions needed for the protection of the public. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via I Cs AG1, FG1 or SG1. This EAL is the hot condition eauivaient of the cold condition EAL CA2.1. Reference(s):

1. USAR Section 8.1 Electric Power Introduction

2. EE-001 AC Startup Electrical Distribution Chart 3. SOP-0046 4.16 KV System 4. AOP-0004 Loss of Offsite Power 5. NEI 99-01 SS1 Page 228 of 296

-::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: S -System Malfunction Subcategory: 1 -Loss of Emergency AC Power Initiating Condition: Prolonged loss of all offsite and all onsite AC power to ENS buses EAL: SG1 .1 General Emergency Loss of all offsite and all onsite AC power capability to DIV I and DIV II 4.16 KV ENS buses

AND EITHER:
Restoration of at least one 4.16 KV ENS bus in < 4 hours is not likely (Note 1)
RPV water level cannot be restored and maintained

> -187 in. Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 1 -Power Operation, 2 -Startup, 3 -Hot Shutdown Definition(s): SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 1 OCFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coo_lant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis: Indication of continuina core cooling degradation is manifested by the inability to restore and maintain RPV water level above the Minimum Steam Cooling Reactor Water Level (-187 in.) (ref. 5). Core submergence is the most desirable means of core cooling, however when RPV level is below TAF. the uncovered portion of the core can be cooled by less reliable means (i.e., steam cooling or spray cooling). This IC addresses a prolonged loss of all power sources to AC ENS emergency buses. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric Page 229 of 296 @ ~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat ~emoval and the ultimate heat sink. Mitigative strategies using other power sources (Division Ill diesel generator. FLEX generators. etc.) may be effective in supplying power to these buses. These power sources must be controlled in accordance with abnormal or emergency operating procedures, or beyond design basis accident response guidelines (e.g., FLEX support guidelines) and must be capable (alone or in combination) of supplying power for long term decay heat removal systems. In particular. suppression pool cooling systems would be essential subsequent to a station blackout. A prolonged loss of these buses will lead to a loss of one or more fission product barriers. In addition, fission product barrier monitoring capabilities may be degraded under these conditio'ns. The EAL should require declaration of a General Emergency prior to meeting the thresholds for IC FG1. This will allow additional time for implementation of offsite protective actions. Escalation of the emergency classification from Site Area Emergency will occur if it is projected that power cannot be restored to at least one AC ENS emergency bus by the end of the analyzed station blackout coping period. Beyond this time, plant responses and event trajectory are subject to greater uncertainty, and there is an i-RG-FeaseEl-greater likelihood of challenges to multiple fission product barriers. 4 hours is the site-specific SBO coping anaiysis -, time (ref. 6). The estimate for restoring at least one ENS emergency bus should be based on a realistic appraisal of the situation. Mitigation actions with a low probability of success should not be used as a basis for delaying a classification upgrade. The goal is to maximize the time available to prepare for, and implement, protective actions for the public. The EAL will also require a General Emergency declaration if the loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core. Reference(s):

1. USAR Section 8.1 Electric Power Introduction

2. EE-001AC Startup Electrical Distribution Chart 3. SOP-0046 4.16 KV System 4. AOP-0004 Loss of Offsite Power 5. EOP-1 RPV Control 6. USAR Appendix 15C Station Blackout 7. NEI 99-01 SG1 Page 230 of 296

==-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

S -System Malfunction Subcategory: 1 -Loss of Emergency AC Power Initiating Condition: Loss of all ENS AC and vital DC power sources for 15 minutes or longer EAL: SG1 .2 General Emergency Loss of all offsite and all onsite AC power capability to DIV I and DIV 11 4.16 KV ENS buses for~ 15 min. (Note 1) AND Indicated voltage is< 105 voe on Safety Related DIV I and DIV 11125 voe buses for 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time 1.imit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 1 -Power Operation, 2 -Startup, 3 -Hot Shutdown Definition(s): SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 1 OCFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: ' (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis: Safety Related DC buses ENB-SWG01A (DIV I) and ENB-SWG01 B (DIV II) feed the Division I and Division II loads respectively. The Division I and Division II batteries each have 60 cells with a specific minimum voltage of 1. 75 volts/cell. These cell voltages yield minimum design bus voltages of 105 voe (ref. 5). This IC addresses a concurrent and prolonged loss of both emergency ENS AC and Vital DC power. A loss of all emergency ENS AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, Page 231 of 296 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. Mitigative strategies using other power sources (Division Ill diesel generator, FLEX generators, etc.) may be effective in supplying power to these buses. These power sources must be controlled in accordance with abnormal or emergency operating procedures, or beyond design basis accident response guidelines (e.g., FLEX support guidelines) and must be capable (alone or in combination) of supplying power for long term decay heat removal systems. In particular, suppression pool cooling systems would be essential subsequent to a station blackout. A loss of vital DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both emergency ENS AC and Vital DC power will lead to multiple challenges to fission product barriers. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. The 15-minute emergency declaration clock begins at the point when both EAL thresholds are met. Reference(s):

1. USAR Section 8.1 Electric Power Introduction

2. EE-001AC Startup Electrical Distribution Chart 3. SOP-0046 4.16 KV System 4. AOP-0004 Loss of Offsite Power 5. Safety Related Battery Specification 244.521 6. USAR 8.3.2 DC Power Systems 7. NEI 99-01 SGS Page 232 of 296

-=::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: S -System Malfunction Subcategory: 2 -Loss of Vital DC Power Initiating Condition: Loss of all vital DC power for 15 minutes or longer EAL: 552.1 Site Area Emergency Indicated voltage is< 105 VDC on Safety Related DIV I and DIV II 125 voe buses for 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 1 -Power Operation, 2 -Startup, 3 -Hot Shutdown Definition(s): SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis: Safety Related DC buses ENB-SWG01A (DIV I) and ENB-SWG01 B (DIV II) feed the Division I and Division II loads respectively. The Division I and Division II batteries each have 60 cells with a specific minimum voltage of 1. 75 volts/cell. These cell voltages yield minimum design bus voltages of 105 voe (ref. 1 ). This IC addresses a loss of vital DC power which compromises the ability to monitor and control SAFETY SYSTEMS. In modes above Cold Shutdown, this condition involves a major failure of plant functions needed for the protection of the public. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via ICs AG1, FG1 or SG1SG-8. This EAL is the hot condition eauivalent of the cold condition EAL CU4.1. Page 233 of 296 /"'.7'\ 7 Entergy Reference(s): River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases 1. Safety Related Battery Specification 244.521 2. USAR 8.3.2 DC Power Systems 3. NEI 99-01 SS8 Page 234 of 296

-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

S -System Malfunction Subcategory: 3 -Loss of Control Room Indications Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer EAL: SU3.1 Unusual Event An UNPLANNED event results in the inability to monitor one or more Table S-2 parameters from within the Control Room for~ 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upoh determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Table 5-2 Safety System Parameters

Reactor power
RPV water level
RPV pressure
Containment pressure
Suppression Pool water level
Suppression Pool temperature Mode Applicability:

1 -Power Operation, 2 -Startup, 3 -Hot Shutdown Definition(s): SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 1 OCFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. 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. Page 235 of 296 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Basis: This IC addresses the difficulty associated with monitoring normal plant conditions without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition is a precursor to a more significant event and represents a potential degradation in the level of safety of the plant. As used in this EAL, an "inability to monitor" means that values for one or more, of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reaqtor power level cannot be determined from any analog, digital and recorder source within the Control Room. An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making. This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, oore coo~i-Ag-fPWRt-1-RPV water level f-BWR}-and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters are lost, the*n the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for reaetor vessel level-fP-VVR] I RPV water level {BWR]-cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication. Escalation of the emergency classification level would be via IG-EAL SA3.1 SA,2. Reference(s):

1. USAR 7.5 Safety-Related Display Instrumentation

2. NEI 99-01 SU2 Page 236 of 296

-~ Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: S -System Malfunction Subcategory: 3 -Loss of Control Room Indications Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress EAL: SA3.1 Alert -An UNPLANNED event results in the inability to monitor one or more Table S-2 parameters from within the Control Room for~ 15 min. (Note' 1) AND Any significant transient is in progress, Table S-3 Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: Table S-2 Safety System Parameters

Reactor power
RPV water level
RPV pressure
Containment pressure
Suppression Pool water level
Suppression Pool temperature Table S-3 Significant Transients
Reactor scram
Runback > 25% thermal reactor power
Electrical load rejection.>

25% full electrical load

ECCS injection
Thermal power oscillations

> 10% 1 -Power Operation, 2 -Startup, 3 -Hot Shutdown Page 237 of 296 i_ ________ --:::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Definition(s): SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 1 OCFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: ( 1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. UNPLANNED -A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

Basis: This IC addresses the difficulty associated with monitoring rapidly changing plant conditions during a transient without the ability to ot;>tain SAFETY SYSTEM parameters from within the Control Room. During this condition, the margin to a potential fission product barrier challenge is reduced. It thus represents a potential substantial degradation in the level of safety of the plant. As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s).

For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room. An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making. This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, RPV water level and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for reactor vessel level [P\AlR] I RPV water level fBl!'/R] cannot be determined from the indications and Page 238 of 296 ~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication. Escalation of the emergency classification level would be via I Cs FS1 or tG-AS1 Reference(s): ' 1. USAR 7.5 Safety-Related Display Instrumentation

2. NEI 99-01 SA2 Page 239 of 296

===* Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

S -System Malfunction Subcategory: 4 -RCS Activity Initiating Condition: RCS activity greater than Technical Specification allowable limits EAL: SU4.1 Unusual Event Offgas Pretreatment radiation monitor high alarm (P601-22A-F03, OFF GAS PRE-TREAT HIGH RADIATION) Mode Applicability: 1 -Power. Operation, 2 -Startup, 3 -Hot Shutdown Definition(s): None Basis: The Offqas Pretreatment monitors radioactivit~he Offqas system downstream of the Offgas condenser. The monitor detects the radiation level that is attributable to the fission gases groduced in the reactor and transported with steam throuah the turbine to the condenser. The High alarm indicates that the radioactivitv oresent at the recombiner effluent discharge is approaching the Technical Specification 3.7.4 limit. The nominal setpoint of 1.5 times the full QOwer process backqround radiation level ensures that the activity will not exceed a value corresponding to the Technical Specification LCO 3.7.4 allowable release rate. (ref. 1) This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant. Escalation of the emergency classification level would be via I Cs FA 1 or the Recognition Category A ICs. Reference(s):

1. TRM section 3.3.7.8.2 Offgas System Radiation Monitoring Instrumentation

2. USAR 11.5 Process and Effluent Radiological Monitoring and Sampling Systems 3. Technical Specification 3.7.4 Main Condenser Offgas 4. NEI 99-01 SU3 Page 240 of 296 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

S -System Malfunction Subcategory: 4 -RCS Activity Initiating Condition: RCS activity greater than Technical Specification allowable limits EAL: SU4.2 Unusual Event Coolant activity > 0.2 µCi/gm dose equivalent 1-131 for> 48 hours OR Coolant activity > 4.0 µCi/gm dose equivalent 1-131 instantaneous Mode Applicability: 1 -*Power Operation, 2 -Startup, 3 -Hot Shutdown Definition(s): None Basis: This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant. Escalation of the emergency classification level would be via ICs FA1 or the Recognition Category A ICs. Reference(s):

1. Technical Specification B 3.4.8, RCS Specific Activity bases 2. USAR Section 15.6.4 Steam System Piping Break Outside Containment

3. NEI 99-01 SU3 Page 241 of 296

"==='" Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: S -System Malfunction Subcategory: 5 -RCS Leakage Initiating Condition: RCS leakage for 15 minutes or longer EAL: SUS.1 Unusual Event RCS unidentified or pressure boundary leakage> 10 gpm for~ 15 min. (Note 1) OR RCS identified leakage> 25 gpm for~ 15 min. (Note 1) OR Leakage from the RCS to a location outside Containment> 25 gpm for~ 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 1 -Power Operation, 2 -Startup, 3 -Hot Shutdown Definition(s): UN/SOLABLE -An open or breached system line that cannot be isolated, remotely or locally. Basis: Failure to isolate the leak (from the Control Room or locallv) within 15 minutes. or if,known that the leak cannot be isolated within 15 minutes. from the start of the leak requires immediate classification. Identified leakage is leakage into the drywell. such as that from pump seals or valve packing, that is g§2tured and conducted to a collecting sump: 9r leakage into the drywell atmosphere from sources that are both specifically located and known either noMo interfere with the operation of leakaqe detection systems or not to be pressure boundary leakage. Unidentified leakage is all leakage into the drvwell that is not identified leakaqe (ref. 2, 3). Pressure boundary leakaae is leakage through a non-isolable fault in a' Reactor Coolant System (RCS) component body, pipe wall, or vessel wall (ref. 2, 3). This IC addresses RCS leakage which may be a precursor to a more significant event. In this case, RCS leakage has been detected and operators, following applicable procedures, have been unable to promptly isolate the leak. This condition is considered to be a potential degradation of the level of safety of the plant. The first and second EAL conditions EAL #1 and EAL #2 are focused on a loss of mass from the RCS due to "unidentified leakage", "pressure boundary leakage" or "identified leakage" (as Page 242 of 296

===* Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases these leakage types are defined in the plant Technical Specifications).

The third condition eAb #-3-addresses an RCS mass loss caused by an UNISOLABLE leak through an interfacing system. These conditions EJ\Ls thus apply to leakage into the containment, a-&eBondary sioo system (e.g., steam-§-en-er-atef-tube leakage in a P-WF?)-or a location outside of containment. The leak rat~ values for each condition E-Ab--were selected because they are usually observable with normal Control Room indications. Lesser values typically require consuming calculations to determine (e.g., a mass balance calculation). The first condition -EAb-#4--uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage. The release of mass from the RCS due to the as-designed/expected operation of a relief valve does not warrant an emergency classification. ~WRs, an-eFAeF§OAB-y-s+assme-ati-en--YvEH:tla be--fe.EtufffiEl-if a mass loss is causeEl--&y-a--FO-Hef..-val-ve-t-haHS-not-+t;l-fl*fte.Ri-A.g-as eesis-Real&xf)BBtee-(-&ft.,&-Fel+e-f-vatve--soo~s--sf3eA--a-R*!-the-l-i-R-e-#ew-0an-Rm-ee-~setatee+.- F-e-F 8-v-V-Rs,a-A stuck-open Safety Relief Valve (SRV) or SRV leakage is not considered either identified or unidentified leakage by Technical Specifications and, therefore, is not applicable to this EAL.

The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible.

Escalation of the emergency classification level would be via I Cs of Recognition Category A or F. Reference(s}:

1. USAR Section 5.2.5 Reactor Coolant Pressure Boundary and ECCS Leakage Detection System 2. Technical Specification Definitions Section 1.1 3. Technical Specification 3.4.5 2. NEI 99-01 SU4 Page 243 of 296 e, . -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

S -System Malfunction Subcategory: 6 -RPS Failure Initiating Condition: Automatic or manual scram fails to shut down the reactor EAL: SU6.1 Unusual Event An automatic scram did not shut down the reactor as indicated by reactor power > 5% after any RPS setpoint is exceeded AND A subsequent automatic scram or manual scram action taken at the reactor control console (Mode Switch, Manual PBs, ARI) is successful in shutting down the reactor as indicated by reactor power~ 5% (APRM downscale) Note 8: A manual scram action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron. injection strategies. Mode Applicability: 1 -Power Operation, 2 -Startup Definition(s): IMMINENT -The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor ~]-I-scram- [B 1 NRJ) that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic fl:Rf}{PWR 1-/ scram-[BVl/R]) is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant. ' The first condition of this EAL identifies the need to cease critical reactor operations by actuation of the automatic Reactor Protection System (RPS) scram function. A reactor scram is automatically initiated by the RPS when certain continuously monitored parameters exceed predetermined setpoints (ref. *1 ). A successful scram has therefore occurred when there is sufficient rod insertion from the trip of RPS to bring the reactor power to or below the APRM downscale setpoint of 5% (ref. 4). For the purposes of emergency classification. successful manual scram actions are those which can be quickly performed from the reactor control console (i.e .. mode switch, manual scram pushbuttons. or ARI initiation). Reactor shutdown achieved by use of alternate control Page 244 of 296 J A ~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases rod insertion methods (i.e., EOP-1 A Enclosure

26) does not constitute a successful manual scram (ref. 2). Following anv automatic RPS scram siqnal, operating procedures (e.g., EOP-1A) prescribe insertion of redundant manual scram signals to back up the automatic RPS scram function and ensure reactor shutdown is achieved.

Even if the first subsequent manual scram signal inserts all control rods to the full-in position irnmediately after the initial failure of the automatic scram .. the lowest level of classification that must be declared is an Unusual Event (ref. 3). Taking the Mode Switch to Shutdown is a manual scram action. When the Mode Switch is taken out of the Run position. however, the nuclear instrumentation scram setooint is lowered. If reactor power remains above the lowered setpoint, an automatic scram is initiated. For the purposes of this EAL, a successful automatic initiation of ARI that reduces reactor power to< 5% is not considered a successful automatic scram. if automatic initiation of ARI has occurred and caused reactor shutdown. the automatic RPS scram must have failed. ARI is a backup means of inserting control rods in the unlikelv event that an automatic RPS scram signal exists but the reactor continues to generate significant power. However. a successful automatic or manual initiation of ,A.RI is an acceptable means of establishing reactor shutdown conditions relative to the EAL threshold in the absence of any required subsequent manual scram actions. In the event that the operator identifies a reactor scram is IMMINENT and initiates a successful manual reactor scram before the automatic scram setpoint is reached, no declaration is reauired. The successful manual scram of the reactor before it reaches its automatic scram setpoint or reactor scram signals caused bv instrumentation channel failures do not lead to a potential fission product barrier loss. If manual reactor scram actions fail to reduce reactor oower below 5%, the event escalates to the Alert under EAL SA6.1. If by procedure, operator actions include the initiation of an immediate manual scram following receipt of an automatic scram signal and there are no clear indications that the automatic scram failed (such as a time delay following indications that a scram setpoint was exceeded), it may be difficult to determine if the reactor was shut down because of automatic scram or manual actions. If a subsequent review of the scram actuation indications reveals that the automatic scram did not cause the reactor to be shut down, then consideration should be given to evaluating the fuel for potential damaoe, and the reporting requirements of 50. 72 should be considered for the transient event. Following the failure on an automatic reactor ftlip[P\IVR] / scram~. operators will promptly initiate man'ual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor ~PINR] I scram [BVVR])). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems. If an initial manual reactor (trip [Pl/VR] / scram [81,A/R]) is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor (-t-r$£-PWR-}+scram [BWR])) using a different switch). Page 245 of 296 ~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Depending upon several factors, the initial or subsequent effort to manually fti:$-f, 0-\,L\/R]+scram fB-VllR-B-the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor ~PVI/F?] I scram -fB-1A,LR-B-signal. If a subsequent manual or automatic ftrip [PWR}-1 scram-{B-WRH-is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems. A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor fmpfPWR}-/-scram [Bl,ll/F?])). This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles". Taking the Reactor Mode Switch to Shutdown is considered to be a manual scram action. [B-lNF?] The plant response to the failure of an automatic or manual reactor (rn-p-f-P-WR}-/-scram fBWRB-will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via G-EAL SA6.1-SA-5-. Depending upon the plant response, escalation is also possible via IC FA 1. Absent the plant conditions needed to meet either IC SA6&Ae-or FA1, an Unusual Event declaration is appropriate for this event. Pffi&e*1-l;lf6-Gffie-r-i&. Should a reactor ft-Fft3-f-P-W~scram-fBWRB-signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied.

If the signal generated as a result o-f plant work causes a plant transient that results in a condition that should have included an automatic reactor (-tf$-[PW-Rtl--scram-fBWRB and the RPS fails to automatically shutdown the reactor, then this IC and tA-e-associatecf EALs are applicable, and should be evaluated.
If the signal generated as a resuit of olant work does not cause a plant transient and the scram failure is determined through other means (e.g., assessment of test results), then this IC and Ute-associated EALs are not applicable and no classification is warranted.

Reference(s):

1. Technical Specification Table 3.3.1.1-1 Reactor Protection System Instrumentation

2. EOP-1A RPV Control, ATWS 3. EOP-1 RPV Control 4. NEI 99-01 SU5 Page 246 of 296

-::::::-Entergy River Bend Station EAL Basis Document Revision XXX .Attachment 1 -Emergency Action Level Technical Bases Category: S -System Malfunction Subcategory: 6 -RPS Failure Initiating Condition: Automatic or manual scram fails to shut down the reactor EAL: SU6.2 Unusual Event A manual scram did not shut down the reactor as indicated by reactor power > 5% after any manual scram action was initiated AND A subsequent automatic scram or manual scram action taken at the reactor cont~ol console (Mode Switch, Manual PBs, ARI) is successful in shutting down the reactor as indicated by reactor power:::; 5% (APRM'downscale) (Note 8) Note 8: A manual scram action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. Mode Applicability: 1 -Power Operatio_n, 2 -Startup Definition(s): SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 1 OCFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: ( 1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram-[-B-WRB that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic (trip [PVVR] I scram [BVl/R]) is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant. Page 24 7 of 296 A -::::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases This EAL addresses a failure of a rnanuallv initiated scram in the absence of having exceeded an automatic RPS trip setpoint and a subsequent automatic or manual scram is successful in shutting down the reactor (reactor power< 5%) (ref. *j ). A successful scram has therefore occurred when there is sufficient rod insertion from the trip of RPS to bring the reactor power to or below the APRM downscale setpoint of 5%_ For the purposes of emerqency classification, successful manual scram actions are those which can be quickly performed from the reactor control console (i.e., Mode Svvitch. manual scram pushbuttons, or ARI initiation). Reactor shutdown achieved by use of alternate control rod insertion methods (Le .. FOP-1A Enciosure

26) does not constitute a successful manual scram (ref. 2). Takinc1 the Mode Switch to Shutdown is a manual scram action. When the Mode Switch is taken out of the Run position, however, the nuclear instrumentation scram setpoint is lowered. If reactor power remains above the lowered setpoint an automatic scram is initiated.

Successful automatic or manual initiation of ARl is an acceptable means of establishing reactor shutdown conditions relative to the EAL threshold in the absence of anv reauired_ subsequent manual scram actions. !f both subsequent automatic and subsequent manual reactor scram actions in the Control Room fail to reduce reactor power below the power associated with the SAFETY SYSTEM design (.:::; 5%) following a failure of an initial manual scram. the event escalates to an Alert h!nder EAL SA6:l. Following the failure on an automatic reactor ft-i:-~Wf?}-1-scram{&W--R-}), operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor (t-Fif)-f-P-WRJ+scram-f&WRB). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems. I If an initial manual reactor ~scram-f&WRB is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor (tlivfP-WRj--1-scram-f&WRB) using a different switch1. Depending upon several factors, the initial or subsequent effort to manually ftr-ip-fP-l4/R-}+scram fB-WR}) the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor (tf$-fPWRt-l-scram-f&WRB signal. If a subsequent manual or automatic ft~ scram-f&WRB is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems. A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor (-tftp fPVVR] I scram [Bl!'/R])). This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles". Page 248 of 296 -~-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Taking the Reactor Mode Switch to Shutdown is considered to be a manual scram action. fBWR] The plant response to the failure of an automatic or manual reactor tIDP-fP-W-Rt+-scram-fBWRJ-} will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC ~SA6. Depending upon the plant response, escalation is also possible via IC FA 1. Absent the plant conditions needed to meet either IC SA-5-SA6 or FA1, an Unusual Event declaration is appropriate for this event. A-feaet8f-Shl.::l-tdewR-i-5-El-eIBrmi+1ee-~A-aeooFElaFis-e--with--app+iB-aele E FAef§-eP.£-y-G-p-er-atifls ProcesuH~-GFiffi.Fi-a-Should a reactor ~RH-scram [BVVR]) signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied.

If the signal generated as a result of plant work causes a plant transient that that results in a condition that should have included an automatic reactor fffiWWR}-,l--scram f8W-RB and the RPS fails to automatically shutdown the reactor, then this IC and tA-e associated EALs are applicable, and should be evaluated.
If the signal aenerated as a result of olant work does not cause a plant transient and the (ffif3-fP-WR]+scram-{BWRH failure is determined through other means (e.g., assessment of test results), then this IC and the-associated EALs are not applicable and no classification is warranted.

Reference(s):

1. Technical Specification Table 3.3.1.1-1 Reactor Protection System Instrumentation

2. EOP-1A RPV Control, ATWS 3. EOP-1 RPV Control 4. NEI 99-01 SUS Page 249 of 296

~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: S -System Malfunction Subcategory: 2 -RPS Failure Initiating Condition: Automatic or manual scram fails to shut down the reactor and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor EAL: SA6.1 Alert An automatic or manual scram fails to shut down the reactor as indicated by reactor power >5% AND Manual scram actions taken at the reactor control console (Mode Switch, Manual PBs, ARI) are not successful in shutting down the reactor as indicated by reactor power> 5% (Note 8) Note 8: A manual scram action is any operator action, cir set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. Mode Applicability: 1 -Power Operation, 2 -Startup Definition(s): _/ SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 1 OCFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor *and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor I ~scram [BV'/R]) that results in a reactor shutdown, and subsequent operator manual actions taken at the reactor control consoles to shutdown the reactor are also unsuccessful. This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if the reactor is Page 250 of 296 e *==-=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases subsequently shutdown by an action taken away from the reactor control consoles since this event entails a significant failure of the RPS. This EAL addresses any automatic or manual reactor scram siqnal that fails to shut down the reactor followed by a subsequent manual scram that fails to shut down the reactor to an extent the reactor is producina enerqy in excess of the heat load for which the SAFETY SYSTEMS were designed (> 5%). For the purposes of emergency classification. successful manual scram actions are those v,;hich can be quickly performed from the reactor control console (i.e., Mode Switch. manual scram pushbuttons, or ARi initiation). Reactor shutdown achieved by use of alternate control rod insertion methods (i.e., EOP-1A Enclosure

26) does not constitute a successful manual scram (ref. 2). For the purposes of this EAL. a successful automatic initiation of ARI that reduces reactor power to or below 5% is not considered a successful automatic scram. lf automatic actuation o'f ARI has occurred and caused reactor shutdmvn, the automatic RPS scram must have failed. ARI is a backup means of inserting control rods in the unlikely event that an automatic RPS scram signal exists but the reactor continues to aenerate significant oovver. However. a successful automatic initiation of ARI is an acceptable means of establishing reactor shutdown conditions relative to the EAL threshold in the absence of anv required subsequent manual scram actions. A manual action at the reactor control console is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor tfF$ [Pll'/R]-1---scram-ffl-WRB).

This action does not include manually driving in control rods or implementation of boron injection strategies. If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control consoles (e.g., locally opening breakers). Actions taken at back panels or other locations withiri the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles". Taking the Reactor Mode Switch to Shutdown is considered to be a manual scram action. *f&WRJ The plant response to the failure of an automatic or manual reactor (tFip [,Pl,41R1+scram--f-BWR}) will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If the failure to shutdown the reactor is prolonged enough to cause a challenge to the coFe-Geoling [PW-Rt+-RPV water level [BWR] or RCS-RGS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC SS5SS6. Depending upon plant responses and symptoms, escalation is also possible via IC FS1. Absent the plant conditions needed to meet either IC SS&-SS6 or FS1, an Alert declaration is appropriate for this event. Page 251 of 296 -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F ICs; however, this IC and EAL are included to ensure a timely emergency declaration. Page 252 of 296 -:-.::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases P-feeeEltlre-Bfi-tefia-Reference(s):

1. Technical Specification Table 3.3.1.1-1 Reactor Protection System Instrumentation

2. EOP-1A RPV Control, ATWS 3. EOP-1 RPV Control 4. NEI 99-01 SA5 Page 253 of 296

-::::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: S -System Malfunction Subcategory: 2 -RPS Failure Initiating Condition: Inability to shut down the reactor causing a challenge to RPV water level or RCS heat removal EAL: 556.1 Site Area Emergency An automatic or manual scram fails to shut down the reactor as indicated by reactor power >5% AND All actions to shut down the reactor are not successful as indicated by reactor power> 5% AND EITHER: RPV water level cannot be restored and maintained > -187 in. OR Heat Capacity Temperature Limit (HCTL) exceeded (EOP Figure 2) Mode Applicability: 1 -Power Operation, 2 -Startup Definition(s): SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 1 OCFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: ' (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis: This IC addresses a failure of the RPS-to initiate or complete an automatic or manual reactor

J (trip [PWR] ,L scram--fBWRB that results in a reactor shutdown, all subsequent operator actions to manually shutdown the reactor are unsuccessful, and continued power generation is challenging the capability to adequately remove heat from the core and/or the RCS. This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration of a Site Area Emergency.

Page 254 of 296 e -::::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases This EAi addresses the follovving: (ll Any automatic reactor scram signal followecl by a manual scram that fails to shut down the reactor to an extent the reactor is producing energy in excess of the heat load for which the SAFETY SYSTEMS were designed (EAL SA6.1 ). and Indications that either core cooling is extremely challenaed or heat removal is extremely challenged. Reactor shutdown achieved by use of control rod insertion methods in EOP-1A Enclosure 26 are also credited as a successful shutdown provided reactor power can be reduced to or below the APRi\/1 downscale trip setpoint before indications of an extreme challenge to either core cooling or heat removal exist (ref. 1) The combination of failure of both front llne and backup protection systems to function in response to a plant transient. along with the continued production of heat poses a direct threat to the Fuel Clad and RCS barriers. indication that core coolinq is extremely chailenqed is manifested bv inability to restore and maintain RPV water level above the Minimum Steam Cooling RPV \Nater Level (iVJSCRWL) (ref. 1 ). The fv1SCRVVL is the lowest RPV level at which the covered portion of the reactor corE;z. will generate sufficient steam to prevent anv clad temperature in the uncovered part of the core from exceedinq 1500°F. This water level is utilized in the EOPs to preclude fuel damage when RPV level is below the too of active fuel. RP\/ level below the MSCRVVL for an extended geriod of time without satisfactorv. core sprav coo lino could be a precursor of a core melt. sequence (ref. 2). The Heat Capacity Temperature Limit (HCTL, FOP Fioure 2) is the highest suopression pool water temperature from which Emergency RPV Depressurization will not raise suppression goo! temoerature above the maximum design suooression pool temperature. The HCTI is a function of RPV pressure and suppression pool water level. It is utilized to preclude failure of the containment and equipment in the containment necessary for the safe shutdown of thqJ2!ant. This threshold is met when the final step of section SPT in EOP-2, Primary Containment Control. is reached (ref. 3). This condition addresses loss of functions required for hot shutdown with the reactor at pressure and temperature. In some instances, the emergency classification resulting from this IGIEAL may be higher than that resulting from an assessment of the plant responses and symptoms against the Recognition Category F ICs/EALs. This is appropriate in that the Recognition Category F ICs/EALs do not address the additional threat posed by a failure to shut_down the reactor. The inclusion of this +G-aF14-EAL ensures the timely declaration of a Site Area Emergency in response to prolonged failure to shutdown the reactor. A-reactor shutdown is detmmined in a~~eae-!&*me-r-gency Pfec-eGlclfe-Br+teri&.- Escalation of the emergency classification level would be via IC AG1 or FG1. Page 255 of 296 -:;::::-Entergy Reference(s): River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases 1. EOP-1A, RPVControl, ATWS 2. EOP-4, RPV Flooding 3. EOP-2, Primary Containment Control 4. NEI 99-01 SS5 Page 256 of 296 ~Entergy River Bend Station EAL Basis Document Revision XXX I Attachment 1 -Emergency Action Level Technical Bases Category: S -System Malfunction Subcategory: 7 -Loss of Communications Initiating Condition: Loss of all onsite or offsite communications capabilities EAL: SU7.1 Unusual Event Loss of all Table S-4 onsite communication methods OR Loss of all Table S-4 State and local agency communication methods OR Loss of all Table S-4 NRC communication methods Table 5-4 Communication Methods System Onsite Plant radio system X Plant Paging System X Sound powered phones X In-plant telephones X Emergency Notification System (ENS) Commercial Telephone System Satellite Phones State of Louisiana Radio State and Local Hotline radio INFORM Notification System Mode Applicability: 1 -Power Operation, 2 -Startup, 3 -Hot Shutdown State/ Local X X X X X L __ Page 257 of 296 NRC X X X -::::::-Entergy Definition(s): None Basis: River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases 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 GRGs-;3tate and local agencies and the NRG. 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 site information via individuals or multiple radio transmission points, individuals being serit to offsite locations, etc.). The first EAL condition EAL #1 addresses a total loss of the communications methods used in support of routine plant operations. The second EAL condition EAL #2 addresses a total loss of the communications methods used to notify all State and local aaencies-GRGs of an emergency declaration. The GRG-s State and local agencies referred to here are the Louisiana Department of Environmental Quality, Governor's Office of Homeland Security and Emergencv Preparedness, Five Local Parishes Office of Homeland Security and Emergency Preparedness and 24 hour notification points, Mississippi Emergency Management Agency and the Mississippi Highway Patrol. (see Geve!Bf3e~ The third EAL condition EAL tt&-addresses a total loss of the communications methods used to notify the NRG of an emergency declaration. This EAL is the hot condition equivalent of the cold condition EAL CU5. *:. t Reference(s):

1. RBS Emergency Plan Section 13.3.6.1.5.4 Communications

2. RBS Emergency Plan Section 13.3.6.2.1 Site Communications

3. NEI 99-01 SU6 Page 258 of 296

~Entergy Riv~r Bend Station EAL Basis Document Revision XXX Attachment 1 -Emerg~ncy Action Level Technical Bases Category: S -System Malfunction Subcategory: 8 -Hazardous Event Affecting Safety Systems Initiating Condition: Hazardous event affecting SAFETY SYSTEMS needed for the current operating mode EAL: SA8.1 Alert The occurrence of any Table S-5 hazardous event AND Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode AND EITHER:

Event damage has caused indications of degraded performance to the second train of the SAFETY SYSTEM needed for the current operating mode
Event damage has resulted in VISIBLE DAMAGE to the second train of the SAFETY SYSTEM needed for the current operating mode (Notes 9, 10) Note 9: If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then emergency classification is not warranted.

Note 10: If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted. Mode Applicability: Table S-5 Hazardous Events I ll========================ll

Seismic event (earthquake)
Internal or external FLOODING event
High winds or tornado strike
FIRE
EXPLOSION
Other events with similar hazard characteristics as determined by the Shift Manager 1 -Power Operation, 2 -Startup, 3 -Hot Shutdown Page 259 of 296
=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Definition(s):

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 require a event inspection to determine if the attributes of an explosion are present. FIRE -Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed. FLOODING -A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area. SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 1 OCFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. I VISIBLE DAMAGE -Damage to a SAFETY SYSTEM train that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the a*ffected SAFETY SYSTEM train. Basis: This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the cur~ent operatinq 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 deqraded 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 SYSTFM that has indications of degraded performance; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events. Page 260 of 296 e *::==* Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Indications of degraded performance addresses damaae to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of deoraded performance should be significant enough to cause concern regarding the operability or reliabillty of the SAFETY SYSTEM train. VISIBLE DAMAGE addresses damage to a SAFETY SYSTEM train that is not in service/ooeration-and that potential iv could cause performance issues. Operators wlll make a determination of VISIBLE DAMAGE based on the totality of available event and darnaqe report information. This is intended to be a brief assessment not requiring lengthy anal:f'.sis or ouantification of the damage. This VlS!BLE DAMAGE should be significant enough to cause concern reaarding the operabilitv or reliability of the SAFETY SYSTEM train. +A-is-tG-aa<lr-esses-a-Ra-2:aFEloos-e-veR4-tl:tat-Bauses-e-awra-g-e-le-a-SAF-e-T-¥-c,~1=*-M,e-r-a s-t-1:icHIB1m-oont-a+F1+R~F-fI+¥....~S-TE M-een"'lpeHBffis,ReeaeEl for th +rns-Ge-Rdi:Hoo-si§H~i+&aRft:yLF-eB-t:l&,C.S-the-rna~ifHe-a-l-e&s-ef-iIBtem~-B-f-a-..fi-&sie A--pffifl-1.:.+&1: BaFFier,anEl--rnerefoFe-m-p

es en ts-ar-1-ast-l::h--'11-e-r-peleffii.a-l-s1.:.+8Sffi-Ati-al-E!eg-FaEl-arf e-F1-vaf....rne-~eve+-ef

~fet-y-sf-U=te-f}l-a-At-:

Ab....'.J+.1-a&'lfe&Se&-Gaffiqte-te-a-&A-F-*+Y-S-Y:S+~M4r-run-mat-i-&-i-R-ser-viBB/ep-eratl&r-1-StRGe inEl+eatte-R-s--f.ef....i.f.w~k-Be-FeaEl-i+y-a-v-a+J.ae1e-;--+!:te-inEHeatie-A-s-e+-Ge§raeeEi-f)BFfon:HaR-eEHS-R-eHkt-,se si§-A~fiBam....eR0ugh-te-eaHse-ee~~-n-reg-ar-effl-§-tRB-ef.IBFaBi-l+t-y-8f-IB1-ia-b+lity-0f-t.he-..SA~-=r¥ g..yg::r~Jl--tra+.'1-:-
A~--aEl-Elfe-&&e&-*1-a~-8AF ETY SYS+*-l\4-eernt3BReffi4!::l:at--i-s-ReHH S6f\ff6-8fef3erati-e-A-0f-reafl.i.ly-ap-,;3-meFl4!:-ifeH§-l=t-+A-fl.i.B-ati-ens-ai-e-Re, or to-a--str-t1-&rure-c-B-P.tai-n~r-:i.g SAF-E:=Pf'....SYST E M-Bemt3e-ne-n-t&--G-peratefs-wfJ.l....m..:'1-ke-tti-i-s-ee-te-ffiti.natte-R-easee-BR-1:~ify eF-aVai+ae-le---e-vent a REl-d-amage-FefJ&FE-ime-rmati0R-:--=R:i-i-s-ts-ii'-1-te-R-aea-te-be-a-brief-assess-m-e-R-1:

RBH6EfHtFin§-l-eRg-EA-y-ana!y-sIB-eF-qua-A-l:ifis-ati0n-ef-i:-he-El-amagec Escalation of the emergency classification level would be via IC FS1 or AS1. This EAL is the hot condition equivalent of cold condition EAL CA6.1. Reference(s):

1. EP FAQ 2016-00 2. NEI 99-01 SA9 Page 261 of 296

-===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases Background NEI 99-01 Revision 6 ICs AA3 and HAS prescribe declaration of an Alert based on IMPEDED access to rooms or areas (due to either area radiation levels or hazardous gas concentrations) where equipment necessary for normal plant operations, cooldown or shutdown is located. These areas are intended to be plant operating mode dependent. Specifically the Developers Notes for AA3 and HAS states: The "site-specific list of plant rooms or areas with entry-related mode applicability identified" should specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, coo/down 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 which entry would be required for each room or area. 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). Further, as specified in IC HAS: The list need not include the Control Room if adequate engineered safety/design features are in place to preclude a Control Room evacuation due to the release of a hazardous gas. Such features may include, but are not limited to, capability to draw air from multiple air intakes at different and separate locations, inner and outer atmospheric boundaries, or the capability to acquire and maintain positive pressure within the Control Room envelope. Page 262 of 296 C"\. ~Entergy River Bend Station EAL Basis Document Revisio'n XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases RBS Table A-3 and H-2 Bases A review of station operating procedures identified the following mode dependent in-plant actions and associated areas that are required for normal plant operation, cooldown or shutdown: GOP / SOP ACTIONS LOCATION MODE NOTES GOP-0005 Power Qperations Perform power maneuvering as directed by the OSM/CRS between 60 MGR 1 and 100% power using the guidance provided in the appropriate Reactivity Maneuvering Plan provided by Reactor Engineering. If possible, notify System Operator prior to changing generator load. MGR 1

Adhere to MVAR vs. MW limits .
WHEN adjusting VARs on the Main Generator, THEN use VAR-1SPGN05 (H13-P680) 9nly. Prior to entry into the Monitored Region of the Power/Flow map verify at MGR 1 least one PBDS Card is operable and begin STP-000-0001 monitoring of PBDS. (TS 3.3.1.3) Adjust pressure setpoint to minimize recirc pump "thrust reversals" as MGR 1 follows:
IF lowering power AND it is desired that pressure set be raised to minimize recirc pump "thrust reversals", THEN prior to lowering core flow to less than 70% rated core flow, raise reactor pressure.
IF raising power AND pressure set was raised to minimize recirc pump "thrust reversals", THEN when core flow is greater than 70% rated core flow, return the reactor pressure to its nominal value. Monitor Reactor Feed Pump vibratio,n and flow. IF necessary to MGR 1 minimize vibration, THEN operate the reactor feed pump Minimum Flow valves per SOP-0009, Long Cycle Clean Up valve, or adjust reactor power. '* IF a reactor feed pump is anticipated to be shut down and Hydrogen MGR/TB 67' 1 Not required : injection will be left in service, THEN install the vent jumper for that for plant pump per SOP-0009, Reactor Feedwater System. shutdown or cooldown Remove from service and/or restart Reactor Feed Pumps as necessary MGR /TB 67' 1 Not required to maintain Reactor Water Level and reactor feed pump flow for plant requirements to minimize vibration.

shutdown or cooldown IF controlling Reactor Power with Reactor Recirculation Flow, THEN MGR 1 refer to SOP-0003. IF power is lowered below 75% AND a Reactor Feed Pump has been MGR 1 secured, THEN, BEFORE power ascension beyond 75% RTP and Page 263 of 296 Q -::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP/ SOP ACTIONS LOCATION MODE NOTES AFTER start of the 3rd feedwater pump verify water chemistry acceptable. IF power is lowered below 75%, THEN when thermal power is above, MCR 1 75% RTP, verify both LEFM Check Plus are Operable. IF the LEFMs are functional per Technical Requirement 3.3.13 AND an MCR 1 LEFM alert is indicated on the ONE heat balance computer screen, THEN reduce reactor power to 3081 MWth or 99.6% rated thermal power in one hour to ensure thermal power limits are not exceeded. Observe the operation MSS-HWCV4 and DTM-AOVSPDV4 for MCR 1 OSP-0102, Turbine Valve Testing. As power is raised, check MSS-HWCV4 open and DTM-l

AOVSPDV4 closed.
As power is lowered, check MSS-HWCV4 closed and DTM-AOVSPDV4 open. When power is raised above 90%, Pressure Set may need to be MCR 1 adjusted as necessary to ensure that the 1st admission main turbine control valves, MSS-HWCV1, 2, and 3 are full open. Monitor turbine vibration bearing temperature and differential expansion MCR 1 per the following:

Turbine Temg & Exgansion RCDR (TMI-NXR102)

Differential Expansion Rotor Long (point 11) between 0.31 inches and 0.69 inches. (Refer To ARP-870-54, GOB, H08)
Rotor Expansion Rotor Long (point 12) between 0.455 inches and 1.545 inches. Turbine Vibration RCDR (TMI-NXR103)
Vibration (points 1 through 10) between O mils and 6 mils . (Refer To ARP-870-54, D08) Tamaris Computer (Display 69, 70)
Bearing oil temperatures ( <setpoint, 180°F) .
Bearing metal temperatures

(<setpoint, 218.?°F) . IF unusual indications are observed, THEN initiate hold in power change until those indications return to normal. WHEN maneuvering power, THEN adhere to the POWER/FLOW maps MCR 1 (avoid the restricted region) in AOP-0024, Thermal Hydraulic Stability Controls and Turbine-Generator loading rate per SOP-0080, Turbine Generator Operation. WHEN in two Recirculation Pump Operation at greater than or equal to MCR 1 70% rated core flow, THEN maintain recirculation flow mismatch less than 5%. Observe the following limitations and precautions: MCR 1

Do not exceed the Turbine Generator normal operating limits. Page 264 of 296 Q -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation

& Shutdown Areas Tables A-3 & H-2 Bases GOP / SOP ACTIONS LOCATION MODE NOTES

Adjust the pressure setpoint to an indicated reactor pressure of between 1035 and 1055 psig for 100% steady state conditions.
IF reactor power exceeds 3091 MWth, THEN take actions using recirculation flow and/or control rod insertion to lower power below 3091 MWth.
IF the core thermal power average for a 2 hour period exceeds the Licensed Power Limit, THEN take timely action to ensure that thermal power is less than or equal to the Licensed Power Limit.
IF reactor thermal power indication becomes unavailable for less than 15 minutes AND steady state operation is expected, THEN note current APRM readings AND verify thermal power does not exceed the noted value.
IF reactor thermal power indication will be unavailable for more than 15 minutes, THEN perform the following:

0 Lower reactor power as indicated on the APRMs such that indicated thermal power does not exceed 100%. (The top of the normal noise band on the chart -recorders should not be above 100%). 0 Reactor Engineering should be contacted for assistance in determining a manual heat balance per REP-0030, Reactor Heat Balance. 0 WHEN performing a manual heat balance AND it is determined that the LEFM signal is not operable, THEN lower reactor power so that the APRMs read less than 98.3% at the top of the normal noise band.

Observe the following restrictions when operating near or above rated core flow as Bi-Stable flow conditions are possible:

-, 0 IF step changes of 60 MWth (2%) or greater are seen in instantaneous CTP, THEN reduce Reactor power using Recirc flow until the step changes in instantaneous power are no longer observed. 0 IF step changes of up to 1.69 MLB/hr (2%) are seen in total core flow, THEN reduce Reactor power using Recirc flow until the step changes in instantaneous power are no longer observed. 0 Notify Reactor Engineering of any power/flow reductions required.

IF any thermal limit exceeds 0.980, THEN notify Reactor Engineering to increase the frequency of monitoring (at least hourly) until a steady state condition is reached or thermal \ limits indicate less than 0.980.
IF any thermal limit exceeds 0.990, THEN notify Reactor Engineering to perform one of the following:

0 Provide instructions for reducing the thermal limit to less than 0.990. 0 Provide a justification for operating with thermal limits ! Page 265 of 296 Q -~ Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP / SOP ACTIONS LOCATION MODE NOTES greater than or equal to 0.990. IF the power change exceeded 15%, THEN perform the following: MCR/TB 95' 1 Not required

Notify Chemistry of the power change to obtain a new for plant Condensate System Oxygen injection flow rate. shutdown or cooldown
Per Chemistry recommendations, adjust the Oxygen flow rate per SOP-0123, Hydrogen Water Chemistry H2 and 02 I System.
IF power ramp rates exceed 15%/hr, THEN notify Chemistry per Technical Requirement.

IF power was lowered below 80%, THEN notify chemistry management MCR 1 when reactor power has been returned to 100%. As power is lowered, at approximately 50% power, transfer Steam MCR 1 Seal Evaporator from Extraction Steam to Main Steam per SOP-0015, Gland Seal Steam System and Exhaust System, if it has not occurred automatically. Transferring Steam Seal Evaporator from Extraction Steam to I . ::r!\\ ':"r"i.( :ti):, /:. Main Steam (SOP-0015). / As power is lowered, at approximately 50% power, if the Steam Seal MCR 1 Evaporator has not already transferred automatically from Extraction Steam to Main Steam, then throttle closed ESS-MOV112, STEAM SEAL EVAPORATOR using the control switch and the STOP pushbutton. IF the pressure controller is operating in automatic AND MCR 1 TME-MOVESFV2 is closed, THEN verify the following:

TME-PIEPR-35, SSE TUBE SIDE PRESSURE indicates less than or equal to 75 psig.
TME-PIEPR-36, SSE SHELL SIDE PRESSURE is stable and indicates less than or equal to 45 psig. WHEN ESS-MOV112, STEAM SEAL EVAPORATOR is full closed, MCR 1 THEN verify DTM-AOV118, EXTR STM TO SSE & RW RBLR opens. As power is raised, at approximately 65-75% power, after checking MCR 1 Annunciator P870-52-E03, 3rd PT EXTR ST AND MAIN STEAM DIFF PRESS LOW is clear, transfer Steam Seal Evaporator from Main Steam to Extraction Steam per SOP-0015, Gland Seal Steam System and Exhaust System. GOP-0002 Power Decrease/Plant Shutdown .* Notify System Operator prior to decreasing generator load. MCR 1 IF the Reference Leg Backfill System is not in service per SOP-0001, MCR 1 Nuclear Boiler Instrumentation (SYS #051 ), THEN have l&C stage equipment, acquire necessary technicians and obtain PMs to backfill Page 266 of 296 Q -::::::* Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation

& Shutdown Areas Tables A-3 & H-2 Bases GOP / SOP ACTIONS LOCATION MODE NOTES -* reactor water level reference legs. (Approximately 12 hours may be needed to prepare for backfilling.) Actual backfilling performance may commence when Operations Shift Manager authorizes. (It is desired to have backfilling completed prior to reactor pressure reaching 450 psig to counter level indication notching possibilities.) Monitor turbine vibration bearing temperature and differential expansion MCR 1 per the following: Turbine Temg & Exgansion RCDR (TMI-NXR102)

Differential Expansion Rotor Long (point 11) between 0.31 inches and 0.69 inches. (Refer To ARP-870-54, GOB, HOB)
Rotor Expansion Rotor Long (point 12) between 0.455 inches and 1.545 inches. Turbine Vibration RCDR (TMI-NXR103)
Vibration (points 1 through 10) between O mils and 6 mils . (Refer To ARP-870-54, DOB) Tamaris Comguter (Disglay 69, 70)
Bearing oil temperatures

(<setpoint, 180°F) .

Bearing metal temperatures

(<setpoint, 218.7°F) . IF unusual indications are observed, THEN initiate hold in power ( change until those indications return to normal. Lower reactor power per the Shutdown/ Emergency Power Reduction MCR 1 reactivity control plan. Contact the on-duty Reactor Engineer. Adjust pressure setpoint to minimize recirc pump "thrust reversals" as MCR 1 follows:

IF lowering power AND it is desired that pressure set be raised to minimize recirc pump "thrust reversals", THEN prior to lowering core flow to less than 70% rated core flow, raise reactor pressure.

IF raising power AND pressure set was raised to minimize recirc pump "thrust reversals", THEN when core flow is greater than 70% rated core flow, return the reactor pressure to its nominal value. At approximately 90% to 80% power observe the following: MCR 1

MSS-HVYCV4 closes
DTM-AOVSPDV4 opens , IF MSRs are to be manually shutdown, THEN at approximately 90% MCR/TB 1 Not required power, start removing the MS Rs from service per SOP-0010, MSR & 123' for plant FW Heaters Extraction Steam and Drains. Remove the MS Rs at a rate shutdown or so as to be completely off line by 760 MWe. Limit rate of change of LP cooldown Turbine inlet steam temperature to 125°F per hour. Monitor Points 6, 7, 8, 9 on TMI-NXR102.

Maximum allowable temperature difference between LP Turbine inlets is 50°F. MS Rs should be gradually valved Page 267 of 296 w *::::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP / SOP ACTIONS LOCATION MODE NOTES out in parallel at essentially the same temperature. IF MSRs are to remain in service with power maintained between 15% and 70%, THEN operate MSRs in accordance with SOP-0010. For power reductions of greater than 15%, notify Chemistry to MCR /TB 95' 1 Not required determine whether the Condensate System oxygen injection is to be for plant secured or flow reduced AND implement the recommendations per shutdown or SOP-0123, Hydrogen Water Chemistry H2 and 02 System. cooldown IF power ramp rates exceed 15%/hr, THEN notify Chemistry per Technical Requirement 3.11.2.1. IF a reactor feed pump is anticipated to be shut down AND Hydrogen TB 67' 1 Not required Injection will be left in service AND a plant shutdown is NOT in for plant progress, THEN install the ventjumper(s) for the pump(s) per SOP-shutdown or 0009, Reactor Feedwater System. cooldown At approximately 70% power, (or with Engineering recommendations) MCR 1 stop one reactor feedwater pump (leave two running) per SOP-0009, Reactor Feedwater System. ,, ' Reactor Feed Pump Shutdown (SOP-0009) '>".. IF securing a Reactor Feed Pump for downpower, THEN monitor the MCR 1 following parameters:

Reactor power should be limited to 85% with only two Reactor Feed Pumps in service.
Normal Feedwater Pump Motor current should be greater than 200 amps and limited to 311 amps. Refer to Precautions and Limitations 2.9 and 2.15.
FWREG position should be limited to less than or equal to 92% open to allow an adequate margin for valve modulation while maintaining reactor level.
Feed pump suction pressure should be maintained above low pressure alarm point of 280 psig. IF NOT already performed to reduce Reactor Feed Pump vibration MCR 1 levels, THEN perform the following for the Reactor Feed Pump being shutdown:
At H13-P680, place CNM-H/A68A(B)(C), RX FWP 1A(B)(C) MIN FLOW FLOW CONTROLLER to MANUAL for the Reactor Feed Pump to be secured.
Open slowly FWR-FV2A(B)(C), RX FWP 1A(B)(C) MIN FLOW VALVE using CNM-H/A68A(B)(C), RX FWP 1A(B)(C) MIN FLOW FLOW CONTROLLER while monitoring Reactor Water Level. IF desired to raise Reactor Water Level, THEN at H13-P680 adjust MCR 1 C33-R600, FW REG VALVES MASTER FLOW CONTROLLER tape set to desired Reactor Water Level within normal level control band. Page 268 of 296

~=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP / SOP ACTIONS LOCATION MODE NOTES IF the HWC System is in service AND the reactor feed pump is not MCR /TB 95' 1 Not required being immediately shut down, THEN at P73-P500, place P73-AOV-for plant F111A(B)(C), HYDROGEN ISOLATION TO FEEDWATER PUMP shutdown or A(B)(C) in CLOSE. cooldown IF the capability of meeting feed flow requirements with the remaining MCR 1 Feedwater Pumps is uncertain, THEN make a determination as follows:

Close FWS-MOV26A(B)(C), RX FWP 1A(B)(C) DISCH VLVfor the pump being shutdown.
Verify the minimum flow valve for the pump being secured is open.
Monitor Feed Flow/Steam Flow mismatch and RPV Level to ensure remaining pump(s) can maintain level.
IF the remaining pump(s) cannot maintain RPV Level, THEN reopen the discharge valve FWS-MOV26A(B)(C), RX FWP 1A(B)(C) DISCH VLV and discontinue this procedure.

IF the last Feedwater Pump is being removed from service, THEN open MCR 1 FWS-MOV109, FEED PUMP BYPASS. Stop FWS-P1A(B)(C), RX FWP P1A(B)(C). MCR 1 Verify CNM-H/A68A(B)(C), RX FWP 1A(B)(C) MIN FLOW FLOW MCR 1 CONTROLLER is in AUTO for the Reactor Feed Pump that was secured. IF Reactor Water Level was intentionally raised in Step 6.1.3, THEN MCR 1 adjust Reactor Water Level to desired level within normal level control band using C33-R600, FW REG VALVES MASTER FLOW CONTROLLER tape set. IF FWS-P1A(B)(C) is to remain in hot standby, THEN maintain seal TB 67' 1 Not required temperatures as follows: for plant Maintain seal water temperature dT less than or equal to 50F AND seal shutdown or water outlet temperature less than or equal to 300F as follows: cooldown

FWS-P1A 0 Throttle CCS-V5003A, RFP FWL-P1A SEAL WATER HX-E4A CCS INLET VALVE, as required.

0 Throttle CCS-V5004A, RFP FWL-P1A SEAL WATER HX-E4B CCS INLET VALVE, as required.

FWS-P1B 0 Throttle CCS-V5003B, RFP FWL-P1B SEAL WATER HX-E4C CCS INLET VALVE, as required.

0 Throttle CCS-V5004B, RFP FWL-P1B SEAL WATER HX-E4D CCS INLET VALVE, as required.

FWS-P1C 0 Throttle CCS-V5003C, RFP FWL-P1C SEAL WATER HX-E4E CCS INLET VALVE, as required.

0 Throttle CCS-V5004C, RFP FWL-P1C SEAL WATER Page 269 of 296 River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP / SOP ACTIONS LOCATION MODE NOTES HX-E4F CCS INLET VALVE, as required. On H13-P870, verify FWL-P5A(B)(C), GEAR !NCR AUX OIL PMP MCR 1 5A(B)(C) auto starts. Verify min flow valve closes 1 -3 minutes after pump shutdown. MCR 1 Verify FWS-MOV26A(B)(C), RX FWP P1A (B)(C) DISCH VLV is MCR 1 closed. On H13-P870, WHEN the 23 minute time delay allowing for pump coast MCR 1 down has passed, THEN verify the following:

FWL-P1A(B)(C), RX FWP A(B)(C) MN OIL PMP 1A(B)(C) auto stops.
FWL-P5A(B)(C), RX FWP A(B)(C) GEAR INC AUX OIL PMP 5A(B)(C) auto stops.
FWL-P2A(B)(C), RX FWP A(B)(C) AUX OIL PMP 2A(B)(C) auto starts on low oil pressure.
IF FWL-P2A(B)(C), RX FWP A(B)(C) AUX OIL PMP 2A(B)(C) does not maintain pressure greater than 4 psi, THEN FWL-P3A(B)(C), RX FWP A(B)(C) AUX OIL PMP 3A(B)(C) auto starts on low oil pressure.

On H13-P870, place FWL-P2A(B)(C), RX FWP A(B)(C) AUX OIL PMP MCR 1 2A(B)(C) control switch in STOP, and verify FWL-P3A(B)(C), RX FWP A(B)(C) AUX OIL PMP 3A(B)(C) auto starts on low oil pressure. On H13-P870, place FWL-P2A(B)(C), RX FWP A(B)(C) AUX OIL PMP MCR 1 2A(B)(C) control switch in AUTO. On H13-P870, place FWL-P3A(B)(C), RX FWP A(B)(C) AUX OIL PMP MCR 1 3A(B)(C) control switch in STOP. On H13-P870, place FWL-P3A(B)(C), RX FWP A(B)(C) AUX OIL PMP MCR 1 3A(B)(C) control switch in AUTO. Locally verify breaker relay trip flags are reset for Reactor Feed Pump NSW 98' 1 Not required stopped in Step 6.1.7. for plant shutdown or cooldown At approximately 70% power, (or with Engineering recommendations) MCR 1 stop one condensate pump (leave two running) per SOP-0007, Condensate System. Shutdown of CNM-P1A(B)(C) CONDENSATE PUMPS (SOP-0007) ltk~ .f\* )t ":'t : ,~,; 1;fifY"~t : Request Aux Control Room remove unnecessary Condensate Filters ACR 1 Not required from service per SOP-0124, Condensate Filtration System. for plant shutdown or cooldown Request Aux Control Room remove unnecessary Condensate Demins ACR 1 Not required from service per SOP-0093, Condensate Demineralizer System. for plant shutdown or Page 270 of 296 w ~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP / SOP ACTIONS LOCATION MODE NOTES cooldown IF shutting down CNM-P1C, CNDS PUMP 1C, THEN secure Oxygen MCR /TB 95' 1 Not required injection per SOP-0123, Hydrogen Water Chemistry H2 and 02 for plant System. shutdown or cooldown IF securing the Condensate System, THEN perform the following: TB 123' 1 Not required

Close all CNM-V3105 A, B, C, D, and E, CNM-FLT1 A, B, C, for plant D, and E BACKWASH AIR SUPPLY valves. shutdown or cooldown
IF desired to isolate and depressurize CNM-TK100, AIR RECEIVING TK, THEN perform the following:

0 CLOSE CNM-V3110, SVCE AIR !SOL VLV INLET SERV. AIR !SOL VLV. 0 Uncap and install hose on CNM-V3112, CNM-TK100 DRAIN ISOLATION VALVE. 0 Open CNM-V3112. Depress the CLOSE pushbutton for CNM-MOV3A(B)(C), CNDS PUMP MCR 1 1A(1B)(1 C) DISCH. WHEN pump motor current lowers below 100 amps, THEN stop MCR 1 CNM-P1A(B)(C), CNDS PUMP 1A(1B)(1C). WHEN CNM-MOV3A(B)(C), CNDS PUMP 1A(1 B)(1C) DISCH is full MCR 1 closed, THEN depress the STOP pushbutton. Verify associated CCS-MOV67A(B)(C), CNDS PMP 1A(1 B)(1C) MOT MCR 1 CLR close for pump stopped. Verify associated CCS-MOV68A(B)(C), CNDS PMP 1A(18)(1 C) BRG MCR 1 CLR close for pump stopped. Locally verify breaker relay trip flags are reset for Condensate Pump NSW 98' 1 Not required stopped in Step 6.1.6. for plant shutdown or cooldown WHEN the Steam Jet Air Ejectors (SJAEs) and Gland Seal and MCR 1 Exhaust System are removed from service, THEN adjust CNM-H/A114 to 10% or to a setpoint determined by the CRS/OSM. As power is reduced, remove FW Reg Valves from service per SOP-MCR 1 0009, Reactor Feedwater System. Removing a FWREG Valve from Service (SOP-0009) ,, *J'i,*, .~,~)": ... */i" Check feedwater flow is within the capability of the remaining FWREGs. MCR 1 ' Station an operator locally at the FWREG Valve to be removed from TB 67' 1 Not required service. for plant shutdown or cooldown Page 271 of 296 -::::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP / SOP ACTIONS LOCATION MODE NOTES Establish communications between the local operator and the Main MCR /TB 67' 1 Not required Control Room (MCR). for plant shutdown or cooldown Place C33-R601A(R613)(R601B), FWREG VALVE A(B)(C) FLOW MCR 1 CONTROLLER in MANUAL. Throttle closed to 10% open the C33-LVF001A(B)(C), FWREG VALVE MCR 1 A(B)(C) to be removed from service while observing that reactor level is being maintained by the remaining FWREGs. IF level is not being maintained by the remaining FWREGs, THEN MCR 1 place the FWREG that was being removed from service back in service as follows:

Open C33-LVF001A(B)(C), FWREG VALVE A(B)(C) to the same position as the in service FWREGs.
Place C33-R601A(R613)(R601B), FWREG VALVE A(B)(C) FLOW CONTROLLER in AUTO. WHEN the FWREG is at 10% open, THEN close the following isolation MCR 1 valve for the FWREG valve that is being removed from service.
For C33-LVF001A close FWS-MOV27A, FWREG VLV 1A INLT Valve.
For C33-LVF001B close FWS-MOV27B, FWREG VLV 1B INLT Valve.
For C33-LVF001C close FWS-MOV27C, FWREG VLV 1C INLT Valve. Fully close the C33-LVF001A(B)(C), FWREG VALVE A(B)(C) that was MCR 1 removed from service. Record the temperature of the feedwater at the reactor feed pumps. TB 67' 1 Not required for plant shutdown or cooldown IF FWS-MOV27A, B, orC, FWREG VLV 1A(1B)(1C)

INLTwere closed MCR 1 with feedwater temperature at the reactor feed pumps greater than 200F, THEN refer to Section 5. 7 for further stroking requirements. WHEN the FWREG Valve is at 0% open, THEN record demanded MCR/TB 67' 1 Not required position in the MCR, position indication in the MCR, and local position for plant indication. shutdown or cooldown ?;~} ; '*~ Within one hour after reactor power is less than or equal to the high MCR 1 STP-500-power setpoint, demonstrate RWL operability by performing STP-500-0704, Rod 0704, Rod Withdrawal Limiter Functional Test (SR 3.3.2.1.2), if not Withdrawal performed within the previous 92 days. Limiter Functional Test is Page 272 of 296 Q ~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases / GOP/ SOP ACTIONS LOCATION Prior to entering the Monitored and/or the Restricted Regions of the MCR Power to Flow map, verify the following indications on the PBDS, Period Based Detection System cards in APRM 'A' and 'B' cabinets:

NORMAL/BYPASS Toggle switch in the NORMAL position.
INOP STATUS LED indication is GREEN. (Depress the INOP STATUS Reset Pushbutton to reset a Red LED inop indication.)
Verify at least one PBDS Card is OPERABLE.
Begin STP-000-0001 monitoring of PBDS. Prior to entry into the Restricted Region of the Power to Flow Map, MCR perform the following:
Verify FCBB is less than or equal to 1.0.(SR 3.2.4.1)
Place the APRM -FCTR, Flow Control Trip Reference cards to the setup trip setpoints by depressing the Normal/Setup pushbutton and verifying the normal/setup LED indication is yellow. At Approximately 50% power, transfer Steam Seal Evaporator from MCR Extraction Steam to Main Steam per SOP-0015, Gland Seal System And Exhaust System, if it has not occurred automatically.

Transferring Steam Seal Evaporator from Extraction Steam to Main Steam (SOP-0015) As power is lowered, at approximately 50% power, if the Steam Seal Evaporator has not already transferred automatically from Extraction Steam to Main Steam, then throttle closed ESS-MOV112, STEAM SEAL EVAPORATOR using the control switch and the STOP pushbutton. IF the pressure controller is operating in automatic AND TME-MOVESFV2 is closed, THEN verify the following:

TME-PIEPR-35, SSE TUBE SIDE PRESSURE indicates less than or equal to 75 psig.
TME-PIEPR-36, SSE SHELL SIDE PRESSURE is stable and indicates less than or equal to 45 psig. WHEN ESS-MOV112, STEAM SEAL EVAPORATOR is full closed, THEN verify DTM-AOV118, EXTR STM TO SSE & RW RBLR opens. At approximately 50% power, shutdown all heater drain pumps per SOP-0010, MSR & FW Heaters Extraction Steam and Drains. Removing the Heater Drain Pumps From Service (SOP-001 O) Page 273 of 296 MCR MCR MCR MCR MODE NOTES 1 1 1 1 1 1 performed only in the MCR

-===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP/ SOP ACTIONS LOCATION MODE NOTES IF it is desired to secure HDL-P1A(B), HTR DR PUMP 1A(B) for Heater MCR 1 String A, THEN at H13-P680, perform the following:

Depress the Close Pushbutton for HDL-MOV55A(B), HTR DR PMP 1A(B) DISCH.
Stop HDL-P1A(B), HTR DR PUMP 1A(B) for Heater String A.
Verify HDL-MOV55A(B), HTR DR PMP 1A(B) DISCH is closed . IF it is desired to secure HDL-P1 C(D), HTR DR PUMP 1 C(D) for Heater MCR 1 String B, THEN at H13-P680, perform the following:
Depress the Close Pushbutton HDL-MOV55C(D), HTR DR PMP 1 C(D) DISCH.
Stop HDL-P1C(D), HTR DR PUMP 1C(D) for Heater String B .
Verify HDL-MOV55C(D), HTR DR PMP 1 C(D) DISCH is closed. ** : ! : ~ : .. * .. At approximately 50% power, perform the following per SOP-0006, MCR 1 Circulating Water, Cooling Tower and Vacuum Priming:
Shut down at least 1 circulating water pump .
Adjust the number of operating cooling tower fans to maintain vacuum and circulating water temperature.

WHEN the recirculation flow control valves are at their minimum MCR 1 position, THEN continue reducing power by inserting control rods in their proper sequence. At about 40% power, transfer both reactor recirculation pumps to MCR 1 SLOW speed per SOP-0003, Reactor Recirculation System. ( .~ .'. .; t; :t ** . !I . .. Transferring from Fast Speed to Slow Speed (SOP-0003)

. . '* . :. *,: ... .. . . Simultaneously depress B33-C001A and B RECIRC PUMP A and B MCR 1 MOTOR BREAKER 5A and 5B XFER TO LFMG pushbuttons.

Observe the following: MCR 1

Both B33-S001A LFMG MOT BRKR 1A and B33-S001B LFMG MOTBRKR 1 B close.
Both B33-C001A RECIRC PUMP A MOTOR BREAKER 5A and B33-C001 B RECIRC PUMP B MOTOR BREAKER 5B open.
WHEN B33-C001A and B, RECIRC PUMP A and B coast down to approximately 360 -470 RPM, THEN B33-S001A and B LFMG A and B GEN BRKR 2A and 2B close and pump speeds stabilize near 450 RPM.
Both B33-K603 A and B, RECIRC LOOP A and B FLOW CONTROL MAN/AUTO stations transfer to MAN ..
Open B33-HVY-F060A(B) to approximately 94% valve position using B33-K603A(B).

Reduce to one reactor feed pump per SOP-0009, Reactor MCR 1 Feedwater System. Page 27 4 of 296 -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP / SOP ACTIONS LOCATION MODE NOTES Reactor Feed Pump Shutdown (SOP-0009)

'**' '.* ' ' ,; 1* IF securing a Reactor Feed Pump for downpower, THEN monitor the MCR 1 following parameters:

Reactor power should be limited to 85% with only two Reactor Feed Pumps in service.
Normal Feedwater Pump Motor current should be greater than 200 amps and limited to 311 amps. Refer To Precautions and Limitations 2.9 and 2.15.
FWREG position should be limited to less than or equal to 92% open to allow an adequate margin for valve modulation while maintaining reactor level.
Feed pump suction pressure should be maintained above low pressure alarm point of 280 psig. IF NOT already performed to reduce Reactor Feed Pump vibration MCR 1 levels, THEN perform the following for the Reactor Feed Pump being shutdown:
At H13-P680, place CNM-H/A68A(B)(C), RX FWP 1A(B)(C) MIN FLOW FLOW CONTROLLER to MANUAL for the Reactor Feed Pump to be secured.
Open slowly FWR-FV2A(B)(C), RX FWP 1A(B)(C) MIN FLOW VALVE using CNM-H/A68A(B)(C), RX FWP 1A(B)(C) MIN FLOW FLOW CONTROLLER while monitoring Reactor Water Level. IF desired to raise Reactor Water Level, THEN at H13-P680 adjust MCR 1 C33-R600, FW REG VALVES MASTER FLOW CONTROLLER tape set to desired Reactor Water Level within normal level control band. IF the HWC System is in service AND the reactor feed pump is not MCR /TB 95' 1 Not required being immediately shut down, THEN at P73-P500, place P73-AOV-for plant F111A(B)(C), HYDROGEN ISOLATION TO FEEDWATER PUMP shutdown or A(B)(C) in CLOSE. cooldown IF the capability of meeting feed flow requirements with the remaining MCR 1 Feedwater Pumps is uncertain, THEN make a determination as follows:
Close FWS-MOV26A(B)(C), RX FWP 1A(B)(C) DISCH VLVfor the pump being shutdown.
Verify the minimum flow valve for the pump being secured is open.
Monitor Feed Flow/Steam Flow mismatch and RPV Level to ensure remaining pump(s) can maintain level.
IF the remaining pump(s) cannot maintain RPV Level, THEN reopen the discharge valve FWS-MOV26A(B)(C), RX FWP 1A(B)(C) DISCH VLVand discontinue this procedure.

IF the last Feedwater Pump is being removed from service, THEN open MCR 1 FWS-MOV109, FEED PUMP BYPASS. Stop FWS-P1A(B)(C), RX FWP P1A(B)(C). MCR 1 Page 275 of 296 -::::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP/ SOP ACTIONS LOCATION MODE NOTES Verify CNM-H/A68A(B)(C), RX FWP 1A(B)(C) MIN FLOW FLOW MCR 3 CONTROLLER is in AUTO for the Reactor Feed Pump that was secured. IF Reactor Water Level was intentionally raised in Step 6.1.3, THEN MCR 3 adjust Reactor Water Level to desired level within normal level control band using C33-R600, FW REG VALVES MASTER FLOW CONTROLLER tape set. IF FWS-P1A(B)(C) is to remain in hot standby, THEN maintain seal TB 67' 1 Not required temperatures as follows: for plant Maintain seal water temperature dT less than or equal to 50F AND seal shutdown or water outlet temperature less than or equal to 300F as follows: cooldown

FWS-P1A 0 Throttle CCS-V5003A, RFP FWL-P1A SEAL WATER HX-E4A CCS INLET VALVE, as required.

,, 0 Throttle CCS-V5004A, RFP FWL-P1A SEAL WATER HX-E4B CCS INLET VALVE, as required.

FWS-P1B 0 Throttle CCS-V5003B, RFP FWL-P1B SEAL WATER HX-E4C CCS INLET VALVE, as required.

0 Throttle CCS-V5004B, RFP FWL-P1B SEAL WATER HX-E4D CCS INLET \,'.'ALVE, as required.

FWS-P1C 0 Throttle CCS-V5003C, RFP FWL-P1C SEAL WATER HX-E4E CCS INLET VALVE, as required.

0 Throttle CCS-V5004C, RFP FWL-P1 C SEAL WATER HX-E4F CCS INLET VALVE, as required. On H13-P870, verify FWL-P5A(B)(C), GEAR !NCR AUX OIL PMP MCR 1 5A(B)(C) auto starts. Verify min flow valve closes 1 -3 minutes after pump shutdown. MCR 1 Verify FWS-MOV26A(B)(C), RX FWP P1A (B)(C) DISCH VLV is MCR 1 closed. On H13-P870, WHEN the 23 minute time delay allowing for pump coast MCR 1 down has passed, THEN verify the following:

FWL-P1A(B)(C), RX FWP A(B)(C) MN OIL PMP 1A(B)(C) auto stops.
FWL-P5A(B)(C), RX FWP A(B)(C) GEAR INC AUX OIL PMP 5A(B)(C) auto stops.
FWL-P2A(B)(C), RX FWP A(B)(C) AUX OIL PMP 2A(B)(C) auto starts on low oil pressure.
IF FWL-P2A(B)(C), RX FWP A(B)(C) AUX OIL PMP 2A(B)(C) does not maintain pressure greater than 4 psi, THEN FWL-P3A(B)(C), RX FWP A(B)(C)AUX OIL PMP 3A(B)(C) auto starts on low oil pressure.

On H13-P870, place FWL-P2A(B)(C), RX FWP A(B)(C) AUX OIL PMP. MCR 1 2A(B)(C) control switch in STOP, and verify FWL-P3A(B)(C), RX FWP Page 276 of 296 a -:::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP/ SOP ACTIONS LOCATION MODE NOTES A(B)(C) AUX OIL PMP 3A(B)(C) auto starts on low oil pressure. On H13-P870, place FWL-P2A(B)(C), RX FWP A(B)(C) AUX OIL PMP MCR 1 2A(B)(C) control switch in AUTO. On H13-P870, place FWL-P3A(B)(C), RX FWP A(B)(C) AUX OIL PMP MCR 1 3A(B)(C) control switch in STOP. On H13-P870, place FWL-P3A(B)(C), RX FWP A(B)(C) AUX OIL PMP MCR 1 3A(B)(C) control switch in AUTO. Locally verify breaker relay trip flags are reset for Reactor Feed Pump NSW 98' 1 Not required stopped in Step 6.1.7. for plant shutdown or cooldown ' *~* ;f. rA( . Decrease reactor power at the rate consistent with generator loading MCR 1 criteria (Attachment 1, MVAR VS MW LIMITS and E!OP-0080, Turbine Generator Operation) using control rod insertion per applicable sequence.

Stop inserting control rods at the low power alarm point (as observed in RC & IS panel) and obtain instruction from the Operations Shift Manager regarding further power reductions/shutdown or continued operation at LPAP.
Decrease reactor power to the LPSP using control rod insertion per applicable sequence.

At 300 MWe load, open the following steam drain valves: MCR 1

DTM-AOV32A, 4TH PT HTR EXTR LINE DR
DTM-AOV32B, 4TH PT HTR EXTR LINE DR
DTM-AOV35A, 3RD PT HTR EXTR LINE DR
DTM-AOV35B, 3RD PT HTR EXTR LINE DR Open or verify open G33-MOVF101, RWGU BOTTOM HEAD DRAIN. MCR 1
Verify drain temperature remains stable using Point #4 on B21 R643 or ERIS computer point B33NA002.

Manually stroke C33-LVF002, STARTUP FWREG VALVE through full MCR 1 travel to verify smooth operation per SOP-0009, Reactor Feedwater System. Manual Stroking of Start Up FWREG (SOP-0009) !tt. \Jlt</.t : .. { ,< )lt: ' Close FWS-MOV105, S/U FW REG VLV ISOL. MCR 1 Station an operator locally to monitor valve position. TB 67' 1 Not required for plant shutdown or cooldown Page 277 of 296 -=::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP/ SOP ACTIONS LOCATION MODE NOTES Establish communications between the local operator and the Main MCR /TB 67' 1 Not required Control Room (MCR). for plant shutdown or cooldown WHEN the FWREG Valve is at 0% open, THEN record demanded MCR /TB 67' 1 Not required position in the MCR, position indication in the MCR, and local position for plant indication in Attachment 7, Calibration Check of FWREG Valves. shutdown or cooldown Use the OPEN and CLOSE Pushbuttons on C33-R602, START UP MCR 1 FWREG VALVE FLOW CONTROLLER to stroke open and then closed . the Start Up FWREG. Check proper valve movement and smooth operation. MCR /TB 67' 1 Not required for plant shutdown or cooldown Check C33-LVF002, START UP FWREG VLVfull closed. MCR 1 Open FWS-MOV105, S/U FW REG VLV !SOL. MCR 1 ' ** *'.li1:r:l ,f> !l . i~{i \ WHEN less than 30% power AND at the direction of the responsible MCR 1 Operations Management, THEN perform the following:

Transfer station loads to preferred source per SOP-0045, 13.8 KV System and SOP-0046, 4.16 KV System.
Verify MVARs are between + 50 and -50 .
At the SRM cabinets, place the Mode Selector Switches to the OPERATE position.
Prior to initiating a Rx Scram, verify the SRM & IRM Channel Functional Tests are current. IF Channel Functional Tests are not current, THEN refer to Tech Spec 3.3.1.1, 3.3.1.2 and TRM TR 3.3.2.1. Secure SPC per SOP-0140, Suppression Pool Cleanup and Alternate MCR 1 Not Decay Heat Removal. required, but system will automaticall y isolate on a level 3 from a RX SCRAM Contact the Auxiliary Control Room to verify that sufficient condensate MCR/ACR 1/3 Not required demineralizers are in service to prevent physical damage to the for plant demineralizers from high feedwater flow transients.

shutdown or cooldown Page 278 of 296 Q -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP/ SOP ACTIONS LOCATION MODE NOTES Cont.act the Auxiliary Control Ro.om to verify that sufficient condensate MCR/ACR 1/3 Not required filtration filters are in service to prevent physical damage to the filters for plant from high feedwater flow transients shutdown or cooldown Reduce the number of FWREG Valves in service to one per SOP-MCR 1 Not required 0009, Reactor Feedwater System. for plant shutdown or cooldown Removing a FWREG Valve from Service (SOP-0009) . ,~; ' ,,1(-)trt }-~Ir/ ij\Jf :: .** Check feedwater flow is within the capability of the remaining FWREGs. MCR 1 Station an operator locally at the FWREG Valve to be removed from TB 67' 1 Not required service. for plant shutdown or cooldown Establish communications between the local operator and the Main MCR /TB 67' 1 Not required Control Room (MCR). for plant shutdown or cooldown Place C33-R601A(R613)(R601B), FWREG VALVEA(B)(C) FLOW MCR 1 CONTROLLER in MANUAL. Throttle closed to 10% open the C33-LVF001A(B)(C), FWREG VALVE MCR 1 A(B)(C) to be removed from service while observing that reactor level is being maintained by the remaining FWREGs. IF level is not being maintained by the remaining FWREGs, THEN MCR 1 place the FWREG that was being removed from service back in service as follows:

Open C33-LVF001A(B)(C), FWREG VALVE A(B)(C) to the same position as the in service FWREGs.
Place C33-R601A(R613)(R601 B), FWREG VALVE A(B)(C) FLOW CONTROLLER in AUTO. WHEN the FWREG is at 10% open, THEN close the following isolation MCR 1 valve for the FWREG valve that is being removed from service
For C33-LVF001A close FWS-MOV27A, FWREG VLV 1A INLT Valve.
For C33-LVF001B close FWS-MOV27B, FWREG VLV 1B INLT Valve.
For C33-LVF001C close FWS-MOV27C, FWREG VLV 1C INLTValve.

I Fully close the C33-LVF001A(B)(C), FWREG VALVE A(B)(C) that was MCR 1 removed from service. Page 279 of 296 Q *-===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bas_es GOP / SOP ACTIONS LOCATION MODE NOTES Record the temperature of the feedwater at the reactor feed pumps. TB 67' 1 IF FWS-MOV27A, B, or C, FWREG VLV 1A(18)(1 C) INLT were closed MCR 1 with feedwater temperature at the reactor feed pumps greater than 200F, THEN refer to Section 5.7 for further stroking requirements. WHEN the FWREG Valve is at 0% open, THEN record demanded MCR /TB 67' 1 Not required position in the MCR, position indication in the MCR, and local position for plant indication. shutdown or cooldown Line up RWCU reject to the main condenser per SOP-0090, Reactor MCR 1 Not Feedwater System. required, but the preferred method to control level if shutdown long term . RWCU blowdown operations (SOP-0090)

, ' .. i*f . ,. ,' Request chemistry sample to verify reactor water quality is within the MCR 1 specifications of Technical Requirement 3.4.13. Notify Radiation Protection prior to rejecting water to the Main MCR 1 Condenser or Radwaste.

IF rejecting to the Main Condenser, THEN open G33 F046, RWCU MCR ' 1 DRAIN TO MN COND. IF rejecting during cold shutdown or refueling, THEN open G33 F031, MCR 1 RWCU DRAIN ORIFICE BYP. IF rejecting with the RWCU HXs isolated, THEN perform the following: MCR 1

Open G33-F107, RWCU REGEN HX BYPASS .
Throttle open G33-PVF033, RWCU REJECT FLOW VALVE to establish reject flow as indicated on G33-R602, RWCU REJECT FLOW.
IF necessary to establish adequate reject flow, THEN close G33-F040, RWCU INBD RETURN VALVE. To establish the reject and maintain RWCU flowrate on G33 R609, MCR 1 RWCU INLET FLOW nearly constant, simultaneously throttle the following:
G33-PVF033, RWCU REJECT FLOW VALVE open using G33 R606, RWCU REJECT FLOW CONTROLLER
G33 F042, RWCU REGEN HX OUTLET closed Observe blowdown flow on G33 R602, RWCU REJECT FLOW. MCR 1 Monitor reactor water level while blowdown is in progress.

MCR 1 Page 280 of 296

-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation

& Shutdown Areas Tables A-3 & H-2 Bases GOP / SOP ACTIONS LOCATION MODE NOTES IF desired, THEN shutdown Reactor Recirculation HPU A(B) per SOP-MCR 1 0003 to prevent unnecessary Flow Control Valve movement. Initiate a Manual Scram per AOP-0001, Reactor Scram. MCR 1/3

Verify the Hydrogen Water Chemistry (HWC) System shuts down on scram si nal. WHEN it is desired to bypass the Feedwater Pump Level 8 Trip, THEN MCR 3 Not required perform Attachment 5, Feedwater Pump Level 8 Trip Jumper for plant Installation/Restoration Step 1. shutdown or cooldown WHEN it is desired to bypass the MSO Level 8 Trip, THEN perform MCR 3 Not required Attachment 8, MSO Level 8 BYPASS Switch Step 1. for plant shutdown or cooldown Monitor Bottom Head Drain Temperature on B21-R643 Point 4 or MCR 3 B33NA002 and take the following actions, as necessary, in a timely controlled manner to prevent an excessive temperature change. (STP-050-0700, RCS Pressure/Temperature Limits Verification).
Reset the Scram .
Reset any FCV runback per ARP-680-04 . Within one hour after THERMAL POWER < 10% RTP in MODE 1, MCR 3 complete the following steps:
Verify/ensure that the RCIS data mode is selected to "CHAN 1 and CHAN 2".
Select and attempt to withdraw an out-of-sequence control rod .
Verify no rod motion occurs .
Ve~ify Annunciator, P680-07A-C01, CONTROL ROD WITHDRAWAL BLOCK is actuated.
Verify WITHDRAWAL BLOCK Status Light is ON and not flashing. (SR 3.3.2.1.4)

Place the APRM FCTR Cards to the Normal trip setpoints by MCR 3 depressing the Normal/Setup pushbutton and verifying the normal/setup LED indication is green After the Main Turbine is tripped, open the Feedwater Heater Vents per MCR 3 SOP-0010, MSR & FW Heaters Extraction Steam and Drains. Establish MSR Steam Blanketing per SOP-0010, MSR & FW MCR 3 Heaters Extraction Steam and Drains. Establishing Steam Blanketing (SOP-0010) IF Aux. Steam is available, THEN perform the following: MCR 3

Throttle ASR-MOV104, MSR STM BLANKET SHUTOFF \ Page 281 of 296 Q -:::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation

& Shutdown Areas Tables A-3 & H-2 Bases GOP / SOP ACTIONS open until both red and green indication is received.

WHEN Aux steam to MSR steam blanketing line is warm as indicated on Computer point ASRTA01, THEN fully open ASR-MOV104.

Verify the following are closed: '

MSS-MOV111, MSR 1 STM SPL Y SHUTOFF
MSS-MOV112, MSR 2 STM SPL Y SHUTOFF
MSS-PVRSHLV1, MSR 1 HIGH LOAD VALVE
MSS-PVRSHLV2, MSR 2 HIGH LOAD VALVE
MSS-PVRSLLV1, MSR 1 LOW LOAD VALVE
MSS-PVRSLLV2, MSR 2 LOW LOAD VALVE
DSR-MOV107, SCAV STM TO 1ST PT HTRA
DSR-MOV109, SCAV STM TO 1ST PT HTR B
DSR-MOV108, SCAV STM TO CONDA
10) DSR-MOV110, SCAV STM TO COND B
DTM-MOV54A, MSL TO MSR 1 COND DR
DTM-MOV54B, MSL TO MSR 2 COND DR IF Aux Steam is available, THEN open the following:
ASR-MOVBSFV1, MSR 1 STM BLANKET SPL Y
ASR-MOVBSFV2, MSR 2 STM BLANKET SPL Y WHEN several minutes have elapsed after opening ASR-MOVBSFV1 and ASR-MOVBSFV2, THEN place the following control switches to CLOSE:
MSS-MOV111, MSR 1 STM SPL Y SHUTOFF
MSS-MOV112, MSR 2 STM SPL Y SHUTOFF IF notching is observed during the depressurization and magnitude is less than six inches, THEN:
Make all possible attempts to maintain reactor pressure .
Have l&C backfill the reference leg in which notching was observed, even if reference leg was overfilled prior to this event. \ IF notching is observed during the depressurization and magnitude is greater than six inches, THEN declare the trip channels associated with that signal inoperable and comply with Technical Specification requirements.

Reduce the number of running condensate pumps to one per SOP-0007, Condensate System. Shutdown of CNM-P1A(B)(C) CONDENSATE PUMPS (SOP-0007) Request Aux Control Room remove unnecessary Condensate Filters from service per SOP-0124, Condensate Filtration System. Page 282 of 296 LOCATION MODE ';\* .. ' MCR 3 MCR 3 ACR NOTES 7 ,** Not required for plant shutdown or cooldown Not required for plant shutdown or cooldown

===* Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation

& Shutdown Areas Tables A-3 & H-2 Bases GOP/ SOP ACTIONS LOCATION MODE NOTES Request Aux Control Room remove unnecessary Condensate Demins ACR 1 Not required from service per SOP-0093, Condensate Demineralizer System. for plant shutdown or cooldown IF shutting down CNM-P1 C, CNDS PUMP 1 C, THEN secure Oxygen MCR /TB 95' 1 Not required injection per SOP-0123, Hydrogen Water Chemistry H2 and 02 for plant System. shutdown or cooldown IF securing the Condensate System, THEN perform the following: TB 123' 1 Not required

Close all CNM-V3105 A, B, C; D, and E, CNM-FLT1 A, B, C, for plant D, and E BACKWASH AIR SUPPLY valves. shutdown or cooldown
IF desired to isolate and depressurize CNM-TK100, AIR RECEIVING TK, THEN perform the following:

\ 0 CLOSE CNM-V3110, SVCE AIR ISOL VLV INLET SERV. AIR ISOL VLV. 0 Uncap and install hose on CNM-V3112, CNM-TK100 DRAIN ISOLATION VALVE. 0 Open CNM-V3112. Depress the CLOSE pushbutton for CNM-MOV3A(B)(C), CNDS PUMP MCR 1 1A(1B)(1C) DISCH. WHEN pump motor current lowers below 100 amps, THEN stop CNM-MCR 1 P1A(B)(C), CNDS PUMP 1A(1B)(1C). WHEN CNM-MOV3A(B)(C), CNDS PUMP 1A(1 B)(1C) DISCH is full MCR 1 closed, THEN depress the STOP pushbutton. Verify associated CCS-MOV67A(B)(C), CNDS PMP 1A(1 B)(1C) MOT MCR 1 CLR close for pump stopped. Verify associated CCS-MOV68A(B)(C), CNDS PMP 1A(1B)(1C) BRG MCR 1 CLR close for pump stopped. ' Locally verify breaker relay trip flags are reset for Condensate Pump NSW 98' Not required stopped in Step 6.1.6 for plant shutdown or cooldown WHEN the Steam Jet Air Ejectors (SJAEs) and Gland Seal and MCR 1 Exhaust System are removed from service, THEN adjust CNM-H/A 114 to 10% or to a setpoint determined by the CRS/OSM. \ *. ;;J/t;;\~~;it~i~ IL** .. -~;oi 1 t;\ *,:~F j~ ,

j\,iJ .,:t:.i WHEN less than or equal to 145 MWT, THEN perform the following:

MCR 1

Secure SJAE per SOP-0092, Offgas System .
Start a mechanical vacuum pump per SOP-0025, Condenser TB 123' & 95' Not required Air Removal System. for plant shutdown or Page 283 of 296

-===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP / SOP ACTIONS LOCATION MODE NOTES cool down WHEN all rods have been fully inserted, THEN initiate RPV cool down MCR 3 at less than or equal to 100°F/hr by one of the following methods:

Initiate an automatic cooldown of the RPVfrom HMI Screen 5532, Pressure Control by performing the following:

0 Depress AUTO RATE in the Cooldown/Heatup Rate section and verify the pushbutton turns cornsilk and disabled. 0 Entered the desired cooldown rate and depress ENTER. 0 Enter the desired Target Pressure in the Throttle Pressure Control section and verify the Target pressure reflects the value entered. 0 Depress the GO pushbutton and verify:

Pressure regulator setpoint is changing automatically.
Bypass valves modulate to control pressure .
RPV temperature lowers in accordance with the selected cooldown rate.
Initiate a cooldown by slowing and periodically reducing turbine pressure regulator setpoint from HMI Screen 5532, Pressure Control as follows: 0 Enter the desired Target pressure into the HMI display and depress ENTER. 0 Depress the GO pushbutton and verify Press Set changes to reflect the Target value.
ON HMI Screed 5537, BPV Jack, control the Bypass valves as follows to establish the required steam flow to the main condenser for cool down: 0 Enter the BPV Jack Rate and depress ENTER. 0 Enter the BPV Target and depress ENTER. 0 Verify the BPVs open to the desired Target value to establish steam flow to the main condenser.

Prior to reaching 600 psig, initiate monitoring of ERIS Narrow Range MCR 3 level for "notching" of one or more level indications (IF ERIS trending is not available, THEN contact l&C to arrange for alternate Narrow Range trending). Down range IRMs to maintain indication between downscale alarm and MCR 3 / upscale alarm. Insert SRMs to maintain SRM counts between 1 x 103 and 1 x 105 cps. MCR 3 Fully insert SRMs before the IRMs are on range 3. Verify overlap between SRM and IRM. (All SRMs reading< 1x105 cps MCR 3 Page 284 of 296 ~=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP / SOP ACTIONS LOCATION MODE NOTES prior to IRMs reading< 5 on range 1.) WHEN all control rods are fully inserted, THEN perform one of the MCR 3 following:

NORMAL SHUTDOWN 0 Place the REACTOR SYSTEM MODE SWITCH to SHUTDOWN.

0 WHEN at least 10 seconds have elapsed, THEN reset the Reactor Scram.

SOFT SHUTDOWN 0 Bypass the REACTOR MODE SWITCH POSITION SCRAM per SOP-0079, Reactor Protection System Attachment

7. 0 Place the REACTOR SYSTEM MODE SWITCH to SHUTDOWN.

0 WHEN at least 10 seconds have elapsed, THEN restore the REACTOR MODE SWITCH POSITION SCRAM per SOP-0079, Reactor Protection System Attachment

7. WHEN the reactor is shutdown AND at the direction of the Operations ow 3 Not required Shift Manager, THEN perform a drywell inspection per Attachment 3, 141'/131'/118 for plant Drywell Inspection Checklist.

'/107'/95'/82' shutdown or cooldown Maintain hot shutdown condition with RPV pressure between 250 psig MCR 3 and 1055 psig. IF RHR Pump warmup and flushing is required, THEN perform MCR 3 warmup/flushing per SOP-0031, Residual Heat Removal. Shutdown Cooling Flush, Warmup, and Startup (SOP-0031) ' '.',:,1' ' '.;'~\' IF RPV pressure is less than 135 psig, THEN have Electrical MCR 3 Maintenance implement the PM Task to re-land the thermal overload/loss of power annunciator leads for E12-F009, RHR SHUTDOWN COOLING !NBD !SOL VALVE. Verify the following ,breakers are ON: MCR 3

EHS-MCC2E BKR 5C, C002A DISCH MIN FLOW VALVE
EHS-MCC2F BKR 7B, C002B DISCH MIN FLOW Shutdown Cooling Flush MCR 3 Not required
Request Chemistry to verify Suppression Pool is within best for plant practice limits of CSP-0006, Chemistry Surveillance and shutdown or Scheduling System. cooldown
IF suppression pool conductivity is NOT within best practice limits of CSP-0006, THEN perform a complete flush.
IF desired and suppression pool conductivity is within best practice limits of CSP-0006, THEN perform the following:

0 Place RHR A(B) in suppression pool cooling. 0 Monitor E12-R61 OA(B), HX A(B) OUTLET CONDUCTIVITY OR Chemistry sample from SST-PNL80, and continue flush until conductivity is less than 2 umho/cm. Page 285 of 296 Q *::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP/ SOP ACTIONS o Vent RHR A(B) HX as follows:

Open E12-F074A(B), RHR A(B) HX UP STREAM VENT VALVE.
Open E12-F0"73A(B), RHR A(B) HX ON STREAM VENT VALVE.
WHEN at least 1 minute has elapsed, THEN close the following:
E12-F073A(B)
E12-F074A(B)
WHEN conductivity is less than 2 umho/cm, THEN perform the following:

o Secure RHR A(B) from suppression pool cooling. o Close E12-VF085A(B), LPCS FILL PUMP STOP CHECK TO RHR A DISCH (DISCH FILL PUMP STOP CHECK TO RHR B DISCH). o Continue with pump warm-up as desired. LOCATION During cooldown, review Attachment 7, High Critical Non-Safety MCR Related MOVs That Are Susceptible to Thermal Binding and carryout actions to prevent thermal binding or actions to unbind the valves if they are closed when the valve temperature is greater than 200°F. IF MS IVs have been closed for pressure control, THEN the following MCR systems may be utilized as necessary to continue a cooldown at less than or equal to 100°F/hr:

RWCU system per SOP-0090, Reactor Water Cleanup System (i.e. RWCU Slowdown Operation)
Main Steam Line Drains MODE NOTES 3 3 IF/WHEN reactor pressure is less than 400 psig, THEN any running MCR / TB 67' 3 Not required for plant shutdown or cooldown reactor feedwater pumps may be shutdown per SOP-0009, Reactor Feedwater System. Prior to reaching 135 psig, initiate monitoring of the following MCR parameters:
RHR Room sump levels (monitor for possible reactor vessel inventory loss from shutdown cooling leakage). (DFR-Ll135 and DFR-Ll138).
Suppression Pool for unexpected level rise (monitor for reactor vessel inventory loss from RHR to the Suppression Pool). WHEN RPV pressure has been lowered to below 135 psig, THEN MCR place one loop of RHR in Shutdown Cooling per SOP-0031, Residual Heat Removal. RHR Pump A Warmup (SOP-0031)

Verify closed E12-F064A, RHR PUMP A MIN FLOW TO SUP PL. Page 286 of 296 3 3 ~-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP/ SOP ACTIONS LOCATION MODE NOTES Verify open E12-F047A, RHR A HX INLET VALVE. MCR 3 Verify closed E12-F004A, RHR PUMP A SUP PL SUCTION VALVE. MCR 3 Verify open E12-F006A, RHR PUMP A SDC SUCTION VALVE. MCR 3 On H13-P870, verify closed SPC-AOV16, SPC HX SW DISCH VLV. MCR 3 On H13-P870, throttle open E12-F068A, RHR HX A SVCE WTR RTN MCR 3 to establish less than or equal to 5800 gpm flow as indicated on E12-R602A, RHR HX A SVCE WTR FLOW. Close E12-F048A, RHR A HX BYPASS VALVE. MCR 3 Verify open E12-F003A, RHR A HX OUTLET VALVE. MCR 3 Verify open E12-F010, RHR SDC MAN ISOL VLV. MCR 3 In the Div 1 RSS Room at C61-PNL001, verify E12-MOVF008 CB 95' Div 1 3 REQUIRED ENABLE/DISABLE Switch is in ENABLE. RSS Room Perform the following: MCR 3

Depress B21H-S33, !NBD ISOLATION SEAL-IN RESET Pushbutton.
Depress B21H-S32, OUTBD ISOLATION SEAL-IN RESET Pushbutton.

At H13-P601, check RHR ISOLATION Status Lights are ON for E12-MCR 3 FOOS and F009. I Note current indicated CNS flow at LWS-PNL 187 on CNS-Fl 116. ACR 3 Not required for plant shutdown or cooldown Slowly open E12-VF020, SHUTDOWN COOLING SUCTION FILL. AB 95' 3 Not required RHR C Pump for plant Room shutdown or cooldown WHEN CNS flow into the shutdown cooling header stops as indicated AB 95' 3 Not required by a lack of flow noise or flow indication of approximately the same RHR C Pump for plant value as previously noted on CNS-Fl116, THEN close E12-VF020. Room shutdown or cooldown Open E12-F008, RHR SHUTDOWN COOLING OUTBD !SOL VALVE. MCR 3 Open E12-F009, RHR SHUTDOWN COOLING INBD ISOL VALVE. MCR 3 Perform STP-204-0204, RHR Shutdown Cooling Piping Fill Verification. Steam 3 Not required Tunnel 114' for plant shutdown or cooldown Notify Radwaste of reactor water flush to the Waste Collector Tanks. MCR 3 Page 287 of 296 River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP / SOP ACTIONS LOCATION MODE NOTES Open E12-F049, RHR A TO RADWASTE UP STREAM !SOL VALVE. MCR 3 Throttle open E12-F040, RHR A TO RADWASTE DN STREAM !SOL MCR 3 VALVE. Monitor E12-R601, RHR TEMPERATURES, Point 1, RHR INLET TO MCR 3 HX1 A-1 (E12-N004A) for temperature rise and throttle E12-F040 to maintain less than or equal to 100°F/hr heatup. Continue the warm-up until E12-R601 Point 1 is within 1 oo*F of reactor MCR 3 water temperature. Close E12-F049, RHR A TO RADWASTE UP STREAM ISOL VALVE. MCR 3 Close E12-F040, RHR A TO RADWASTE DN STREAM ISOL VALVE. MCR 3 Open E12-F048A, RHR HXA BYPASS VALVE. MCR 3 I RHR Pump B Warm-up (SOP-0031) ii; ? *. *'it /'.;< . ,, . :( Verify closed E12-F064B, RHR PUMP B MIN FLOW TO SUP PL. MCR 3 Verify open E12-F047B, RHR B HX INLET VALVE. MCR 3 Verify closed E12-F004B, RHR PUMP B SUP PL SUCTION VALVE. MCR 3 Verify open E12-F006B, RHR PUMP B SDC SUCTION VALVE. MCR 3 IF Standby Service Water is supplying service water loads, THEN on MCR 3 H13-P870, verify closed SPC-AOV16, SPC HX SW DISCH VLV. On H13-P870, throttle open E12-F068B, RHR HX B SVCE WTR RTN MCR 3 to establish less than or equal to 5800 gpm flow as indicated on E12-R602B, RHR HX B SVCE WTR FLOW. Verify open E12-F010, RHR SOC MAN ISOL VLV. MCR 3 Verify closed E12-F049, RHR A TO RADWASTE UP STREAM ISOL MCR 3 VALVE. Verify closed E12-F040, RHR A TO RADWASTE DN STREAM !SOL MCR 3 VALVE. In the Div 1 RSS Room at C61-PNL001, verify E12-MOVF008 CB 95' Div 1 3 REQUIRED ENABLE/DISABLE Switch is in ENABLE. RSS Room Perform the following: MCR 3

Depress B21H-S32, OUTBD ISOLATION SEAL-IN RESET Pushbutton.
Depress B21H-S33, INBD ISOLATION SEAL-IN RESET Pushbutton.

At H13-P601, check RHR ISOLATION Status Lights are ON for E12-MCR 3 F008 and F009. Note current indicated CNS flow at LWS-PNL 187 on CNS-Fl 116. ACR 3 Not required for plant shutdown or Page 288 of 296 River Bend Station EAL Basis Document Revisipn XXX Attachment 2 -Safe Operation

& Shutdown Areas Tables A-3 & H-2 Bases GOP / SOP ACTIONS LOCATION MODE NOTES GOoldown Slowly open E12-VF020, SHUTDOWN COOLING SUCTION FILL. AB 95' RHR 3

Not required C Pump for plant Room shutdown or cooldown WHEN CNS flow into the shutdown cooling header stops as indicated AB 95' RHR 3 Not required by a lack of flow noise or flow indication of approximately the same C Pump for plant value as previously noted on CNS-Fl116, THEN close E12-VF020.

Room shutdown or cooldown Open E12-F008, RHR SHUTDOWN COOLING OUTBD ISOL VALVE. MCR 3 Open E12-F009, RHR SHUTDOWN COOLING INBD ISOL VALVE. MCR 3 Perform STP-204-0204, RHR Shutdown Cooling Piping Fill Verification. Steam 3 Not required Tunnel 114'

for plant shutdown or cooldown Notify Radwaste of reactor water flush to the Waste Collector Tanks. MCR 3 Unlock and open E12-VF072B, RHR B DISCH LINE FLUSH. *AB 70' RHR 3 REQUIRED B Pump Room Unlock and throttle open E12-VF070, RHR DR TO RADWASTE.

AB 80' RHR 3 REQUIRED A Pump Room Monitor E12-R601, RHR TEMPERATURES, Point 11, RHR DISCH TO MCR 3 RADWASTE (E12-N024) and throttle E12-VF070 to maintain less than . or equal to 1 OO'F/hr heatup. Continue the warm-up until E12-R601 Point 11 is within 100'F of MCR 3 reactor water temperature. Close and lock E12-VF070, RHR DR TO RADWASTE. AB 80' RHR 3 REQUIRED A Pump Room Close and lock E12-VF072B, RHR B DISCH LINE FLUSH. AB 70' RHR 3 REQUIRED B Pump Room Startup of Shutdown Cooling (SOP-0031) ~H F s):; ,, ;: IF any of the following RHR Shutdown Cooling interlocks are to be MCR 3 bypassed, THEN obtain senior plant management review and approval and verify contingency methods are in place to supply sufficient makeup water if a draining event occurs while the SOC interlocks are bypassed:

Low reactor water level isolation of E12-F008, RHR SHUTDOWN COOLING OUTBD ISOL VALVE and E12-F009, RHR SHUTDOWN COOLING !NBD ISOL VALVE. Page 289 of 296 Q ~=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation

& Shutdown Areas Tables A-3 & H-2 Bases GOP / SOP ACTIONS LOCATION MODE NOTES

Interlocks between E12-F004, RHR PUMP SUP PL SUCTION VALVE and E12-F006, RHR PUMP SOC SUCTION. On H13-P601, verify less than 135 psig Reactor Pressure as indicated MCR 3 on B21-R623A(B), RX LEVEUPRESSURE RECORDER A(B). Verify closed the following:

MCR 3

E12-F004A(B), RHR PUMP A(B) SUP PL SUCTION VALVE
E12-F064A(B), RHR PUMP A(B) MIN FLOW TO SUP PL
E12-F024A(B), RHR PUMP A(B) TEST RTN TO SUP PL
E12-F037A(B), RHR A(B) TO UPPER POOL FPC ASSIST
E12-F048A(B), RHR A(B) HX BYPASS VALVE .
E12-F011A(B), RHR A(B) HX CNDS FLUSH TO SUP PL. Place in OFF and initiate administrative controls for EHS-MCC2E(2F)

AB 114' West 3 REQUIRED BKR 5C(7B), C002A(B) DISCH MIN FLOW VALVE. IF Standby Service Water is supplying service water loads, THEN on MCR 3 H13-P870, verify closed SPC-AOV16, SPC HX SW DISCH VLV. On H13-P870, throttle open E12-F068A(B), RHR HX A(B) SVCE WTR MCR 3 RTN to establish less than or equal to 5800 gpm flow as indicated on H13-P601, E12-R602A(B), RHR HX A(B) SVCE WTR FLOW. Verify Step 4.4.2 has been performed. MCR 3 At H13-P601, depress B21H-S32, OUTBD ISOLATION SEAL-IN MCR 3 RESET Pushbutton. At H13-P601, depress B21 H-S33, INBD ISOLATION SEAL-IN RESET MCR 3 Pushbutton. At H13-P601, check RHR ISOLATION Status Lights are ON for E12-MCR 3 FOOS and E12-F009. In the Div 1 RSS Room at C61-PNL001, verify E12-MOVFOOB CB 95' Div 1 3 REQUIRED ENABLE/DISABLE Switch is in ENABLE. RSS Room Verify open the following: MCR 3

E12-F010, RHR SOC MAN !SOL VLV
E12-F009, RHR SHUTDOWN COOLING INBD ISOL VALVE
E12-F008, RHR SHUTDOWN COOLING OUTBD ISOL VALVE
E12-F006A(B), RHR PUMP A(B) SOC SUCTION VALVE
E12-F047A(B), RHR A(B) HX INLET VALVE Verify open one of the following:

MCR 3 .

E12-F053A(B), RHR PUMP A(B) SOC INJECTION VALVE
E12-F037A(B), RHR A(B) TO UPPER POOL FPC ASSIST Close E12-F003A(B), RHR A(B) HX OUTLET VALVE. MCR 3 Start E12-C002A(B), RHR PUMP A(B) and IMMEDIATELY throttle MCR 3 open E12-F048A(B), RHR A(B) HX BYPASS VALVE to obtain greater Page 290 of 296

--Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP / SOP ACTIONS LOCATION MODE NOTES than or equal to 2000 gpm and less than or equal to 3000 gpm. Establish a stable flow of greater than or equal to 4000 gpm and less MCR 3 than or equal to 5000 gpm by throttling E12-F048A(B), RHR A(B) HX BYPASS VALVE. Throttle open E12-F003A(B), RHR A(B) HX OUTLET VALVE to MCR 3 approximately 10 PERCENT as indicated on E12-R611 A(B), HX A(B) OUTLET VLV POS. Establish a cool down rate of less than 1 OO'F/hr as follows: MCR 3

Slowly throttle open E12-F003A(B)

RHR A(B) HX OUTLET VALVE and monitor the cooldown rate.

Throttle E12-F003A(B), RHR A(B) HX OUTLET VALVE and E12-F048A(B), RHR A(B) HX BYPASS VALVE to obtain the desired cooldown rate or maintain the desired coolant temperature while maintaining a constant RHR loop flow.
IF shifting divisions of Shutdown Cooling per Section 5.6, THEN in the other RHR loop, throttle E12-F003B(A), RHR B(A) HX OUTLET VALVE and E12-F048B(A), RHR B(A) HX BYPASS VALVE to maintain the desired cooldown rate or coolant temperature while maintaining a constant RHR loop flow.
Close FWS-MOV?A(B), A(B) FW OUTBD ISOL. IF it is desired to establish RHR Shutdown Cooling Protection, THEN MCR/ AB 95' Not required Go To Section 4.5. & 115' for plant shutdown or cooldown \1; ,:;;~ }; ): i ' . ', . ),'? J~f :' .*' WHEN RHR Shutdown Cooling is established and adequate RPV MCR 3 makeup is assured via CRD or Feedwater, THEN close FWS-MOV?A(B), A(B) FW OUTBD ISOL valve on the Feedwater Header supporting RHR Shutdown Cooling. 1 WHEN RPV cooldown is being conducted using RHR Shutdown MCR 3 Cooling, THEN stop discharging steam to the main condenser, break condenser vacuum and continue to shutdown the turbine plant as follows:
Place CONDENSER LOW VACUUM BYPASS Switches to BYPASS.
Close all turbine bypass valves and steam drain valves . * . Open CNM-AOWB, CNDS VAC BRKR.
WHEN condenser vacuum reaches approximately O" Hg, THEN shutdown steam seals per SOP-0015, Gland Seal System and Exhaust System. WHEN condenser vacuum reaches O" Hg AND mechanical vacuum TB 67' 3 Not required pump operation is no longer required, THEN align Alternate Hotwell for plant Level tygon tubing per SOP-0008, Condensate Storage, Makeup and shutdown or Transfer.

cooldown Page 291 of 296 -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP / SOP ACTIONS LOCATION MODE NOTES At less than 190°F, perform the following: MCR 3

Open B21-MOVF001, RX DN STREAM HEAD VENT TO DW EQPT DR SUMP.
Open B21-MOVF002, RX UP STREAM HEAD VENT TO DW EQPT DR SUMP.
Close B21-MOVF005, RX HEAD VENT TO MSL A.
Establish administrative controls to maintain the established vessel vent path until vessel head piping is disassembled.

IF required/desired, THEN close the MSIVs by placing the following in MCR 3 CLOSE:

B21-F028B, MSL B OUTBD MSIV
B21-F028D, MSL D OUTBD MSIV
B21-F028A, MSL A OUTBD MSIV
B21-F028C, MSL C OUTBD MSIV
B21-F022B, MSL B INBD MSIV
B21-F022D, MSLD INBD MSIV
B21-F022A, MSL A INBD MSIV
B21-F022C, MSL C INBD MSIV ' . .* * ..* * . . . , . . At less than 200°F, Mode 4, perform the following:

MCR 3 At H13-P632, place the following switches to BYPASS:

E31A-S1A, RWCU ISOLATION BYPASS DIV 1
E31A-S2A, RCIC ISOLATION BYPASS DIV 1
E31A-S4A, RHR ISOLATION BYPASS DIV 1 At H13-P642, place the following switches to BYPASS: MCR 3
E31A-S1B, RWCU ISOLATION BYPASS DIV 2
E31A-S2B, RCIC ISOLATION BYPASS DIV 2
E31A-S4B, RHR ISOLATION BYPASS DIV 2 Implement Shutdown Cooling Protection per SOP-0031, Residual Heat MCR/ AB 95' 3 Not required Removal. & 115' for plant shutdown or cooldown Bypass RPS trip logic using EOP-0005 Enclosure 12 Bypass Switches MCR 3 per SOP-0079, Reactor Protective System. Bypass ARI logic trips per SOP-0079.

MCR 3 Bypass Backup Scram Valve trips per SOP-0079. MCR 3 At less than 200°F, Mode 4, perform the following to prevent isolating MCR 3 Not required Breathing Air: for plant

Verify open SAS-MOV102, SVCE AIR OUTBD ISOL shutdown or Page 292 of 296 Q ~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation

& Shutdown Areas Tables A-3 & H-2 Bases GOP / SOP ACTIONS LOCATION MODE NOTES cooldown

Open EHS-MCC2J Bkr 3C, SAS-MOV102 CONTAINMENT AB 141' West 3 Not required SERVICE AIR ISOLATION.

for plant shutdown or cooldown

Hang the following SAS-MOV102 Caution Tags stating that MCR 3 Not required before closing the breaker or valve, verify that Breathing Air is for plant not in use.: shutdown or 0 EHS-MCC2J Bkr 3C cooldown 0 SAS-MOV102 local handwheel 0 SAS-MOV102 MCR control switch
Hang "Breathing Air in Use" sign in the MCR. MCR 3 Not required for plant shutdown or cooldown
Make an "Open Item" Narrative Log entry by checking the MCR 3 Not required "Open Item" box stating that "Breathing Air Is in Use" to carry for plant over until Breathing Air is no longer in use. shutdown or cooldown At less than 200°F, notify Chemistry to consider securing the Durability AB 114' 3 Not required Monitor. Crescent for plant Area shutdown or cooldown IF required/desired, THEN shutdown Reactor Recirculation System per MCR 3 Not required SOP-0003, Reactor Recirculation System and raise reactor water level for plant to at least 75 inches on shutdown range level instrumentation.

shutdown or cooldown As necessary, reduce the number of operating Turbine Building Chillers TB 67' 3 Not required per SOP-0064 to prevent the chillers from tripping on low load. for plant shutdown or cooldown GOP-0003 Scram Recovery Verify/establish on-scale neutron monitoring on the SRMs and IRMs MCR 3

VERIFY the SRM Channel Functional Tests are current. If Channel Functional Tests are not current, refer to Tech Spec 3.3.1.2. Maintain RPV pressure to prevent excessive cooldown rates or RPV MCR 3 overpressurization by:
Use of Main Turbine Bypass System .
Use of Normal Plant Steam Loads/Steam Line Drains .
RCIC in CST to CST Mode per SOP-0035 Reactor Core Isolation Cooling System. Page 293 of 296

~-----~-----~-~~---- -Q ~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP/ SOP ACTIONS LOCATION MODE NOTES

Alternate opening of SRVs as needed (only on RPV isolation).

Maintain RPV Level using: MGR 3

Condensate/Feedwater
RCIC
HPCS
CRD/RWCU If tripped, restart Reactor Recirculation Pumps on LFMG per MGR/ LFMG 3 Not required SOP-0003, Reactor Recirculation:

Room for plant

Open flow control valves to the full open position . shutdown or cooldown Verify Main Turbine steam seals are being maintained at approximately MGR 3 4 psig. Start air removal pumps per SOP-0025 Condenser Air Removal MGR /TB 67' 3 Not required System. for plant
Maintain condenser vacuum between 23" Hg and 28" Hg . shutdown or cooldown IF the Steam Jet Air Ejectors have been lost, THEN Secure Offgas MGR /TB 95' 3 Not required System per SOP-0092, to establish purge air flow in order to prevent & 123' for plant system reverse flow. shutdown or -cooldown Notify Chemistry Department to operate the Offgas Hydrogen MGR/TB 3 Not required Analyzers per COP-0227, Operation of the' Offgas Hydrogen Analyzers.

123' for plant shutdown or cooldown Record the highest vessel pressure indicated by tracking pointer on RB 114' 3 Not required B21-PIR004A and B21-PIR004B (114' Containment). for plant shutdqwn or Reset the tracking pointer. cooldown Inspect all CRD HCUs for leakage due to piping cracks. RB 114' 3 Not required

Notify Engineering NOE that visu_.§1 inspections of HCU for plant charging water piping are required.

shutdown or cooldown WHEN FWREG Valves are removed from service, THEN perform MGR/TB 67' 3 Not required SOP-0009 Attachment for Calibration Check of FWREG Valve. for plant shutdown or cooldown Go To GOP-0002 -Start at the beginning of GOP-0002 and complete MGR 3 Page 294 of 296 Q -~ Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP / SOP ACTIONS LOCATION MODE NOTES all steps required to place the plant in the desired mode after scram. IF the plant tripped while connected to the grid, THEN notify Site MCR 3 Design Engineering to notify the TOP personnel of the event per ENS-DC-201, ENS Transmission Grid Monitoring Attachment 9.3 Step 3.0[1]. Engineering perform a review of post scram cooldown data and MCR 3 compare to PT Curves provided in STP-050-0700 Attachment

3. Also verify the cool down rate is bounded by analyzed thermal cycles. Shift Manager to perform a Post SCRAM crew critique identifying all MCR 3 Human Performance issues and Equipment Malfunctions.

Document each item on separate CRs. Attach Crew Critique to this procedure. AOM/Shift Manager review equipment malfunctions and recommend to MCR 3 OSRC required repairs prior to restart. This should include an evaluation of risk mitigating Non TRM structures, systems, and components. Control Room ventilation systems have adequate engineered safety/design features in place to preclude a Control Room evacuation due to the release of a hazardous gas. Therefore, the Control Room is not included in this assessment or in Table H-2. Page 295 of 296 Q -~ Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases Table A-3 & H-2 Results Table A-3 & H-2 Safe Operation & Shutdown Rooms/Areas Room/Area Mode Auxiliary Building 70' RHR B Pump Room 3 Auxiliary Building 80' RHR A Pump Room 3 Auxiliary Building 114' West 3 Control Building 95' Div 1 RSS Room 3 Mode 3 is included above for SOC-related activities because the procedures begin alignment in Mode 3; however, these actions could be delayed until Mode 4, if necessary. In order to ensure adequate guidance to emergency response personnel, the above areas are added to the EAL in order to provide prompt operator guidance for EAL declaration. Control Room ventilation systems have adequate engineered safety/design features in place to preclude a Control Room evacuation due to the release of a hazardous gas. Therefore, the Control Room is not included in this assessment or in Table H-2. Page 296 of 296 ENCLOSURE 3 RBG-47847 PROPOSED EAL TECHNICAL BASIS DOCUMENT (CLEAN) -:::::::-Entergy River Bend Station EAL Basis Document Revision XXX River Bend Station EAL Technical Basis Page 1 of 278 River Bend Station EAL Basis Document Revision XXX Table of Contents 1.0 I INTRODUCTION ........................................................ Error! Bookmark not defined. 2.0 DISCUSSION ............................................................................................................. 3 2'.1 Background ..................................................................................................... 3 2.2 Fission Product Barriers ............................................................................. , .... 4 2.3 Fission Product Barrier Classification Criteria .............................

...................

4 2.4 EAL Organization ............................................................................................ 5 2.5 Technical Bases Information ........................................................................... 7 2.6 Operating Mode Applicability ...... '. .................................................................... 8 3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS .................................. 9 3.1 General Considerations ................................................................................... 9 3.2 Classification Methodology ................................................................

..........

10

4.0 REFERENCES

............. _ ............................................................................................ 14 4.1 Developmental .............................................................................................. 14

4.2 Implementing

................................................................................................. 14 5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS ........................................

. ........ 15 5.1 Definitions (ref. 4.1.1 except as noted) ..........................................................

15 5.2 Abbreviations/Acronyms ................................................................................ 20 6.0 RBS-TO-NEI 99-01 Rev. 6 EAL CROSS-REFERENCE .......................................... 23 7.0 ATTACHMENTS ...................................................................................................... 27 7.1 Attachment 1, Emergency Action Level Technical Bases ............................. 27 ) I Category A-Abnormal Rad Levels/ Rad Effluent ...... : ................................ 28 Category C -Cold Shutdown / Refueling System Malfunction ...................... 65 Category E -Independent Spent Fuel Storage Installation (ISFSI) ............ 106 ' Category F -Fission Product Barrier Degradation ...................................... 109 Table F-1 Fission Product Barrier Threshold Matrix & Bases 122 Category H -Hazards and Other Conditions Affecting Plant Safety ........... 168 Category S -System Malfunction ............................................................... 204 7 1 2 Attachment 2, Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases244 Page 2 of 278 L_____ ____ -- -::::::-Entergy River Bend Station EAL Basis Document Revision XXX

1.0 INTRODUCTION

This document provides an explanation and rationale for each Emergency Action Level (EAL) included in the EAL Upgrade Project for River Bend Station (RBS). It should be used to facilitate review of the RBS EALs and provide historical documentation for future reference. Decision-makers responsible for implementation of EIP-2-001, Classification of Emergencies, may use this document as a technical reference in support of EAL interpretation. This information may assist the Emergency Director in making classifications, particularly those involving judgment or multiple events. The basis information may also be useful in training and for explaining event classifications to off-site officials. The expectation is that emergency classifications are to be made as soon as conditions are present and recognizable for the classification, but within 15 minutes or less in all cases of conditions present. Use of this document for assistance is not intended to delay the emergency classification. Because the information in a basis document can affect emergency classification making (e.g., the Emergency Director refers to it during an event), the NRC staff expects that changes to the basis document will be evaluated in accordance with the provisions of 10 CFR 50.54(q). 2.0 DISCUSSION 2.1 Background EALs are the plant-specific indications, conditions or instrument readings that are utilized to classify emergency conditions defined in the RBS Emergency Plan. In 1992, the NRC endorsed NUMARC/NESP-007 "Methodology for Development of Emergency Action Levels" as an alternative to NUREG-0654 EAL guidance. NEI 99-01 (NUMARC/NESP-007) Revisions 4 and 5 were subsequently issued for industry implementation. Enhancements over earlier revisions included: ,i

Consolidating the system malfunction initiating conditions and example emergency action levels which address conditions that may be postulated to occur during plant shutdown conditions.
Initiating conditions and example emergency action levels that fully address conditions that may be postulated to occur at permanently Defueled Stations and Independent .Spent Fuel Storage Installations (ISFSls).
Simplifying the fission product barrier EAL threshold for a Site Area Emergency.

Subsequently, Revision 6 of NEI 99-01 has been issued which incorporates resolutions to numerous implementation issues including the NRC EAL Frequently Asked Questions (FAQs). Using NEI 99-01 Revision 6, "Methodology for the Development of Emergency Action Levels for Non-Passive Reactors," November 2012 (ref. 4.1.1), RBS conducted an EAL implementation upgrade project that produced the EALs discussed herein. Page 3 of 278 River Bend Station EAL Basis Document Revision XXX 2.2 Fission Product Barriers Fission product barrier thresholds represent threats to the defense in depth design concept that precludes the release of radioactive fission products to the environment. This concept relies on multiple physical barriers, any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment. Many of the EALs derived from the NEI methodology are fission product barrier threshold based. That is, the conditions that define the EALs are based upon thresholds that represent the loss or potential loss of one or more of the three fission product barriers. "Loss" and "Potential Loss" signify the relative damage and threat of damage to the barrier. A "Loss" threshold means the barrier no longer assures containment of radioactive materials. A "Potential Loss" threshold implies a greater* probability of barrier loss and reduced certainty of maintaining the barrier. The primary fission prodLJct barriers are: A. Fuel Clad Barrier (FCB): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets. B. Reactor Coolant System Barrier (RCS): The RCS Barrier is the reactor coolant system pressure boundary and includes the RPV and all reactor coolant system piping out to and including the isolation valves. ) C. Containment Barrier (CNS): The Primary Containment Barrier includes the drywell, the containment, their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves. Containment Barrier thresholds are used as criteria for escalation of the Emergency Classification Level (EGL) from Alert to a Site Area Emergency or a General Emergency. 2.3 Fission Product Barrier Classification Criteria The following criteria are the bases for event classification related to fission product barrier loss or potential loss: Alert: Any loss or any potential loss of either Fuel Clad or.RCS Barrier Site Area Emergency: Loss or potential loss of any two barriers General Emergency: Loss of any two barriers and loss or potential loss of the third barrier 2.4 EAL Organization The RBS EAL scheme includes the following features:

Division of the EAL set into three broad groups: Page 4 of 278

-===-Entergy River Bend Station EAL Basis Document Revision XXX o EALs applicable under any plant operating modes -This group would be reviewed by the EAL-user any time emergency classification is considered. o *-EALs applicable only under hot operating modes -This group would only be reviewed by the EAL-user when the plant is in Hot Shutdown, Startup, or Power Operation mode. o EALs applicable only under cold operating modes -This group would only be reviewed by the EAL-user when the plant is in Cold Shutdown, Refueling or Defueled mode. The purpose of the groups is to avoid review of hot condition EALs when the plant is in a cold condition and avoid review of cold condition EALs when the plant is in a hot condition. This approach significantly.minimizes the total number of EALs that must be reviewed by the EAL-user for a given plant condition, reduces EAL-user reading burden and, thereby, speeds identification of the EAL that applies to the emergency.

Within each group, assignment of EALs to categories and subcategories:

Category and subcategory titles are selected to represent conditions that are operationally significant to the EAL-user. The RBS EAL categories are aligned to and represent the NEI 99-. 01" Recognition Categories." Subcategories are used in the RBS scheme as necessary to further divide the EALs of a category into logical sets of possible emergency classification thresholds. The RBS EAL categories and subcategories are listed below. The primary tool for determining the emergency classification level is the EAL Classification Matrix. The user of the EAL Classification Matrix may (but is not required to) consult the EAL technical bases in order to obtain additional information concerning the EALs under classification consideration. The user should consult Section 3.0 and Attachments 1 & 2 of this document for such information. Page 5 of 278 Q ' ~Entergy River Bend Station EAL Basis Document Revision XXX EAL Groups, Categories and Subcategories EAL Group/Category Any Operating Mode: A -Abnormal Rad Levels/ Rad Effluent H -Hazards and Other Conditions Affecting Plant Safety E -Independent Spent Fuel Storage Installation (ISFSI) Hot Conditions: S -System Malfunction I EAL Subcategory 1 -Radiological Effluent 2 -Irradiated Fuel Event 3 -Area Radiation Levels 1 -Security 2 -Seismic Event 3 -Natural or Technological Hazard 4-Fire 5 -Hazardous Gas 6 -Control Room Evacuation 7 -Emergency Director Judgment 1 -Confinement Boundary 1 -Loss of Emergency AC Power 2 -Loss of Vital DC Power 3 -Loss of Control Room Indications 4-RCS Activity 5 -RCS Leakage 6 -RPS Failure 7 -Loss of Communications 8 -Hazardous Event Affecting Safety Systems F -Fission Product Barrier Degradation None Cold Conditions: C -Cold Shutdown / Refueling System Malfunction 1 -RPV Level 2 -Loss of Emergency AC Power 3-RCS Temperature 4 -Loss of Vital DC Power 5 -Loss of Communications 6 -Hazardous Event Affecting Safety Systems Page 6 of 278 River Bend Station EAL Basis Document Revision XXX 2.5 Technical Bases Information EAL technical bases are provided in Attachment 1 for each EAL according to EAL group (Any, Hot, Cold), EAL category (A, C, E, F, H and S) and EAL subcategory. A summary explanation of each category and subcategory is given at the beginning of the technical bases discussions of the EALs included in the category. For each EAL, the following information is provided: Category Letter & Title Subcategory Number & Title Initiating Condition (IC) Site-specific description of the generic IC given in NEI 99-01 Rev. 6. EAL Identifier (enclosed in rectangle) Each EAL is assigned a unique identifier to support accurate communication of the emergency classification to onsite and offsite personnel. Four characters define each EAL identifier:

1. First character (letter):

Corresponds to the EAL category as described above (A, C, E, F, H orS) 2. Second character (letter): The emergency classification (G, S, A or U) G = General Emergency S = Site Area Emergency A= Alert U = Unusual Event 3. Third character (number): Subcategory number within the given category. Subcategories are sequentially numbered beginning with the number one (1 ). If a category does not have a subcategory, this character is assigned the number one ( 1 ). 4. Fourth character (number): The numerical sequence of the EAL within the EAL subcategory. If the subcategory has only one EAL, it is given the number one (1 ). Classification (enclosed in rectangle): Unusual Event (U), Alert (A), Site Area Emergency (S) or General Emergency (G) EAL (enclosed in rectangle) Exact wording of the EAL as it appears in the EAL Classification Matrix. Page 7 of 278 r River Bend Station EAL Basis Document Revision XXX Mode Applicability One or more of the following plant operating conditions comprise the mode to which each EAL is applicable: 1 -Power Operation, 2 -Startup, 3 -Hot Shutdown, 4 -Cold Shutdown, 5 -Refueling, DEF -Defueled, or All. (See Section 2.6 for operating mode definitions) Definitions: If the EAL wording contains a defined term, the definition of the term is included in this section. These definitions can also be found in Section 5.1.-Basis: An EAL basis section that provides RBS~relevant information concerning the EAL as well as a description of the rationale for the EAL as provided in NEI 99-01 Rev. 6. , Reference(s): Source documentation from which the EAL is derived 2.6 Operating Mode Applicability 1 Power Operation Reactor is critical and the mode switch is in RUN , 2 Startup The mode switch is in REFUEL (with all reactor vessel head closure bolts fully tensioned) or STARTUP/HOT STANDBY 3 Hot Shutdown The mode switch is in SHUTDOWN and average reactor coolant temperature is >200°F 4 1 Cold Shutdown The mode switch is in SHUTDOWN and average reactor coolant temperature is ::=; 200°F 5 Refueling l, The mode switch is in REFUEL or SHUTDOWN with one or more reactpr vessel head closure bolts are less than fully tensioned DEF Defueled RPV contains no irradiated fuel The plant operating mode that exists at the time that the event occurs (prior to any protective system or operator action being initiated in response to the condition) should be compared to the mode applicability of the EALs. If a lower or higher plant operating mode is reached before the emergency classification is made, the declaration shall be based on the mode that existed at the time the event occurred. Page 8 of 278 River Bend Station EAL Basis Document Revisioh XXX 3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS 3.1 General Considerations When making an emergency classification, the Emergency Director must consider all information having a bearing on the proper assessment of an,lnitiating 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 based on loss or potential loss of Fission Product Barrier Thresholds. EAL matrices should be read from left to right, from General Emergency to Unusual Event, and top to bottom. Declaration decisions should be independently verified before declaration is made except when gaining this verification would exceed the 15 minute declaration requirement. Place keeping should be used on all EAL matrices. 3.1.1 Classification Timeliness NRG regulations require the licensee to establish and maintain the capability to assess, classify, and declare an emergency condition within 15 minutes after the availability of indications to plant operators that an emergency action level has been exceeded and to promptly declare the emergency condition as soon as possible following identification of the appropriate emergency classification level. The NRG staff has provided guidance on implementing this requirement in NSIR/DPR-ISG-01, "Interim Staff Guidance, Emergency Planning for Nuclear Power Plants" (ref. 4.1.8). 3.1.2 Valid Indications All emergency classification assessments shall 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, verification could be accomplished through an instrument channel check, response on related or redundant indicators, or direct observation by plant personnel. An indication, report, or condition is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, *or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. 3.1.3 Imminent Conditions For ICs and EALs that have a stipulated time duration (e.g., 15 minutes, 30 minutes, etc.), the Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time. If an ongoing radiological release is detected and the release start time is

unknown, it should be assumed that the release duration specified in the IC/EAL has been exceeded, absent data to the contrary.)

Page 9 of 278 -::::==* Entergy River Bend Station EAL Basis Document Revision XXX 3.1.4 Planned vs. Unplanned Events A planned work activity that res_ults in an expected event or condition which meets or exceeds an EAL does not warrant an emergency declaration provided that: 1) the activity proceeds as planned, and 2) the plant remains within the limits imposed by the operating license. Such activities include planned work to test, manipulate, repair, maintain or modify a system or component. In these cases, the controls associated with the planning, preparatia,n and execution of the work will ensure that compliance is maintained with all aspects of the operating license provided that the activity proceeds and concludes as expected. Events or conditions of this type may be subject to the reporting requirements of 10 § CFR 50. 72 (ref. 4.1.4). 3.1.5 Classification Based on Analysis 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.). For these.EALs, the EAL wording or the associated basis discussion will identify the necessary analysis. In these cases, the 15-minute declaration period starts with the availability of the analysis results that show the threshold to be exceeded (i.e., this is the time that the EAL information is first available). The NRC expects licensees to establish the capability to initiate and complete EAL-related analyses within a reasonable period of time (e.g., maintain the necessary expertise on-shift). 3.1.6 Emergency Director Judgment 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 EAL 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 in the Fission Product Barrier Tables; judgment may be used to determine the status of a fission product barrier. 3.2 Classification Methodology To make an emergency classification, the user will compare an event or condition (i.e., th'e relevant plant indications and reports) to an EAL(s) and determine if the EAL has been met or exceeded. The evaluation of an EAL must be consistent with the related Operating Mode Applicability and Notes. If an EAL has been met or exceeded, the associated IC is likewlse met, the emergency classification process "clock" starts, and the ECL must be declared in accordance with plant procedures no later than fifteen minutes after the process "clock" started. When assessing an EAL that specifies a time duration for the off-normal condition, the "clock" for the EAL time duration runs concurrently with the emergency classification process "clock." For a full discussion of this timing requirement, refer to NSIR/DPR-ISG-01 (ref. 4.1.8). Page 10 of 278 -===-Entergy River Bend Station EAL Basis Document Revision XXX 3.2.1 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 units, a Site Area Emergency should be declared.

There is no "additive" effect from multiple EALs meeting the same ECL. For example:

If two Alert EALs are met, whether at one unit or at two units, an Alert should be declared.

3.2.2 Consideration of Mode Changes During Classification The mode in effect at the time that an event or condition occurred, and prior to any plant or operator response, is the mode that determines whether or not an IC is applicable. If an event or condition occurs, and results in a mode change before the emergency is declared, the emergency classification level is still based on the mode that existed at the time that the event or condition was initiated (and not when it was declared). Once a different mode is reached, any new event or condition, not related to the original event or condition, requiring emergency classification should be evaluated against the ICs and EALs applicable to the operating mode at the time of the new event or condition. For events that occur in Cold Shutdown or Refueling; escalation is via EALs that are applicable in the Cold Shutdown or Refueling modes, even if Hot Shutdown (or a high~r mode) is entered during the subsequent plant response. In particular, the fission product barrier EALs are applicable only to events that initiate in the Hot Shutdown mode or higher. 3.2.3 Classification of Imminent Conditions* Although EALs provide specific thresholds, the Emergency Director must remain alert to events or conditions that could lead to meeting or exceeding an EAL within a relatively short period of time (i.e., a change in the ECL is IMMINENT). If, in the judgment of the Emergency Director, meeting an EAL is IMMINENT, the emergency classification should be made as if the EAL has been met. While applicable to all emergency classification levels, this approach is particularly important at the higher emergency classification levels since it provides additional time for implementation of protective measures. 3.2.4 .Emergency Classification Level Upgrading and Termination An ECL may be terminated when the event or condition that meets the classified IC and EAL no longer exists, and other site-specific termination requirements are met. 3.2.5 Classification of Short-Lived Events Event-based I Cs and EALs define a variety of specific occurrences that have potential or actual safety significance. By their nature, some of these events may be short-lived and, thus, over before the emergency classification assessment can be completed. If an event occurs that meets or exceeds an EAL, the associated ECL must be declared regardless of its Page 11 of 278 Q -::::~ Entergy River Bend Station EAL B.asis Document Revision XXX continued presence at the time of declaration. Examples of such events include an earthquake or a failure of the reactor protection system to automatically scram the reactor followed by a successful manual scram. 3.2.6 Classification of Transient Conditions Many of the I Cs and/or EALs employ time-based criteria. These criteria will require that the IC/EAL conditions be present for a defined period of time before an emergency declaration is warranted. In cases where no time-based criterion is specified, it is recognized that some transient conditions may cause an EAL to be met for a brief period of time (e.g., a few seconds to a few minutes). The following guidance should be applied to the classification of these conditions.

EAL momentarily met during expected plant response -In instances in which 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 ATWS occurs and the high pressure ECCS systems fail to automatically start. The plant enters an inadequate core cooling condition (a potential loss of both the Fuel Clad and RCS Barriers). If an operator manually starts a high pressure ECCS system in accordance with an EOP step and clears the inadequate core cooling 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 (process clock) 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 when an operator is able to take a successful corrective action prior to the Emergency Director completing the review and steps necessary to make the emergency declaration. This provision is included to ensure that any public protective actions resulting from the emergency classification are truly warranted by the plant conditions. Page 12 of 278 ~-Entergy River Bend Station EAL Basis Document Revision XXX 3.2. 7 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 (ref. 4.1.3) is applicable. Specifically, the event should be reported to the NRC in accordance with 10 CFR § 50. 72 (ref. 4.1.4) 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. 3.2.8 Retraction of an Emergency Declaration Guidance on the retraction of an emergency declaration reported to the NRC is discussed in NUREG-1022 (ref. 4.1.3). Page 13 of 278 ' -:::::::-Entergy River Bend Station EAL Basis Document Revision XXX

4.0 REFERENCES

4.1 Developmental 4.1.1 NEI 99-01 Revision 6, Methodology for the Development of Emergency Action Levels for Non-Passive Reactors, ADAMS Accession Number ML 12326A805 4.1.2 RIS 2007-02 Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events, February 2, 2007. 4.1.3 NUREG-1022 Event Reporting Guidelines: 1 OCFR50. 72 and 50. 73 4.1.4 10 § CFR 50. 72 Immediate Notification Requirements for Operating Nuclear Power Reactors 4.1.5 10 § CFR 50. 73 License Event Report System 4.1.6 RBS Technical Specifications Table 1.1-1, Modes 4.1. 7 RBS USAR Section 2.1 Geography and Demography 4.1.8 NSIR/DPR-ISG-01 Interim Staff Guidance, Emergency Planning for Nuclear Power Plants 4.1.9 RBS Emergency Plan 4.1.10 RBS USAR 9.1.2.5 Holtec HI-STORM Dry Spent Fuel Storage System 4.1.11 RBS USAR 9.1.4.2.3.11 Fuel Transfer System 4.1.12 OSP-0037 Shutdown Operations Protection Plan (SOPP) 4.1.13 RBS Security Plan 4.2 Implementing 4.2.1 EIP-2-001 Classification of Emergencies 4.2.2 NEI 99-01 Rev. 6 to RBS EAL Comparison Matrix 4.2.3 RBS EAL Matrix Page 14 of 278 River Bend Station EAL Basis Document Revision XXX 5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS 5.1 Definitions (ref. 4.1.1 except as noted) Selected terms used in Initiating Condition and Emergency Action Level statements are set in all capital letters (e.g., ALL CAPS). These words are defined terms that have specific meanings as used in this document. The definitions of these terms are provided below. 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 small fractions of the EPA Protective Action Guideline exposure levels. Confinement Boundary The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the RBS ISFSI, Confinement Boundary is defined as the Holtec System Multi-Purpose Canister (MPC) (ref. 4.1 :10). ' Containment Closure The actions taken to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. Containment Closure is established when the containment requirements of OSP-0037 (ref. 4.1.12) are met and at least one integral barrier to the release of radioactive material is provided, within the specified limits using STP-057-3804. Emergency Action Level (EAL) A pre-determined, site-specific, observable threshold for an Initiating Condition that, when met or exceeded, places the plant in a given emergency classification level. Emergency Classification Level (ECL) One of a set of names or titles established by the US Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions according to (1) potential or actual effects or consequences, and (2) resulting onsite and offsite response actions. The emergency classification levels, in ascending order of severity, are:

Unusual Event (UE)
Alert
Site Area Emergency (SAE)
General Emergency (GE) Page 15 of 278
-===-Entergy River Bend Station EAL Basis Document Revision XXX 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 require a post-event inspection to determine if the attributes of an explosion are present. Fire Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed. Fission Product Barrier Threshold A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier. Flooding A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area. General Emergency Events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area. Hostage A person(s) held as leverage against the station to-ensure that demands will be met by the station. Hostile Action c_ An act toward RBS or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on RBS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA). Hostile Force One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction. Page 16 of 278 -===' Entergy River Bend Station'EAL Basis Document Revision XXX Imminent The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. lmpede(d) Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed). Independent Spent Fuel Storage Installation (ISFSI) A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage.

Initiating Condition (IC) An event or condition that aligns with the definition of one of the four emergency classification levels by virtue of the potential or actual effects or consequences.

Owner Controlled Area (OCA) I

For the purposes of classification, the Security Owner Controlled Area (SOCA) or the area between the SOCA Fence and the PROTECTED AREA Boundary (ref 4.1.13). Projectile An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety. Protected Area The area within the perimeter of the RBS security fence. (ref. 4.1.9). Refueling Pathway Reactor cavity (well), containment spent fuel pool, fuel transfer canal, and fuel building spent fuel pools, but not including_

the reactor vessel, comprise the refueling pathway (ref. 4.1.11 ). Restore Take the appropriate action required to return the value of an identified parameter to the applicable limits. Safety System A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as related (as defined in 1 OCFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: ( 1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; Page 17 of 278 River Bend Station EAL Basis Document Revision XXX (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Security Condition Any security event as listed in the approved security confingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A security condition does not involve a HOSTILE ACTION. Site Area Emergency Events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; (1) toward site personnel or equipment that could lead to the likely failure of or; (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA PAG exposure levels beyond the SITE BOUNDARY. Site Boundary For classification and dose projection purposes, the Site Boundary is the area defined as the exclusion area or.exclusion zone in 10CFR100.3 (a) which is a boundary of approximately 3,000 feet (or 0.5748 mile) from the RBS reactor centerline. (ref. 4.1.7) Unisolable An open or breached system line that cannot be isolated, remotely or locally. Unplanned A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Unusual Event Events are in progress or have occurred which indicate a potential degradation in the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of SAFETY SYSTEMS occurs. Valid An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. Page 18 of 278 Entergy River Bend Station EAL Basis Document Revision XXX Visible Damage Damage to a SAFETY SYSTEM train that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected SAFETY SYSTEM train. Page 19 of 278

-:::::::-

Entergy River Bend Station EAL Basis Document Revision XXX 5.2 Abbreviations/Acro,nyms ~F ....................................................................................................... Degrees Fahrenheit 0 ****"**********:*********************************************************************************************************** Degrees AB ........................................................................................................... Auxiliary Building AC ........................................................................................................ Alternating Current AOP ................................................................................. Abnormal Operating Procedure APRM .-................................................................................. Average Power Range Meter A TWS ...................................................................... Anticipated Transient Without Scram BWR ................................................................................................ Boiling Water Reactor BWR0(3 .................................................................. Boiling Water Reactor Owners Group COE ....................................................................................... Committed Dose Equivalent CFR ..................................................................................... Code of Federal Regulations CNB .............................................................................................................. Containment Barrier CS ...............

..............................................................................................................

Core Spray CTMT ..... :* ....................................................................................................... Containment DEF ..................................................................................................................... Defueled OBA ............................................................................................... Design Basis Accident DC ............................................................................................................... Direct Current DIG ......................................................................................................... Diesel Generator ORMS ....................................................................... Digital Radiation Monitoring System EAL ............................................................................................. Emergency Action Level ECCS ......................................................................

..... Emergency Core Cooling System ECL ..................................................................................

Emergency Classifipation Level EOF .: ................................................................................ Emergency Operations Facility EOP ............................................................................... Emergency Operating Procedure EPA .............................................................................. Environmental Protection Agency EPG ............................................................................... Emergency Procedure Guideline EPP ....................................................................................... Emergency Plan Procedure ERO ............................

..............................................

Emergency Response Organization ESF .....................................................

...................................

Engineered Safety Feature FAA .........................................................

........................

Federal Aviation Administration FBI ............................... 1 ................................................... Federal Bureau of Investigation I FEMA. .................................................. _ ............ Federal Emergency Management Agency FSAR ........ '. ........................................................................... Final Safety Analysis Report GE ..................................... _ ................................................................ General Emergency HCTL ............................................................................ Heat Capacity Temperature Limit Page 20 of 278 River Bend Station EAL Basis Document Revision XXX HPCS ....................................................................................... High Pressure Core Spray IC ......................................................................................................... 1.lnitiating Condition IPEEE ................. Individual Plant Examination of External Events (Generic Letter 88-20) ISFSI. ........................................................... Independent Spent Fuel Storage Installation Kett ......................................................................... Effective Neutron Multiplication Factor LCO .................................................................................. Limiting Condition of Operation LER ................................................................................................ Licensee Event Report LOCA ......................................................................................... Loss of Coolant Accident LPCS ..................... , ........................................................ , ......... Low Pressure Core Spray LRW ........................................................................................................ Liquid Radwaste LWR ................................................................................................... Light Water Reactor MPC ................. .-................. Maximum Permissible Concentration/Multi-Purpose Canister MPH ........................................................................................................... Miles Per Hour mR, mRem, mrem, mREM .............................................. mi Iii-Roentgen Equivalent Man MSCRWL. ........................................................ Minimum Steam Colling RPV Water Level MSIV ....................................................................................... Main Steam Isolation Valve MSL ........................................................................................................ Main Steam Line MW .................................................................................................................... Megawatt NEI. .............................

.................................................................

Nuclear Energy Institute NEIC ..........................

........................................

National Earthquake Information Center NESP ................................................................... National Environmental Studies Project NORAD ................................................... North American Aerospace Defense Command (NO)UE ......................................................... '. ...................... Notification of Unusual Event NPP .................................................................................................. Nuclear .Power Plant NRC ................................................................................ Nuclear Regulatory Commission NSSS ................................................................................ Nuclear Steam Supply System OBE ...................................................................................... Operating Basis.,Earthquake ODCM ............................................................................. Offsite Dose Calculation Manual ORO ................................................................................. Offsite Response Organization PA ...................

...............................................................................

.". .......... Protected Area PAG ............................................................................ ........... Protective Action Guideline PB .................................................................................................................... Pushbutton PCIS ..................................................................... Primary Containment Isolation System PRA/PSA ..................... Probabilistic Risk Assessment / Probabilistic Safety Assessment PSID ......................................................................... Pounds Per Square Inch Differential PSIG ...................................................................... '. ......... Pounds per Square Inch Gauge Page 21 of 278 Q . *::::=*Entergy River Bend Station EAL Basis Document Revision XXX R ........................................................................................................................ Roentgen RCB ..........

.....................................................................................................

RCS Barrier RCIC ................................................................................. Reactor Core Isolation Cooling RCS ............................................................................................ Reactor Coolant System Rem, rem, REM ....................................................................... Roentgen Equivalent Man (' RETS ......................................................... Radiological Effluent Technical Specifications RHR ............................................................................................. Residual Heat Removal RPS ........................................................................................ Reactor Protection System RPV ........................................................................................... Reactor Pressure Vessel RWCU .......................................................................................... Reactor Water Cleanup SAG ......................................................................................... Severe Accident Guideline SAP ....................................................................................... Severe Accident Procedure SAR ............................................................................................... Safety Analysis Report SBO ......................................................................................................... Station Blackout SCBA ....................................................................... Self-Contained Breathing Apparatus SOCA .............................................................................. Security Owner Controlled Area SPDS ............................................................................ Safety Parameter Display System SRO ............................................................................................ Senior Reactor Operator SRV ..................................................................................................... Safety Relief Valve SSE .............................................................................

.........

Safe Shutdown Earthquake TEDE ............................................................................... Total Effective Dose Equivalent TAF ....................................................................................................... Top of Active Fuel TSC .......................................................................................... Technical Support Center UFSAR ................................................................... Updated Final Safety Analysis Report USGS ........................................................... , ................ United States Geological Survey Page 22 of 278

--Entergy River Bend Station EAL Basis Document Revision XXX 6.0 RBS-TO-NEI 99-01 Rev. 6 EAL CROSS-REFERENCE This cross-reference is provided to facilitate association and location of an RBS EAL within the NEI 99-01 IC/EAL identification scheme. Further information regarding the development of the RBS EALs based on the NEI guidance can be found in the EAL Comparison Matrix. RBS NEI 99-01 Rev. 6 EAL IC Example EAL AU1.1 AU1 1, 2 AU1.2 AU1 3 AU2.1 AU2 1 AA1.1 AA1 1 AA1.2 AA1 2 AA1.3 AA1 3 AA1.4 AA1 4 AA2.1 AA2 1 AA2.2 AA2 2 AA2.3 AA2 3 AA3.1 AA3 1 AA3.2 AA3 2 AS1.1 AS1 1 AS1.2 AS1 2 AS1.3 AS1 3 AS2.1 AS2 1 AG1.1 AG1 1 AG1.2 AG1 2 AG1.3 AG1 3 Page 23 of 278 I

~==-Entergy River Bend Station EAL Basis Document Revision XXX RBS NEI 99-01 Rev. 6 EAL IC Example EAL AG2.1 AG2 1 CU1.1 CU1 1 CU1.2 CU1 2 CU2.1 CU2 1 CU3.1 CU3 1 CU3.2 CU3 2 CU4.1 CU4 1 CU5.1 CU5 1,2,3 CA1.1 CA1 1 CA1.2 CA1 2 CA2.1 CA2 1 CA3.1 CA3 1, 2 CA6.1 CA6 1 CS1.1 CS1 1 CS1.2 CS1 2 CS1.3 CS1 3 CG1.1 CG1 1 CG1.2 CG1 2 EU1.1 EU1 1 FA1.1 FA1 1 FS1 .1 FS1 1 FG1.1 FG1 1 HU1.1 HU1 1, 2, 3 Page 24 of 278 --Entergy River Bend Station EAL Basis Document Revision XXX RBS NEI 99-01 Rev. 6 EAL IC Example EAL HU2.1 HU2 1 HU3.1 HU3 1 HU3.2 HU3 2 , HU3.3 HU3 3 HU3.4 HU3 4 HU4.1 HU4 1 HU4.2 HU4 2 HU4.3 HU4 3 HU4.4 HU4 4 HU7.1 HU? 1 HA1.1 HA1 1, 2 HA5.1 HAS 1 HA6.1 HA6 1 HA7.1 HA? 1 HS1.1 HS1 1 HS6.1 HS6 1 HS7.1 HS7 1 J HG7.1 HG? 1 SU1.1 SU1 1 SU3.1 SU2 1 SU4.1 SU3 1 SU4.2 SU3 2 SU5.1 SU4 1, 2, 3 Page 25 of 278 -::::=* Entergy River Bend Station EAL Basis Document Revision XXX RBS NEI 99-01 Rev. 6 EAL IC Example EAL SU6.1 SU5 1 SU6.2 SU5 2 SU7.1 SU6 1, 2, 3 N/A SU? 1, 2 SA1.1 SA1 1 SA3.1 SA2 1 SA6.1 SA5 1 SA8.1 SA9 1 SS1.1 SS1 1 SS2.1 SS8 1 SS6.1 SS5 1 SG1.1 SG1 1 SG1.2 SGS 1 Page 26 of 278

===-Entergy River Bend Station EAL Basis Document Revision XXX 7.0 ATTACHMENTS 7.1 Attachment 1, Emergency Action Level Technical Bases 7.2 Attachment 2, Safe Op~ration

& Shutdown Areas Tables A-3 & H-2 Bases Page 27 of 278 River Bend Station EAL Basis DocumentRevision XXX -Attachment 1 -Emergency Action Level Technical Bases Category A-Abnormal Rad Levels / Rad Effluent EAL Group: ANY (EALs in this category are applicable to / any plant condition, hot or cold.) Many EALs are based on actual or potential degradation of fission product barriers because of the elevated potential for offsite radioactivity release. Degradation of fission product' barriers though is not always apparent via non-radiological symptoms. Therefore, direct indication of elevated radiological effluents or area radiation levels are appropriate symptoms for emergency classification. At lower levels, abnormal radioactivity releases may be indicative of a failure of containment systems .or precursors to more significant releases. At higher release rates, offsite radiological conditions may result which require offsite protective actions. Elevated area radiation level~ in plant may also be indicative of the failure of containment systems or preclude access to plant vital equipment necessary to ensure plant safety. Events of this category pertain to the following subcategories:

1. Radiological Effluent I Direct indication of effluent radiation monitoring systems provides a rapid assessment mechanism to determine releases in excess of classifiable limit~. Projected offsite doses, actual offsite field measurements or measured release rates via sampling indicate doses or dose rates above classifiable limits. 2. Irradiated Fuel Event Conditions indicative of a loss of adequate shielding or damage to irradiated fuel may preclude access to vital plant areas or result in radiological releases that warrant emergency classification.*

3. Area Radiation Levels Sustained general area radiation levels which may preclude access to areas requiring continuous occupancy also warrant emergency classification.

Page 28 of 278 ~Entergy River Bend Station EAL Basis Document Revision XXX Category: Subcategory: Attachment 1 -Emergency Action Level Technical Bases A -Abnormal Rad Levels / Rad Effluent 1 -Radiological Effluent Initiating Condition:

Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer EAL: AU1.1 Unusual Event Reading on any Table A-1 effluent radiation monitor> column "UE" for 60 min. , (Notes 1, 2, 3) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.

The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is e~ceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Note 3: 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. Table A-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE Main Plant Vent -Primary RE125 9.56E+08 µCi/sec 9.56E+07 µCi/sec 9.63E+06 µCi/sec 1.01 E+05 µCi/sec Main Plant Vent -RE126 1.66E-01 µCi/ml 1.74E 0 03 µCi/ml Secondary

II) Fuel Bldg Vent -Primary RE5A 7.75E+08 µCi/sec 7.75E+07 µCi/sec 7.75E+06 µCi/sec 6.50E+03 µCi/sec :, 0 Q) II) Fuel Bldg Vent -Secondary RE5B 1.72E-01 µCi/ml 1.38E-03 µCi/ml cu --------c., Radwaste Bldg Vent -RE6A 8.03E+08 µCi/sec 8.03E+07 µCi/sec 8.03E+06 µCi/sec 6.96E+04 µCi/sec Primary Radwaste Bldg Vent -RE6B --------2.12E-01 µCi/ml 1. 71 E-04 µCi/ml Secondary

!! :, Liquid Radwaste RE107 ------------2 x Alarm Setpoint CT :::i Page 29 of 278 -::::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases _ Mode Applicability: All Definition(s): VALID -An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt rel_ated to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. Basis: This IC addresses a potential reduction in the level of safety of the plant as indicated by a level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared. Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALi D for classification purposes. Releases should not be prorated or averaged. For ex,ample, a release exceeding 4 times release limits for 30 minutes does not meet the EAL. This EAL addresses normally occurdng continuous radioactivity releases from monitored gaseous or liquid effluent pathways as well as radioactivity releases that cause effluent radiation monitor readings to exceed 2 times the limit established by a radioactivity discharge permit. Such, releases are typically associated with planned batch releases from non-* continuous release pathways (e.g., radwaste, waste gas). Escalation of the emergency classification level would be via IC AA 1. Page 30 of 278 ... , / River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Reference(s):

1. SOP-0086 Digital Radiation Monitoring System 2. RSP-0008 Offsite Dose Calculation Manual 3. EP-CALC-RBS~1801 Radiological Effluent EAL Threshold Values 4. NEI 99-01 AU1 Page 31 of 278

~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: A -Abnormal Rad Levels / Rad Effluent Subcategory: 1 -Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer. EAL: AU1.2 Unusual Event Sample analysis for a gaseous or liquid release indicates a concentration or release rate > 2 x ODCM limits for~ 60 min. (Notes 1, 2) Note 1: Note 2: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. ' If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. I Mode Applicability: All Definition(s): None Basis: This IC addresses a potential reduction in the level of safety of the plant as indicated by a level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid raaiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared. Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL. Page 32 of 278 8 . -::::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases This EAL addresses uncontrolled gaseous or liquid releases that are detected by sample 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 1. Reference(s):

1. RSP-0008 Offsite Dose Calculation Manual 2. NEI 99-01 AU1 Page 33 of 278 Q ~E -::==-n t ergy River Bend Station EAL Basis Document.

Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: A -Abnormal Rad Levels/ Rad Effluent Subcategory: 1 -Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid COE EAL: AA1.1 Alert Reading on any Table A-1 effluent radiation monitor> column "ALERT" for 15 min. (Notes 1, 2, 3, 4) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceede:id. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Note 3: 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. Note 4 The pre-calculated effluent monitor values presented in EALs AA 1.1, AS 1.1 and AG1 .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Table A-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE Main Plant Vent -Primary RE125 9.56E+08 µCi/sec 9.56E+07 µCi/sec 9.63E+06 µCi/sec 1.01 E+05 µCi/sec Main Plant Vent -RE126 1.66E-01 µCi/ml 1.74E-03 µCi/ml Secondary

1/) Fuel Bldg Vent -Primary RE5A 7.75E+08 µCi/sec 7.75E+07 µCi/sec 7.75E+06 µCi/sec 6.50E+03 µCi/sec ::s 0 Cl) 1/) Fuel Bldg Vent -Secondary RE5B 1.72E-01 µCi/ml 1.38E-03 µCi/ml ns --------(!) Radwaste Bldg Vent -RE6A 8.03E+08 µCi/sec 8.03E+07 µCi/sec 8.03E+06 µCi/sec 6.96E+04 µCi/sec Primary Radwaste Bldg Vent -RE6B --------2.12E-01 µCi/ml 1.71 E-04 µCi/ml Secondary

E ::s Liquid Radwaste RE107 C" ------------2 x Alarm Setpoint :J Page 34 of 278

~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Mode Appli_cability: All Definition(s): VALID -An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. Basis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual . offsite doses greater than or equal to 1 % of the EPA Protective Action Guides (PAGs). It . . includes both monitored and un-monitored releases. Releases of this magnitude represent an I actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release). 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 anc:i conditions. The TEDE dose is set at 1 % of the EPA PAG of 1,000 mrem while the 50 mrem thyroid'CDE was established in consideration of the 1 :5 ratio of the EPA PAG for TEDE and thyroid COE. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for:_ classification purposes.

Escalation of the emergency classification level would be via IC AS 1. Reference(s):

1. SOP-0086 Digital Radiation Monitoring System 2. RSP-0008 Offsite Dose Calculation Manual 3. EP-CALC-RBS-1801 Radiological Effluent EAL Threshold Values 4. NEI 99-01 AA1 Page 35 of 278

~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: A -Abnormal Rad Levels / Rad Effluent Subcategory: 1 -Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid COE EAL: AA1.2 Alert Dose assessment using actual meteorology indicates doses > 10 mrem TEDE or 50 mrem thyroid COE at or beyond the SITE BOUNDARY (Note 4) Note 4: The pre-calculated effluent mqnitor values presented in EALs AA 1.1, AS 1.1 and AG 1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Mode Applicability: All -Definition(s): SITE BOUNDARY -For classification and dose projection purposes, the Site Boundary is the area defined as the exclusion area or exclusion zone in 1OCFR100.3 (a) which is a boundary of approximately 3,000 feet (or 0.5748 mile) from the RBS reactor centerline. Basis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1 % of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release). 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 event~ and conditions. The TEDE dose is set at 1 % of the EPA PAG of 1,000 mrem while the 50 mrem thyroid COE was established in consideration of the 1 :5 ratio of the EPA PAG for TEDE and thyroid COE. Escalation of the emergency classification level would be via IC AS1. Reference(s):

1.

EIP-2-024 Offsite Dose Calculations

2. NEI 99-01 AA1 Page 36 of 278 I

~-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: A -Abnormal Rad Levels / Rad Effluent Subcategory: 1 -Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid COE EAL: AA1.3 Alert Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses > 10 mrem TEDE or 50 mrem thyroid COE at or beyond the SITE BOUNDARY for 60 min. of exposure (Notes 1, 2) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability: All Definition(s): SITE BOUNDARY -For classification and dose projection purposes, the Site Boundary is the area defined as the exclusion area or exclusion zone in 1OCFR100.3 (a) which is a boundary of approximately 3,000 feet (or 0.5748 mile) from the RBS reactor centerline. Basis: ) This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1 % of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release). Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 1 % of the EPA PAG of 1,000 mrem while the 50 mrem thyroid COE was established in consideration of the 1 :5 ratio of the EPA PAG for TEDE and thyroid COE. This EAL is assessed per the ODCM (ref. 2) I Escalation of the emergency classification level would be via IC AS1. Page 37 of 278 -::::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Reference(s):

1. EIP-2-024 Offsite Dose Calculations

2. RSP-0008 Offsite Dose Calculation Manual 3. NEI 99-01 AA1 Page 38 of 278

-:;::!:'" Entergy River Bend Station EAL Basis Docur:nent Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: A -Abnormal Rad Levels/ Rad Effluent Subcategory: 1 -Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid COE EAL: AA1.4 Alert Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:

Closed window dose rates > 10 mR/hr expected to continue for~ 60 min.
Analys*es of field survey samples indicate thyroid COE> 50 mrem for 60 min. of inhalation.

' (Notes 1, 2) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability: All Definition(s): SITE BOUNDARY -For classification and dose projection purposes, the Site Boundary is the area defined as the exclusion area or exclusion zone in 1OCFR100.3 (a) which is a boundary of approximately 3,000 feet (or 0.5748 mile) from the RBS reactor centerline. Basis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1 % of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release). Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. '--The TEDE dose is set at 1 % of the EPA PAG of 1,000 mrem while the 50 mrem thyroid COE was established in consideration of the 1 :5 ratio of the EPA PAG for TEDE and thyroid COE. Page 39 of 278

=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Escalation of the emergency classification level would be via IC AS1. Reference(s):

1. EIP-2-014 Offsite Radiological Monitoring

2. NEI ~9-01 AA1 Page 40 of 278 Q *::::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

A -Abnormal Rad Levels / Rad Effluent Subcategory: 1 -Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid COE EAL: AS1.1 Site Area Emergency Reading on any Table A-1 effluent radiation monitor> column "SAE"_ for 15 min. (Notes 1, 2, 3, 4) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Note 3: 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. Note 4: The pre-calculated effluent monitor values presented in EALs AA 1.1, AS 1.1 and AG 1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Table A-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE Main Plant Vent -Primary RE125 9.56E+08 µCi/sec 9.56E+07 µCi/sec 9.63E+06 µCi/sec 1.01 E+05 µCi/sec Main Plant Vent -RE126 1.66E-01 µCi/ml 1.74E-03 µCi/ml Secondary

U) Fuel Bldg Vent -Primary RE5A 7.75E+08 µCi/sec 7.75E+07 µCi/sec 7.75E+06 µCi/sec 6.50E+03 µCi/sec :, 0 QI U) Fuel Bldg Vent -Secondary RE5B 1.72E-01 µCi/ml 1.38E-03 µCi/ml Cll --------(!) Radwaste Bldg Vent -RE6A 8.03E+08 µCi/sec 8.03E+07 µCi/sec 8.03E+06 µCi/sec 6.96E+04 µCi/sec Primary Radwaste Bldg Vent -RE6B --------2.12E-01 µCi/ml 1.71 E-04 µCi/ml Secondary

E :, Liquid Radwaste RE107 ------------2 x Alarm Setpoint C" :J Page 41 of 278

-::::::" Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Mode Applicability: All Definition(s): VALID -An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public. 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 COE was established in consideration of the 1 :5 ratio of the EPA PAG for TEDE and thyroid COE. Classification based on effluent monitor readings assumes that a release path to the environment is established.

If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. Escalation of the emergency classification level would be via IC AG1. Reference(s):

1. SOP-0086 Digital Radiation Monitoring System 2. RSP-0008 Offsite Dose Calculation Manual 3. EP-CALC-RBS-1801 Radiological Effluent EAL Threshold Values 4. NEI 99-01 AS1 Page 42 of 278

-:-.::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: A -Abnormal Rad Levels / Rad Effluent Subcategory: 1 -Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem T~DE or 500 mrem thyroid COE EAL: AS1.2 Site Area Emergency Dose assessment using actual meteorology indicates doses> 100 mrem TEDE or 500 mrem thyroid COE at or beyond the SITE BOUNDARY (Note 4) ' Note 4: The pre-calculated effluent monitor values presented in EALs AA 1.1, AS1 .1 and AG1 .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Mode Applicability: All Definition(s): SITE BOUNDARY -For classification and dose projection purposes, the Site Boundary is the area defined as the exclusion area or exclusion zone in 10CFR100.3 (a) which is a boundary of approximately 3,000 feet (or 0.5748 mile) from the RBS reactor centerline. Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public. 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 COE was established in consideration of the 1 :5 ratio of the EPA PAG for TEDE and thyroid COE. Escalation of the emergency classification level would be via IC AG1. Page 43 of 278 -::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Reference(s):

1. EIP-2-024 Offsite Dose Calculations

2. NEI 99-01 AS1 Page 44 of 278

-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

A-Abnormal Rad Levels/ Rad Effluent Subcategory: 1 -Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose g[eater than 100 mrem TEDE or 500 mrem thyroid COE EAL: , AS1 .3 Site Area Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:

Closed window dose rates> 100 mR/hr expected to continue for~ 60 min.
Analyses of field survey samples indicate thyroid COE> 500 mrem for 60 min. of inhalation. (Notes 1, 2) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.

The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. ) Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability: All Definition(s): \ SITE BOUNDARY -For classification and dose projection purposes, the Site Boundary is the area defined as the exclusion area or exclusion zone in 1OCFR100.3 (a) which is a boundary of approximately 3,000 feet (or 0.5748 mile) from the RBS reactor centerline. Basis: -This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with

the failure of plant systems needed for the protection of the public. 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 COE was established in consideration of the 1 :5 ratio of the EPA PAG for TEDE and thyroid COE. I . Escalation of the emergency classification level would be via IC AG1. Page 45 of 278

River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Reference(s}:

1. EIP-2-014 Offsite Radiological Monitoring

2. NEI 99-01 AS1 Page 46 of 278 I L --::::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

A -Abnormal Rad Levels / Rad Effluent Subcategory: 1 -Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid COE EAL: AG1 .1 General Emergency Reading on any Table A-1 effluent radiation monitor> column "GE" for 15 min. (Notes 1, 2, 3, 4) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Note 3: 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. Note 4: The pre-calculated effluent monitor values presented in EALs AA 1.1, AS 1.1 and AG 1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Table A-1 Effluent Monitor Classification Thresholds Release Point I Monitor I GE I SAE I Alert I UE Main Plant Vent -Primary RE125 9.56E+08 µCi/sec 9.56E+07 µCi/sec 9.63E+06 µCi/sec 1.01 E+05 µCi/sec Main Plant Vent -RE126 1.66E-01 µCi/ml 1.74E-03 µCi/ml Secondary

en Fuel Bldg Vent -Primary RE5A 7.75E+08 µCi/sec 7.75E+07 µCi/sec 7.75E+06 µCi/sec 6.50E+03 µCi/sec :::i 0 Q) en Fuel Bldg Vent -Secondary RE5B 1.72E-01 µCi/ml 1.38E-03 µCi/ml C"CI --------(!) Radwaste Bldg Vent -RE6A 8.03E+08 µCi/sec 8.03E+07 µCi/sec 8.03E+06 µCi/sec 6.96E+04 µCi/sec Primary Radwaste Bldg Vent -RE6B --------2.12E-01 µCi/ml 1. 71 E-04 µCi/ml Secondary

!:! :::i Liquid Radwaste RE107 ----------2 x Alarm Setpoint C" :J Page 4 7 of 278 I

-~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Mode Applicability: All Definition(s): VALID -An indication, report, or condition, is considered to be valid wh~n it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public. 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 mre_m thyroid COE was established in consideration of the 1 :5 ratio of the EPA PA~ for TEDE and thyroid COE. 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 VALi D for classification purposes. Reference(s):

1. SOP-0086 Digital Radiation Monitoring System 2. RSP-0008 Offsite Dose Calculation Manual 3. EP-CALC-RBS-1801 Radiological Effluent EAL Threshold Values 4. NEI 99-01 AG1 Page 48 of 278

. -:;::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: A-Abnormal Rad Levels/ Rad Effluent Subcategory: 1 -Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid COE EAL: AG1.2 General Emergency Dose assessment using actual meteorology indicates doses> 1,000 mrem TEDE or 5,000 mrem thyroid COE at or beyond the SITE BOUNDARY (Note 4) Note 4: The pre-calculated effluent monitor values presented in EALs AA1 .1, AS1 .1 and AG1 .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Mode Applicability:

All Definition(s): SITE BOUNDARY -For classification and dose projection purposes, the Site Boundary is the area defined as the exclusion area or exclusion zone in 1OCFR100.3 (a) which is a boundary of approximately 3,000 feet (or 0.5748 mile) from the RBS reactor centerline. Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public. 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 COE was established in consideration of the 1 :5 ratio of the EPA PAG for TEDE and thyroid COE. Reference(s):

1. EIP-2-024 Offsite Dose Calculations

2. NEI 99-01 AG1 Page 49 of 278 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

A -Abnormal Rad Levels / Rad Effluent Subcategory: 1 -Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid COE EAL: AG1 .3 General Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY: '

Closed window dose rates > 1,000 mR/hr expected to continue for~ 60 min.
Analyses of field survey samples indicate thyroid COE> 5,000 mrem for 60 min. of inhalation. (Notes 1, 2) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.

The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability: All Definition(s): SITE BOUNDARY -For classification and dose projection purposes, the Site Boundary is the area defined as the exclusion area or exclusion zone in 1OCFR100.3 (a) which is a boundary of approximately 3,000 feet (or 0.5748 mile) from the RBS reactor centerline. Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public. 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 COE was established in consideration of the 1 :5 ratio of the EPA PAG for TEDE and thyroid COE. Page 50 of 278 I

*:::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Reference(s):

1. EIP-2-014 Offsite Radiological Monitoring

2. NEI 99-01 AG1 Page 51 of 278

-===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: A -Abnormal Rad Levels / Rad Effluent Subcategory: 2 -Irradiated Fuel Event Initiating Condition: UNPLANNED loss of water level above irradiated fuel EAL: AU2.1 Unusual Event UNPLANNED water level drop in the REFUELING PATHWAY as indicated by level instrumentation, low water level alarm or visual observation. AND UNPLANNED rise in corresponding area radiation levels as indicated by any of the following radiation monitors:

RMS-RE140 Refueling Floor Near North Entrance
RMS-RE141 Refueling Floor Near South Entrance
RMS-RE192 Fuel Building Operating Floor -South
RMS-RE193 Fuel Building Operating Floor -North Mode Applicability:

All'-Definition(s): UNPLANNED-. A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. REFUELING PATHWAY-Reactor cavity (well), containment spent fuel pool, fuel transfer canal, and fuel building spent fuel pools, but not including the reactor vessel, comprise the refueling pathway. Basis: This IC addresses a drop in water level above irradiated fuel sufficient to cause elevated radiation levels. This condition could be a precursor to a more serious event and is also indicative of a minor loss in the ability to control radiation levels within the plant. It is therefore a potential degradation in the level of safety of the plant. A water level drop will be primarily determined by indications from available level instrumentation. Other sources of level indications may include reports from plant personnel (e.g., from a refueling crew) or video camera observations (if available). A significant drop in the water level may also cause a rise in 'th~ radiation levels of adjacent areas that can be detected by monitors in those locations. Page 52 of 278 ~-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases The effects of planned evolutions should be considered. For example, a refueling bridge area radiation monitor reading may rise 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. The following low level alarms on P870 are related to low level in the REFUELING PATHWAY (ref. 4):

SFP low alarm #0111 (H13-P870

/ 56A / E02)

Upper Transfer Pool Low alarm #0336 (H13-P870

/ 56A / E03)

Cask Pool Low alarm #0337 (H13-P870

/ 56A / 003)

Lower Transfer Pool low alarm #0335 (H13-P870

/ 56A / F03)

Rx Bldg Storage Pool low alarm #0112 (H13-P870

/ 56A / H03) 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 AA2. Reference(s): 1 . AOP-0027 Fuel Handling Mishaps 2. USAR 12.3 Table 12.3-1 Area Direct Radiation Monitor Locations

3. USAR 9.1.4.2.3.11 Fuel Transfer System 4. ARP-870-0034 P870-56 Alarm Response 5. NEI 99-01 AU2 Page 53 of 278 /

River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: Subcategory: Initiating Condition: EAL: AA2.1 Alert A -Abnormal Rad Levels / Rad Effluent 2 -Irradiated Fuel Event Significant lowering of water level above, or damage to, irradiated fuel \ IMMINENT uncovery of irradiated fuel in the REFUELING PATHWAY Mode Applicability: All l;)efinition(s ): CONFINEMENT BOUNDARY-The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the RBS ISFSI, Confinement Boundary is defined as the Holtec System Multi-Purpose Canister (MPC).

IMMINENT -The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. REFUELING PATHWAY-Reactor cavity (well), containment spent fuel pool, fuel transfer canal, and fuel building spent fuel pools, but not including the reactor vessel, comprise the refueling pathway. 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 REFUELING PATHWAY. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment.

As such, they represent an actual or potential substantial degradation of the level of safety of the plant. 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 ofthe CONFINEMENT BOUNDARY is classified in accordance with IC EU1. This EAL escalates from AU2.1 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 off curve). Classification of an event using this EAL should be based on the totality of available indications, reports and observations. While an area radiation monitor could detect a rise 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 Page 54 of 278 Q *===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases should be considered in combination with other available indications of inventory loss. A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance with Recognition Category C during the Cold Shutdown and Refueling modes. Escalation of the emergency classification level would be via IC AS1. Reference(s):

1. NEI 99-01 AA2 Page 55 of 278 i I L __ -River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

A -Abnormal Rad Levels / Rad Effluent Subcategory: 2 -Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL: AA2.2 Alert Damage to irradiated fuel resulting in a release of radioactivity AND High alarm on any Table A-2 radiation monitor Table A-2 Fuel Damage Radiation Monitors

RMS-RE140 Refueling Floor Near North Entrance
RMS-RE141 Refueling Floor Near South Entrance
RMS-RE192 Fuel Building Operating Floor -South
RMS-RE193 Fuel Building Operating Floor -North
RMS-RE5A(B)

Fuel Building Ventilation Exhaust Mode Applicability: All Definition(s ): CONFINEMENT BOUNDARY-The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the RBS ISFSI, the Confinement Boundary is comprised the Holtec System Multi-Purpose Canister (MPG). Basis: This EAL addresses events that have caused actual damage to an irradiated fuel assembly. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant. This EAL 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 EU1. This EAL addresses a release of radioactive material caused by mechanical damage to irradiated fuel. Damaging events may include the dropping, bumping or binding of an Page 56 of 278

===* Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases assembly, or dropping a heavy load onto an assembly.

A rise in readings on 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). Escalation of the emergency classification level would be via IC AS1. Reference(s):

1. AOP-0027 Fuel Handling Mishaps 2. USAR 12.3 Table 12.3-1 Area Direct Radiation Monitor Locations

3. USAR 12.3 Table 12.3-2 Airborne Process and Effluent Radiation Monitors 4. NEI 99-01 AA2 Page 57 of 278

=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

A -Abnormal Rad Levels / Rad Effluent Subcategory: 2 -Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL: AA2.3 Alert Lowering of spent fuel pool level to 108.0 ft. (Level 2) on SFC-Ll29A/B Mode Applicability: All Definition(s): None . Basis: This EAL addresses events that have caused a significant lowering of water level within the spent fuel pool. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant. Spent fuel pool water level at this value is within the lower end of the level range necessary to prevent significant dose consequences from direct gamma radiation to personnel performing operations in the vicinity of the spent fuel pool. This condition reflects a significant loss of spent fuel pool water inventory and thus it is also a precursor to a loss of the ability to adequately cool the irradiated fuel assembles stored in the pool. Escalation of the emergency classification level would be via IC AS1 or AS2. Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication (SFC-Ll29A/B) capable of identifying normal level (Level 1 ), SFP level approximately 23 fl. above the top of the fuel racks, (Level 2) 107 ft. 10 5/16 in. (rounded to 108.0 ft.) which is that level adequate to provide substantial radiation shielding for a person standing on the SFP operating deck, and SFP level at the top of the fuel racks (Level 3) 85 fl. 10 5/16 in. (rounded to 86.0 ft.) (ref. 1). RBS uses a Level 3 of approximately one foot above the highest point of any fuel rack providing added margin. Spent Fuel Pool Level indicators SFC-Ll29A and B are read on the 98 fl. elevation Control Building on the interior of Jhe West exterior wall (ref. 2). Page 58 of 278 -:::::::-Entergy Refe re nee( s): River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases -, 1. RBG-47570 Completion of Required Action by NRC Order EA-12-051 Reliable SFP Instrumentation

2. RBS-FSG-011 Alternate Spent Fuel Pool Makeup and Cooling 3. NEI 99-01 AA2 ) Page 59 of 278 River Bend Station EAL Basis Document Revision XXX / Attachment 1 -Emergency Action Level Technical Bases Category:

A -Abnormal Rad Levels / Rad Effluent Subcategory: 2 -Irradiated Fuel Event Initiating Condition: Spent fuel pool level at the top of the fuel racks EAL: AS2.1 Site Area Emergency Lowering of spent fuel pool level to 86.0 ft. (Level 3) on SFC-Ll29A/B Mode Applicability: All Definition(s): IMMINENT -The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. Basis: This EAL addresses a significant loss of spent fuel pool inventory control and makeup capability leading to IMMINENT fuel damage. This condition entails major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.

It is recognized that this IC would likely not be met until well after another Site Area Emergency IC was met; however, it is included to provide classification diversity.

Escalation of the emergency classification level would be via IC AG1 or AG2. Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication (SFC-Ll29A/B) capable of identifying normal level (Level 1 ), SFP level approximately 23 ft. above the top of the fuel racks, (Level 2) 107 ft. 10 5/16 in. (rounded to 108.0 ft.) which is that level adequate to provide substantial radiation shielding for a person standing on the SFP operating deck, and SFP level at the top of the fuel racks (Level 3) 85 ft. 10 5/16 in. (rounded to 86.0 ft.) (ref. 1 ). RBS uses a Level 3 of approximately one foot above the highest point of any fuel rack providing added margin. Spent Fuel Pool Level indicators SFC-Ll29A and B are read on the 98 ft. elevation Control Building on the interior of the West exterior wall (ref. 2).

Reference(s):

1. RBG-47570 Completion of Required Action by NRC Order EA-12-051 Reliable SFP Instrumentation

2. RBS-FSG-011 Alternate Spent Fuel Pool Makeup and Cooling 3. NEI 99-01 AS2 Page 60 of 278 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

A -Abnormal Rad Levels / Rad Effluent Subcategory: 2 -Irradiated Fuel Event Initiating Condition: Spent fuel pool level cannot be restored to at least the top of the fuel racks for 60 minutes or longer EAL: AG2.1 General Emergency Spent fuel pool level cannot be restored to at least 86.0 ft. (Level 3) on SFC-Ll29A/B for 60 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: All Definition(s ): None Basis: This EAL 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. Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication (SFC-Ll29A/B) capable of identifying normal level (Level 1 ), SFP level approximately 23 ft. above the top of the fuel racks, (Level 2) 107 ft. 10 5/16 in. (rounded to 108.0 ft.) which is that level adequate to provide substantial radiation shielding for a person standing on the SFP operating deck, and SFP level at the top of the fuel racks (Level 3) 85 ft. 10 5/16 in. (rounded to 86.0 ft.) (ref. 1 ). RBS uses a Level 3 of approximately one foot above the highest point of any fuel rack providing added margin. Spent Fuel Pool Level indicators SFC-Ll29A and B are read on the 98 ft. elevation Control Building on the interior of the West exterior wall (ref. 2). Reference(s): ,. 1. RBG-47570 Completion of Required Action by NRC Order EA-12-051 Reliable SFP Instrumentation

2. RBS-FSG-011 Alternate Spent Fuel Pool Makeup and Cooling 3. NEI 99-01 AG2 / Page 61 of 278

-::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: A -Abnormal Rad Levels / Rad Effluent Subcategory: 3 -Area Radiation Levels Initiating Condition: Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown EAL: AA3.1 . Alert Dose rate > 15 mR/hr in EITHER of the following areas:

Control Room (RMS-RE170)
Central Alarm Station (by survey) Mode Applicability:

All Definition(s): IMPEDE(D) -Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed). Basis: Areas that meet this threshold include the Control Room (CR) and the Central Alarm Station (CAS). The Control Room is monitored for excessive radiation by one detector, RMS-RE170 (ref. 1 ). The CAS is included in this EAL because of its importance to permitting access to areas required to assure safe plant operations. There are no permanently installed area radiation monitors in CAS that may be used to assess this EAL threshold. Therefore, this threshold is evaluated using local radiation survey for this area. 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 rise in radiation levels and determine if another IC may be applicable. Escalation of the emergency classification level would be via Recognition Category A, C or F ICs. Reference(s):

1. USAR 12.3 Table 12.3-1 ,Area Direct Radiation Monitor Locations

2. NEI 99-01 AA3 Page 62 of 278 I_ ( '"=::='" Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

A -Abnormal Rad Levels / Rad Effluent Subcategory: 3 -Area Radiation Levels Initiating Condition: Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown EAL: AA3.2 Alert An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to any Table A-3 room or area (Note 5) Note 5: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted. Table A-3 Safe Operation & Shutdown Rooms/Areas Room/Area Ayxiliary Building 70' RHR B Pump Room Auxiliary Building 80' RHR A Pump Room Auxiliary Building 114' West Control Building 95' Div 1 RSS Room Mode Applicability: 3 -Hot Shutdown Definition(s): Mode 3 3 3 \ 3 IMPEDE(D) -Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed). 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 rise in radiation levels and determine if another IC may be applicable. Page 63 of 278 -~ Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases For AA3.2, 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 r:,ot contingent upon whether entry js actually necessary at the time of the higher 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 rise occurs, and the procedures used for normal operation, cooldown_

and shutdown do not require entry into the affected room until Mode 3.

The higher 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 access control measures are of a conservative or precautionary nature, and would not actually prevent or IMPEDE a required action. Escalation of the emergency classification level would be via Recognition Category A, C or F ICs. 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. The list of plant rooms or areas with entry-related mode applicability identified specify those 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.

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) are not included. In addition, the list specifies the plant mode(s) during which entry would be required for each room or area (ref. 1 ). EAL AA3.2 mode applicability has been limited to the mode limitations of Table A-3 (Mode 3 only). Reference(s):

1. Attachment 2 Safe Operation

& Shutdown Areas Tables A-3 & H-2 Bases 2. NEI 99-01 AA3 Page 64 of 278 -~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category C -Cold Shutdown / Refueling System Malfunctio'n EAL Group: Cold Conditions (RCS temperature

,; 200°F); EALs in this category are applicable only in one or more cold operating modes. Category C EALs are directly associated with cold shutdown or refueling system safety functions.

Given the variability ofplant configurations (e.g., systems out-of-service for maintenance, containment open, reduced AC power redundancy, time since shutdown) during . these periods, the consequences of any given initiating event can vary greatly. For example, a loss of decay heat removal capability that occurs at the end of an extended outage has less significance than a similar loss occurring during the first week after shutdown. Compounding these events is the likelihood that instrumentation necessary for assessment may also be inoperable. The cold shutdown and refueling system malfunction EALs are based on performance capability to the extent possible with consideration given to RCS integrity, CONTAINMENT CLOSURE, and fuel clad integrity for the applicable operating modes (4 -Cold Shutdown, 5 -Refueling, DEF -Defueled). The events of this category pertain to the following subcategories:

1. RPV Level RPV water level is directly related to the status of adequate core cooling and, therefore, fuel clad integrity.

2. Loss of Emergency AC Power Loss of vital plant electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity.

Th.is category includes loss of onsite and offsite power sources for 4.16 KV ENS buses.

3. RCS Temperature' Uncontrolled or inadvertent temperature or pressure rises are indicative of a potential loss of safety functions.

4. Loss of Vital DC Power Loss of emergency plant electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and emergency core cooling systems which may . be necessary to ensure fission product barrier integrity.

This category includes loss of power to or degraded voltage on the 125V DC vital buses.

r 5. Loss of Communications Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the planrwarrant emergency classification.

Page 65 of 278 Q -::::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases \ 6. Hazardous Event Affecting Safety Systems Certain hazardous natural and technological events may result in VISIBLE DAMAGE to or degraded performance of safety systems warranting classification. Page 66 of 278

River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

C -Cold Shutdown / Refueling System Malfunction Subcategory: 1 -RPV Level Initiating Condition: UNPLANNED loss of RPV inventory EAL: CU1.1 Unusual Event UNPLANNED loss of reactor coolant results in RPV water level less than a required lower limit for~ 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 4 -Cold Shutdown, 5 -Refueling Definition(s): 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: River Bend is equipped with multiple RPV water level instruments including: Wide Range, Fuel Zone, Shutdown Range, Upset Range, and Narrow Range (ref. 1 ). Multiple instruments on different reference and variable legs should be monitored. The Upset Range and Shutdown Range instruments share a common reference leg; therefore, Narrow Range instruments should be routinely monitored when relying on Shutdown or Upset Range instrument as the primary indication. With the plant in Cold Shutdown, RPV water level is normally maintained above the RPV low level scram setpoint of 9.7 in. (ref. 2). However, if RPV level is being controlled below the RPV low level scram setpoint, or if level is being maintained in a designated band in the reactor vessel it is the inability to maintain level above the low end of the designated control band due to a loss of inventory resulting from a leak in the RCS that is the concern. With the plant in Refueling mode, RPV water level is normally maintained at or above the reactor vessel flange. Technical Specifications require at least 23 ft of water above the top of the reactor vessel flange in the refueling cavity during refueling operations (ref. 3). The RPV flange is at approximately 200 in. on the Shutdown Range. (ref. 4). This EAL addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor RPV level concurrent Page 67 of 278 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant. Refueling evolutions that lower RCS water inventory are carefully planned and controlled. An UNPLANNED event that results in water level lowering 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. This EAL recognizes that the minimum required RPV level can change several times during the course of a refueling outage as different plant configurations and system lineups are implemented. This EAL is met if the minimum level, specified for the current plant conditions, cannot be maintained for 15 minutes or longer. The minimum level is typically specified in the applicable operating procedure but may be specified in another controlling document. The 15-minute threshold duration allows sufficient time for prompt operator actions to restore and maintain the expected water level. This criterion excludes transient conditions causing a brief lowering of water level. Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA1 or CA3. Reference(s):

1. EOP-1 RPV Control 2. Technical Requirements Table 3.3.1.1-1 RPS Instrumentation

3. Technical Specification 3.9.6 Reactor Pressure Vessel (RPV) Water Level -Irradiated Fuel 4. GMP-0102 Reactor Vessel Disa 9 sembly 5. NEI 99-01 CU1 Page 68 of 278

-:::::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: C -Cold Shutdown / Refueling System Malfunction Subcategory: 1 -RPV Level Initiating Condition: UNPLANNED loss of RPV inventory EAL: CU1.2 Unusual Event RPV water level cannot be monitored AND EITHER

UNPLANNED rise in any Table C-1 sump or pool level due to a loss of RPV inventory
Visual observation of UNISOLABLE RCS leakage Mode Applicability:

Table C-1 Sumps/Pool

Drywell equipment drain sump
Drywell floor drain sump
Pedestal floor drain sump
CTMT equipment drain sump
CTMT floor drain sump
Suppression Pool
RHR A, B, C, HPCS, LPCS, RCIC room sumps
Auxiliary Building floor drain sump 4 -Cold Shutdown, 5 -Refueling Definition(s):

UNISOLABLE -An open or breached system line that cannot be isolated, remotely or locally. UNPLANNED-. A parameter changes 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. Page 69 of 278* -

~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Basis: In Cold Shutdown mode, the RCS will normally be intact and standard RPV level monitoring means are available.

RPV level in the Refueling mode is normally monitored using the Shutdown Range instrument. . In this EAL, all water level indication is unavailable and the RPV inventory loss must be detected by the leakage indications listed in Table C-1. Level rises must be evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure they are indicative of RPV leakage. Rise in dryw$11 equipment drain sump level and drywell floor sump level is the normal method of monitoring and calculating leakage from the RPV (ref. 1, 2). An Auxiliary Building sump level rise may also be indicative of RCS inventory losses external to the Containment from systems connected to the RPV. With RHR System operating in the Shutdown Cooling mode, an unexplained rise in suppression pool water level could be indicative of RHR valve misalignment or leakage. If the make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS that cannot be isolated could be indicative of a loss of RPV inventory. This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor RPV level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant. Refueling evolutions that lower RCS water inventory are carefully planned and controlled. An UNPLANNED event that results in water level lowering 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. This EAL addresses a condition where all means to determine RPV 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 (Table C-1 ). Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV. Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA 1 or CA3. Reference(s):

1. SOP-0104 Floor and Equipment Drains System 2. SOP-0033 Drywell and Containment Leak Detection System 3. EOP-3 Secondary Containment and Radioactivity Release Control 4. NEI 99-01 CU1 Page 70 of 278

-::==-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: Subcategory: C -Cold Shutdown / Refueling System Malfunction 1 -RPV Level Initiating Condition: Significant Loss of RPV inventory EAL: CA1.1 Alert Loss of RPV inventory as indicated by RPV water level < -43 in. (Level 2) Mode Applicability: 4 -Cold Shutdown, 5 -Refueling Definition(s): None Basis: The threshold RPV water level of -43 in. is the Level 2 actuation setpoint for HPCS and RCIC. Although RCIC cannot restore RPV inventory in the cold condition, the Level 2 actuation setpoint is pperationally significant and is indicative-of a loss of RPV inventory significantly below the low RPV water level scram setpoint specified in CU1 .1 (ref. 1, 2). 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 this EAL, a lowering of RPV water level below the specified level indicates that operator actions have not been successful in restoring and maintaining RPV water level. The heat-up rate of the coolant will rise as the available water inventory is reduced. A continuing drop in water level will lead to core uncovery. Although related, this EAL is concerned with the loss of RPV inventory and not the potential concurrent effects on systems needed for decay heat removal (e.g., loss of a Decay Heat Removal suction point). A rise in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3. If RPV water level continues to lower, then escalation to Site Area Emergency would be via IC CS1. Reference(s):

1. Technical Requirements Table 3.3.5.1-1, Emergency Core Cooling System Instrumentation

2. Technical Requirements Table 3.3.5.2-1, RCIC Instrumentation

3. NEI 99-01 CA 1 ' Page 71 of 278 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

Subcategory: C -Cold Shutdown / Refueling System Malfunction 1 -RPV Level Initiating Condition: Significant Loss of RPV inventory EAL: CA1.2 Alert RPV water level cannot be monitored for~ 15 min. (Note 1) AND EITHER

UNPLANNED rise in any Table C-1 sump or pool levels due to a loss of RPV inventory
Visual observation of UNISOLABLE RCS leakage Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.

The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: Table C-1 Sumps/Pool

Drywell equipment drain sump
Drywell floor drain sump
Pedestal floor drain sump
CTMT equipment drain sump
CTMT floor drain sump
Suppression Pool
RHR A, B, C, HPCS, LPCS, RCIC room sumps
Auxiliary Building floor drain sump 4 -Cold Shutdown, 5 -Refueling Definition(s):

UN/SOLABLE -An open or breached system line that cannot be isolated, remotely or locally. , UNPLANNED -A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Page 72 of 278 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Basis: In Cold Shutdown mode, the RCS will normally be intact and standard RPV level monitoring means are available. RPV level in the Refueling mode is normally monitored using the Shutdown Range instrument. In this EAL, all water level indication is unavailable and the RPV inventory loss must be detected by the leakage indications listed in Table C-1. Level rises must be evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure they are indicative of RPV leakage. Rise in drywell equipment drain sump level and drywell floor sump level is the normal method of monitoring and calculating leakage from the RPV (ref. 1, 2). An Auxiliary Building sump level rise may also be indicative of RCS inventory losses external to the Containment from systems connected to the RPV. With RHR System operating in the Shutdown Cooling mode, an unexplained rise in suppression pool water level could be indicative of RHR valve misalignment or leakage. If the make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS that cannot be isolated could be indicative of a loss of RPV inventory. This IC addresses conditJons 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 substantia*I reduction in the level. of plant safety. For this EAL, the inability to monitor RPV level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV. The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CS1. If the RPV inventory level continues to lower, then escalation to Site Area Emergency would be via IC CS1. Reference(s):

1. SOP-0104 Floor and Equipment Drain~ System 2. SOP-0033 Drywell and Containment Leak Detection System 3. EOP-3 Secondary Containment and Radioactivity Release Control 4. NEI 99-01 CA 1 Page 73 of 278 Q -===-Entergy River Bend Station EAL Basis Document Revision XXX ' Attachment 1 -Emergency Action Level Technical Bases Category:

C -Cold Shutdown / Refueling System Malfunction Subcategory: 1 -RPVLevel Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability EAL: CS1 .1 Site Area Emergency CONTAINMENT CLOSURE not established AND RPV water level < -143 in. (Level 1) Mode Applicability: 4 -Cold Shutdown, 5 -Refueling Definition(s): CONTAINMENT CLOSURE -The actions taken to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. Containment Closure is established when the containment requirements of OSP-0037 are met and at least one integral barrier to the release of radioactive material is provided, within the specified limits using STP-057-3804. IMMINENT -The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. Basis: The threshold RPV water level of-143 in. is the low-low-low ECCS actuation setpoint (Level 1 ). The magnitude of this loss of water indicates that makeup systems have not been effective and may not be capable of preventing further lowering of RPV water level and potential core uncovery. The inability to restore and maintain level after reaching this setpoint infers a failure of the RCS barrier and Potential Loss of the Fuel Clad barrier (ref. 1, 2). This IC addresses a significant and prolonged loss of RPV inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.

Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RPV level cannot be restored, fuel damage is probable.

Page 7 4 of 278

~------------, -::::::a Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases J 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 RPV levels of EALs CS1 .1 and CS1 .2 reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment. This EAL addresses concerns raised by Generic Letter 88-17, Loss of DecayHeat 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 CG1 or AG1. Reference(s):

1. Technical Requirements Table 3.3.5.1-1, Emergency Core Cooling System Instrumentation

2. NEI 99-01 CS1 Page 75 of 278

~* -::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: C -Cold Shutdown / Refueling System Malfunction Subcategory: 1 -RPV Level Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability EAL: CS1 .2 Site Area Emergency CONTAINMENT CLOSURE estc1blished AND RPV water level< -162 in. (TAF) Mode Applicability: 4 -Cold Shutdown, 5 -Refueling Definition(s): CONTAINMENT CLOSURE -The actions taken to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown c~nditions. Containment Closure is established when the containment requirements of OSP-0037 are met and at least one integral barrier to the release of radioactive material is provided, within the specified limits using STP-057-3804. IMMINENT-The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. Basis: When RPV level drops to the top of active fuel (TAF) (an indicated RPV level of-162 in.), core uncovery starts to occur (ref. 1 ). This IC addresses a significant and prolonged loss of RPV inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coplant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration. Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RPV 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 RPV levels of EALs CS1 .1 and CS1 .2 reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment. Page 76 of 278 \ River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases This EAL addresses 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 leve*1 would be via IC CG1 or AG1. Reference( s ): 1. EPSTG*0002 Appendix B EOP and SAP Bases 2. NEI 99-01 CS1 Page 77 of 278 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: C -Cold Shutdown / Refueling System Malfunction Subcategory: 1 -RPV Level Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability EAL: CS1.3 Site Area Emergency RPV level cannot be monitored for~ 30 min. (Note 1) AND Core uncovery is indicated by any of the following:

UNPLANNED rise in any Table C-1 sump or pool levels of sufficient magnitude to indicate core uncovery
Visual observation of UNISOLABLE RCS leakage of sufficient magnitude to indicate core uncovery
RMS-RE140 Refueling Floor Near North Entrance, RMS,-RE141 Refueling Floor Near South Entrance or RMS-RE16 A/B Primary containment

-PAM A/B reading > 9 R/hr Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: Table C-1 Sumps/Pool

Drywell equipment drain sump
Drywell floor drain sump
Pedestal floor drain sump
CTMT equipment drain sump
CTMT floor drain sump
Suppression Pool
RHR A, B, C, HPCS, LPCS, RCIC room sumps
Auxiliary Building floor drain sump
4 -Cold Shutdown, 5 -Refueling Page 78 of 278 ;

-:::::=~ Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Definition(s): IMMINENT -The trajectory of events otconditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. UNISOLABLE -An open or breached system line that cannot be isolated, remotely or locally. UNPLANNED -A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: In Cold Shutdown mode, the RCS will normally be intact and standard RPV level monitoring means are available. RPV level in the Refueling mode is normally monitored using the Shutdown Range instrument. In this EAL, all water level indication is unavailable and the RPV inventory loss must be detected by the leakage indications listed in Table C-1. Level rises must be evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure they are indicative of RPV leakage. Rise in drywell equipment drain sump level and drywell floor sump level is the normal method of monitoring and calculating leakage from the RPV (ref. 1, 2). An Auxiliary Building sump level rise may also be indicative of RCS inventory losses external to the Containment from systems connected to the RPV. With RHR System operating in the Shutdown Cooling mode, an unexplained rise in suppression pool water level could be indicative of RHR valve misalignment or leakage. If the make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS that cannot be isolated could be indicative of a loss of RPV inventory. In the Refueling Mode, as water level in the RPV lowers, the dose rate above the core will rise. The dose rate due to this core shine should result in *indications on area radiation monitors. The 9 R/hr value is selected for this EAL because it is 90% of the scale for RMS-RE140 and 141 (lower range monitors) and on scale for the higher range monitors. This value represents a reading that is higher than that likely to be attributable to normal refuel floor operations. These monitors are located in the Containment on the refuel floor.

This IC addresses a significant and prolonged loss of RPV inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.

Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RPV level cannot be restored, fuel damage is probable. Page 79 of 278 --::::::.-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases In this EAL, the 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to , account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring. The inability to monitor RPV level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV. This EAL addresses 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 NU MARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management. Escalation of the emergency classification level would be via IC CG1 or AG1 Reference(s):

1. SOP-0104 Floor and Equipment Drains System 2. SOP-0033 Drywell and Containment Leak Detection System 3. EOP-3 Secondary Containment and Radioactivity Release Control 4. NEI 99-01 CS1 Page 80 of 278

-:::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: C -Cold Shutdown / Refueling System Malfunction Subcategory: 1 -RPV Level Initiating Condition: Loss of RPV inventory affecting fuel clad integrity with Containment challenged EAL: CG1 .1 General Emergency RPV level < -162 in. (TAF) for~ 30 min. (Note 1) AND Any Containment Challenge indication, Table C-2 Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required. Table C-2 Containment Challenge Indications

CONTAINMENT CLOSURE not established (Note 6)
Drywell or containment hydrogen concentration>

4%

UNPLANNED rise in containment pressure
Exceeding one or more Secondary Containment Control MAX SAFE area radiation levels: Area DRMS Grid 2 Max. Safe Operating Value RHR Equip Rm A 1213 9.5E+03 mR/hr RHR Equip Rm B 1214 9.5E+03 mR/hr RHR Equip Rm C 1215 9.5E+03 mR/hr Mode Applicability:

4 -Cold Shutdown, 5 -Refueling Page 81 of 278 . ~Entergy River Bend*Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Definition(s): CONTAINMENT CLOSURE -The actions taken to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. Containment Closure is established when the containment requirements of OSP-0037 are met and at least one integral barrier to the release of radioactive material is provided, within the specified limits using STP-057-3804. IMMINENT -The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. 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: When RPV level drops below-162 in., core uncovery starts to occur (ref. 1). \ Four conditions are associated with a challenge to Containment integrity:

CONTAINMENT CLOSURE is not established.
In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive mixture of dissolved gases in the . containment.

However, containment monitoring and/or sampling should be performed to verify this assumption and a General Emergency declared if it is determined that hydrogen concentration has exceeded the minimum necessary to support a hydrogen burn (4%). The Igniter System is designed to prevent hydrogen accumulation by locally burning hydrogen in a controlled manner as soon as the hydrogen enters the containment atmosphere and reaches the igniters. For high rates of hydrogen production, ignition. occurs at the lowest concentration that can support ignition. Following ignition, hydrogen is consumed through formation of diffusion flames where the gas enters the containment, thus controlling hydrogen concentration at approximately 4% (ref. 2).

Any UNPLANNED rise in containment pressure in the Cold Shutdown or Refueling mode indicates a potential loss of C,ONTAINMENT CLOSURE capability.

UNPLANNED containment pressure rise indicates.CONTAINMENT CLOSURE cannot be assured and _ the containment cannot be relied upon as a barrier to fission product release.

Secondary Containment radiation monitors should provide indication of a larger release that may be indicative of a challenge to CONTAINMENT CLOSURE. The MAX SAFE radiation levels are indicative of problems, in the secondary containment that are spreading.

The locations into which the primary system discharge is of concern correspond to the areas addressed in Table SC-2 of EOP-3, Secondary Containment Page 82 of 278 Q * ~=-Entergy River Bend Station f::AL Basis Document Revision XXX . ,, Attachment 1 -Emergency Action Level Technical Bases and Radioactivity Release Control that are in service under Cold Shutdown conditions (ref. 3). This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site ar~a. 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 RPV 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 mixturEl means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration

limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity.

It therefore represents a challenge to Containment integrity. In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment. If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration. reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containm.ent is challenged. --This EAL a9dresses 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 NU MARC 91-06, Guidelires for Industry Actions to Assess Shutdown Management. Reference(s):

1. EOP-1 RPV Control 2. EPSTG*0002 Appendix B sOP and SAP Bases 3. EOP-3 Secondary Containment and Radioactivity Release Control 4. NEI 99-01 CG1 Page 83 of 278 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action *Level Technical Bases Category:

C -Cold Shutdown / Refueling System Malfunction Subcategory: 1 -RPV Level Initiating Condition: Loss of RPV inventory affecting fuel clad integrity with Containment challenged EAL: CG1.2 General Emergency RPV water l~vel cannot be monitored for~ 30 min. (Note 1) AND Core uncovery is indicated by any of the following:

UNPLANNED rise in any Table C-1 sump or pool levels of sufficient magnitude to indicate core uncovery
Visual observation of UNISOLABLE RCS leakage of sufficient magnitude to indicate core uncovery
RMS-RE140 Refueling Floor Near North Entrance, RMS-RE141 Refueling Floor Near South Entrance or RMS-RE16 A/B Primary containment

-PAMA/Breading > 9 R/hr AND Any Containment Challenge indication, Table C-2 Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required. Page 84 of 278 . -::::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Table C-1 Sumps/Pool

Drywell equipment drain sump
Drywell floor drain sump
Pedestal floor drain sump
CTMT equipment drain sump
CTMT floor drain sump
Suppression Pool
RHR A, B, C, HPCS, LPCS, RCIC room sumps
Auxiliary Building floor drain sump Table C-2 Containment Challenge Indications
CONTAINMENT CLOSURE not established (Note 6)
Drywell or containment hydrogen concentration>

4%

UNPLANNED rise in containment pressure
Exceeding one or more Secondary Containment Control MAX SAFE area radiation levels: Area DRMS Grid 2 Max. Safe Operating Value RHR Equip Rm A 1213 9.5E+03 mR/hr RHR Equip Rm B 1214 9.5E+03 mR/hr RHR Equip Rm C 1215 9.5E+03 mR/hr Mode Applicability:

4 -Cold Shutdown, 5 -Refueling Definition(s): CONTAINMENT CLOSURE -The actions taken to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. Page 85 of 278

=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Containment Closure is established when the Containment requirements of OSP-0037 (rer 4.1.12) are met with the foJlowing exception:

a functional barrier must exist at the time of the event (i.e., cannot rely on contingency methods to establish a functional barrier). IMMINENT -The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. UNISOLABLE -An open or breached system line that cannot be isolated, remotely or locally. UNPLANNED -A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: In Cold Shutdown mode, the RCS will normally be intact and standard RPV level monitoring means are available. RPV level in the Refueling mode is normally monitored using the Shutdown Range instrument. In this EAL, all water level indication is unavailable and the RPV inventory loss must be detected by the leakage indications listed in Table C-1. Level rises must be evaluated against other potential sources of leakage such as cooling wate 1 r sources inside the drywell to ensure they are indicative of RPV leakage. Rise in drywell equipment drain sump level and drywell floor sump level is the normal method of monitoring and calculating leakage from the RPV (ref. 1, 2). An Auxiliary Building sump level rise may also be indicative of RCS inventory losses external to the Containment from systems connected to the RPV. With RHR System operating in the Shutdown Cooling mode, an unexplained rise in suppression pool water level could be indicative of RHR valve misalignment or leakage. If the make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS that cannot be isolated could be indicative of a loss of RPV inventory. In the Refueling Mode, as water level in the RPV lowers, the dose rate above the core will rise. The dose rate due to this core shine should result in indications on area radiation monitors. The 9 R/hr value is selected for this EAL because it is 90% of the scale for RMS-RE140 and 141 (lower range monitors) and on scale for the higher range monitors. This value represents a reading that is higher than that likely to be attributable to normal refuel floor operations. These monitors are located in the Containment on the refuel floor. Four conditions are associated with a challenge to Containment integrity:

CONTAINMENT CLOSURE is not established.
In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive mixture of dissolved gases in the containment.

However, containment monitoring and/or sampling should be performed to verify this assumption and a General Emergency declared if it is determined that hydrogen concentration has exceeded the minimum necessary to support a hydrogen ' Page 86 of 278 -~ Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Lev~I Technical Bases burn (4%). The Igniter System is designed to prevent hydrogen accumulation by locally burning hydrogen in a controlled manner as soon as the hydrogen enters the containment atmosphere and reaches the igniters. For high rates of hydrogen production, ignition occurs at the lowest concentration that can support ignition. Following ignition, hydrogen is consumed through formation of diffusion flames where the gas enters the containment, thus controlling hydrogen concentration at approximately 4% (ref. 3).

Any UNPLANNED rise in containment pressure in the Cold Shutdown or Refueling mode indicates a potential loss of CONTAINMENT CLOSURE capability.

UNPLJ;\NNED containment pressure rise indicates CONTAINMENT CLOSURE cannot be assured and the containment cannot be relied upon as a barrier to fission product release.

Secondary Containment radiation monitors should provide indication of a larger release . that may be indicative of a challenge to CONTAINMENT CLOSURE. The MAX SAFE radiation levels are indicative of problems in the secondary containment that are spreading.

The locations into which the primary system discharge is of concern correspond to the areas addressed in Table SC-2 of EOP-3, Secondary Containment and Radioactivity Release Control that are in service under Cold Shutdown conditions (ref. 4). This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceea EPA PAG exposure levels offsite for more than the immediate site area. Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RPV 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 mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a challenge to Containment integrity. In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to core uncovery could result in an explosive gas mixture in containment. If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed Page 87 of 278 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged. The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring. The inability to monitor RPV level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV. This EAL addresses 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. Reference(s):

1. SOP-0104 Floor and Equipment Drains System 2. SOP-0033 Drywell and Containment Leak Detection System 3. EPSTG*0002 Appendix B EOP and SAP Bases 4. EOP-3. Secondary Containment and Radioactivity Release Control 5. NEI 99-01 CG1 Page 88 of 278 l

-=::-::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: C -Cold Shutdown / Refueling System Malfunction Subcategory: 2 -Loss of Emergency AC Power Initiating Condition: Loss of all but one AC-power source to ENS buses for 15 minutes or longer EAL: CU2.1 Unusual Event AC power capability, Table C-3, to DIV I and DIV II 4.16 KV ENS buses reduced to a single power source for~ 15 min. (Note 1) AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS ' Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Table C-3 AC Power Sources Offsite

1RTX-XSR1C
1 RTX-XSR 1 D Onsite
EGS-EG1A
EGS-EG1B Mode Applicability:

4 -Cold Shutdown, 5 -Refueling, DEF -Defueled Definition(s): SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 1 OCFR50.2): Those structures, systems and components that are relied upoffto remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; / Page 89 of 278 ~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis: The DIV Ill bus (1 E22*S004) is not credited because it only supplies power to the HPCS pump and associated loads, not any long term decay heat removal systems. In particular, suppression pool cooling mechanisms would be essential subsequent to a station blackout. This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of . safety-related equipment. When in the cold shutdown, refueling, or defueled mode, this condition is not classified as an Alert because of the greater time available to restore another power source to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition is considered to be a potential degradation of the level of safety of the plant. An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an ENS bus. Some examples of this condition are presented below.

A loss of all offsite power with a concurrent failure of all but one emergency ENS power source (e.g., an onsite diesel generator).
A loss of all offsite power and loss of all emergency ENS power sources (e.g., onsite diesel generators) with a single train of emergency ENS buses being back-fed from the unit main generator.
A loss of emergency ENS power sources (e.g., onsite diesel generators) with a single train of emergency ENS buses being fed from an offsite power source. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power. The subsequent loss of the remaining single power source would escalate the event to an Alert in accordance with IC CA2. This EAL is the cold condition equivalent of the hot condition EAL SA 1.1. Reference(s):
1. USAR Section 8.1 Electric Power Introduction

2. EE-001 AC Startup ~lectrical Distribution Chart 3. SOP-0046 4.16 KV System 4. AOP-0004 Loss of Offsite Power 5. NEI 99-01 CU2 Page 90 of 278

-===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases I Category: C -Cold Shutdown / Refueling System Malfunction Subcategory: 2 -Loss of Emergency AC Power Initiating Condition: Loss of all offsite and all onsite AC power to ENS buses for 15 minutes or longer EAL: CA2.1 Alert Loss of all offsite and all onsite AC power capability to DIV I and DIV 11 4.16 KV ENS buses for 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 4 -Cold Shutdown, 5 -Refueling, DEF -Defueled Definition(s): SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 1 OCFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. 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. Mitigative strategies using other power sources (Division Ill diesel generator, FLEX generators, etc.) may be effective in supplying power to these buses. These power sources must be controlled in accordance with abnormal or emergency operating procedures, or beyond design basis accident response guidelines (e.g., FLEX support guidelines) and must be capable (alone or in combination) of supplying power for long term decay heat removal systems. In particular, suppression pool cooling systems would be essential subsequent to a station blackout. Page 91 of 278 _J ~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases l When in the-cold shutdown, refueling, or defueled mode, this condition is not classified as a Site Area Emergency because of the greater time available to restore an ENS bus to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition represents an actual or potential substantial degradation of the level of safety of the plant. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via IC CS1 or AS1. This EAL is the cold condition equivalent of the hot condition EAL SS1 .1. Reference( s): 1. USAR Section 8.1 Electric Power Introduction

2. EE-001AC Startup Electrical Distribution Chart 3. SOP-0046 4.16 KV System 4. AOP-0004 Loss of Offsite Power 5. NEI 99-01 CA2 Page 92 of 278 J J River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

C -Cold Shutdown / Refueling System Malfunction Subcategory: 3 -RCS Temperature -) Initiating Condition: UNPLANNED rise in RCS temperature EAL: CU3.1 Unusual Event UNPLANNED rise in RCS temperature to> 200°F due to loss of decay heat removal capability Mode Applicability: 4 -Cold Shutdown, 5 -Refueling Definition(s): CONTAINMENT CLOSURE -The actions taken to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. Containment Closure is established when the containment requirements of OSP-0037 are met and at least one integral barrier to the release of radioactive material is provided, within the specified limits using STP-057-3804. 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 an UNPLANNED rise in RCS temperature above the Technical Specification cold shutdown temperature limit, and 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 EAL CA3.1. A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a'classification. This EAL involves a loss of decay heat removal capability, or an addition of heat to the RCS in excess of that which can currently be removed, such that reactor coolant temperature cannot be maintained below the cold shutdown temperature limit specified in Technical Specifications. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation. During an outage, the level in the reactor vessel will normally be maintained at or 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 Page 93 of 278 {~ ~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases inventory may result in a rapid rise in reactor coolant temperature depending on the time after shutdown. Escalation to Alert would be via IC CA 1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria. Reference(s):

1. Technical Specification Table 1.1-1 Modes 2. STP-050-0700 RCS Pressure and Temperature Limits Verification

3. AOP-0051 Loss of Decay Heat Removal 4. NEI 99-01 CU3 Page 94 of 278 River Bend Station EAL Basis Document Revision XXX Attachment 1 -EmE:1rgency Action Level Technical Bases Category:

C -Cold Shutdown / Refueling System Malfunction Subcategory: 3 -RCS Temperature Initiating Condition: UNPLANNED rise in RCS temperature EAL: CU3.2 Unusual Event Loss of all RCS temperature and RPV water level indication for~ 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will I ikely be exceeded. The Emergency Director is not all owed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 4 -Cold Shutdown, 5-Refueling Definition{s ): CONTAINMENT CLOSURE -The actions taken to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. Containment Closure is established when the containment requirements of OSP-0037 are met and at least one integral barrier to the release of radioactive material is provided, within the specified limits using STP-057-3804. ., 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 EAL addresses the inability to determine RCS temperature and level, and 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. This EAL reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor RCS conditions and operators would be unable to monitor key parameters necessary to assure core decay heat removal. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication. Page 95 of 278 Q *:.:::::*Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Escalation to Alert would be via IC CA1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria. Refe re nee( s): 1. EOP-1 RPV Control 2. Technical Specifications Table 1.1-1 Modes 3. STP-050-0700 RCS Pressure and Temperature Limits Verification

4. AOP-0051 Loss of Decay Heat Removal 5. NEI 99-01 CU3 Page 96 of 278

-::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: C -Cold Shutdown / Refueling System Malfunction Subcategory: 3 -RCS Temperature Initiating Condition: Inability to maintain plant in cold shutdown EAL: CA3.1 Alert UNPLANNED rise in RCS temperature to> 200°F for> Table C-4 duration (Note 1) OR UNPLANNED RPV pressure rise > 10 psig Note 1: 'I The Emergency Director should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Table C-4 RCS Heat-up Duration Thresholds RCS Status CONTAINMENT Heat-up Duration CLOSURE Status -Intact N/A ( 60 min.* Not intact Established 20 min.* Not established 0 min.

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

Mode Applicability: 4 -Cold Shutdown, 5 -Refueling Definition(s ): CONTAINMENT CLOSURE -The actions taken to secure primary containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. Containment Closure is established when the containment requirements of OSP-0037 are met and at least one integral barrier to the release of ,radioactive material is provided, within the specified limits using STP-057-3804. 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. Page 97 of 278 -:.::=-Entergy River Bend Station EAL Basis Document Revision XXX ,) Attachment 1 -Emergency Action Level Technical Bases Basis: In the absence of reliable RCS temperature indication caused by the loss of decay heat removal capability, classification should be based on the RCS pressure rise criteria when the RCS is intact in Mode 4 or based on time to boil data when in Mode 5 or the RCS is not intact in Mode 4. This EAL addresses conditions involving a loss of decay heat removal capability or an addition of heat to the RCS in excess of that which can currently be removed. Either condition represents an actual or potential substantial degradation of the level of safety of the plant. A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification. The RCS Heat-up Duration Thresholds table addresses a rise in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact. The 20-minute criterion was included to allow time fqr operator action to address the temperature rise. The RCS Heat-up Duration Thresholds table also addresses a rise 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 rise without a substantial degradation in plant safety. Finally, in the case where there is a rise in RCS temperature, the RCS is not intact and CONTAINMENT CLOSURE is not established, no heat-up duration is allowed (i.e., 0 minutes). This is because 1) the evaporated reactor coolant may be released directly into the containment atmosphere and subsequently to the environment, and 2) there is reduced reactor coolant inventory above the top of irradiated fuel. The RCS pressure rise threshold provides a pressure-based indication of RCS heat-up in the absence of RCS temperature monitoring capability. Escalation of the emergency classjfication level would be via IC CS1 or AS1. Refe re nee( s): 1 Technical Specifications Table 1.1-1 Mods 2. STP-050-0700 RCS Pressure and Temperature Limits Verification

3. AOP-0051 Loss of Decay Heat Removal 4. NEI 99-01 CA3 Page 98 of 278

-:;::~ Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: C -Cold Shutdown / Refueling System Malfunction Subcategory: Loss of Vital DC Power Initiating Condition: Loss of Vital DC power for 15 minutes or longer EAL: CU4.1 Unusual Event Indicated voltage is < 105 VDC on required Safety Related DIV I and DIV 11 125 voe buses for;::: 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 4 -Cold Shutdown, 5 -Refueling Definition(s): SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 1 OCFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a saf~ shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis Safety Related DC buses ENB-SWG01A (DIV I) and ENB-SWG01 B (DIV II) feed the Division I and Division II loads respectively. The Division I and Division II batteries each have 60 cells with a specific minimum voltage of 1. 75 volts/cell. These cell voltages yield minimum design bus voltages of 105 VDC (ref. 1 ). This IC addresses a loss of vital DC power which compromises the ability to monitor and , control operable SAFETY SYSTEMS when the plant is in the cold shutdown or refueling mode. In these modes, the core decay heat load has been significantly reduced, and coolant system temperatures and pressures are lower; these conditions raise 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 Page 99 of 278 L_ Q -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Train A is out-of-service (inoperable) for scheduled outage maintenance work and Train B is in-service (operable), then a loss of Vital DC power affecting Train B would require the declaration of an Unusual Event. A loss of Vital DC power to Train A would not warrant an emergency classification. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Depending upon the event, escalation of the emergency classification level would be via IC CA 1 or CA3, or an IC in Recognition Category A. This EAL is the cold condition equivalent of the hot condition EAL SS2.1. Reference(s):

1. Safety Related Battery Specification 244.521 2. USAR 8.3.2 DC Power Systems 3. NEI 99-01 CU4 Page 100 of 278 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases . . Category:

C -Cold Shutdown / Refueling System Malfunction Subcategory:

5 -Loss of Communications Initiating Condition: Loss of all onsite or offsite communications capabilities EAL: CU5.1 Unusual Event Loss of all Table C-5 onsite communication methods OR Loss of all Table C-5 State and local agency communication methods OR Loss of all Table C-5 NRG communication methods TableC-5 Communication Methods System Onsite Plant radio system X Plant Paging System I X I Sound powered phones X In-plant telephones X Eme_rgency Notification System (ENS) Commercial Telephone System Satellite Phones State of Louisiana Radio State and Local Hotline radio INFORM Notification System Mode Applicability: 4 -Cold Shutdown, 5 -Refueling, DEF -Defueled Page 101 of 278 State/ Local X X X X X NRC X X X I ~::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Definition(s): None 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 State and local agencies and the NRC. This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.). The first EAL condition addresses a total loss of the communications methods used in support of routine plant operations. The second EAL condition addresses a total loss of,the communications methods used to notify all State and local agencies of an emergency declaration. The State and local agencies referred to here are the Louisiana Department of Environmental Quality, Governor's Office of Homeland Security and Emergency Preparedness, five Local Parishes Office of Homeland Security and Emergency Preparedness and 24 hour notification points, Mississippi Emergency Management Agency and the Mississippi Highway P 9 trol. The third EAL condition addresses a total loss of the communications methods used to notify the NRC of an emergency declaration. This EAL is the cold condition equivalent of the hot condition EAL SU? .1. Reference(s):

1. RBS Emergency Plan Section 13.3.6.1.5.4 Communications

2. RBS Emergency Plan Section 13.3.6.2.1 Site Communications

3. NEI 99-01 CU5 Page 102 of 278

-~*Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: C -Cold Shutdown / Refueling System Malfunction Subcategory: 6 -Hazardous Event Affecting Safety Systems Initiating Condition: Hazardous event affecting SAFETY SYSTEMS needed for the current operating mode -EAL: CA6.1 Alert The occurrence of any Table C-6 hazardous event AND Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode AND EITHER:

Event damage has caused indications of degraded performance to the second train of the SAFETY SYSTEM needed for the current operating mode
Event damage has resulted in VISIBLE DAMAGE to the second train of the SAFETY SYSTEM needed for the current operating mode (Notes 9, 10) Note 9: If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then emergency classification is not warranted.

Note 10: If the hazardous .event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted. Table C-6 Hazardous Events

Seismic event (earthquake)
Internal or external FLOODING event
High winds or tornado strike
FIRE
EXPLOSION
Other events with similar hazard characteristics as determined by the Shift Manager Page 103 of 278

@ ~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Mode Appli~ability: 4 -Cold Shutdown, 5'-Refueling Definition(s): 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 require a event inspection to determine if the attributes of an explosion are present. FIRE -Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed. FLOODING -A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area. SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): ' Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: ( 1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or. mitigate the consequences of accidents which could result in potential offsite exposures. VISIBLE DAMAGE -Damage to a SAFETY SYSTEM train that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected SAFETY SYSTEM train. 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; Page 104 of 278 Q *===* Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases 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. Escalation of the emergency classification level would be via IC CS1 or AS1. This EAL is the cold condition equivalent of hot condition EAL SA8.1. Reference(s):

1. EP FAQ 2016-002 2. NEI 99-01 CA6 Page 105 of 278

~"Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category E -Independent Spent Fuel Storage Installation (ISFSI) EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.) An independent spent fuel storage installation (ISFSI) is a complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. A significant amount of the radioactive material contained within a canister must escape its packaging and enter the biosphere for there to be a significant environmental effect resulting from an accident involving the dry storage of spent nuclear fuel. An Unusual Event is declared on the basis of the occurrence of an event of sufficient magnitude that a loaded cask CONFINEMENT BOUNDARY is damaged or violated. , *. The RBS ISFSI is located wholly within the plant PROTECTED AREA. Therefore any security event related to the ISFSI are classified under Category H1 security event related EALs. / Page 106 of 278 -::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: ISFSI Subcategory: Confinement Boundary Initiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARY EAL: EU1.1 Unusual Event Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading on the surface of a loaded spent fuel cask (HI-STORM overpack) > EITHER of the following:

60 mRem/hr (T + ri) on the top of the overpack
600 mRem/hr (T + ri) on side of the overpack (excluding inlet and outlet ducts) Mode Applicability:

All Definition(s ): CONFINEMENT BOUNDARY-The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the RBS ISFSI, Confinement Boundary is defined as the Holtec System Multi-Purpose Canister (MPG). INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI): A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. Basis: This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of a storage cask containing spent fuel. It applies to irradiated fuel that is licensed for dry storage beginning at the point that the loaded storage cask is sealed. The issues of concern are the creation of a potential or actual release path to the environment, degradation of one or more fuel assemblies due to environmental factors, and configuration changes which could cause challenges in removing the cask or fuel from storage. , The existence of "damage" is determined by radiological survey. The specified EAL threshold values correspond to 2 times the cask technical specification values. The technical specification (licensing bases document) multiple of "2 times", which is also used in Recognition Category A IC AU1, is used here to distinguish between non-emergency and emergency conditions (ref. 2). 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 Page 107 of 278 ' --===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases "on-contact" dose rate limit is exceeded may be determined based on measurement of a dose rate at some distance from the cask. Security-related events for ISFSls are covered under !Cs HU1 and HA1. Refe re nee( s): 1. USAR 9.1.2.5 Holtec HI-STORM Dry Spent Fuel Storage System 2. RBS HI-STORM 100 SYSTEM Certificate of Compliance for Spent Fuel Storage Casks Amendment 5, Appendix A Technical Specifications for the HI-STORM 100 Cask System Section 5. 7.4 3. NEI 99-01 E-HU1 Page 108 of 278 .I River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category F -Fission Product Barrier Degradation EAL Group: Hot Conditions (RCS temperature > 200°F); EALs in this category are applicable only in one or more hot operating modes. EALs in this category represent threats to the defense in depth design concept that precludes the release of highly radioactive fission products to the environment. This concept relies on multiple physical barriers any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment. The primary fission product barriers are: A. Fuel Clad Barrier (FCB): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets. B. Reactor Coolant System Barrier (RCB): The RCS Barrier is the reactor coolant system pressure boundary and includes the RPV and all reactor coolant system piping out to and including the isolation valves. ' C. Containment Barrier (CNB): The Containment Barrier includes the drywell, the containment, their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves. Containment Barrier thresholds are used as criteria for escalation of the Emergency Classification Level (EGL) from an Alert to a Site Area Emergency or a General Emergency. The EALs in this category require evaluation of the loss and potential loss thresholds listed in the fission product barrier matrix of Table F-1. "Loss" and "Potential Loss" signify the relative damage and threat of damage to the barrier. "Loss" means the barrier no longer assures containment of radioactive materials. "Potential Loss" means integrity of the barrier is threatened and could be lost if conditions continue to degrade. The number of barriers that are lost or potentially lost and the following criteria determine the appropriate emergency classification level: Alert: Any loss or any potential loss of either Fuel Clad or RCS Barrier Site Area Emergency: Loss or potential loss of any two barriers General Emergency: Loss of any two barriers and loss or potential loss of third barrier Page 109 of 278 ~Entergy River Bend Station EAL Basis Document Revision XXX ,) Attachment 1 -Emergency Action Level Technical Bases The logic used for emergency classification based on fission product barrier monitoring should reflect the following considerations:

The Fuel Clad Barrier and the RCS Barrier are weighted more heavily than the Containment Barrier.
Unusual Event ICs associated with RCS and Fuel Clad Barriers are addressed under System Malfunction ICs.
For accident conditions involving a radiological release, evaluation of the fission product barrier thresholds will need to be performed in conjunction with dose assessments to ensure correct and timely escalation of the emergency classification.

For example, an evaluation of the fission .product barrier thresholds may result in a Site Area Emergency classification while a dose assessment may indicate that an EAL for General Emergency IC AG1 has been exceeded.

The fission product barrier thresholds specified within a scheme reflect plant-specific RBS design and operating characteristics.
As used in this category, the term RCS leakage encompasses not just those types defined in Technical Specifications but also includes the loss of RCS mass to any location-inside the primary containment, an in'terfacing system, or outside of the primary containment.

The release of liquid or steam mass from the RCS due to the designed/expected operation of a relief valve is not consi.dered to be RCS leakage.

At the Site Area Emergency level, EAL users should maintain cognizance of how far present conditions are from meeting a threshold that would require a General Emergency declaration.

For example, if the Fuel Clad and RCS fission product barriers were both lost, then there should be frequent assessments of containment radioactive inventory and integrity. Alternatively, if both the Fuel Clad and RCS fission product barriers were potentially lost, the Emergen.cy Director would have more assurance that there was no immediate need to escalate to a General Emergency. I Page 110 of 278 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: Subcategory: Fission Product Barrier Degradation N/A Initiating Condition: Any loss or any potential loss of either Fuel Clad or RCS EAL: FA1.1 Alert Any loss or any potential loss of either Fuel Clad or RCS barrier (Table F-1) Mode Applicability: 1 -Power Operation, 2 -Startup, 3 -Hot Shutdown Definition(s): None Basis: Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 lists the fission product barrier thresholds, bases and references. At the Alert classification level, Fuel Clad and RCS barriers are weighted more heavily than the. Containment barrier. Unlike the Containment barrier, loss or potential loss of either the Fuel Clad or RCS barrier may result in the relocation of radioactive materials or degradation of core cooling capability. Note that the loss or potential loss of Containment barrier in combination with loss or potential loss of either Fuel Clad or RCS barrier results in declaration of a Site Area Emergency under EAL FS1 .1 Reference(s):

1. NEI 99-01 FA1 ) Page 111 of 278 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss or potential loss of any two barriers EAL: FS1 .1 Site Area Emergency Loss or potential loss of any two barriers (Table F-1) Mode Applicability: 1 -Power Operation, 2 -Startup, 3 -Hot Shutdown Definition(s): IMMINENT -The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. Basis: Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 lists the fission product barrier thresholds, bases and references. At the Site Area Emergency classification level, each barrier is weighted equally. A Site Area Emergency is therefore appropriate for any combination of the following conditions:

One barrier loss and a second barrier loss (i.e., loss -loss)
One barrier loss and a second barrier potential loss (i.e., loss -potential loss)
One barrier potential loss and a second barrier potential loss (i.e., potential loss -potential loss) At the Site Area Emergency.classification level, the ability to dynamically assess the proximity of present conditions with respect to the threshold for a General Emergency is important.

For example, the existence of Fuel Clad and RCS Barrier loss thresholds in addition to offsite dose assessments would require continual assessments of radioactive inventory and Containment integrity in anticipation of reaching a General Emergency classification. Alternatively, if both Fuel Clad and RCS .potential loss thresholds existed, the Emergency Director would have greater assurance that escalation to a General Emergency is less IMMINENT. Reference(~):

1. NEI 99-01 FS1 Page 112 of 278

~ *====-Entergy River Ber,d Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss of any two barriers and loss or potential loss of third barrier EAL: FG1.1 General Emergency Loss of any two barriers AND Loss or potential loss of the third barrier (Table F-1) Mode Applicability: 1 -Power Operation, 2 -Startup, 3 -Hot Shutdown Definition(s): None Basis: Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 lists the fission product barrier thresholds, bases and references. At the General Emergency classification level each barrier is weighted equally. A General Emergency is therefore appropriate for any combination of the following conditions:

Loss of Fuel Clad, RCS and Containment Barriers
Loss of Fuel Clad and RCS Barriers with potential loss of Containment Barrier
Loss of RCS and Containment Barriers with potential loss of Fuel Clad Barrier
Loss of Fuel Clad and Containment Barriers with potential loss of RCS Barrier ) Reference(s):

1. NEI 99-01 FG1 Page 113 of 278 V I

-===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Table F-1 Fission Product Barrier Threshold Matrix & Bases Table F-1 lists the threshold conditions that define the Loss and Potential Loss of the three fission product barriers (Fuel Clad, Reactor Coolant System, and Containment). The table is structured so that each of the three barriers occupies adjacent columns. Each fission product barrier column is further divided into two columns; one _for Loss thresholds and one for Potential Loss thresholds. The first column of the table (to the left of the Fuel Clad Barrier Loss column) lists the categories (types) of fission product barrier thresholds. The fission product barrier categories ,are: A. RPV Water Level B. RCS Leak Rate C. Containment Conditions D. Containment Radiation/ RCS Activity E. Containment Integrity or Bypass F. Emergency Director Judgment Each category occupies a row in Table F-1 thus forming a matrix defined by the categories. The intersection of each row with each Loss/Potential Loss column forms a cell in which one or more fission product barrier thresholds appear. If NEI 99-01 does not define a threshold for a barrier Loss/Potential Loss, the word "None" is entered in the cell. Thresholds are assigned sequential numbers within each barrier column beginning with number one (ex., FCB1, FCB2 ... FC86). If a cell in Table F-1 contains more than one numbered threshold, each of the numbered thresholds, if exceeded, signifies a Loss or Potential Loss of the barrier. It is not necessary to exceed all of the thresholds in a category before declaring a barrier Loss/Potential Loss. Subdivision of Table F-1 by category facilitates association of plant conditions to the applicable fission product barrier Loss and Potential Loss thresholds. This structure promotes a systematic approach to assessing the classification status of the fission product barriers. When equipped with knowledge of plant conditions related to the fission product barriers, the EAL-user first scans down the category column of Table F-1, locates the likely category and then reads across the fission product barrier Loss and Potential Loss thresholds in that category to determine if a threshold has been exceeded. If a threshold has not been exceeded, the EAL-user proceeds to the next likely category and continues review of the thresholds in the new category If the EAL-user determines that any threshold has been exceeded, by definition, the barrier is lost or potentially lost -even if multiple thresholds in the same barrier column are exceeded, only that one barrier is lost or potentially lost. The EAL-user must examine each of the three fission product barriers to determine if other barrier thresholds in the category are lost or Page 114 of 278

-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases potentially lost. For example, if containment radiation is sufficiently high, a Loss of the Fuel Clad and RCS Barriers and a Potential Loss of the Containment Barrier can occur. Barrier Losses and Potential Losses are then applied to the algorithms given in EALs FG1 .1, FS1 .1, and FA1 .1 to determine the appropriate emergency classification.

In the remainder of this Attachment, the Fuel Clad Barrier threshold bases appear first, followed by the RCS Barrier and finally the Containment Barrier threshold bases. In each barrier, the bases are given according to category Loss followed by category Potential Loss beginning with Category A, then 8, ... ,F. Page 115 of 278 -c::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Table F-1 Fission Product Barrier Threshold Matrix Fuel Clad Barrier (FCB) Reactor Coolant System Barrier (RCB) Containment Barrier (CNB) Category Loss Potential Loss Loss Potential Loss Loss Potential Loss A FCB2 RPV water level cannot be RCB1 RPV water level cannot be restored and maintained> restored and maintained> CNB1 SAP entry is required RPVWater FCB1 SAP entry is required -162 in. (TAF) -162 in. (TAF) None None Level or cannot be determined or cannot be determined RCB2 UNISOLABLE break in any of CNB2 UNISOLABLE primary RCB4 UNISOLABLE primary system system leakage that results the following: leakage that results in in exceeding EITHER: . Main steam lines exceeding EITHER:

One or more EOP-;J Max B RCIC steam Line
One or more EOP-3 Max Safe area radiation None None . Normal area radiation operating value that can be None RCS Leak Rate . RWCU operating value (Table F-2) read in the Control Room . Feedwater
One or more Isolation (Table F-2) RCB3 Emergency Depressurization Temperature alarms
One or more EOP-3 Max is required (Table F-2) Safe area tern perature operating value (Table F-2) CNBS Containment pressure>

15 psig CNB3 UNPLANNED rapid drop in C containment pressure following CNB6 Drywell or containment RCBS Drywell pressure > 1.68 psid containment pressure rise hydrogen concentration > 4% None None None CTMT due to RCS leakage CNB4 Containment pressure CNB7 Parameters cannot be restored Conditions response not consistent with and maintained within the safe LOCA conditions zone of the HCTL curve (EOP Figure 2) D FCB3 Containment radiation (RMS-RE16) > 3,000 R/hr RCB6 Drywell radiation (RMS-RE20) C > 30 R/hr CNB8 Containment radiation CTMTRad/ FCB4 Coolant activity None None None (RMS-RE16) > 12,000 R/hr RCS > 300 µCi/gm dose Activity equivalent 1-131 CNB9 UNISOLABLE direct E downstream pathway to the environment exists after CTMT None None None None Containment isolation signal None Integrity or CNB101ntentional Containment Bypass venting per EOPs F FCBS Any condition in the opinion FCB6 Any condition in the opinion of RCB7 Any condition in the opinion of RCB8 Any condition in the opinion of CNB11 Any condition in the opinion of CNB12Any condition in the opinion of Emergency of the Em erg ency Director the Emergency Director that the Emergency Director that the Emergency Director that the Emergency Director that the Emergency Director that that indicates loss of the fuel indicates potential loss of the indicates loss of the RCS indicates potential loss of the indicates loss of the indicates potential loss of the Director clad barrier fuel clad barrier barrier RCS barrier Containment barrier Containment barrier Judgment / Page 116 of 278 -::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Fuel Clad Category: A. RPV Water Level Degradation Threat: Loss Threshold: I FCB1 SAP entry is required' Definition(s): None Basis: Emergency Operating Procedure (EOPs) specify entry to the Severe Accident Procedures (SAPs) when core cooling is severely challenged. These EOPs provide instructions to ensure adequate core cooling by maintaining RPV water level above prescribed limits or operating sufficient RPV injection sources when level cannot be determined (ref. 1, 2). The EOP conditions requiring SAP entry represent a challenge to core cooling and are the minimum values to assure core cooling without further degradation of the clad. This threshold is also a Potential Loss of the Containment barrier (CNB1 ). Since SAP entry occurs after core uncovery has occurred a Loss of the RCS barrier exists (RCB1 ). SAP entry, therefore, represents a Loss of two barriers and a Potential Loss of a third, which requires a General Emergency classification. ' The Loss threshold represents the EOP requirement for entry into the SAPs. This is identified in the BWROG EPGs/SAGs when adequate core cooling cannot be assured. Reference(s):

1. EOP-1 RPV Control 2. EOP-4 RPV Flooding 3. EP FAQ 2015-004 4. NEI 99-01, RPV Water Level Fuel Clad Loss 2.A Page 117 of 278

-::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Fuel Clad Category: A. RPV Water Level Degradation Threat: Potential Loss Threshold: FCB2 RPV water level cannot be restored and maintained> -162 in. (TAF) or cannot be determined Definition(s): None Basis: An RPV water level instrument reading of -162 in. indicates RPV level is at the top of active fuel (TAF) (ref. 1 ). When RPV level is at or above the TAF, the core is completely' submerged. Core submergence is the most desirable means of core cooling. When RPV level is below TAF, the uncovered portion of the core must be cooled by less reliable means (i.e., steam cooling or spray cooling). If core uncovery is threatened, the EOPs specify alternate, more extreme, RPV water level control measures in order to restore and maintain adequate core cooling. Since core uncovery begins if RPV level drops to TAF, the level is indicative of a challenge to core cooling and the Fuel Clad barrier. When RPV water level cannot be determined, EOPs require entry to EOP-4, RPV Flooding. RPV water level indication provides the primary means of knowing if adequate core cooling is being maintained (ref. 2). When all means of determining RPV water level are unavailable, the fuel clad barrier is threatened and reliance on alternate means of assuring adequate core cooling must be attempted. The instructions in EOP-4 specify these means, which include emergency depressurization of the RPV and injection into the RPV at a rate needed to flood to the elevation of the main steam lines or hold RPV pressure above .the Minimum Steam Cooling Pressure (in scram-failure events). If RPV water level cannot be determined with respect to the top of active fuel, a potential loss of the fuel clad barrier exists. Note that EOP-1A, RPV Control, ATWS, may require intentionally lowering RPV water level to -162 in. and control' level between the Minimum Steam Cooling RPV Water Level (MSCRWL) and the top of active fuel (ref. 3). Under these conditions, a high-power ATWS event exists and requires at least a Site Area Emergen-cy classification in accordance with the System Malfunction -RPS Failure EALs. In high-power A TWS/failure to scram events, EOPs may direct the operator to deliberately_ lower RPV water level to the top of active fuel in order to reduce reactor power. RPV water Page 118 of 278 ~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases level is then controlled between the top of active fuel and the Minimum Steam Cooling RPV Water Level (MSCRWL). Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority. For such events, EALs SA6.1 or SS6.1 will dictate the need for emergency classification. This water level corresponds to the top of the active fuel and is used in the EOPs to indicate a challenge to core cooling. The RPV water level threshold is the same as RCS barrier Loss threshold RCB1. Thus, this threshold indicates a Potential Loss of the Fuel Clad barrier and a Loss of the RCS barrier that appropriately escalates the emergency classification level to a Site Area Emergency. This threshold is considered to be exceeded when, as specified in the site-specific EOPs, RPV water cannot be restored and maintained above the specified level following depressurization of the RPV ( either manually, automatically or by failure of the RCS barrier) or when procedural guidance or a lack of low pressure RPV injection sources preclude Emergency RPV depressurization. EOPs allow the operator a wide choice of RPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurization of the RPV in order to facilitate RPV water level control with low-pressure injection sources. In some events, elevated RPV pressure may prevent restoration of RPV water level until pressure drops below the shutoff heads of available injection sources. Therefore, this Fuel Clad barrier Potential Loss is met only after either: 1) the RPV has been depressurized , 1 or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory. The term "cannot be restored and maintained above" means the value of RPV water level is not able to be brought above the specified limit (top of active fuel). The determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the

current value is below the top of active fuel, but does not permit extended operation below the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained.

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. Reference(s):

1. EOP-1 RPV Control 2. EOP-4 RPV Flooding 3. EOP-1A RPV Control, ATWS 4 NEI 99-01 RPV Water Level Potential Loss 2.A Page 119 of 278.

=-Entergy River Bend Station EAL Basis Document Revision XXX
Attachment 1 -Emergency Action Level Technical Bases Barrier: Fuel Clad Category:

B. RCS Leak Rate Degradation Threat: Loss Threshold: Page 120 of 278 -c::::=-Entergy Barrier: Category: River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Fuel Clad B. RCS Leak Rate Degradation Threat: Potential Loss Threshold: I None Page 121 of278 Barrier: Category: River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Fuel Clad C. CTMT Conditions Degradation Threat: Loss Threshold: None Page 122 of 278 Barrier: Category: River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Fuel Clad C. CTMT Conditions Degradation Threat: Potential Loss Threshold: None Page 123 of 278

-=::::s Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Fuel Clad Category:

D. CTMT Radiation I RCS Activity Degradation Threat: Loss

Threshold:

FCB3 Containment radiation (RMS-RE16) > 3,000 R/hr Definition(s ): None Basis: The containment radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals 300 µCi/gm dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to significant fuel clad damage (ref. 1 ). 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 RCB6 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 EGL to a Site Area EmerQency. There is no Fuel Clad barrier Potential Loss threshold associated with CTMT Radiation/RCS Activity. Reference(s):

1. Calculation G13.18.9.4-045 Containment Doses for Emergency Action Levels (EALs) 2. NEI 99-01 Primary Containment Radiation Fuel Clad Loss 4.A Page 124 of 278 j

-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Fuel Clad Category:

D. CTMT Radiation/ RCS Activity Degradation Threat: Loss Threshold: FCB4 Coolant activity> 300 µCi/gm dose equivalent 1-131 Definition(s): None Basis: This threshold indicates that RCS radioactivity concentration is greater than 300 µCi/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. It is recognized that sample collection* and analysis of reactor coolant with highly elevated _activity levels could require several hours to complete. Nonetheless, a sample-related threshold is included as a backup to other indications. There is no Fuel Clad barrier Potential Loss threshold associated with CTMT Radiation/RCS Activity. Reference(s):

1. NEI 99-01 RCS Activity Fuel Clad Loss 1.A Page 125 of 278

-::::::-Entergy Barrier: *category: River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Fuel Clad D. CTMT Radiation / RCS Activity Degradation Threat: Potential Loss Threshold: None Page 126 of 278

=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Fuel Clad Category:

E. CTMT Integrity or Bypass Degradation Threat: Loss Threshold: Page 127 of 278 ./ -:::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Fuel Clad Category: E. CTMT Integrity or Bypass Degradation Threat: Potential Loss Threshold: Page 128 of 278 @ -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Fuel Clad Category: F. Emergency Director Judgment Degradation Threat: Loss Threshold: FCBS Any condition in the opinion of the Emergency Director that indicates loss of the Fuel Clad barrier Definition(s): None Basis: This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad barrier is lost. Reference(s):

1. NEI 99-01 Emergency Director Judgment Fuel Clad Loss 6.A Page 129 of 278 I"\ -:-:::::-Entergy Barrier: Category:

River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Fuel Clad F. Emergency Director Judgment Degradation Threat: Potential Loss Threshold: FCB6 Any condition in the opinion of the Emergency Director that indicates potential loss of the Fuel Clad barrier Definition(s): None Basis: This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored. Refe re nee( s): 1. NEI 99-01 Emergency Director Judgment Fuel Clad Potential Loss 6.A Page 130 of 278 -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: A. RPV Water Level Degradation Threat: Loss Threshold: RCB1 RPV water level cannot be restored and maintained> -162 in. (TAF) or cannot be determined Definition(s): None Basis: An RPV water level instrument reading of -162 in. indicates level is at the top of active fuel (TAF) (ref. 1 ). TAF is significantly lower than the normal operating RPV level control band. To reach this level, RPV inventory loss would have previously required isolation of the RCS and Containment barriers, and initiation of all ECCS. If RPV water level cannot be maintained above TAF, ECCS and other sources of RPV injection have been ineffective or incapable of reversing the lowering level trend. The cause of the loss of RPV inventory is therefore assumed to be a LOCA. By definition, a LOCA event is a Loss of the RCS barrier. When RPV water level cannot be determined, EOPs require entry to EOP-4, RPV Flooding. RPV water level indication provides the primary means of knowing if adequate core cooling is being maintained (ref. 2). The instructions in EOP-4 specify emergency depressurization of the RPV, which is defined to be a Loss of the RCS barrier (RCS Loss RC83). Note that EOP-1A, RPV Control, ATWS, may require intentionally lowering RPV water level to -162 in. and control level between the Minimum Steam Cooling RPV Water Level (MSCRWL) and the top of active fuel (ref. 3). Under these conditions, a high-power ATWS event exists and requires at least a Site Area Emergency classification in accordance with the System Malfunction -RPS Failure EALs. In high-power A TWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level to the top of active fuel in order to reduce reactor power. RPV water level is then controlled between the top of active fuel and the Minimum Steam Cooling RPV Water Level (MSCRWL). Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority. For such events, ICs SA6 or SS6 will dictate the need for emergency classification. Page 131 of 278 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases This water level corresponds to the top of active fuel and is used in the EOPs to indicate challenge to core cooling. The RPV water level threshold is the same as Fuel Clad barrier Potential Loss threshold FCB2. Thus, this threshold indicates a Loss of the RCS barrier and Potential Loss of the Fuel Clad barrier and that appropriately escalates the emergency classification level to a Site Area Emergency. This threshold is considered to be exceeded when, as specified in the site-specific EOPs, RPV water cannot be restored and maintained above the specified level 1 following de pressurization of the RPV ( either manually, automatically or by failure of the RCS barrier) or when procedural guidance or a lack of low pressure RPV injection sources preclude Emergency RPV depressurization EOPs allow the operator a wide choice of RPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurization of the RPV in order to facilitate RPV water level control with low-pressure injection sources. In some events, elevated RPV pressure may prevent restoration of RPV water level until pressure drops below the shutoff heads of available injection sources. Therefore, this RCS barrier Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory. The term, "cannot be restored and maintained above," means the value of RPV water level is not able to be brought above the specified limit (top of active fuel). The determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the current value is below the top of active fuel, but does not permit extended operation beyond the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained. There is no RCS barrier Potential Loss threshold associated with RPV Water Level. Reference(s):

1. EOP-1 RPV Central 2. EOP-4 RPV Flooding 3. EOP-1A RPV Control, ATWS 4. NEI 99-01 RPV Water Level RCS Loss 2.A Page 132 of 278 Barrier: Category:

River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Reactor Coolant System A. RPV Water Level Degradation Threat: Potential Loss Threshold: None Page 133 of 278 -~Entergy River Bend Station EAL Basis Document Revision XXX J Attachment 1 -Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category:

8. RCS Leak Rate Degradation Threat: Loss Threshold:

RCB2 UNISOLABLE break in any of the following:

Main steam line
RCIC steam line
RWCU
Feedwater Definition(s):

UN/SOLABLE -An open or breached system line that cannot be isolated, remotely or locally. Basis: Failure to isolate the leak (from the Control Room or locally), within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification. The conditions of this threshold include required containment isolation failures allowing a flow path to the environment. A release pathway outside primary containment exists when flow is not prevented by downstream isolations. In the case of a failure of both isolation valves to close but in which no downstream flowpath exists, emergency declaration under this threshold would not be required. Similarly, if the emergency response requires the normal process flow of a system outside containment (e.g., EOP requirement to bypass MSIV low RPV water level interlocks and maintain the main condenser as a heat sink using main turbine bypass valves), the threshold is not met. The combination of these threshold conditions represent the loss of both the RCS and Containment (see Loss CNB9) barriers and justifies declaration of a Site Area Emergency (i.e., Loss or Potential Loss of any two barriers). Even though RWCU and Feedwater systems do not contain steam, they are included in the list because an UNISOLABLE break could result in the high-pressure discharge of fluid that is flashed to steam from relatively large volume systems directly connected to the RCS. Page 134 of 278 @ ~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Large high-energy lines that rupture outside primary containment can discharge significant amounts of inventory and jeopardize the pressure-retaining capability of the RCS until they are isolated. If it is determined that the ruptured line cannot be promptly isolated, remotely or locally, the RCS barrier Loss threshold is met. Reference(s):

1. NEI 99-01 RCS Leak Rate RCS Loss 3.A / Page 135 of 278

~"Entergy Barrier: Category: River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Reactor Coolant System B. RCS Leak Rate Degradation Threat: Loss Threshold: RCB3 Emergency Depressurization is required Definition(s): None Basis: Plant symptoms requiring Emergency R.PV Depressurization per the EOPs are indicative of a loss of the RCS barrier (ref. 1, 2). Emergency RPV Depressurization in accordance with the EOPs is indicative of a loss of the RCS barrier. If Emergency RPV Depressurization is performed, the plant operators are directed to open safety relief valves (SRVs). Even though the RCS is being vented into the suppression pool, a Loss of the RCS barr\er exists due to the diminished effectiveness of the RCS to retain fission products within its boundary. EOP-1 Emergency Depressurization allows terminating the depressurization if necessary to maintain RCIC as an injection source. This would require closing the SRVs. Even though the SRVs may be rE3closed, this threshold is still met due to the requirement for an Emergency Depressurization having been met (ref. 2). Reference( s): 1. EOP-1 RPV Control Emergency Depressurization

2. EP FAQ 2015-003 3. NEI 99-01 RCS Leak Rate RCS Loss 3.8 Page 136 of 278

~=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases

Barrier: Reactor Coolant System Category:

B. RCS Leak Rate Degradation Threat: Potential Loss Threshold: RCB4 UNISOLABLE primary system leakage that results in exceeding EITHER:

One or more EOP-3 Max Normal area radiation operating value (Table F-2)
One or more Isolation Temperature alarms (Table F-2) Table F-2 Secondary Containment Operating Values Area Temperatures Parameter Isolation Temperature Max Safe Main Steam Line Tunnel 173°F (P601-19A-A1/A3/B1/B3) 200°F RHR Equipment Area 1 (A) 117°F (P601-20A-B4) 200°F RHR Equipment Area 2 (B) 117°F (P601-20A-B4) 200°F RCIC Equipment Area 182°F (P601-21A-B6) 200°F RWCU Pump Room 1 (A)/ 2 (B) 165°F (P680-1 A-A2/B2) 200°F Area Radiation Levels Parameter Max Normal Max Safe HPCS Area (1212) Grid 2 8.20E+01 mR/hr 9.5E+03 mR/hr RHR Equipment Room A (1213) Grid 2 8.20E+01 mR/hr 9.5E+03 mR/hr RHR Equipment Room B (1214) Grid 2 8.20E+01 mR/hr 9.5E+03 mR/hr RHR Equipment Room C (1215) Grid 2 8.20E+01 mR/hr 9.5E+03 mR/hr LPCS Equipment Room (1216) Grid 2 8.20E+01 mR/hr 9.5E+03 mR/hr HPCS Penetration Area (1217) Grid 2 8.20E+01 mR/hr 9.5E+03 mR/hr LPCS Penetration Area (1218) Grid 2 8.20E+01 mR/hr 9.5E+03 mR/hr RCIC Equipment Room (1219) Grid 2 1.20E+02 mR/hr 9.5E+03 mR/hr Definition(s):

FLOODING -A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area. UNISOLABLE -An open or breached system line that cannot be isolated, remotely or locally. Page 137 of 278 -::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Basis: Failure to isolate the leak (from the Control Room or locally), within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification. The presence of elevated general area temperatures or radiation levels in the Secondary Containment may be indicative of UNISOLABLE primary system leakage outside the containment. The EOP-3 Max Normal and Isolation Temperature alarm setpoint values in Table F-2 define this RCS threshold because they are the maximum normal operating/ Technical Specification Isolation values and signify the onset of abnormal system operation. When parameters reach this level, equipment failure or mis-operation may be occurring. Elevated 'parameters may also adversely affect the ability to gain access to or operate equipment within the affected area. The locations into which the primary system discharge is of concern correspond to the areas addressed in EOP-3, Secondary Containment and Radioactivity Release Control (ref. 1 ). In general, multiple indications should be used to determine if a primary system is discharging outside Containment. For example, a high area radiation condition does not necessarily indicate that a primary system is discharging into the Secondary Containment since this may be caused by radiation shine from nearby steam lines or the movement of radioactive materials. Conversely, a high area radiation condition in conjunction with other indications (e.g. room FLOODING, high area temperatures, reports of steam in the secondary containment, an unexpected rise in feedwater flowrate, or unexpected main turbine control valve closure) may

indicate that a primary system is discharging into the Secondary Containment.

Potential loss of RCS based on primary system leakage outside the primary containment is determined from EOP temperature or radiation Max Normal Operating values in areas such as main steam line tunnel, RCIC, etc., which indicate a direct path from the RCS to areas outside primary containment. A Max Normal Operating value is the highest value of the identified parameter expected to occur during normal plant operating conditions with all directly associated support and control systems functioning properly. An Isolation Temperature value is indicative of an UNISOLABLE leak when temperatures do not begin to recover as a result of the isolation actions following the alarm and represents a Technical Specification limiting value. The indicators reaching the threshold barriers and confirmed to be caused by RCS leakage from a primary system warrant an Alert classification. A primary system is defined to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a reduction in the steam or water being discharged through an unisolated break in the system. Reference(s):

1. EOP-3 Secondary Containment and Radioactivity Release Control 2. NEI 99-01 RCS Leak Rate RCS Potential Loss 3.A Page 138 of 278 Q ~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category:

C. CTMT Conditions Degradation Threat: Loss Threshold: RCBS Drywell pressure> 1.68 psid due to RCS leakage Definition(s): None Basis: The drywell high pressure scram setpoint is an entry condition to EOP-1, RPV Control. A high Containment pressure of greater than 0.3 psig is an entry condition to EOP-2, Primary Containment Control (ref. 1, 2). Normal containment pressure control functions (e.g., operation of drywell and containment cooling, vent using containment vessel purge, etc.) are specified in EOP-2 in advance of less desirable but more effective functions (e.g., Emergency Depressurization, etc.). In the design basis, containment pressures above the drywell high pressure scram setpoint are assumed to be the result of a high-energy release into the containment for which normal pressure control systems are inadequate or incapable of reversing the rising pressure trend. Pressures of this magnitude, however, can be caused by non-LOCA events such as a loss of drywell cooling or inability to control containment vent/purge (ref. 3). The threshold phrase" ... due to RCS leakage" focuses the barrier failure on the RCS instead of the non-LOCA malfunctions that may adversely affect containment pressure. Drywell pressure greater than 1.68 psid with corollary indications (e.g., drywell temperature, indications of loss of RCS inventory) should therefore be considered a Loss of the RCS barrier. Loss of drywell cooling that results in pressure greater than 1.68 psid should not be considered an RCS barrier Loss. The 1.68 psid value is the drywell high pressure setpoint which indicates a LOCA by automatically initiating the ECCS or equivalent makeup system. There is no RCS barrier Potential Loss threshold associated with CTMT Conditions. Page 139 of 278 -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Reference(s):

1. EOP-1 RPV Control 2. EOP-2 Primary Containment Control 3. USAR Section 6.2.1 Containment Functional Design 4. NEI 99-01 Primary Containment Pressure RCS Loss 1.A Page 140 of 278

-~-Entergy Barrier: Category: River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Reactor Coolant System C. CTMT Conditions Degradation Threat: Potential Loss Threshold: I None Page 141 of 278 ----* ---1 I River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: D. CTMT Radiation/ RCS Activity Degradation Threat: Loss Threshold: RCB6 Drywell radiation (RMS-RE20) > 30 R/hr Definition(s): N/A Basis: Under post-LOCA conditions coaxial cables used on the drywell post-accident monitors RE20A/B) are susceptible to Thermally Induced Currents (TIC). These currents may cause the drywell PAMs to read falsely high (-469 R/hr) on a rapid temperature rise and read falsely low on a rapid temperature drop. When accident temperature conditions stabilize indicated radiation dose rates would be more accurate. The duration of the spurious signal would last approximately 15 minutes. During the period of false readings operators should rely on other indications of RCS leakage including a rise in drywell temperature and pressure (RCB5). The drywell radiation monitor reading (38 R/hr rounded to 30 R/hr for readability) corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower. than that specified for Fuel Clad Barrier Loss threshold FCB3 since it indicates a loss of the RCS Barrier only (ref. 1 ). There is no RCS barrier Potential Loss threshold associated with CTMT Radiation/RCS Activity. Reference( s): 1. Calculation G13.18.9.4-045 Containment Doses for Emergency Action Levels (EALs) 2. NEI 99-01 Primary Containment Radiation RCS Loss 4.A Page 142 of 278 Barrier: Category: River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Reactor Coolant System D. CTMT Radiation/ RCS Activity Degradation Threat: Potential Loss Threshold: None Page 143 of 278 \

-Entergy River Bend Station EAL Basis Document Revision XXX v Attachment 1 -Emergency Action Level Technical Bases ,_ Barrier: Reactor Coolant System Category:

E. CTMT Integrity or Bypass Degradation Threat: Loss Threshold: Page 144 of 278 Barrier: Category: River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Reactor Coolant System E. CTMT Integrity or Bypass Degradation Threat: Potential Loss Threshold: [ None Page 145 of 278 Q ~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: F. Emergency Director Judgment Degradation Threat: Loss Threshold: RCB7 Any condition in the opinion of the Emergency Director that indicates loss of the RCS barrier Definition(s): None Basis: This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is lost. Reference(s):

NEI 99-01 Emergency Director Judgment RCS Loss 6.A Page 146 of 278

~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: F. Emergency Director Judgment Degradation Threat: Potential Loss Threshold: RCBB Any condition in the opinion of the Emergency Director that indicates potential loss of the RCS barrier Definition(s): None Basis: This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored. Reference(s):

1. NEI 99-01 Emergency Director Judgment RCS Potential Loss 6.A Page 147 of 278

~=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Containment Category: A. RPV Water Level Degradation Threat: Loss Threshold: None Page 148 of 278 / -~ Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Containment Category: A. RPV Water Level Degradation Threat: Potential Loss Threshold: CNB1 SAP entry is required Definition(s): None Basis: EOPs specify entry to the SAPs when core cooling is severely challenged. These EOPs provide instructions to ensure adequate core cooling by maintaining RPV water level above prescribed limits or operating sufficient RPV injection sources when level cannot be determined (ref. 1, 2). The EOP conditions requiring SAP entry represent a challenge to core cooling and are the ,I minimum values to assure core cooling without further degradation of the clad. This threshold is also a Loss of the Fuel Clad barrier (Loss FCB1 ). Sine~ SAP entry occurs after core uncovery has occurred a Loss of the RCS barrier exists (Loss RCB1 ). SAP entry, ther~fore, represents a Loss of two barriers and a Potential Loss of a third, which requires a General Emergency classification. " The Potential Loss threshold is identical to the Fuel Clad Loss RPV Water Level threshold FCB1. The Potential Loss requirement for entry into the SAPs indicates adequate core cooling cannot be assured and that core damage is possible. BWR EPGs/SAGs (RBS term SAPs) specify the conditions when the EPGs are exited and SAGs are entered. Entry into SAGs is a logical escalation in response to the inability to assure adequate core cooling. PRA studies indicate that the condition of this Potential Loss threshold could be a core melt sequence which, if not corrected, could lead to RPV failure and greater potential for primary containment failure. In conjunction with the RPV water level Loss thresholds in the Fuel Clad and RCS barrier columns, this threshold results in the declaration of a General Emergency. There is no Containment barrier Loss threshold associated with RPV Water Level. \ Page 149 of 278 -::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Reference(s):

1. EOP-1 RPV Control 2. EOP-4 RPV Flooding 3. EP FAQ 2015-004 4. NEI 99-01 RPV Water Level PC Potential Loss 2.A Page 150 of 278 Q -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Containment Category:

B. RCS Leak Rate Degradation Threat: Loss Threshold: CNB2 UNISOLABLE primary system leakage that results in exceeding EITHER:

One or more EOP-3 Max Safe area radiation operating value (Table F-2)
One or more EOP-3 Max Safe area temperature operating value (Table F-2) Table F-2 Secondary Containment Operating Values ~rea Temperatures . Parameter Isolation Temperature Max Safe Main Steam Line Tunnel 173°F (P601-19A-A1/A3/81/83) 200°F RHR,Equipment Area 1 (A) 117°F (P601-20A-84) 200°F RHR Equipment Area 2 (8) 117°F (P601-20A-84) 200°F RCIC Equipment Area 182°F (P601-21A-B6) 200°F RWCU Pump Room 1 (A)/ 2 (8) 165°F (P680-1A-A2/82) 200°F Area Radiation Levels Parameter Max Normal Max Safe HPCS Area (1212) Grid 2 8.20E+01 mR/hr 9.5E+03 mR/hr RHR Equipment Room A (1213) Grid 2 8.20E+01 mR/hr 9.5E+03 mR/hr RHR Equipment Room 8 (1214) Grid 2 8.20E+01 mR/hr 9.5E+03 mR/hr RHR Equipment Room C (1215) Grid 2 8.20E+01 mR/hr 9.5E+03 mR/hr LPCS Equipment Room (1216) Grid 2 8.20E+01 mR/hr 9.5E+03 mR/hr HPCS Penetration Area (1217) Grid 2 8.20E+01 mR/hr 9.5E+03 mR/hr LPCS Penetration Area (1218) Grid 2 8.20E+01 mR/hr 9.5E+03 mR/hr RCIC Equipment Room (1219) Grid 2 1.20E+02 mR/hr 9.5E+03 mR/hr Page 151 of 278

. ~:: Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Adlon Level Technical Bases Definition(s): FLOODING -A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area. UN/SOLABLE -An open or breached system line that cannot be isolated, remotely or locally. Basis: Failure to isolate the leak (from the Control Room or locally), within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification. The presence of elevated general area temperatures or radiation levels in the secondary containment may be indicative of UNISOLABLE primary system leakage outside th.e containment. The Max Safe conditions define this Containment barrier threshold because they are indicative of problems in the secondary containment that are spreading and pose a threat to achieving a safe plant shutdown. This threshold addresses problematic discharges outside containment that may not originate from a high-energy line break. The locations into which the primary system discharge is of concern correspond to the areas addressed in EOP-3, Secondary Containment and Radioactivity Release Control (ref. 1 ). In general, multiple indications should be used to determine if a primary system is discharging outside containment. For example, a high area radiation condition does not necessarily indicate that a primary system is discharging into the secondary containment since this may be caused by radiation shine from nearby steam lines or the movement of radioactive materials. Conversely, a high area radiation condition in conjunction with other indications (e.g. room FLOODING, high area temperatures, reports of steam in the secondary containment, an unexpected rise in feedwater flowrate, or unexpected main turbine control valve closure) may indicate that a primary system is discharging into the secondary containment. I The Max Safe Operating Temperature and the Max Safe Operating Radiation Level are each the highest value of these parameters at which neither: (1) equipment necessary for the safe shutdown of the plant will fail, nor (2) personnel access necessary for the safe shutdown of the plant will be precluded. EOPs utilize these temperatures and radiation levels to establish conditions under which RPV depressurization is required. There is no Containment barrier Potential Loss threshold associated with RCS Leak Rate. Reference(s):

1. EOP-3 Secondary Containment and Radioactivity Release Control 2. NEI 99-01 RCS Leak Rate PC Loss 3.C Page 152 of 278 Barrier: Category:

River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Containment B. RCS Leak Rate Degradation Threat: Potential Loss Threshold: None Page 153 of 278 -~ Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Containment Category: C. CTMT Conditions Degradation Threat: Loss Threshold: CNB3 UNPLANNED rapid drop in containment pressure following containment pressure rise Definition(s): 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: Rapid UNPLANNED loss of containment pressure (i.e., not attributable to containment cooling or condensation effects) following an initial pressure rise indicates a loss of primary containment integrity. This threshold relies on operator recognition of an unexpected response for the condition and therefore a specific value is not assigned. The unexpected (UNPLANNED) response is important because it is the indicator for a containment bypass condition.

Reference(s):

1. NEI 99-01 Primary Containment Conditions PC Loss 1.A Page 154 of 278

-~ Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Containment Category: C. CTMT Conditions . Degradation Threat: Loss Threshold: CNB4 Containment pressure response not consistent with LOCA conditions Definition(s): 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: Primary containment pressure should rise as a result of mass and energy release into the primary containment from a LOCA. Thus, primary containment pressure not rising under these conditions indicates a loss of primary containment integrity. These thresholds rely or;i operator recognition of an unexpected response for the condition and therefore a specific value is not assigned. The unexpected (UNPLANNED) response is important because it is the indicator for a containment bypass condition. Reference(s):

1. USAR Table 6.2-7, Results of Containment Response Analysis 2. USAR Table 6.2-1, Containment Design Parameters

3. NEI 99-01 Primary Containment Conditions PC Loss 1.8 Page 155 of 278

~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Containment Category: C. CTMT Conditions Degradation Threat: Potential Loss Threshold: CNBS Containment pressure > 15 psig Definition(s): VALID -An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the *condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. Basis: When the containment pressure exceeds the maximum allowable value (15 psig) (ref. 1 ), containment venting may be required even if offsite radioactivity release rate limits will be exceeded (ref. 2). This pressure is based on the containment design pressure as identified in the accident analysis. If this threshold is exceeded, a challenge to the containment structure has occurred because assumptions used in the accident analysis are no longer VALID and an unanalyzed condition exists. This constitutes a Potential Loss of the Containment barrier even if a containment breach has not occurred. The threshold pressure is the primary containment internal design pressure. Structural acceptance testing demonstrates the capability of the primary containment to resist pressures greater than the internal design pressure. A pressure of this magnitude is greater than those expected to result from any design basis accident and, thus, represent a Potential Loss of the Containment barrier. Reference(s):

1. USAR Table 6.2-1, Containment Design Parameters

2. EOP-2 Primary Containment Control 3. NEI 99-01, Primary Containment Conditions PC Potential Loss 1 .A Page 156 of 278

/ ~=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Containment Category: C. CTMT Conditions Degradation Threat: Potential Loss Threshold: CNB6 Drywell or containment hydrogen concentration > 4% Definition(s): None Basis: In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive mixture of dissolved gases in the containment. However, containment monitoring and/or sampling should be performed to verify this assumption and a General Emergency declared if it is determined that hydrogen concentration has exceeded the minimum necessary to support a hydrogen burn (4%). The Igniter System is designed to prevent hydrogen accumulation by locally burning hydrogen in a controlled manner as soon as the hydrogen enters the containment atmosphere and reaches the igniters. For high rates of hydrogen production, ignition occurs at the lowest concentration that can support ignition. Following ignition, hydrogen is consumed through formation of diffusion flames where the gas enters the containment, thus controlling hydrogen concentration at approximately 4% 1ref. 1 ). If hydrogen concentration reaches or exceeds the lower flammability limit, as defined in plant EOPs, in an oxygen rich environment, a potentially explosive mixture exists. If the combustible mixture ignites inside the containment, loss of the Containment barrier could occur. Reference(s):

1. EPSTG*0002 Appendix 8 EOP and SAP Bases 2. NEI 99-01 Primary Containment Conditions PC Potential Loss 1.8 Page 157 of 278

--Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Containment Category: C. CTMT Conditions Degradation Threat: Potential Loss Threshold: CNB7 Parameters cannot be restored and maintained within the safe zone of the HCTL curve (EOP Figure 2) . Definition(s): None Basis: The Heat Capacity Temperature Limit (HCTL) is the highest suppression pool temperature from which Emergency RPV Depressurization will not raise:

Suppression chamber temperature above the maximum temperature capability of the suppression chamber and equipment within the suppression chamber which may be required to operate when the RPV is pressurized, OR
Suppression chamber pressure above Primary Containment Pressure Limit, while the rate of energy transfer from the RPV to the containment is greater than the capacity of the containment vent. The HCTL is a function of RPV pressure, suppression pool temperature and suppression pool water level. It is utilized to preclude failure of the containment and equipment in the containment necessary for the safe shutdown of the plant.and therefore, the inability to maintain plant parameters below the limit constitutes a potential loss of containment.

The term "cannot be restored and maintained above" means the parameter value(s) is not able to be brought within the specified limit. The determination requires an evaluation of system performance and availability in relation to the parameter value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained within a specified limit does not require imm~diate action simply because the current value is outside . the limit, but does not permit extended operation outside the limit; the threshold must be considered reached as soon as it is apparent that operation within the limit cannot be attained. Page 158 of 278 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Reference(s):

1. EOP-2 Primary Containment Control 2. NEI 99-01 Primary Containment Conditions PC Potential Loss 1 .C i, ' Page 159 of 278

-::::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Containment Category: D. CTMT Radiation/RCS Activity Degradation Threat: Loss Threshold: None Page 160 of 278 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Containment Category: C. CTMT Radiation/RCS Activity Degradation Threat: Potential Loss J Threshold: CNBS Containment radiation (RMS-RE16) > 12,000 R/hr

Definition(s):

None Basis: In order to reach this Containment barrier Potential Loss threshold, a loss of the RCS barrier (Loss RCB6) and a loss of the Fuel Clad barrier (Loss FCB3) have already occurred. This threshold, therefore, represents a General Emergency classification. The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that 20% of the fuel cladding has failed (ref. 1 ). This level of fuel clad failure is well above that used to determine the analogous Fuel Clad Barrier Loss and RCS Barrier Loss thresholds. NUREG-1228, Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents, indicates the fuel clad failure must be greater than approximately 20% in order for there to be a major release of radioactivity requiring offsite protective actions. For this condition to exist, there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential loss of containment which would then escalate the emergency classification level to a General Emergency. There is no Containment barrier Loss threshold associated with CTMT Radiation/RCS Activity. I 'i Reference(s):

1. Calculation G13.18.9.4-045 Containment Doses for Emergency Action Levels (EALs) 2. NEI 99-01 NEI 99.:01 Primary Containment Radiation Potential Loss 4.A ! ,' Page 161 of 278

-::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Containment Category: E. CTMT Integrity or Bypass Degradation Threat: Loss Threshold: CNB9 UNISOL..ABLE direct downstream pathway to the environment exists after Containment isolation signal Definition(s): UN/SOLABLE -An open or breached system line that cannot be isolated, remotely or locally. VALID -An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. Basis: Failure to isolate the leak (from the Control Room or locally), within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification. This threshold addresses failure of open isolation devices which should close upon receipt of a manual or automatic containment isolation signal resulting in a significant radiological release pathway directly to the environment. The concern is the UNISOLABLE open pathway to the environment. A failure of the ability to isolate any one line indicates a breach of containment integrity. This threshold also applies to a containment bypass due to a HPCS or LPCS line break outside containment with injection check valve failure allowing an UNISOLABLE direct pathway for RCS release to the environment. The use of the modifier "direct" in defining the release path discriminates against release paths through interfacing liquid systems or minor release pathways, such as instrument lines, not protected by the Primary Containment Isolation System (PCIS). Leakage into a closed system is to be considered only if the closed system is breached and thereby creates a significant pathway to the environment. Examples include UNISOLABLE main steam line or RCIC steam line breaks, UNISOLABLE RWCU system breaks, and UNISOLABLE containment atmosphere vent paths. If the main condenser is available with an UNISOLABLE main steam line, there may be releases through the steam jet air ejectors and gland seal exhausters. These pathways Page 162 of 278 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases t are monitored, however, and do not meet the intent of an UNISOLABLE release path to the environment. These minor releases are assessed using the Category A, Abnormal Rad Levels / Rad Effluent, EALs. The existence of a filter is not considered in the threshold assessment. Filters do not remove fission product noble gases. In addition, a filter could become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream. EOP-2 Primary Containment Control, may specify primary containment venting and intentional bypassing of'the containment isolation valve logic, even if offsite radioactivity release rate limits are exceeded (ref. 1 ). Under these conditions with a VALID containment isolation signal, the Containment barrier should be considered lost. Refer to CNB10. Following the leakage of RCS mass into primary containment and a rise in primary containment pressure, there may be minor radiological releases associated with allowable primary containment leakage through various penetrations or system components. Minor releases may also occur if a primary containment isolation valve(s) fails to close but the primary containment atmosphere escapes to an enclosed system. These releases do not constitute a loss or potential loss of primary containment but should be evaluated using the Recognition Category A ICs. There is no Containment barrier Potential Loss threshold associated with CTMT Integrity or Bypass. Reference(s): \ 1. EOP-2 Primary Containment Control 2. NEI 99-01 Primary Containment Isolation Failure PC Loss 3.A Page 163 of 278 -::::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: . Containment Category: E. CTMT Integrity or Bypass Degradation Threat: Loss Threshold: CNB10 Intentional Containment venting per EOPs Definition(s): VALID -An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related* or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. Basis: EOP-2, Primary Containment Control, may specify containment venting and intentional bypassing of the containment isolation valve logic, even if offsite radioactivity release rate limits are exceeded. Under these conditions, with a VALID primary containment isolation signal, the threshold is met when the operator begins venting the containment in accordance with Enclosure 21, not when actions are taken to bypass interlocks prior to opening the vent valves (ref. 1 ). -Intentional venting of primary containment for primary containment pressure or combustible gas control to the secondary containment and/or the environment is a Loss of the Containment. Venting for primary containment pressure control when not in an accident situation (e.g., to control pressure below the drywell high pressure scram setpoint) does not meet the threshold condition. There is no Containment barrier Potential Loss threshold associated with CTMT Integrity or Bypass. Refe re nee( s): 1. EOP-2 Primary Containment Control 2. NEI 99-01 Primary Containment Isolation Failure PC Loss 3.8 Page 164 of 278 Barrier: Category: River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Containment E. CTMT Integrity or Bypass Degradation Threat: Potential Loss Threshold: None Page 165 of 278 -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Containment Category: F. Emergency Director Judgment Degradation Threat: Loss Threshold: CNB11 Any condition in the opinion of the Emergency Director that indicates loss of the Containment barrier Definition{s): None Basis: This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Containment Barrier is lost. Reference{s):

1. NEI 99-01 Emergency Director Judgment PC Loss 6.A Page 166 of 278

--====-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Barrier: Containment Category: E. Emergency Director Judgment Degradation Threat: Potential Loss Threshold: CNB12 Any condition in the opinion of the Emergency Director that indicates potential loss of the Containment barrier Definition(s): None Basis: This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Containment Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the everit that barrier status cannot be monitored. Reference(s):

1. NEI 99-01 Emergency Director Judgment PC Potential Loss 6.A Page 167 of 278

. -=:::;:-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category H -Hazards and Other Conditions Affecting Plant Safety EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.) . -Hazards are non-plant, system-related events that can directly or indirectly affect plant operation, reactor plant safety or personnel safety. \ 1. Security Unauthorized entry attempts into the PROTECTED AREA, bomb threats, sabotage attempts, and actual security compromises threatening loss of physical control of the plant. 2. Seismic Event Natural events such as earthquakes have potential to cause plant structure or equipment damage of sufficient magnitude to threaten personnel or plant safety.

3. Natural or Technological Hazard Other natural and non-naturally occurring events that can cause damage to plant facilities include tornados, FLOODING, hazardous material releases and events restricting site access warranting classification.

4. Fire FIRES can pose significant hazards to personnel and reactor safety. Appropriate for classification are FIRES within the plant PROTECTED AREA or which may affect operability of equipment needed for safe shutdown 5. Hazardous Gas Toxic, corrosive, asphyxiant or flammable gas leaks can affect normal plant operations or preclude access to plant areas required to safely shutdown the plant. 6. Control Room Evacuation Events that are indicative of loss of Control Room habitability.

If the Control Room must be evacuated, additional support for monitoring and controlling plant functions is necessary through the emergency response facilities.

7. Emergency Director Judgment The EALs defined in other categories specify the predetermined symptoms or events that are indicative of emergency or potential emergency conditions and thus warrant classification.

While these EALs have been developed to address the full spectrum of possible emergency conditions which may warrant classification and subsequent implementation of the Emergency Plan, a provision for classification of emergencies based on operator/management experience and judgment is still necessary. The EALs of this category provide the Emergency Director the latitude to classify emergency conditions consistent with the established classification criteria based upon Emergency Director judgment. Page 168 of 278

~ -::::::-Entergy RiverBend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

H -Hazards Subcategory: 1 -Security Initiating Condition: Confirmed SECURITY CONDITION or threat EAL: HU1.1 Unusual Event A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by RBS Security Shift Supervision OR Notification of a credible security threat directed at the site QR A validated notification from the NRG providing information of an aircraft threat Mode Applicability: All Definition(s): HOSTAGE -A person(s) held as leverage against the station to ensure that demands will be met by the station. HOSTILE ACTION -An act toward RBS or its personnel that includes the use of violent force to destroy eq6ipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile Action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on RBS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA). OWNER CONTROLLED AREA (OCA) -For the purposes of classification, the Security Owner Controlled Area (SOGA) or the area between the SOGA Fence and the PROTECTED AREA Boundary. PROJECTILE -An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety. PROTECTED AREA -The area within the perimeter of the RBS security fence. SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: Page 169 of 278

River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition;

-J ' (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. SECURITY CONDITION -Any security event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A security condition does not involve a HOSTILE ACTION. : Basis: This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR § 73.71 or 10 CFR § 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HA1 and HS1. Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event. Classification of these events will initiate appropriate threat-related notifications to plant personnel and Offsite Response Organizations. Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program]. The first threshold references the Security Shift Supervision because these are the individuals trained to confirm that a security event is occurring or has occurred. Training on security event confirmation and classification is controlled due to the nature of Safeguards and 10 CFR § 2.39 information. The second threshold addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with the Security Plan for RBS. The third threshold addresses the threat from the impact of an aircraft on the plant. The NRG Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may also be provided by NORAD through the NRG. Validation of the threat is performed in accordance with AOP-0063 Outside Threats (ref. 2). Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Security Plan for RBS (ref. 1 ). Escalation of the emergency classification level would be via IC HA 1. Page 170 of 278 ~Entergy River Bend Station EAL Basis Document Revision XXX I Attachment 1 -Emergency Action Level Technical Bases Reference(s):

1. RBS Security Plan 2. AOP-0063 Outside Threats 3. AOP-0054 Security Events 4. NEI 99-01 HU1 Page 171 of 278

-::::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: H -Hazards Subcategory: 1 -Security Initiating Condition: HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes EAL: HA1.1 Alert A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by RBS Security Shift Supervision OR A validated notification from NRC of an aircraft attack threat within 30 min. of the site Mode Applicability: All Definition(s): HOSTAGE -A person(s) held as leverage against the station to ensure that demands will be met by the station. HOSTILE ACTION -An act toward RBS or its personnel that includes the use of violent-force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILES, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile Action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on RBS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA). HOSTILE FORCE -One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction. OWNER CONTROLLED AREA (OCA) -For the purposes of classification, the Security Owner Controlled Area (SOGA) or the area between the SOGA Fence and the PROTECTED AREA Boundary.

-PROJECTILE

-An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety. *

PROTECTED AREA -The area within the perimeter of the RBS security fence. Page 172 of 278
--Entergy River Bend Station EAL Basis Document Revision XXX I Attachment 1 -Emergency Action Level Technical Bases Basis

This IC addresses the occurrence of a HOSTILE ACTION within the SECURITY OWNER CONTROLLED AREA or notification of an aircraft attack threat. This event will require rapid response and assistance due to the possibility of the attack progressing to the PROTECTED AREA, or the need to prepare the plant and staff for a potential aircraft impact. Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event. Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].

As time and conditions allow, these events require a heightened state of readiness by the plant *J staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Alert declaration will also heighten the awareness of Offsite Response Organizations (OROs), allowing them to be better prepared should it be necessary to consider further actions. This EAL does not apply to incidents that are accidental events, acts of civil disobedience, or 1 otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73. 71 or 10 CFR § 50. 72. The first threshold is applicable for any HOSTILE ACTION occurring, or that has occurred, in the OWNER CONTROLLED AREA. . The second threshold addresses the threat from the impact of an aircraft on the plant, and the anticipated arrival time is within 30 minutes. The intent of this EAL is to ensure that related notifications are made in a timely manner so that plant personnel and OROs are in a heightened state of readiness. This EAL is met when the threat-related information has been validated in accordance with AOP-0063 Outside Threats (ref. 2). The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may be provided by NORAD through the NRC. In some cases, it may not be readily apparent if an aircraft impact within the SECURITY OWNER CONTROLLED AREA was intentional (i.e., a HOSTILE ACTION). It is expected, although not certain, that notification by an appropriate Federal agency to the site would clarify this point. In this case, the appropriate federal agency is intended to be NORAD, FBI, FAA or NRC. The emergency declaration, including one based on other ICs/EALs, should not be \ unduly delayed while awaiting notification by a Federal agency. Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or Page 173 of 278 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases threat location. Security-sensitive information should be contained in non-public documents such as the Security Plan for RBS (ref. 1 ). Escalation of the emergency classification level would be via IC HS1. Reference(s):

1. RBS Security Plan 2. AOP-0063 Outside Threats 3. AOP-0054 Security Events 4. NEI 99-01 HA 1 Page 174 of 278

-===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: H -Hazards Subcategory: 1 -Security J Initiating Condition: HOSTILE ACTION within the PROTECTED AREA EAL: HS1 .1 Site Area Emergency A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by RBS Security Shift Supervision Mode Applicability: All Definition(s): HOSTAGE -A person(s) held as leverage against the station to ensure that demands will be met by the station. HOSTILE ACTION -An act toward RBS or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile Action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on RBS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between indiviquals in the OWNER CONTROLLED AREA). HOSTILE FORCE -One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction. OWNER CONTROLLED AREA (OCA) -For the purposes of classification, the Security Owner Controlled Area (SOGA) or the area between the SOGA Fence and the PROTECTED AREA Boundary. PROJECTILE -An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety. PROTECTED AREA -The area within the perimeter of the RBS security fence. Basis: This IC addresses the occurrence of a HOSTILE ACTION within the PROTECTED AREA. This event will require rapid response and assistance due to the possibility for damage to plant equipment. Timely arid accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event (ref. 1, 2, 3). Page 175 of 278 .~ -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program]. As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Site Area Emergency declaration will mobilize Offsite Response Organization (ORO) resources and have them available to develop and implement public protective actions in the unlikely event that the attack is successful in impairing multiple safety functions. This EAL does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72. Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Security Plan for RBS (ref. 1 ). Reference(s):

1. RBS Security Plan 2. AOP-0063 Outside Threats 3. AOP-0054 Security Events 4. NEI 99-01 HS1 Page 176 of 278

' I L__ ____ _ ---Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: Subcategory: H -Hazards and Other Conditions Affecting Plant Safety 2 -Seismic Event Initiating Condition: Seismic event greater than OBE levels EAL: HU2.1 Unusual Event Seismic event > OBE as indicated by EITHER of the following:

Annunciator P680-02A-C06, SEISMIC EVENT HIGH
Annunciator P680-02A-B06, SEISMIC EVENT HIGH/HIGH and amber lights illuminated on H13-P869 ERS-NBl101 Mode Applicability:

All Definition(s): None Basis: I This IC addresses a seismic event that results in accelerations at the plant site greater than those specified for an Operating Basis Earthquake (QBE). An earthquake greater than an QBE but less than a Safe Shutdown Earthquake (SSE) should have no significant impact on safety-related systems, structures and components; however, some time may be required for the plant staff to ascertain the actual post-event condition of the plant (e.g., performs downs and post-event inspections). Given the time necessary to perform walk-downs and inspections, and fully understand any impacts, this event represents a potential degradation of the level of safety of the plant. '-Event verification with external sources should not be necessary during or following an OBE. Earthquakes of this magnitude should be readily felt by on-site personnel and recognized as a seismic event. The Shift Manager or Emergency Director may seek external verification if deemed appropriate (e.g., a call to the U.S. Geological Survey (USGS), check internet news sources, etc.); however, the verification action must not preclude a timely emergency declaration. Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via EAL CA6.1 or SAB.1. -. To avoid inappropriate emergency classification resulting from spurious actuation of the seismic instrumentation or felt motion not attributable to seismic activity, an offsite agency (USGS, National Earthquake Information Center (NEIC)) can confirm that an earthquake has occurred in the area of the plant. Such confirmation should not, however, preclude a timely Page 177 of 278 Q *:::::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases emergency declaration based on receipt of the OBE alarm. If requested, provide the analyst with the following RBS coordinates: 30° 45' 26" north latitude, 91° 19' 54" west longitude (ref. 3). Alternatively, near real-time seismic activity can be accessed via the NEIC website. Reference(s): 1 . ARP-680-02 P680-02 Alarm Response 2. AOP-0028 Seismic Event 3. USAR section 2.1.1.1 Specification of Location 4. NEI 99-01 HU2 Page 178 of 278 -::::::-c Entergy Category: Subcategory: River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases

H -Hazards and Other Conditions Affecting Plant Safety 3 -Natural or Technological Hazard Initiating Condition:

Hazardous event EAL: HU3.1 Unusual Event A tornado strike within the PROTECTED AREA Mode Applicability: All Definition(s): PROTECTED AREA -The area within the perimeter of the RBS security fence. Basis: / This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant. This EAL addresses a tornado striking (touching down) within the PROTECTED AREA. Escalation of the emergency classification level would be based on I Cs in Recognition Categories A, F, S or C.

If damage is confirmed visually or by other in-plant indications, the event may be escalated to an Alert under EAL CA6.1 or SA8.1. A tornado striking (touching down) within the PROTECTED AREA warrants declaration of an Unusual Event regardless of the measured wind speed at the meteorological tower. A tornado is defined as a violently rotating column of air in contact with the ground and extending from the base of a thunderstorm.

Reference(s):

1. AOP-0029 Severe Weather Operation

2. NEI 99-01 HU3 Page 179 of 278

--===-Entergy River Bend Station EAL Basis Document Revision XXX I Attachment 1 -Emergency Action Level Technical Bases Category: H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 -Natural or Technological Hazard Initiating Condition: Hazardous event EAL: HU3.2 Unusual Event Internal room or area FLOODING of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component required by Technical Specifications for the current operating mode

Mode Applicability:

All Definition(s): FLOODING -A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area. SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related ( as defined in 1 OCFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: * (1) The integrity of the reac:::tor coolant pressure boundary; (2) The capability to shut down th~ reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant. This EAL addresses FLOODING of a building room or area that results in operators isolating power to a SAFETY SYSTEM component due to water level or other wetting concerns. Classification is also required if the water level or related wetting causes an automatic isolation of a SAFETY SYSTEM component from its power source (e.g., a breaker or relay trip). To warrant classification, operability of the affected component must be required by Technical Specifications for the current operating mode. Escalation of the emergency classification level would be based on I Cs in Recognition Categories A, F, S or C. Page 180 of 278

Q -===* Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Refer to EAL CA6.1 or SA8.1 for internal FLOODING affecting one or more SAFETY SYSTEM trains. Reference(s):

1. NEI 99-01 HU3 Page 181 of 278

-:::::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 -Natural or Technological Hazard Initiating Condition: Hazardous event EAL: HU3.3 Unusual Event Movement of personnel within the PROTECTED AREA is IMPEDED due to an event external to the PROTECTED AREA involving hazardous materials (e.g., an offsite chemical spill or toxic gas release) Mode Applicability: All Definition(s): IMPEDE(D) -Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed). PROTECTED AREA -The area within the perimeter of the RBS security fence. Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant. This EAL addresses a hazardous materials event originating at a location outside the PROTECTED AREA and of sufficient magnitude to IMPEDE the movement of personnel within the PROTECTED AREA. Escalation of the emergency classification level would be based on I Cs in Recognition Categories A, F, Sor C. Reference(s):

1. NEI 99-01 HU3 Page 182 of 278

, River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 -Natural or Technological Hazard Initiating Condition: Hazardous event EAL: HU3.4 Unusual Event A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles (Note 7) ' Note 7: This EAL does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents. Mode Applicability: All Definition(s): FLOODING -A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area. Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant. This EAL addresses a hazardous event that causes an on-site impediment to vehicle movement and significant enough to prohibit the plant staff from accessing the site using personal vehicles. Examples of such an event include site FLOODING caused by a hurricane, heavy rains, up-river water releases, dam failure, etc., or an on-site train derailment blocking the access road. This EAL is not intended apply to routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the FLOODING around the Cooper Station during the Midwest floods of 1993, or the FLOODING armmd Ft. Calhoun Station in 2011. Escalation of the emergency classification level would be based on I Cs in Recognition Categories A, F, S or C. , Reference(s):

1. NEI 99-01 HU3 Page 183 of 278

-::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 -Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL: HU4.1 Unusual Event A FIRE is not extinguished within 15 min. of any of the following FIRE detection indications (Note 1 ):

Report from the field (i.e., visual observation)
Receipt of multiple (more than 1) fire alarms or indications
Field verification of a single fire alarm AND The FIRE is located within any Table H-1 area Note 1: The Emergency Director should declare the event promptly upon determining that the time limit.has been exceeded, or will likely be exceeded.

The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded .. Mode Applicability: All Definition(s}.: Table H-1 Fire Areas

Reactor Building
Auxiliary Building
Fuel Building
Control Building
Standby Cooling Tower*
Diesel Generator Building
Tunnels (8, D,E, F, G) 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 ~ame is preferred but is not required if large quantities of smoke and heat are observed.

VALID -An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the Page 184 of 278 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases condition's existence, or the report's accuracy is removed. Implicit in'this definition is the need

for timely assessment.

Basis: This IC addresses the magnitude and extent of FIRES that may be in.dicative of a potential degradation of the level of safety of the plant. The intent of the 15-minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., smoldering waste paper basket). In addition to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation of a suppression system, etc. Upon receipt, operators will take prompt actions to confirm the validity of an initial fire alarm, indication, or report. For EAL assessment purposes, the emergency declaration clock starts at the time that the initial alarm, indication, or report was received, and not the time that a subsequent verification action was performed. Similarly, the fire duration clock also starts at the time of receipt of the initial alarm, indication or report. Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via EAL CA6.1 or SA8.1. The 15 minute requirement begins with a credible notification that a FIRE is occurring, or receipt of multiple VALID fire detection system alarms or field validation of a single fire alarm. The alarm is to be validated using available Control Room indications or alarms to prove that it is not spurious, or by reports from the field. Table H-1 Fire Areas are those areas that contain equipment necessary for safe operation and shutdown of the plant (ref. 1 ). Reference(s):

1. AOP-0052 Fire Outside the Main Control Room in Areas Containing Safety Related Equipment

2. NEI 99-01 HU4 Page 185 of 278

"===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 -Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL: HU4.2 Unusual Event Receipt of a single fire alarm (i.e., no other indications of a FIRE) AND The fire alarm is ,indicating a FIRE within any Table H-1 area AND The existence of a FIRE is not verified within 30 min. of al 9 rm receipt (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: All Definition(s): Table H-1 Fire Areas

Reactor Building
Auxiliary Building
Fuel Building
Control Building
Standby Cooling Tower
Diesel Generator Building
Tunnels (B, D,E, F, G) 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.

VALID -An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment. Page 186 of 278 ~Entergy River Bend Station EAL Basis Document Revision XXX Attachment1 -Emergency Action Level Technical Bases Basis: This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant. This EAL addresses receipt of a single fire alarm, and the existence of a FIRE is not verified (i.e., proved or disproved) within 30-minutes of the alarm. Upon receipt, operators will take prompt actions to confirm the validity of a single fire alarm. For EAL assessment purposes, the 30-minute clock starts at the time that the initial alarm was received, and not the time that a subsequent verification action was performed. A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failure or a spurious activation, and not an actual FIRE. For this reason, additional time is allowed to verify the validity of the alarm. The 30-minute period is a reasonable amount of time to determine if an actual FIRE exists; however, after that time, and absent information to the contrary, it is assumed that an' actual FIRE is in progress. If an actual FIRE is verified by a report from the field, then HU4.1 is immediately applicable, and the emergency must be declared if the FIRE is not extinguished within 15-minutes of the report. If the alarm is verified to be due to an equipment failure or a spurious activation, and this verification occurs within 30-minutes of the receipt of the alarm, then this EAL is not applicable and no emergency declaration is warranted. Appendix R to 10 CFR 50, states in part: Criterion 3 of Appendix A to this part specifies that "Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions." When considering the effects of fire, those systems associated with achieving and maintaining safe shutdown conditions assume major importance to safety because damage to them can lead to core damage resulting from loss of coolant through boil-off. Because fire may affect safe shutdown systems and because the loss of function of systems used to mitigate the consequences of design basis accidents under post-fire conditions does not per se impact public safety, the need to limit fire damage to systems required to achieve and maintain safe shutdown conditions is greater than the need to limit fire damage to those systems required to mitigate the consequences of design basis accidents. In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of 1-hour fire barriers for the enclosure of cable and equipment and associated non-safety circuits of one redundant train (G.2.c).,, As used in HU4.2, the 30-minutes to verify a single alarm is well within this worst-case 1-hour time period. Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via EAL CA6.1 or SA8.1. Page 187 of 278 -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases The 30 minute requirement begins upon receipt of a single VALID fire detection system alarm. The alarm is to be validated using available Control Room indications or alarms to prove that it is not spurious, br by reports from the field. Actual field reports must be made within the 30 minute time limit or a classification must be made. If a fire is verified to be occurring by field report, classification shall be made based on EAL HU4.1, with the 15 minute requirement beginning with the verification of the fire by field report. Table H-1 Fire Areas are those areas that contain equipment necessary for safe operation and shutdown of the plant (ref. 1 ). Reference(s):

1. AOP-0052 Fire Outside the Main Control Room in Areas Containing Safety Related Equipment

2. NEI 99-01 HU4 Page 188 of 278 Q -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 -Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL: HU4.3 Unusual Event A FIRE within the PROTECTED AREA not extinguished within 60 min. of the initial report, alarm or indication (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: All Definition(s): FIRE -Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities-of smoke and heat are observed. PROTECTED AREA -The area within the perimeter of the RBS security fence. Basis: This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant. In addition to a FIRE addressed by EAL HU4.1 or HU4.2, a FIRE within the plant PROTECTED AREA not extinguished within 60-minutes may also potentially degrade the level of plant safety. Depending upon the plant mode at the time of the event, escalation of the emerg*ency classification level would be via EAL CA6.1 or SA8.1. Reference(s):

1. NEI 99-01 HU4 Page 189 of 278 River Bend Station EAL Basis Document Revision XXX ' . Attachment 1 -Emergency Action Level Technical Bases Category:

H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 -Fire ' Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL: HU4.4 Unusual Event A FIRE within the PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish Mode Applicability: All Definition(s): FIRE -Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed. PROTECTED AREA -The area within the perimeter of the RBS security fence. Basis: This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant. If a FIRE within the plant PROTECTED AREA is of sufficient size to require a response by an offsite firefighting.agency (e.g., a local town Fire Department), then the level of plant safety is potentially degraded. The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish. Declaration is not necessary if the agency resources are placed on stand-by, or supporting post-extinguishment recovery or investigation actions. Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via EAL CA6.1 or SA8.1. Reference(s):

1. NEI 99-01 HU4 Page 190 of 278

-~ Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 5 -Hazardous Gas Initiating Condition: Gaseous release IMPEDING access to equipment necessary for normal plant operations, cooldown or shutdown EAL: HAS.1 Alert Release of a toxic, corrosive, asphyxiant or flammable gas into any Table H-2 room or area AND Entry into the room or area is prohibited or IMPEDED (Note 5) Note 5: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted.

Table H-2 Safe Operation

& Shutdown Rooms/Areas Room/Area Auxiliary Building 70' RHR B Pump Room Auxiliary Building 80' RHR A Pump Room Auxiliary Building 114' West Control Building 95' Div 1 RSS Room\ Mode Applicability: 3 -Hot Shutdown Definition(s): ' Mode 3 3 3 3 IMPEDE(D) -Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed). Basis: This IC addresses an event involving a release of a hazardous gas that precludes or IMPEDES access to equipment necessary to maintain normal plant operation, or required for a normal plant cooldown and shutdown. This condition represents an actual or potential substantial degradation of the level of safety of the plant. 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 gaseous release. The Page 191 of 278 -::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases emergency classification is not contingent upon whether entry is actually necessary at the time of the release. Evaluation of the IC and EAL does not require atmospheric sampling; it only requires the Emergency Director's judgment that the gas concentration in the affected room/area is sufficient to preclude or significantly IMPEDE procedurally required access. This judgment may be based on a variety of factors including an existing job hazard analysis, report of ill effects on personnel, advice from a subject matter expert or operating experience with the same or similar hazards. Access should be considered as IMPEDED if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed). 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 gaseous release).

For example, the plant is in Mode 1 when the gaseous release occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 3.

The gas release is a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., fire suppression system testing).

I

The access control measures are of a conservative or precautionary nature, and would not actually prevent or IMPEDE a required action.
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 s*pecifications at the time of the event. An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels. Most commonly, asphyxiants work by merely displacing air in an enclosed environment.

This reduces the concentration of oxygen below the normal level of around 19%, which can lead to breathing difficulties, unconsciousness or even death. This EAL does not apply to firefighting activities that generate smoke and that automatically or manually activate a fire suppression system in'-an area. Escalation of the emergency classification level would be via Recognition Category A, C or F ICs. The list of plant rooms or areas with entry-related mode applicability identified specify those 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. 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) are not included. In addition, the list specifies the plant

mode(s) during which entry would be required for each room or area (ref. 1 ). Page 192 of 278 Q *:;::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases EAL HA5.1 mode applicability has been limited to the mode limitations of Table H-2 (Mode 3 only). l* Reference(s):

1. Attachment 2 Safe Operation

& Shutdown Areas Tables A-3 & H-2 Bases 2: NEI 99-01 HA5 Page 193 of 278 -:;::::-, Entergy I River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 6 -Control Room Evacuation Initiating Condition: Control Room evacuation resulting in transfer of plant control to alternate locations EAL: HA6.1 Alert An event has resulted in plant control being transferred from the Control Room to the Remote Shutdown Panels Mode Applicability: All Definition(s ): None Basis: This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations outside the Control Room. The loss of the ability to control the plant from the Control Room is considered to be a potential substantial degradation in the level of plant safety. Following a Control Room evacuation, control of the plant will be transferred to alternate shutdown locations. The necessity to control a plant shutdown from outside the Control Room, in addition to responding to the event that required the evacuation of the Control Room, will present challenges to plant operators and other on-shift personnel. Activation of the ERO and emergency response facilities will assist in responding to these challenges. Transfer of plant control begins when the last licensed operator leaves the Control Room. Escalation of the emergency classification level would be via IC HS6. Reference(s):

1. AOP-0031 Shutdown from Outside the Main Control Room 2. NEI 99-01 HA6 Page 194 of 278

-===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 6 -Control Room Evacuation Initiating Condition: Inability to control a key safety function from outside the Control Room EAL: HS6.1 Site Area Emergency An event has resulted in plant control being transferred from the Control Room to the Remote Shutdown Panels AND Control of any of the following key safety functions is not re-established within 15 min. (Note 1 ):

Reactivity (Modes 1 and 2 only)
RPV water level
RCS heat removal Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.

The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 1 -Power Operation, 2 -Startup, 3 -Hot Shutdown, 4 -Cold Shutdown, 5 -Refueling Definition(s): None Basis: This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations, and the control of a key safety function cannot be reestablished in a timely manner. The failure to gain control of a key safety function following a transfer of plant control to alternate locations is a precursor to a challenge to one or more fission product barriers within a relatively short period of time. The determination of whether or not "control" is established at the remote safe shutdown location(s) i~ based on Emergency Director judgment. The Emergency Director is expected to make a reasonable, informed judgment within 15 minutes whether or not the operating staff has control of key safety functions. from the remote safe shutdown location(s). Transfer of plant control and the time period to establish control begins when the last licensed operator leaves the Control Room. Escalation of the emergency classification level would be via IC FG1 or CG1 Page 195 of 278 ~Entergy Reference(s}: River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases 1. AOP-0031 Shutdown from Outside the Main Control Room 2. EP FAQ 2015-014 3. NEI 99-01 HS6 ) Page 196 of 278 a::::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 -Emergency Director Judgment Initiating Condition: Other conditions exist that in the judgment of the Emergency Director warrant declaration of a UE EAL: HU7.1 Unusual Event Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response -or monitoring are expected unless further degradation of SAFETY SYSTEMS occurs. Mode Applicability: All Definition(s): SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically"systems classified as safety-related (as defined in 1 OCFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: ' (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the e.mergency classification level description for an UNUSUAL EVENT. Reference(s):

1. NEI 99-01 HU7 Page 197 of 278 I -::::::-, Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 -Emergency Director Judgment Initiating Condition: Other conditions exist that in the judgment of the Emergency Director warrant declaration of an ALERT EAL: HA7.1 Alert Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels. Mode Applicability: All Definition(s): HOSTAGE -A person(s) held as leverage against the station to ensure that demands will be met by the station. HOSTILE ACTION -An act toward RBS or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile Action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on RBS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA). OWNER CONTROLLED AREA (OCA) -For the purposes of classification, the Security Owner Controlled Area (SOCA) or the area between the SOCA Fence and the PROTECTED AREA Boundary. PROJECTILE -An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety. PROTECTED AREA -The area within the perimeter of the RBS security fence. Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the' emergency classification level description for an ALERT. Page 198 of 278 -:;::::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Reference(s):

1. NEI 99-01 HA? Page 199 of 278 Q --::::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 -Emergency Director Judgment Initiating Condition: Other conditions exist that in the judgment of the Emergency Director warrant declaration of a SITE AREA EMERGENCY EAL: HS7.1 Site Area Emergency Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts, (1) toward site personnel or equipment that could lead to the likely failure of or, (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action ~uideline exposure levels beyond the SITE BOUNDARY Mode Applicability: All Definition(s): HOSTILE ACTION -An act toward RBS or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an erid. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile Action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on RBS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA). . OWNER CONTROLLED AREA (OCA) -For the purposes of classification, the Security Owner Controlled Area (SOGA) or the area between the SOGA Fence and the PROTECTED AREA . Boundary. PROJECTILE -An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety. PROTECTED AREA -The area within the,perimeter of the RBS security fence. SITE BOUNDARY -For classification and dose projection purposes, the Site Boundary is the area defined as the exclusion area or exclusion zone in 1OCFR100.3 (a) which is a boundary of approximately 3,000 feet (or 0.5748 mile) from the RBS reactor centerline. Page 200 of 278 -I ~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a SITE AREA EMERGENCY. Reference(s):

1. NEI 99-01 HS? I Page 201 of 278 (

~Entergy River Bend Station EAL Basis Document Revis/on XXX Attachment 1 -Emergency Action Level Technical Bases Category: H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 -Emergency Director Judgment Initiating Condition: Other conditions exist that in the judgment of the Emergency Director warrant declaration of a GENERAL EMERGENCY EAL: HG7.1 General Emergency Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area : Mode Applicability: All Definition(s):

-HOSTAGE -A person(s) held as leverage against the station to ensure that demands will be met by the station. HOSTILE ACTION -An act toward RBS. or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.

Hostile Action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on RBS. Non-terrorism-based EALs sho,uld be used to address such activities (i.e., this may include violent acts between . individuals in the OWNER CONTROLLED AREA). ' IMMINENT -The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. OWNER CONTROLLED AREA (OCA) -For the purposes of classification, the Security Owner Controlled Area (SOGA) or the area between the SOGA Fence and the PROTECTED AREA Boundary. PROJECTILE -An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety. PROTECTED AREA -The area within the perimeter of the 'RBS security fence. Page 202 of 278 Q ~::c Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a GENERAL EMERGENCY. Reference(s):

1. NEI 99-01 HG? Page 203 of 278

'-River Bend Station EAL Basis Document Revision XXX -Attachment 1 -Emergency Action Level Technical Bases Category S -System Malfunction ' . EAL Group: Hot Conditions (RCS temperature > 200°F); EALs in this category are applicable only in one or more hot operating modes. Numerous system-related equipment failure events that warrant emergency classification have been identified in this category. They may pose actual or potential threats to plant safety. The events of this category pertain to the following subcategories: 1 . Loss of Emergency AC Power Loss of emergency plant electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of onsite and offsite power sources for 4.16 KV ENS buses. 2. Loss of Vital DC Power Loss of emergency plant electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and emergency core cooling systems which may be nE~cessary to ensure fission product barrier integrity. This category includes loss of power to or degraded voltage on the 125V DC vital buses. 3. Loss of Control Room Indications Certain events that degrade plant operator ability to effectively assess plant conditions within the plant warrant emergency classification. Losses of indicators are in this subcategory.

4. RCS Activity During normal operation, reactor coolant fission product activity is very low. Small concentrations of fission products in the coolant are primarily from the fission of tramp uranium in the fuel clad or minor perforations in the clad itself. Any significant rise from these base-line levels (2% -5% clad failures) is indicative offuel failures and is covered under the Fission Product Barrier Degradation category.

However, lesser amounts of clad damage may result in coolant activity exceeding Technical Specification limits. These fission products will be circulated with the reactor coolant and can be detected by coolant sampling.

5. RCS Leakage The reactor pressure vessel provides a volume for the coolant that covers the reactor core. The reactor pressure vessel and associated pressure piping (reactor coolant system) together provide a barrier to limit the release of radioactive material should the reactor fuel clad integrity fail. Excessive RCS leakage greater than Technical Specification limits indicates potential pipe cracks that may propagate to an extent threatening fuel clad, RCS and containment integrity.

Page 204 of 278 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases 6. RPS Failure This subcategory includes events related to failure of the Reactor Protection System (RPS) to initiate and complete reactor scrams. In the plant licensing basis, postulated failures of the RPS to complete a reactor scram comprise a specific set of analyzed events referred to as Anticipated Transient Without Scram (ATWS) events. For EAL classification, however, ATWS is intended to mean any scram failure event that does not achieve reactor shutdown. If RPS actuation fails to assure reactor shutdown, positive control of reactivity is at risk and could cause a threat to fue! clad, RCS and containment integrity.

7. Loss of Communications Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification.

8. Hazardous Event Affecting Safety Systems Various natural and technological events that result in degraded plant SAFETY SYSTEM performance or significant VISIBLE DAMAGE warrant emergency classification under this subcategory.

Page 205 of 278 '--:::::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: S -System Malfunction Subcategory: 1 -Loss of Emergency AC Power Initiating Condition: Loss of all offsite AC power capability to ENS buses for 15. minutes or longer EAL: SU1.1 Unusual Event Loss of all offsite AC power capability, Table S-1, to DIV I and DIV II 4.16 KV ENS buses for~ 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining tl)at the time limit has* been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Table 5-1 AC Power Sources Offsite

1RTX-XSR1C

,* 1RTX-XSR1D Onsite

EGS-EG1A
EGS-EG18 Mode Applicability:

1 -Power Operation, 2 -Startup, 3 -Hot Shutdown Definition(s): None Basis: The DIV Ill bus (1 E22*S004) is not credited because it only supplies power to the HPCS pump and associated loads, not any long term decay heat removal systems. In particular, suppression pool cooling mechanisms would be essential subsequent to a station blackout. This IC addresses a prolonged foss of offsite power. The loss of offsite power sources renders the plant more vulnerable to a complete loss of power to AC ENS buses. This condition represents a potential reduction in the level of safety of the plant. For emergency classification purposes, "capability" means that an offsite AC power source(s) is available to the ENS buses, whether or not the buses are powered from it. Page 206 of 278

===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 ,_ Emergency Action Level Technical Bases Fifteen minutes was selected as a threshold to exclude transient or momentary losses of offsite power. Escalation of the emergency classification level would be via IC SA 1. Refe re nee( s): 1. USAR Section 8.1 Electric Power Introduction

2. EE-001AC Startup Electrical Distribution Chart 3. SOP-0046 4.16 KV System 4. AOP-0004 Loss of Offsite Rower 5. NEI 99-01 SU1 Page 207 of 278

\ __ I -:.:==-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: S -System Malfunction Subcategory: 1 -Loss of Emergency AC Power Initiating Condition: Loss of all but one AC power source to ENS buses for 15 minutes or longer EAL: SA1.1 Alert AC power capability, Table S-1, to DIV I and DIV II 4.16 KV ENS buses reduced to a single power source for~ 15 min. (Note 1)

AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.

The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Table S-1 AC Power Sources Offsite

1 RTX-XSR1 C
1 RTX-XSR1 D Onsite
EGS-EG1A
EGS-EG1 B Mode Applicability:

1 -Power Operation, 2 -Startup, 3 -Hot Shutdown Definition(s): SAFETY SYSTEM -A system required for safe plantoperation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 1 OCFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; Page 208 of 278 I I -==-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis: The DIV Ill bus (1 E22*S004) is not credited because it only supplies power to the HPCS pump and associated loads, not any long term decay heat removal systems. In particular, suppression pool cooling mechanisms would be essential subsequent to a station blackout. This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment. This IC provides an escalation path from IC SU1. An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an ENS bus. Some examples of this condition are presented below.

A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).
A loss of all offsite power and loss of all ENS emergency power sources (e.g., onsite diesel generators) with a single train of ENS buses being back-fed from the unit main generator.
A loss of ENS emergency power sources (e.g., onsite diesel generators) with a single train of ENS emergency buses being fed from an offsite power source_. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power. Escalation of the emergency classification level would be via IC SS1. This EAL is the hot condition equivalent of the cold condition EAL CU2.1 . . Reference(s):

1. USAR Section 8.1 Electric Power Introduction

2. EE-001AC Startup Electrical Distribution Chart 3. SOP-0046 4.16 KV System 4. AOP-0004 Loss of Offsite Power 5. NEI 99-01 SA1 Page 209 of 278

\ *===-Entergy River Bend Station EAL Basis Document Revision XXX -Attachment 1 -Emergency Action Level Technical Bases Category: S -System Malfunction Subcategory: 1 -Loss of Emergency AC Power ) Initiating Condition: Loss of all offsite and all onsite AC power to ENS buses for 15 minutes or longer EAL: SS1 .1 Site Area Emergency Loss of all offsite and all onsite AC power capability to DIV I and DIV II 4.16 KV ENS buses for~ 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 1 -Power Operation, 2 -Startup, 3 -Hot Shutdown Definition(s): SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 1 OCFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. 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. Mitigative strategies using other power sources (Division Ill diesel generator, FLEX generators, etc.) may be effective in supplying power to these buses. These power sources must be controlled in accordance with abnormal or emergency operating procedures, or beyond design basis accident response guidelines (e.g., FLEX support guidelines) and must be capable * (alone or in combination) of supplying power for long term decay heat removal systems. In particular, suppression pool cooling systems would be essential subsequent to a station blackout. In addition, fission product barrier monitoring capabilities may be degraded under Page 210 of 278 -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases these conditions. This IC represents

a condition that involves actual or likely major failures of plant functions needed for the protection of the public. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be _via IC AG1, FG1 or SG1. This EAL is the hot condition equivalent of the cold condition EAL CA2.1. Reference(s):

1. USAR Section 8.1 Electric Power Introduction

2. EE-001AC Startup Electrical Distribution Chart 3. SOP-0046 4.16 KV System 4. AOP-0004 Loss of Offsite Power 5. NEI 99-01 SS1 / Page 211 of 278

-::::.::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -: Emergency Action Level Technical Bases Category: S -System Malfunction Subcategory: 1 -Loss of Emergency AC Power Initiating Condition: Prolonged loss of all offsite and all onsite AC power to ENS buses EAL: SG1 .1 General Emergency Loss of all offsite and all onsite AC power capability to DIV I and DIV II 4.16 KV ENS buses AND EITHER:

Restoration of at least one 4.16 KV ENS bus in < 4 hours is not likely (Note 1)
RPV water level cannot be restored and maintained>

-187 in. Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 1 -Power Operation, 2 -Startup, 3 -Hot Shutdown Definition(s): SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECG~. These are typically systems classified as safety-related (as defined in 10CFR50.2): ' Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: ( 1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis: Indication of continuing core cooling degradation is manifested by the inability to restore and maintain RPV water level above the Minimum Steam Cooling Reactor Water Level (-187 in.). (ref. 5). Core submergence is the most desirable means of core cooling, however when RPV level is below TAF, the uncovered portion of the core can be cooled by less reliable means I (i.e., steam cooling or spray cooling). This IC addresses a prolonged loss of all power sources to AC ENS emergency buses. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric Page 212 of 278 -===--Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 ..:.... Emergency Action Level Technical Bases power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. Mitigative strategies using other power sources (Division Ill diesel generator, FLEX generators, etc.) may be effective in supplying power to these buses. These power sources must be controlled in accordance with abnormal or emergency operating procedures, or beyond design basis accident response guidelines (e.g., FLEX support guidelines) and must be capable (alone or in combination) of supplying power for long term decay heat removal systems. In particular, suppression pool cooling systems would be essential subsequent to a station blackout. A prolonged loss of these buses will lead to a loss of one or more fission product barriers. In addition, fission product barrier monitoring capabilities may be degraded under these conditions. The EAL should require declaration of a General Emergency prior to meeting the thresholds for IC FG1. This will allow additional time for implementation of offsite protective actions. Escalation of the emergency classification from Site Area Emergency will occur if it is projected that power cannot be restored to at least one AE ENS emergency bus by the end of the analyzed station blackout coping period., Beyond this time, plant responses and event . trajectory are subject to greater: uncertainty, and there is a greater likelihood of challenges to multiple fission product barriers. 4 hours is the site-specific SBO coping analysis time (ref. 6). . ' The estimate for restoring at least one ENS emergency bus should be based on a realistic appraisal of the situation. Mitigation actions with a low probability of success should not be used as a basis for delaying a classification upgrade. The goal is to maxin:,ize the time available to prepare for, and implement, protective actions for the public. The EAL will also require a General Emergency declaration if the loss of AC power results in parameters that indicate an inability to adequately remoye decay heat from the core. Reference(s):

1. USAR Section 8.1 Electric Power Introduction

2. EE-001AC Startup Electrical Distribution Chart 3. SOP-0046 4.16 KV System 4. AOP-0004 Loss of Offsite Power 5. EOP-1 RPV Control 6. USAR Appendix 15C Station Blackout 7. NEI 99-01 SG1 Page 213 of 278

-:::::--Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: S -System Malfunction Subcategory: 1 -Loss of Emergency AC Power Initiating Condition: Loss of all ENS AC and vital DC power sources for 15 minutes or longer EAL: SG1 .2 General Emergency I Loss of all offsite and all onsite AC power capability to DIV rand DIV II 4.16 KV ENS buses for~ 15 min. (Note 1) AND Indicated voltage is< 105 VDC on Safety Related DIV I and DIV II 125 VDC buses for 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 1 -Power Operation, 2 -Startup, 3 -Hot Shutdown Definition(s): SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis: Safety Related DC buses ENB-SWG01A (DIV I) and ENB-SWG01 B (DIV II) feE;d the Division I and Division II loads respectively. The Division I and Division II batteries each have 60 cells with a specific minimum voltage of 1. 75 volts/cell. These cell voltages yield minimum design bus voltages of 105 VDC (ref. 5). This IC addresses a concurrent and prolonged loss of both emergency ENS AC and Vital DC power. A loss of all emergency ENS AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, Page 214 of 278

===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. Mitigative strategies using other power sources (Division Ill diesel generator, FLEX generators, etc.) may be effective in supplying power to these buses. These power sources must be controlled in accordance with abnormal or emergency operating procedures, or beyond design basis accident response guidelines (e.g., FLEX support guidelines) and must be capable (alone or in combination) of supplying power for long term decay heat removal systems. In particular, suppression pool cooling systems would be essential subsequent to a station blackout.

A loss of vital DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both emergency ENS AC and Vital DC power will lead to multiple challenges to fission product barriers. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. The 15-minute emergency declaration clock begins at the point when both EAL thresholds are met. Reference(s):

1. USAR Section 8.1 Electric Power Introduction

2. EE-001AC Startup Electrical Distribution Chart 3. SOP-0046 4.16 KV System 4. AOP-0004 Loss of Offsite Power 5. Safety Related Battery Specification 244.521 6. USAR 8.3.2 DC Power Systems 7. NEI 99-01 SGS Page 215 of 278 I River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

S -System Malfunction Subcategory: 2 -Loss of Vital DC Power Initiating Condition: Loss of all vital DC power for 15 minutes or longer EAL: SS2.1 Site Area Emergency Indicated voltage is< 105 VDC on Safety Related DIV I and DIV II 125 VDC buses for 15 min. (Note 1) ' Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 1 -Power Operation, 2 -Startup, 3 -Hot Shutdown Definition(s): SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the told shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 1 OCFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential o,:ffsite expo~ures. Basis: Safety Related DC buses ENB-SWG01A (DIV I) and ENB-SWG01 B (DIV II) feed the Division I and Division II loads respectively. The Division I and Division II batteries each have 60 cells with a specific minimum voltage of 1. 75 volts/cell. These cell voltages yield minimum design bus voltages of 105 voe (ref. 1 ).

This IC addresses a loss of vital DC power which compromises the ability to monitor and control SAFETY SYSTEMS. In modes above Cold Shutdown, this condition involves a major failure of plant functions needed for the protection of the public. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. ' Escalation o(the emergency classification level would be via IC AG1, FG1 or SG1. This EAL is the hot condition equivalent of the cold condition EAL CU4.1. Page 216 of 278
===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Reference(s):

1. Safety Related Battery Specification 244.521 2. USAR 8.3.2 DC Power Systems 3. NEI 99-01 SS8 Page 217 of 278

~=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: S -System Malfunction Subcategory: 3 -Loss of Control Room Indications Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer EAL: SU3.1 Unusual Event An UNPLANNED event results in the inability to monitor one or more Table S-2 parameters from within the Control Room for~ 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Table S-2 Safety System Parameters*

Reactor power
RPV water level
RPV pressure
Containment pressure
Suppression Pool water level
Suppression Pool temperature Mode Applicability:

1 -Power Operation, 2 -Startup, 3 -Hot Shutdown Definition(s ): SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 1 OCFR50.2): . Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. 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. Page 218 of 278 Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -E_mergency Action Level Technical Bases Basis: This IC addresses the difficulty associated with monitoring normal plant conditions without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition is a precursor to a more significant event and represents a potential degradation in the level of safety of the plant. As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room. An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making. This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, RPV water level and RCS heat removal. The loss-of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for RPV water level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well. Fifteen minutes was selected as a threshold to exclude transiel')t or momentary losses of indication. Escalation of the emergency classification level would be via EAL SA3.1. Refe re nee( s): 1. USAR 7.5 Safety-Related Display Instrumentation

2. NEI 99-01 SU2 Page 219 of 278 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

S -System Malfunction Subcategory: 3 -Loss of Control Room Indications Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress EAL: SA3.1 Alert An UNPLANNED event results in the inability to monitor one or more Table S-2 parameters from within the Control Room for~ 15 min. (Note 1) AND Any significant transient is in progress, Table S-3 Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: Table 5-2 Safety System Parameters

Reactor power
RPV water level
RPV pressure
Containment pressure
Suppression Pool water level
Suppression Pool temperature Table 5-3 Significant Transients
Reactor scram
Runback > 25% thermal reactor power
Electrical load rejection

> 25% full electrical load

ECCS injection
Thermal power.oscillations>

10% 1 -Power Operation, 2 -Startup, 3 -Hot Shutdown Page 220 of 278 Q . -~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Definition(s): SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems .,classified as safety-related (as defined in 1 OCFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reacto(coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. 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 tr~nsient. The cause of the parameter change or event may be known or unknown. Basis: This IC addresses the difficulty associated with monitoring rapidly changing plant conditions during a transient without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. During this condition, the margin to a potential fission product barrier challenge is reduced. It thus represents a potential substantial degradation in the level of safety of the plant. As used in this EAL, an "inability to monitor means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room. An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRG event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making. This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, RPV water level and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for RPV water level cannot be determined from the indications and recorders on a main control board, Page 221 of 278 -::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases the SPDS or the plant computer, the availability of other parameter values may be compromised as well. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication. Escalation of the emergency classification level would be via IC FS1 or AS1 Reference(s):

1. USAR 7.5 Safety-Related Display Instrumentation

2. NEI 99-01 SA2 Page 222 of 278 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

S -System Malfunction Subcategory: 4 -RCS Activity Initiating Condition: RCS activity greater than Technical Specification allowable limits EAL: SU4.1 Unusual Event Offgas Pretreatment radiation monitor high alarm (P601-22A-F03, OFF GAS PRE-TREAT HIGH RADIATION) Mode Applicability: 1 -Power Operation, 2 -Startup, 3 -Hot Shutdown Definition(s): None Basis: The Offgas Pretreatment monitors radioactivity in the Offgas system downstream of the Offgas condenser. The monitor detects the radiation level that is attributable to the fission gases produced in the reactor and transported with steam through the turbine to the condenser. The High alarm indicates that the radioactivity present at the recombiner ~ffluent discharge is approaching the Technical Specification 3.7.4 limit. The nominal setpoint of 1.5 times the full power process background radiation level ensures that the activity will not exceed a value corresponding to the Technical Specification LCO 3. 7.4 allowable release rate. (ref. 1) This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant. Escalation of the emergency classification level would be via IC FA1 or the Recognition Category A ICs. Reference(s):

1. TRM section 3.3.7.8.2 Offgas System Radiation Monitoring Instrumentation 2 .. USAR 11.5 Process and Effluent Radiological Monitoring and Sampling Systems 3. Technical Specification 3.7.4 Main Condenser Offgas 4. NEI 99-01 SU3 Page 223 of 278 )

Q ~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: S -System Malfunction Subcategory: 4 -RCS Activity Initiating Condition: RCS activity greater than Technical Specification allowable limits EAL: SU4.2 Unusual Event Coolant activity> 0.2 µCi/gm dose equivalent 1-131 for> 48 hours OR Coolant activity> 4.0 µCi/gm dose equivalent 1-131 instantaneous Mode Applicability: 1 -Power Operation, 2 -Startup, 3 -Hot Shutdown Definition(s): None Basis: This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant. Escalation of the emergency classification level would be via IC FA 1 or the Recognition Category A ICs. Reference(s):

1. Technical Specification B 3.4.8, RCS Specific Activity bases 2. USAR Section 15.6.4 Steam System Piping Break Outside Containment

3. NEI 99-01 SU3 Page 224 of 278 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

S -System Malfunction Subcategory: 5 -RCS Leakage Initiating Condition: RCS leakage for 15 minutes or longer EAL: SUS.1 Unusual Event RCS unidentified or pressure boundary leakage> 10 gpm for;;:: 15 min. (Note 1) OR RCS identified leakage> 25 gpm for;;:: 15 min. (Note 1) OR Leakage from the RCS to a location outside Containment> 25 gpm for;;:: 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Mode Applicability: 1 -Power Operation, 2 -Startup, 3 -Hot Shutdown Definition(s): UN/SOLABLE -An open or breached system line that cannot be isolated, remotely or locally. Basis: Failure to isolate the leak (from the Control Room or locally) within 15 minutes, or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification. Identified leakage is leakage into the drywell, such as that from pump seals or valve packing, that is captured and conducted to a collecting sump; or leakage into the drywell atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary leakage. Unidentified leakage is all leakage into the drywell that is not identified leakage (ref. 2, 3). Pressure boundary leakage is leakage through a non-isolable fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall (ref. 2, 3). This IC addresses RCS leakage which may be a precursor to a more significant event. In this case, RCS leakage has been detected and operators, following applicable procedures, have been unable to promptly isolate the leak. This condition is considered to be a potential degradation of the level of safety of the plant. The first and second EAL conditions are focused on a loss of mass from the RCS due to "unidentified leakage", "pressure boundary leakage" or "identified leakage" (as these leakage Page 225 of 278 -::::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases types are defined in the plant Technical Specifications). The third condition addresses an RCS mass loss caused by an UNISOLABLE leak through an interfacing system. These conditions thus apply to leakage into the containment, or a location outside of containment. The leak rate values for each condition were selected because they are usually observable with normal Control Room indications. Lesser values typically require time-consuming calculations to determine (e.g., a mass balance calculation). The first condition uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage. The release of mass from the RCS due to the as-designed/expected operation of a relief valve does not warrant an emergency classification. A stuck-open Safety Relief Valve (SRV) or SRV leakage is not considered either identified or unidentified leakage by Technical Specifications and, therefore, is not applicable to this EAL. The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible. Escalation of the emergency classification, level would be via I Cs of Recognition Category A or F. ' Reference(s):

1. USAR Section 5.2.5 Reactor Coolant Pressure Boundary and ECCS Leakage Detection System 2. Technical Specification Definitions Section 1.1 3. Technical Specification 3.4.5 2. NEI 99-01 SU4 \/ Page 226 of 278 \

-===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: Subcategory: S -System Malfunction 6 -RPS Failure Initiating Condition: Automatic or manual scram fails to shut down the reactor EAL: SU6.1 Unusual Event An automatic scram did not shut down the reactor as indicated by reactor power > 5% after any RPS setpoint is exceeded AND A subsequent automatic scram or manual scram action taken at the reactor control console (Mode Switch, Manual PBs, ARI) is successful in shutting down the reactor as indicated by reactor powers; 5% (APRM downscale) Note 8: A manual scram action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. Mode Applicability: 1 -Power Operation, 2-Startup Definition(s): IMMINENT -The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. Basis: This IC addresses a failure of the RPS to_ initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic scram is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant. The first condition of this EAL identifies the need to cease critical reactor operations by actuation of the automatic Reactor Protection System (RPS) scram function. A reactor scram is automatically initiated by the RPS when certain continuously monitored parameters exceed predetermined setpoirits (ref. 1 ). A successful scram. has therefore occurred when there is sufficient rod insertion from the trip of RPS to bring the reactor power to or below the APRM downscale setpoint of 5% (ref. 4). For the purposes of emergency classification, successful manual scram actions are those which can be quickly performed from the reactor control console (i.e., mode switch, manual scram pushbuttons, or ARI initiation). Reactor shutdown achieved by use of alternate control rod insertion methods (i.e., EOP-1A Enclosure

26) does not constitute a successful manual scram (ref. 2). Page 227 of 278

~=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Following any automatic RPS scram signal, operating procedures (e.g., EOP-1A) prescribe insertion of redundant manual scram signals to back up the automatic RPS scram ful")ction and ensure reactor shutdown is achieved. Even if the first subsequent manual scram signal inserts all control rods to the full-in position immediately after the initial failure of the automatic scram, the lowest level of classification that must be declared is an Unusual Event (ref. 3). Taking the Mode Switch to Shutdown is a manual-scram action. When the Mode Switch is taken out of the Run position, .however, the nuclear instrumentation scram setpoint is lowered. If reactor power remains above the lowered setpoint, an automatic scram is initiated. For-the purposes of this EAL, a successful automatic initiation of ARI that reduces reactor power to< 5% is not considered a successful automatic scram. If automatic initiation of ARI has occurred and caused reactor shutdown, the automatic RPS scram must have failed. ARI is a backup means of inserting control rods in the unlikely event that an automatic RPS scram signal exists but the reactor continues to generate significant power. However, a successful automatic or manual initiation of ARI is an acceptable means of establishing reactor shutdown conditions relative to the EAL threshold in the absence of any required subsequent manual scram actions. In the event that the operator identifies a reactor scram is IMMINENT and initiates a successful manual reactor scram before the automatic scram setpoint is reached, no declaration is required. The successful manual scram of the r~actor before it reaches its automatic scram setpoint or reactor scram signals caused by instrumentation channel failures do not lead to a potential fission product barrier loss. If manual reactor scram actions fail to reduce reactor power below 5%, the event escalates to the Alert under EAL SA6.1. , If by procedure, operator actions include the initiation of an immediate manual scram following receipt of an automatic scram signal and there are no clear indications that the automatic scram failed (such as a time delay following indications that a scram setpoint was exceeded), it may be difficult to determine if the reactor was shut down because of automatic scram or manual actions. If a subsequent review of the scram actuation indications reveals that the automatic scram did not cause the reactor to be shut down, then consideration should be given to evaluating the fuel for potential damage, and the reporting requirements of 50. 72 should be considered for the transient event.

Following the failure on an automatic reactor scram, operators will promptly initiate manual _ actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor
scram). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.
If an initial manual reactor scram is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram using a different switch). Depending upon several factors, the initial or subsequent effort to manually scram the reactor, or a concurrent plant condition, may lead to 1 the generation of an automatic reactor scram signal. If a subsequent manual or automatic Page 228 of 278

-::::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases scram is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems. A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram). This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-*panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles". Taking the Reactor Mode Switch to Shutdown is considered to be a manual scram action. The plant response to the failure of an automatic or manual reactor scram will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will -escalate to an Alert via EAL SA6.1. Depending upon the plant response, escalation is also possible via IC FA1. Absent the plant conditions needed to meet either IC SA6 or FA1, an Unusual Event declaration is appropriate for this event. Should a reactor scram signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied.

If the signal generated as a result of plant work causes a plant transient that results in a condition that should have included an automatic reactor scram and the RPS fails to automatically shutdown the reactor, then this IC and associated EALs are applicable, and should be evaluated.
If the signal generated as a result of plant work does not cause a plant transient and the scram failure is determined through other means (e.g., assessment of test results), then this IC and associated EALs are not applicable and no classification is warranted.

Reference(s):

1. Technical Specification Table 3.3.1.1-1 Reactor Protection System Instrumentation

2. EOP-1A RPV Control, ATWS 3. EOP-1 RPV Control 4. NEI 99-01 SU5 Page 229 of 278

~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: Subcategory: S -System Malfunction 6 -RPS Failure Initiating Condition: Automatic or manual scram fails to shut down the reactor EAL: SU6.2 Unusual Event A manual scram did not shut down the reactor as indicated by reactor power > 5% after any manual scram action was initrated AND A subsequent automatic scram or manual scram action taken at the reactor control console (Mode Switch, Manual PBs, ARI) is successful in shutting down the reactor as indicated by reactor power :s; 5% (APRM downscale) (Note 8) Note 8: A manual scram action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. Mode Applicability: 1 -Power Operation, 2 -Startup Definition(s): SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 1 OCFR50.2): ' Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic scram is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant. This EAL addresses a failure of a manually initiated scram in the absence of having exceeded an automatic RPS trip setpoint and a subsequent automatic or manual scram is successful in shutting down the reactor (reactor power :s; 5%) (ref. 1 ).

Page 230 of 278

*-:::==-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases A successful scram has therefore occurred when there is sufficient rod insertion from the trip of RPS to bring the reactor power to or below the APRM downscale setpoint of 5%. , For the purposes of emergency classification, successful manual scram actions are those which can be quickly performed from the reactor' control console (i.e., Mode Switch, manual scram pushbuttons, or ARI initiation).

Reactor shutdown achieved by use of alternate control rod insertion methods (i.e., EOP-1A Enclosure

26) does not constitute a successful manual scram (ref. 2). Taki~g the Mode Switch to Shutdown is a manual scram action. When the Mode Switch is taken out of the Run position, however, the nuclear instrumentation scram setpoint is lowered. If reactor power remains above the lowered setpoint, an automatic scram is initiated.

Successful automatic or manual initiation of ARI is an acceptable means of establishing reactor shutdown conditions relative tb the EAL threshold in the absence of any required subsequent manual scram actions. If both subsequent automatic and subsequent manual reactor scram actions in the Control Room fail to reduce reactor power below the power associated with the SAFETY SYSTEM design (:s; 5%) following a failure of an initial manual scram, the event escalates to an Alert under EAL SA6.1. Following the failure on an automatic reactor scram, operators will promptly initiate manual actions at the reactor control cqnsoles to shutdown the reactor (e.g., initiate a manual reactor scram). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems. If an initial manual reactor scram is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram) using a different switch. Depending upon several factors, the initial or subsequent effort to manually scram the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor scram signal. If a subsequent manual or automatic scram is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems. A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram). This action does-not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles". Taking the Reactor Mode Switch to Shutdown is considered to be a manual scram action. The plant response to the failure of an automatic or manual reactor scram will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant Page 231 of 278 ~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases conditions, etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC SA6. Depending upon the plant response, escalatio'n is also possible via IC FA1. Absent the plant conditions needed to meet either IC SA6 or FA1, an Unusual Event declaration is appropriate for this event. Should a reactor scram signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied.

If the signal generated as a result of plant work causes a plant transient that that results in a condition that should have included an automatic reactor scram and the RPS fails to automatically shutdown the reactor, then this IC and associated EALs are applicable, and should be evaluated.
If the signal generated as a result of plant work does not cause a plant transient and the scram failure is determined through other means (e.g., assessment of test results), then this IC and associated EALs are not applicable and no classification is warranted.

Refe re nee( s): 1. Technical Specification Table 3.3.1.1-1 Reactor Protection System Instrumentation

2. EOP-1A RPV Control, ATWS 3. EOP-1 RPV Control 4. NEI 99-01 SUS Page 232 of 278

-:::::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: Subcategory: S -System Malfunction 2 -RPS Failure Initiating Condition: Automatic or manual scram fails to shut down the reactor and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor EAL: SA6.1 Alert An automatic or manual scram fails to shut down the reactor as indicated by reactor power >5% AND Manual scram actions taken at the reactor control console (Mode Switch, Manual PBs, ARI) are not successful in shutting down the reactor as indicated by reactor power> 5% (Note 8) ' Note 8: A manual scram action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. Mode Applicability: 1 -Power Operation, 2 -Startup Definition(s): SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 1 OCFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and subsequent operator manual actions taken at the reactor control consoles to shutdown the reactor are also unsuccessful. This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if the react6r is subsequently shutdown by an action \ Page 233 of 278 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases taken away from the reactor control consoles since this event entails a significant failure of the RPS. This EAL addresses any automatic or manual reactor scram signal that fails to shut down the reactor followed by a subsequent manual scram that fails to shut down the reactor to an extent the reactor is producing energy in excess of the heat load. for which the SAFETY SYSTEMS were designed (> 5%). For the purposes of emergency classification, successful manual scram actions are those which can be quickly performed from the reactor control console (i.e., Mode Switch, manual . scram pushbuttons, or ARI initiation). Reactor shutdown achieved by use of alternate control rod insertion methods (i.e., EOP-1A Enclosure

26) does not constitute a successful manual scram (ref. 2). For the purposes of this EAL, a successful automatic initiation of ARI that reduces reactor power to or below 5% is not considered a successful automatic scram. If automatic actuation of ARI has occurred and caused reactor shutdown, the automatic RPS scram must have failed. ARI is a backup* means of inserting control rods in the unlikely event that an automatic RPS scram signal exists but the reactor continues to generate significant power. However, a successful automatic initiation of ARI is an acceptable means of establishing reactor shutdown conditions relative to the EAL threshold in the absence of any required subsequent manual scram actions. A manual action at the reactor control console is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram). This action does not include manually driving in control rods or implementation of boron injection strategies.

If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control consoles (e.g., locally opening breakers). Actions taken at back panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor . control consoles". Taking the Reactor Mode Switch to Shutdown is considered. to be a manual scram action. The plant response to the failure of an automatic or manual reactor scram) will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If the failure to shutdown the reactor is prolonged enough to cause a challenge to RPV water leveL or RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC SS6. Depending upon plant responses and symptoms, escalation is also possible via IC FS1. Absent the plant conditions needed to meet either IC SS6 or FS1, an Alert declaration is appropriate for this event. It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F ICs; however, this IC and EAL are incl~ded to ensure a timely emergency declaration. Page 234 of 278 f --Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Refe re nee( s): 1. Technical Specification Table 3.3.1.1-1 Reactor Protection System Instrumentation

2. EOP-1A RPV Control, ATWS 3. EOP-1 RPV Control 4. NEI 99-01 SA5 Page 235 of 278 River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases

Category:

S -System Malfunction Subcategory: 2 -RPS Failure Initiating Condition: Inability to shut down the reactor causing a challenge to RPV water level or RCS heat removal EAL: 556.1 Site Area Emergency An automatic or manual scram fails to shut down the reactor as indicated by reactor power >5% ' AND All actions to shut down the reactor are not successful as indicated by reactor power> 5% AND EITHER: RPV water level cannot be restored and maintained> -187 in. OR Heat Capacity Temperature Limit (HCTL) exceeded (EOP Figure 2) Mode Applicability: 1 -Power Operation, 2 -Startup Definition(s): SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, all subsequent operator actions to manually shutdown the reactor are unsuccessful, and continued power generation is challenging the capability to adequately remove heat from the core and/or the RCS.' This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration

of a Site Area Emergency.

! Page 236 of 278 'i ~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases This EAL addresses the following:

Any automatic reactor scram signal followed by a manual scram that fails to shut down the reactor to an extent the reactor is producing energy in excess of the heat load for which the SAFETY SYSTEMS were designed (EAL SA6.1 ), and
Indications that ei.ther core cooling is extremely challenged or heat removal is extremely challenged.
Reactor shutdown achieved by use of control rod insertion methods in EOP-1A Enclosure 26 . are also credited as a successful shutdown provided reactor power can be reduced to or below the APRM downscale trip setpoint before indications of an extreme challenge to either core cooling or heat removal exist. (ref. 1) The combination of failure of both front line and backup protection systems to function in response to a plant transient, along with the continued production of heat, poses a direct threat to the Fuel Clad and RCS barriers.

Indication that core cooling is extremely challenged is manifested by inability to restore and maintain RPV water level above the Minimum Steam Cooling RPV Water Level (MSCRWL) (ref. 1 ). The MSCRWL is the lowest RPV level at which the covered portion of the reactor core will generate sufficient steam to prevent any clad temperature in the uncovered part of the core from exceeding 1500°F. This water level is utilized in the EOPs to preclude fuel damage when RPV level is below the top of active fuel. RPV level below the MSCRWL for an extended period of time without satisfactory core spray cooling could be a precursor of a core melt sequence (ref. 2). The Heat Capacity Temperature Limit (HCTL, EOP Figure 2) is the highest suppression pool water temperature from which Emergency RPV Depressurization will not raise suppression pool temperature above the maximum design suppressfon pool temperature. The HCTL i~-a function of RPV pressure and suppression pool water level. It is utilized to preclude failure of the containment and equipment in the containment necessary for the safe shutdown of the plant. This threshold is met when the final step of section SPT in EOP-2, Primary Containment Control, is reached (ref. 3). This condition addresses loss of functions required for hot shutdown with the reactor at pressure and temperature. In some instances, the emergency classification resulting from this EAL may be higher than I that resulting from an assessment of the plant responses and symptoms against the Recognition Category F ICs/EALs. This is appropriate in that the Recognition Category F ICs/EALs do not address the additional threat posed by a failure to shut down the reactor. The ** inclusion of this EAL ensures the timely declaration of a Site Area Emergency in response to prolonged failure to shutdown the reactor. Escalation of the emergency classification

level would be via IC AG1 or FG1. Page 237 of 278

~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Reference(s):

1. EOP-1A, RPV Control, ATWS 2. EOP-4, RPV Flooding 3. EOP-2, Primary Containment Control 4. NEI 99-01 SS5 Page 238 of 278 Q -~=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category:

S -System Malfunction Subcategory: 7 -Loss of Communications Initiating Condition: Loss of all onsite or offsite communications capabilities EAL: SU7.1 Unusual Event Loss of all Table s:..4 onsite communication methods OR Loss of all Table S-4 State and local agency communication methods OR Loss of all Table S-4 NRC communication methods Table S-4 Communication Methods System Onsite Plant radio system X Plant Paging System X Sound powered phones X In-plant telephones X Emergency Notification System (ENS) Commercial Telephone System Satellite Phones State of Louisiana Radio State and Local Hotline radio INFORM Notification System Mode Applicability: 1 -Power Operation, 2 -Startup, 3 -Hot Shutdown Page 239 of 278 State/ Local X X X X X NRC X X X -===-Entergy Definition(s): None Basis: River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases 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 State and local agencies and the NRG. 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 site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.). The first EAL condition addresses a total loss of the communications methods used in support of routine plant operations. The second EAL condition addresses a total loss of the communications methods used to notify all State and local agencies of an emergency declaration. The State and local agencies referred to here are the Louisiana Department of Environmental Quality, Governor's Office of Homeland Security and Emergency Preparedness,, Five Local Parishes Office of Homeland Security and Emergency Preparedness and 24 hour notification points, Mississippi Emergency Management Agency and the Mississippi Highway Patrol. The third EAL condition addresses a total loss of the communications methods used to notify the NRG of an emergency declaration. This EAL is the hot condition equivalent of the cold condition EAL CU5.1. Reference(s):

1. RBS Emergency Plan Section 13.3.6.1.5.4 Communications

2. RBS Emergency Plan Section 13.3.6.2.1 Site Communications

3. NEI 99-01 SU6 Page 240 of 278

/ ~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases Category: S -System Malfunction Subcategory: 8 -Hazardous Event Affecting Safety Systems Initiating Condition: Hazardous event affecting SAFETY SYSTEMS needed for the current operating mode EAL: SAB.1 Alert The occurrence of any Table S-5 hazardous event AND Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode AND EITHER:

Event damage has caused indications of degraded performance to the second train of the SAFETY SYSTEM needed for the current operating mode
Event damage has resulted in VISIBLE DAMAGE to the second train of the SAFETY SYSTEM needed for the current operating mode (Notes 9, 10) Note 9: If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then emergency classification is not warranted.

Note 10: If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted. Mode Applicability: Table 5-5 Hazardous Events

Seismic event (earthquake)
Internal or external FLOODING event
High winds or tornado strike
FIRE
EXPLOSION
Other events with similar hazard characteristics as determined by the Shift Manager 1 -Power Operation, 2 -Startup, 3 -Hot Shutdown Page 241 of 278

~--=----r------ 1 ~Entergy River Bend Station EAL Basis Document Revision XXX

Attachment 1 -Emergency Action Level Technical Bases Definition(s):

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 require a event inspection to determine if the attributes of an explosion are present. FIRE -Combustion characterized by heat and light. Sources of smoke such as slipping drive .belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed. FLOODING -A condition where water is entering'a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area. SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 1 OCFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. VISIBLE DAMAGE -Damage to a SAFETY SYSTEM train that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected SAFETY SYSTEM train. 1* 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; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events. \ Page 242 of 278 ' ' . ~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 1 -Emergency Action Level Technical Bases 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 a determination of VISIBLE DAMAGE based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. Escalation of the emergency classification level would be via IC FS1 or AS1. This EAL is the hot condition equivalent of cold condition EAL CA6.1. Reference(s):

1. EP FAQ 2016-00 2. NEI 99-01 SA9 Page 243 of 278 River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation

& Shutdown Areas Tables A-3 & H-2 Bases Background NEI 99-01 Revision 6 I Cs AA3 and HA5 prescribe declaration of an Alert based on IMPEDED access to rooms or areas (due to either area radiation levels or hazardous gas concentrations) where equipment necessary for normal plant operations, cooldown or shutdown is located. These areas are intended to be plant operating mode dependent. Specifically the Developers Notes for AA3 and HA5 states: The "site-specific list of plant rooms or areas with entry-related mode applicability identified" should specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, coo/down 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 which entry would be required for each room or area. 'i 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). Further, as specified in IC HA5: The list need not include the Control Room if adequate engineered safety/design features are in place to preclude a Control Room evacuation due to the release of a hazardous gas. Such features may include, but are not limited to, capability to draw air from multiple air intakes at different and separate locations, inner and outer atmospheric boundaries, or the capability to acquire and maintain positive pressure within the Control Room envelope. / Page 244 of 278 Q -:;:::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases RBS Table A-3 and H-2 Bases A review of station operating procedures identified the following mode dependent in-plant actions and associated areas that are required for normal plant operation, cooldown or shutdown: GOP / SOP ACTIONS LOCATION MODE NOTES GOP-0005

Power Operations Perform power maneuvering as directed by the OSM/CRS between 60 MCR 1 and 100% power using the guidance provided in the appropriate Reactivity Maneuvering Plan provided by Reactor Engineering.

If possible, notify System Operator prior to changing generator load. MCR 1

Adhere to MVAR vs. MW limits .
WHEN adjusting VARs on the Main Generator, THEN use VAR-1SPGN05 (H13-P680) only. Prior to entry into the Monitored Region of the Power/Flow map verify at MCR 1 least one PBDS Card is operable and begin STP-000-0001 monitoring of PBDS. (TS 3.3.1.3) Adjust pressure setpoint to minimize recirc pump "thrust reversals" as MCR 1 follows:
IF lowering power AND it is desired that pressure set be raised to minimize recirc pump "thrust reversals", THEN prior to lowering core flow to less than 70% rated core flow, raise reactor pressure.
IF raising power AND pressure set was raised*to minimize recirc pump "thrust reversals", THEN when core flow is greater than 70% rated core flow, return the reactor pressure to its nominal value. Monitor Reactor Feed Pump vibration and flow. IF necessary to MCR 1 minimize vibration, THEN operate the reactor feed pump Minimum Flow valves per SOP-0009, Long Cycle Clean Up valve, or adjust reactor power. IF a reactor feed pump is anticipated to be shut down and Hydrogen MCR/TB 67' 1 Not required injection will be left in service, THEN install the vent jumper for that for plant pump per SOP-0009, Reactor Feedwater System. shutdown or cooldown Remove from service and/or restart Reactor Feed Pumps as necessary MCR /TB 67' 1 Not required to maintain Reactor Water Level and reactor feed pump flow for plant requirements to minimize vibration.

shutdown or cooldown IF controlling Reactor Power with Reactor Recirculation Flow, THEN MCR 1 refer to SOP-0003. IF power is lowered below 75% AND a Reactor Feed Pump has been MCR 1 secured, THEN, BEFORE power ascension beyond 75% RTP and Page 245 of 278 ,J -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP/ SOP ACTIONS LOCATION MODE NOTES AFTER start of the 3rd feedwater pump verify water chemistry acceptable. IF power is lowered below 75%, THEN when thermal power is above MCR 1 75% RTP, verify both LEFM Check Plus are Operable. IF the LEFMs are functional per Technical Requirement 3.3.13 AND an MCR 1 LEFM alert is indicated on the-ONE heat balance computer screen, THEN reduce reactor power to 3081 MWth or 99.6% rated thermal power in one hour to ensure thermal power limits are not exceeded. Observe the operation MSS-HVYCV4 and DTM-AOVSPDV4 for MCR 1 OSP-0102, Turbine Valve Testing.

As power is raised, check MSS-HVYCV4 open and DTM-AOVSPDV4 closed.
As power is lowered, check MSS-HVYCV4 closed and DTM-AOVSPDV4 open. When power is raised above 90%, Pressure Set may need to be MCR 1 adjusted as necessary to ensure that the 1st admission main turbine control valves, MSS-HVYCV1, 2, and 3 are full open. Monitor turbine vibration bearing temperature and differential expansion MCR 1 per the following:

Turbine Temg & Exgansion RCDR (TMI-NXR102)

Differential Expansion Rotor Long (point 11) between 0.31 inches and 0.69 inches. (Refer To ARP-870-54, GOS, HOS)
Rotor Expansion Rotor Long (point 12) between 0.455 inches and 1.545 inches. Turbine Vibration RCDR (TMI-NXR103)
Vibration (points 1 through 10) between O mils and 6 mils . (Refer To ARP-870-54, DOS) Tamaris Comguter (Disglay 69, 70)
Bearing oil temperatures

(<setpoint, 180°F) .

Bearing metal temperatures

(<setpoint, 218.?°F) . IF unusual indications are observed; THEN initiate hold in power change until those indications return to normal. WHEN maneuvering power, THEN adhere to the POWER/FLOW maps MCR 1 (avoid the restricted region) in AOP-0024, Thermal Hydraulic Stability Controls and Turbine-Generator loading rate per SOP-0080, Turbine Generator Operation. WHEN in two Recirculation Pump Operation at greater than or equal to MCR 1 70% rated core flow, THEN maintain recirculation flow mismatch less than 5%. Observe the following limitations and precautions: MCR 1

Do not exceed the Turbine Generator normal operating limits. Page 246 of 278 Q -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation

& Shutdown Areas Tables A-3 & H-2 Bases GOP / SOP ACTIONS LOCATION MODE NOTES

Adjust the pressure setpoint to an indicated reactor pressure of between 1035 and 1055 psig for 100% steady state conditions.
IF reactor power exceeds 3091 MWth, THEN take actions using recirculation flow and/or control rod insertion to lower power below 3091 MWth.
IF the core thermal power average for a 2 hour period exceeds the Licensed Power Limit, THEN take timely action to ensure that thermal power is less than or equal to the Licensed Power Limit.
IF reactor thermal power indication becomes unavailable for less than 15 minutes AND steady state operation is expected, THEN note current APRM readings AND verify thermal power does not exceed the noted value.
IF reactor thermal power indication will be unavailable for more than 15 minutes, THEN perform the following:

0 Lower reactor power as indicated on the APRMs such that indicated thermal power does not exceed 100%. (The top of the normal noise band on the chart recorders should not be above 100%). ' 0 Reactor Engineering should be contacted for assistance in determining a manual heat balance per REP-0030, Reactor Heat Balance. 0 WHEN performing a manual heat balance AND it is determined that the LEFM signal is not operable, THEN lower reactor power so that the APRMs read less than 98.3% at the top of the normal noise band.

Observe the following restrictions when operating near or above rated core flow as Bi-Stable flow conditions are possible:

0 IF step changes of 60 MWth (2%) or greater are seen in instantaneous CTP, THEN reduce Reactor power using Recirc flow until the step changes in instantaneous power are no longer observed. 0 IF step changes of up to 1.69 MLB/hr (2%) are seen in total core flow, THEN reduce Reactor power using . Recirc flow until the step changes in instantaneous power are no longer observed. 0 Notify Reactor Engineering of any power/flow reductions required.

IF any thermal limit exceeds 0.980, THEN notify Reactor Engineering to increase the frequency of monitoring (at least hourly) until a steady state condition is reached or thermal limits indicate less than 0.980.
IF any thermal limit exceeds 0.990, THEN notify Reactor Engineering to perform one of the following:

0 Provide instructions for reducing the thermal limit to less than 0.990. 0 Provide a justification for operating with thermal limits Page 24 7 of 278 River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP / SOP ACTIONS LOCATION MODE NOTES greater than or equal to 0.990. IF the power change exceeded 15%, THEN perform the following: MCR /TB 95' 1 Not required

Notify Chemistry of the power change to obtain a new for plant Condensate System Oxygen injection flow rate.
shutdown or cooldown
Per Chemistry recommendations, adjust the Oxygen flow rate per SOP-0123, Hydrogen Water Chemistry H2 and 02 System.
IF power ramp rates exceed 15%/hr, THEN notify Chemistry per Technical Requirement.

IF power was lowered below 80%, THEN notify chemistry management MCR 1 when reactor power has been returned to 100%. As power is lowered, at approximately 50% power, transfer Steam MCR 1 Seal Evaporator from Extraction Steam to Main Steam per SOP-0015, Gland Seal Steam System and Exhaust System, if it has not occurred automatically. Transferring Steam Seal Evaporator from Extraction Steam to '~[!:'. t,, , ,ri:,, *,, ;~;' 'l~ Main Steam (SOP-0015). ' .,.) ' .* ., ' ' ; As power is lowered, at approximately 50% power, if the Steam Seal MCR 1 Evaporator has not already transferred automatically from Extraction Steam to Main Steam, then throttle closed ESS-MOV112, STEAM SEAL EVAPORATOR using the control switch and the STOP pushbutton. IF the pressure controller is operating in automatic AND MCR 1 TME-MOVESFV2 is closed, THEN verify the following:

TME-PIEPR-35, SSE TUBE SIDE PRESSURE indicates less than or equal to 75 psig.
TME-PIEPR-36, SSE SHELL SIDE PRESSURE is stable and indicates less than or equal to 45 psig. WHEN ESS-MOV112, STEAM SEAL EVAPORATOR is full closed, MCR 1 THEN verify DTM-AOV118, EXTR STM TO SSE & RW RBLR opens. As power is raised, at approximately 65-75% power, after checking MCR 1 Annunciator P870-52-E03, 3rd PT EXTR ST AND MAIN STEAM DIFF PRESS LOW is clear, transfer Steam Seal Evaporator from Main Steam to Extraction Steam per*soP-0015, Gland Seal Steam System and Exhaust System. ' ., ' GOP-0002 Power Decrease/Plant Shutdown Notify System Operator prior to decreasing generator load. MCR 1 IF the Reference Leg Backfill System is not in service per SOP-0001, MCR 1 Nuclear Boiler Instrumentation (SYS #051 ), THEN have l&C stage equipment, acquire necessary technicians and obtain PMs to backfill Page 248 of 278

-::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP / SOP ACTIONS LOCATION MODE NOTES reactor water level reference legs. (Approximately 12 hours may be needed to prepare for backfilling.) Actual backfilling performance may commence when Operations ShiftlManager authorizes. (It is desired to have backfilling completed prior to reactor pressure reaching 450 psig to counter level indication notching possibilities.) Monitor turbine vibration bearing temperature and differential expansion MCR 1 per the following: Turbine Temg & Exgansion RCDR (TMI-NXR102)

Differential Expansion Rotor Long (point 11) between 0.31 inches and 0.69 inches. (Refer To ARP-870-54, GOS, HOS)
Rotor Expansion Rotor Long (point 12) between 0.455 inches and 1.545 inches. Turbine Vibration RCDR {TMI-NXR103)
Vibration (points 1 through 10) between O mils and 6 mils . (Refer To ARP-870-54, 008) Tamaris Comguter {Disglay 69, 70)
Bearing-oil temperatures

(<setpoint, 180°F) .

Bearing metal temperatures

(<setpoint, 218.?°F) . IF unusual indications are observed, THEN initiate hold in power change until those indications return to normal. Lower reactor power per the Shutdown/ Emergency Power Reduction MCR 1 reactivity control plan. Contact the on-duty Reactor Engineer. Adjust pressure setpoint to minimize recirc pump "thrust reversals" as MCR 1 follows:

IF lowering power AND it is desired that pressure set be raised to minimize recirc pump "thrust reversals", THEN prior to lowering core flow to less than 70% rated core flow, raise reactor pressure.

IF raising power AND pressure set was raised to minimize recirc pump "thrust reversals", THEN when core flow is greater than 70% rated core flow, return the reactor pressure to its nominal value. At approximately 90% to 80% power observe the following: MCR 1

MSS-HVYCV4 closes
DTM-AOVSPDV4 opens IF MSRs are to be manually shutdown, THEN at approximately 90% MCR/TB 1 Not required power, start removing the MS Rs from service per SOP-0010, MSR & 123' for plant FW Heaters Extraction Steam and Drains. Remove the MS Rs at a rate shutdown or so as to be completely off line by 760 MWe. Limit rate of change of LP cooldown Turbine inlet steam temperature to 125°F per hour. Monitor Points 6, 7, 8, 9 on TMI-NXR102.

Maximum allowable temperature difference between LP Turbine inlets is 50°F. MSRs should be gradually valved Page 249 of 278 a ~=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 ...:.. Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP/ SOP ACTIONS LOCATION MODE NOTES out in parallel at essentially the same temperature. IF MSRs are to remain in service with power maintained between 15% and 70%, THEN operate MSRs in accordance with SOP-0010. For power reductions of greater than 15%, notify Chemistry to MCR/TB 95' 1 Not required determine whether the Condensate System oxygen injection is to be for plant secured or flow reduced AND implement the recommendations per shutdown or SOP-0123, Hydrogen Water Chemistry H2 and 02 System. cooldown IF,power ramp rates exceed 15%/hr, THEN notify Chemistry per Technical Requirement 3.11.2.1. IF a reactor feed pump is anticipated to be shut down AND Hydrogen TB 67' 1 Not required Injection will be left in service AND a plant shutdown is NOT in for plant progress, THEN install the ventjumper(s) for the pump(s) per SOP-shutdown or 0009, Reactor Feedwater System. cooldown At approximately 70% power, (or with Engineering recommendations) MCR 1 stop one reactor feedwater pump (leave two running) per SOP-0009, Reactor Feedwater System. Reactor Feed Pump Shutdown (SOP-0009) .** ; *, . ,';.. ' . ' \_ ; *;~. IF securing a Reactor Feed Pump for downpower, THEN monitor the MCR 1 following parameters:

Reactor power should be limited to 85% with only two Reactor Feed Pumps in service.
Normal Feedwater Pump Motor current should be greater than 200 amps and limited to 311 amps. Refer to Precautions and Limitations 2.9 and 2.15.
FWREG position should be limited to less than or equal to 92% open to allow an adequate margin for valve modulation while maintaining reactor level.
Feed pump suction pressure should be maintained above low pressure alarm point of 280 psig. IF NOT already performed to reduce Reactor Feed Pump vibration MCR 1 levels, THEN perform the following for the Reactor Feed Pump being shutdown:
At H13-P680, place CNM-H/A68A(B)(C), RX FWP 1A(B)(C) MIN FLOW FLOW CONTROLLER to MANUAL for the Reactor Feed Pump to be secured.
Open slowly FWR-FV2A(B)(C), RX FWP 1A(B)(C) MIN FLOW VALVE using CNM-H/A68A(B)(C), RX FWP 1A(B)(C) MIN FLOW FLOW CONTROLLER while monitoring Reactor Water Level. IF desired to raise Reactor Water Level, THEN at H13-P680 adjust MCR 1 C33-R600, FW REG VALVES MASTER FLOW CONTROLLER tape set to desired Reactor Water Level within normal level control band. Page 250 of 278

-===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP I SOP ACTIONS LOCATION MODE NOTES IF the HWC System is in service AND the reactor feed pump is not MCR/TB 95' 1 Not required being immediately shut down, THEN at P73-P500, place P73-AOV-for plant F111A(B)(C), HYDROGEN ISOLATION TO FEEDWATER PUMP shutdown or A(B)(C) in CLOSE. cooldown IF the capability of meeting feed flow requirements with the remaining MCR 1 Feedwater Pumps is uncertain, THEN make a determination as follows:

Close FWS-MOV26A(B)(C), RX FWP 1A(B)(C) DISCH VLV for the pump being shutdown.
Verify the minimum flow valve for the pump being secured is open.
Monitor Feed Flow/Steam Flow mismatch and RPV Level to ensure remaining pump(s) can maintain level.
IF the remaining pump(s) cannot maintain RPV Level, THEN reopen the discharge valve FWS-MOV26A(B)(C), RX FWP 1A(B)(C) DISCH VLV and discontinue this procedure.

IF the last Feedwater Pump is being removed from service, THEN open MCR 1 FWS-MOV109, FEED PUMP BYPASS. Stop FWS-P1A(B)(C), RX FWP P1A(B)(C). MCR 1 ' Verify CNM-H/A68A(B)(C), RX FWP 1A(B)(C) MIN FLOW FLOW MCR 1 CONTROLLER is in AUTO for the Reactor Feed Pump that was secured. IF Reactor Water Level was intentionally raised in Step 6.1.3, THEN MCR 1 adjust Reactor Water Level to desired level within normal level control band using C33-R600, FW REG VALVES MASTER FLOW CONTROLLER tape set. IF FWS-P1A(B)(C) is to remain in hot standby, THEN maintain seal TB 67' 1 Not required temperatures as follows: for plant Maintain seal water temperature dT less than or equal to 50F AND seal shutdown or water outlet temperature less than or equal to 300F as follows: cooldown

FWS-P1A 0 Throttle CCS-V5003A, RFP FWL-P1A SEAL WATER HX-E4A CCS INLET VALVE, as required.

0 Throttle CCS-V5004A, RFP FWL-P1A SEAL WATER HX-E4B CCS INLET VALVE, as required.

FWS-P1B 0 Throttle CCS-V5003B, RFP FWL-P1B SEAL WATER HX-E4C CCS INLET VALVE, as required.

0 Throttle CCS-V5004B, RFP FWL-P1B SEAL WATER HX-E4D CCS INLET VALVE, as required.

FWS-P1C 0 Throttle CCS-V5003C, RFP FWL-P1C SEAL WATER HX-E4E CCS INLET VALVE, as required.

0 Throttle CCS-V5004C, RFP FWL-P1 C SEAL WATER Page 251 of 278 Q ~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP / SOP ACTIONS LOCATION MODE NOTES HX-E4F CCS INLET VALVE, as required. On H13-P870, verify FWL-P5A(B)(C), GEAR INCR AUX OIL PMP MGR 1 5A(B)(C) auto starts. Verify min flow valve closes 1 -3 minutes after pump shutdown. MGR 1 Verify FWS-MOV26A(B)(C), RX FWP P1A (B)(C) DISCH VLV is MGR 1 closed. On H13-P870, WHEN the 23 minute time delay allowing for pump coast MGR 1 down has passed, THEN verify the following:

FWL-P1A(B)(C), RX FWP A(B)(C) MN OIL PMP 1A(B)(C) auto stops.
FWL-P5A(B)(C), RX FWP A(B)(C) GEAR INC AUX OIL PMP 5A(B)(C) auto stops.
FWL-P2A(B)(C), RX FWP A(B)(C) AUX OIL PMP 2A(B)(C) auto starts on low oil pressure.
IF FWL-P2A(B)(C), RX FWP A(B)(C) AUX OIL PMP 2A(B)(C) does not maintain pressure greater than 4 psi, THEN FWL-P3A(B)(C), RX FWP A(B)(C) AUX OIL PMP 3A(B)(C) auto starts on low oil pressure.

On H13-P870, place FWL-P2A(B)(C), RX FWP A(B)(C) AUX OIL PMP MGR 1 2A(B)(C) control switch in STOP, and verify FWL-P3A(B)(C), RX FWP A(B)(C) AUX OIL PMP 3A(B)(C) auto starts on low oil pressure. On H13-P870, place FWL-P2A(B)(C), RX FWP A(B)(C) AUX OIL PMP MGR 1 2A(B)(C) control switch in AUTO. On H13-P870, place FWL-P3A(B)(C), RX FWP A(B)(C) AUX OIL PMP MGR 1 \ 3A(B)(C) control switch in STOP. On H13-P870, place FWL-P3A(B)(C), RX FWP A(B)(C) AUX OIL PMP MGR 1 3A(B)(C) control switch in AUTO. Locally verify breaker relay trip flags are reset for Reactor Feed Pump NSW 98' 1 Not required stopped in Step 6.1. 7. for plant shutdown or cooldown At approximately 70% power, (or with Engineering recommendations) MGR 1 stop one condensate pump (leave two running) per SOP-0007, Condensate System. Shutdown of CNM-P1A(B){C) CONDENSATE PUMPS (SOP-0007) ) .; Jr; :: .. lf,Jttl Request Aux Control Room remove unnecessary Condensate Filters ACR 1 Not required from service per SOP-0124, Condensate Filtration System. for plant shutdown or cooldown Request Aux Control Room remove unnecessary Condensate Dem ins ACR 1 Not required from service per SOP-0093, Condensate Demineralizer System. for plant shutdown or Page 252 of 278 r--,, *::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP / SOP ACTIONS LOCATION MODE NOTES cooldown IF shutting down CNM-P1 C, CNDS PUMP 1 C, THEN secure Oxygen MCR /TB 95' 1 Not required injection per SOP-0123, Hydrogen Water Chemistry H2 and 02 for plant System. shutdown or cooldown IF securing the Condensate System, THEN perform the following: TB 123' 1 Not required

Close all CNM-V3105 A, B, C, D, and E, CNM-FLT1 A, B, C, for plant D, and E BACKWASH AIR SUPPLY valves. shutdown or cooldown
IF desired to isolate and depressurize CNM-TK100, AIR RECEIVING TK, THEN perform the following:

0 CLOSE CNM-V3110, SVCE AIR !SOL VLV INLET SERV. AIR !SOL VLV. 0 Uncap and install hose on CNM-V3112, CNM-TK100 DRAIN ISOLATION VALVE. 0 Open CNM-V3112. Depress the CLOSE pushbutton for CNM-MOV3A(B)(C), CNDS PUMP MCR 1 1A(1B)(1C) DISCH. WHEN pump motor current lowers below 100 amps, THEN stop MCR 1 CNM-P1A(B)(C), CNDS PUMP 1A(1B)(1C). WHEN CNM-MOV3A(B)(C), CNDS PUMP 1A(1B)(1C) DISCH is full MCR 1 closed, THEN depress the STOP pushbutton. Verify associated CCS-MOV67A(B)(C), CNDS PMP 1A(1B)(1C) MOT MCR 1 CLR close for pump stopped. Verify associated CCS-MOV68A(B)(C), CNDS PMP 1A(1 B)(1C) BRG MCR 1 CLR close for pump stopped. Locally verify breaker relay trip flags are reset for Condensate Pump NSW 98' 1 Not required stopped in Step 6.1.6. for plant shutdown or cooldown WHEN the Steam Jet Air Ejectors (SJAEs) and Gland Seal and MCR i 1 Exhaust System are removed from service, THEN adjust CNM-H/A114 to 10% or to a setpoint determined by the CRS/OSM. As power is reduced, remove FW Reg Valves from service per SOP-MCR 1 0009, Reactor Feedwater System. Removing a FWREG Valve from Service (SOP-0009)

~: i~{ . l!:t'1 l1::'!I,11~11lt\t;~t1,**.*f Check feedwater flow is within the capability of the remaining FWREGs. MCR 1 I Station an operator locally at the FWREG Valve to be removed from TB 67' 1 Not required service. for plant shutdown or cooldown Page 253 of 278

/7"'\ -:::==-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP / SOP ACTIONS LOCATION MODE NOTES Establish communications between the local operator and the Main MCR / TB 67' 1 Not required Control Room (MCR). for plant shutdown or cooldown Place C33-R601A(R613)(R601B), FWREG VALVEA(B)(C) FLOW MCR 1 CONTROLLER in MANUAL. Throttle closed to 10% open the C33-LVF001A(B)(C), FWREG VALVE MCR 1 A(B)(C) to be removed from service while observing that reactor level is being maintained by the remaining FWREGs. IF level is not being maintained by the remaining FWREGs, THEN MCR 1 place the FWREG that was being removed from service back in service as follows:

Open C33-LVF001A(B)(C), FWREG VALVE A(B)(C) to the same position as the in service FWREGs.
Place C33-R601A(R613)(R601 B), FWREG VALVE A(B)(C) FLOW CONTROLLER in AUTO. WHEN the FWREG is at 10% open, THEN close the following isolation MCR 1 valve for the FWREG valve that is being removed from service.
For C33-LVF001A close FWS-MOV27A, FWREG VLV 1A INLT Valve.
For C33-LVF001 B close FWS-MOV27B, FWREG VLV 1 B INL T Valve.
For C33-LVF001C close FWS-MOV27C, FWREG VLV 1C INLT Valve. Fully close the C33-L VF001 A(B)(C), FWREG VALVE A(B)(C) that was MCR 1 removed from service. \ Record the temperature of the feedwater at the reactor feed pumps. TB 67' 1 Not required for plant shutdown or cooldown IF FWS-MOV27A, B, or C, FWREG VLV 1A(1B)(1C)

INLTwere closed MCR 1 with feedwater temperature at the reactor feed pumps greater than 200F, THEN refer to Section 5.7 for further stroking requirements. WHEN the FWREG Valve is at 0% open, THEN record demanded MCR/TB 67' 1 Not required position in the MCR, position indication in the MCR, and local position for plant indication. shutdown or cooldown -j . '.~. ,;:~ ~:1,.:1;; m '.,:*-1ft. ' . Within one hour after reactor power is less than or equal to the high MCR 1 STP-500-power setpoint, demonstrate RWL operability by performing STP-500-0704, Rod 0704, Rod Withdrawal Limiter Functional Test (SR 3.3.2.1.2), if not Withdrawal performed within the previous 92 days. Limiter Functional Test is Page 254 of 278 Q -::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP / SOP ACTIONS LOCATION MODE NOTES performed only in the MCR Prior to entering the Monitored and/or the Restricted Regions of the MCR 1 Power to Flow map, verify the following indications on the PBDS, Period Based Detection System cards in APRM 'A' and 'B' cabinets:

NORMAL/BYPASS Toggle switch in the NORMAL position .
INOP STATUS LED indication is GREEN. (Depress the INOP STATUS Reset Pushbutton to reset a Red LED inop indication.)
Verify at least one PBDS Card is OPERABLE .
Begin STP-000-0001 monitoring of PBDS . Prior to entry into the Restricted Region of the Power to Flow Map, MCR 1 perform the following:
Verify FCBB is less than or equal to 1.0.(SR 3.2.4.1)
Place the APRM -FCTR, Flow Control Trip Reference cards to the setup trip setpoints by depressing the Normal/Setup pushbutton and verifying the normal/setup LED indication is yellow. At Approximately 50% power, transfer Steam Seal Evaporator from MCR 1 Extraction Steam to Main Steam per SOP-0015, Gland Seal System And Exhaust System, if it has not occurred automatically.

Transferring Steam Seal Evaporator from Extraction Steam to Main Steam (SOP-0015) As power is lowered, at approximately 50% power, if the Steam Seal MCR 1 Evaporator has not already transferred automatically from Extraction Steam to Main Steam, then throttle closed ESS-MOV112, STEAM SEAL EVAPORATOR using the control switch and the STOP pushbutton. IF the pressure controller is operating in automatic AND MCR 1 TME-MOVESFV2 is closed, THEN verify the following:

TME-PIEPR-35, SSE TUBE SIDE PRESSURE indicates less than or equal to 75 psig.
TME-PIEPR-36, *ssE SHELL SIDE PRESSURE is stable and indicates less than or equal to 45 psig. WHEN ESS-MOV112, STEAM SEAL EVAPORATOR is full closed, MCR 1 THEN verify DTM-AOV118, EXTR STM TO SSE & RW RBLR opens. At approximately 50% power, shutdown all heater drain pumps per MCR 1 SOP-0010, MSR & FW Heaters Extraction Steam and Drains. Removing the Heater Drain Pumps From Service (SOP-0010)

Page 255 of 278 Q ~Entergy RiverBend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP / SOP ACTIONS LOCATION MODE NOTES IF it is desired to secure HDL-P1A(B), HTR DR PUMP 1A(B) for Heater MCR 1 String A, THEN at H13-P680, perform the following: .

Depress the Close Pushbutton for HDL-MOV55A(B), HTR DR PMP 1A(B) DISCH.
Stop HDL-P1A(B), HTR DR PUMP 1A(B) for Heater String A.
Verify HDL-MOV55A(B), HTR DR PMP 1A(B) DISCH is closed . IF it is desired to secure HDL-P1 C(D), HTR DR PUMP 1 C(D) for Heater MCR 1 String B, THEN at H13-P680, perform the following:
Depress the Close Pushbutton HDL-MOV55C(D), HTR DR PMP 1 C(D) DISCH.
Stop HDL-P1 C(D), HTR DR PUMP 1 C(D) for Heater String B .
Verify HDL-MOV55C(D), HTR DR PMP 1C(D) DISCH is closed. *:, 'J' :, At approximately 50% ,power, perform the following per SOP-0006, MCR 1 Circulating Water, Cooling Tower and Vacuum Priming:
Shut down at least 1 circulating water pump .
Adjust the number of operating cooling tower fans to maintain vacuum and circulating water temperature.

WHEN the recirculation flow control valves are at their minimum MCR 1 position, THEN continue reducing power by inserting control rods in their proper sequence. At about 40% power, transfer both reactor recirculation pumps to MCR 1 SLOW speed per SOP-0003, Reactor Recirculation System. ~.:' ' ~-Transferring from Fast Speed to Slow Speed (SOP-0003) I . : Simultaneously depress B33-C001A and B RECIRC PUMP A and B MCR 1 MOTOR BREAKER 5A and 58 XFER TO LFMG pushbuttons. ObseNe the following: MCR 1

Both B33-S001A LFMG MOT BRKR 1A and B33-S001 B LFMG MOTBRKR 1 B close.
Both B33-C001A RECIRC PUMP A MOTOR BREAKER 5A and B33-C001 B REC I RC PUMP B MOTOR BREAKER 58 open.
WHEN B33-C001A and B, RECIRC PUMP A and B coast down to approximately 360 -470 RPM, THEN B33-S001A and B LFMG A and B GEN BRKR 2A and 28 close and pump speeds stabilize near 450 RPM.
Both B33-K603 A and B, RECIRC LOOP A and B FLOW CONTROL MAN/AUTO stations transfer to MAN.
Open B33-HVY-F060A(B) to approximately 94% valve position using B33-K603A(B).

Reduce to one reactor feed pump per SOP-0009, Reactor MCR 1 Feedwater System. Page 256 of 278 *.* . -===-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP / SOP ACTIONS LOCATION MODE NOTES .. '~f tt J~f \ ;~"t;' .. * : i* . *'£1r* '{ . Reactor Feed Pump Shutdown (SOP-0009) ... .* IF securing a Reactor Feed Pump for downpower, THEN monitor the MCR 1 following parameters:

Reactor power should be limited to 85% with only two Reactor Feed Pumps in service.
Normal Feedwater Pump Motor current should be greater than 200 amps and limited to 311 amps. Refer To Precautions and Limitations 2.9 and 2.15.
FWREG position should be limited to less than or equal to 92% open to allow an adequate margin for valve modulation while maintaining reactor level.
Feed pump suction pressure should be maintained above low pressure alarm point of 280 psig. IF NOT already performed to reduce Reactor Feed Pump vibration MCR 1 levels, THEN perform the following for the Reactor Feed Pump being shutdown:
At H13-P680, place CNM-H/A68A(B)(C), RX FWP 1A(B)(C) MIN FLOW FLOW CONTROLLER to MANUAL for the Reactor Feed Pump to be secured.
Open slowly FWR-FV2A(B)(C), RX FWP 1A(B)(C) MIN FLOW VALVE using CNM-H/A68A(B)(C), RX FWP 1A(B)(C) MIN FLOW FLOW CONTROLLER while monitoring Reactor Water Level. IF desired to raise Reactor Water Level, THEN at H13-P680 adjust MCR 1 C33-R600, FW REG VALVES MASTER FLOW CONTROLLER tape set to desired Reactor Water Level within normal level control band. IF the HWC System is in service AND the reactor feed pump is not MCR/TB 95' 1 Not required being immediately shut down, THEN at P73-P500, place P73-AOV-for plant F111A(B)(C), HYDROGEN ISOLATION TO FEEDWATER PUMP shutdown or A(B)(C) in CLOSE. cooldown IF the capability of meeting feed flow requirements with the remaining MCR 1 Feedwater Pumps is uncertain, THEN make a determination as follows:
Close FWS-MOV26A(B)(C), RX FWP 1A(B)(C) DISCH VLVfor the pump being shutdown.
Verify the minimum flow valve for the pump being secured is open.
Monitor Feed Flow/Steam Flow mismatch and RPV Level to ensure remaining pump(s) can maintain level.
IF the remaining pump(s) cannot maintain RPV Level, THEN reopen the discharge valve FWS-MOV26A(B)(C), RX FWP 1A(B)(C) DISCH VLV and discontinue this procedure.

IF the last Feedwater Pump is being removed from service, THEN open MCR 1 FWS-MOV109, FEED PUMP BYPASS. Stop FWS-P1A(B)(C), RX FWP P1A(B)(C). MCR 1 Page 257 of 278 -~--Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP / SOP ACTIONS LOCATION MODE NOTES Verify CNM-H/A68A(B)(C), RX FWP 1A(B)(C) MIN FLOW FLOW MCR 3 CONTROLLER is in AUTO for the Reactor Feed Pump that was secured. IF Reactor Water Level was intentionally raised in Step 6.1.3, THEN MCR 3 adjust Reactor Water Level to desired level within normal level control band using C33-R600, FW REG VALVES MASTER FLOW CONTROLLER tape set. IF FWS-P1A(B)(C) is to remain in hot standby, THEN maintain seal TB 67' 1 Not required temperatures as follows: for plant Maintain seal water temperature dT less than or equal to 50F AND seal shutdown or water outlet temperature less than or equal to 300F as follows: cooldown 1

FWS-P1A 0 Throttle CCS-V5003A, RFP FWL-P1A SEAL WATER HX-E4A CCS INLET VALVE, as required.

0 Throttle CCS-V5004A, RFP FWL-P1A SEAL WATER HX-E4B CCS INLET VALVE, as required.

FWS-P1B 0 Throttle CCS-V5003B, RFP FWL-P1B SEAL WATER HX-E4C CCS INLET VALVE, as required.

0 Throttle CCS-V5004B, RFP FWL-P1B SEAL WATER HX-E4D CCS INLET VALVE, as required.

FWS-P1C 0 Throttle CCS-V5003C, RFP FWL-P1C SEAL WATER HX-E4E CCS INLET VALVE, as required.

0 Throttle CCS-V5004C, RFP FWL-P1C SEAL WATER HX-E4F CCS INLET VALVE, as required. On H13-P870, verify FWL-P5A(B)(C), GEAR !NCR AUX OIL PMP MCR 1 5A(B)(C) auto starts. Verify min flow valve closes 1 -3 minutes after punip shutdown. MCR 1 Verify FWS-MOV26A(B)(C), RX FWP P1A (B)(C) DISCH VLV is MCR 1 closed. On H13-P870, WHEN the 23 minute time delay allowing for pump coast MCR 1 down has passed, THEN verify the following:

FWL-P1A(B)(C), RX FWP A(B)(C) MN OIL PMP 1A(B)(C) auto stops.
FWL-P5A(B)(C), RX FWP A(B)(C) GEAR INC AUX OIL PMP 5A(B)(C) auto stops.
FWL-P2A(B)(C), RX FWP A(B)(C) AUX OIL PMP 2A(B)(C) auto starts on low oil pressure.
IF FWL-P2A(B)(C), RX FWP A(B)(C) AUX OIL PMP 2A(B)(C) does not maintain pressure greater than 4 psi, THEN FWL-P3A(B)(C), RX FWP A(B)(C) AUX OIL PMP 3A(B)(C) auto starts on low oil pressure.

On H13-P870, place FWL-P2A(B)(C), RX FWP A(B)(C) AUX OIL PMP MCR 1 2A(B)(C) control switch in STOP, and verify FWL-P3A(B)(C), RX FWP Page 258 of 278 ~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP/ SOP ACTIONS LOCATION MODE NOTES A(B)(C) AUX OIL PMP 3A(B)(C) auto starts on low oil pressure. On H13-P870, place FWL-P2A(B)(C), RX FWP A(B)(C) AUX OIL PMP MCR 1 2A(B)(C) control switch in AUTO. On H13-P870, place FWL-P3A(B)(C), RX FWP A(B)(C) AUX OIL PMP MCR 1 3A(B)(C) control switch in STOP. On H13-P870, place FWL-P3A(B)(C), RX FWP A(B)(C) AUX OIL PMP MCR 1 3A(B)(C) control switch in AUTO. Locally verify breaker relay trip flags are reset for Reactor Feed Pump NSW 98' 1 Not required stopped in Step 6.1.7. for plant shutdown or cooldown Decrease reactor power at the rate consistent with generator loading MCR 1 criteria (Attachment 1, MVAR VS MW LIMITS and SOP-0080, Turbine Generator Operation) using control rod insertion per applicable sequence.

Stop inserting control rods at the low power alarm point (as observed in RC & IS panel) and obtain instruction from the Operations Shift Manager regarding further power reductions/shutdown or continued operation at LPAP.
Decrease reactor power to the LPSP using control rod insertion per applicable sequence.

At 300 MWe load, open the following steam drain valves: MCR 1

DTM-AOV32A, 4TH PT HTR EXTR LINE DR
DTM-AOV32B, 4TH PT HTR EXTR LINE DR
DTM-AOV35A, 3RD PT HTR EXTR LINE DR
DTM-AOV35B, 3RD PT HTR EXTR LINE DR Open or verify open G33-MOVF101, RWCU BOTTOM HEAD DRAIN. MCR 1
Verify drain temperature remains stable using Point #4 on B21 R643 or ERIS com uter oint B33NA002.

Manually stroke C33-LVF002, STARTUP FWREG VALVE through full MCR 1 travel to verify smooth operation per SOP-0009, Reactor Feedwater System. Manual Stroking of Start Up FWREG (SOP-0009) Close FWS-MOV105, S/U FW REG VLV ISOL. MCR 1 Station an operator locally to monitor valve position. TB 67' 1 Not required for plant shutdown or cooldown Page 259 of 278

===* Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation

& Shutdown Areas Tables A-3 & H-2 Bases GOP / SOP ACTIONS LOCATION MODE NOTES Establish communications between the local operator and the Main MCR /TB 67' 1 Not required Control Room (MCR). for plant shutdown or cooldown WHEN the FWREG Valve is at 0% open, THEN record demanded MCR /TB 67' 1 Not required position in the MCR, position indication in the MCR, and local position for plant indication in Attachment 7, Calibration Check of FWREG Valves. shutdown or cooldown Use the OPEN and CLOSE Pushbuttons on C33-R602, START UP MCR 1 FWREG VALVE FLOW CONTROLLER to stroke open and then closed the Start Up FWREG. ( Check proper valve movement and smooth operation. MCR /TB 67' 1 Not required for plant shutdown or cooldown Check C33-LVF002, START UP FWREG VLVfull closed. MCR 1 Open FWS-MOV105, S/U FW REG VLV ISOL. MCR 1 I, I I I I , \ ;/ \ : '"' If '1 WHEN less than 30% power AND at the direction of the responsible MCR 1 Operations Management, THEN perform the following:

Transfer station loads to preferred source per SOP-0045, 13.8 KV System and SOP-0046, 4.16 KV System.
Verify MVARs are between + 50 and -50 .
At the SRM cabinets, place the Mode Selector Switches to the OPERATE position.
Prior to initiating a Rx Scram, verify the SRM & IRM Channel Functional Tests are current. IF Channel Functional Tests are not current, THEN refer to Tech Spec 3.3.1.1, 3.3.1.2 and TRM TR 3.3.2.1. Secure SPC per SOP-0140, Suppression Pool Cleanup and Alternate MCR 1 Not Decay Heat Removal. required, but system will automaticall y isolate on a level 3 from a RX SCRAM Contact the Auxiliary Control Room to verify that sufficient condensate MCR/ACR 1/3 Not required demineralizers are in service to prevent physical damage to the for plant demineralizers from high feedwater flow transients.

shutdown or cooldown Page 260 of 278 -~ Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP / SOP ACTIONS LOCATION MODE NOTES Contact the Auxiliary Control Room to verify that sufficient condensate MCR/ACR 1/3 Not required filtration filters are in service to prevent physical damage to the filters for plant from high feedwater flow transients shutdown or cooldown Reduce the number of FWREG Valves in service to one per SOP-MCR 1 Not required 0009, Reactor Feedwater System. for plant shutdown or cooldown Removing a FWREG Valve from Service (SOP-0009) ,., :ri " J '"

I 1;l:* **.*. . I Check feedwater flow is within the capability of the remaining,FWREGs.

MCR 1 Station an operator locally at the FWREG Valve to be removed from TB 67' 1 Not required service. for plant shutdown or cooldown Establish communications between the local operator and the Main MGR/TB 67' 1 Not required Control Room (MCR). for plant shutdown or cooldown Place C33-R601A(R613)(R601B), FWREG VALVE A(B)(C) FLOW MCR 1 CONTROLLER in MANUAL. Throttle closed to 10% open the C33-LVF001A(B)(C), FWREG VALVE MCR 1 A(B)(C) to be removed from service while observing that reactor level is being maintained by the remaining FWREGs. IF level is not being maintained by the remaining FWREGs, THEN MCR 1 place the FWREG that was being removed from service back in service as follows:

Open C33-LVF001A(B)(C), FWREG VALVE A(B)(C) to the same position as the in service FWREGs.
Place C33-R601A(R613)(R601B), FWREG VALVE A(B)(C) FLOW CONTROLLER in AUTO. WHEN the FWREG is at 10% open, THEN close the following isolation MCR 1 valve for the FWREG valve that is being removed from service
For C33-LVF001A close FWS-MOV27A, FWREG VLV 1A INLT Valve.
For C33-LVF001B close FWS-MOV27B, FWREG VLV 1B INLT Valve.
For C33-LVF001C close FWS-MOV27C, FWREG VLV 1C INLT Valve, Fully close the C33-LVF001A(B)(C), FWREG VALVE A(B)(C) that was *MCR 1 removed from service. Page 261 of 278 Q ~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation

& Shutdown Areas Tables A-3 & H-2 Bases GOP / SOP ACTIONS LOCATION MODE NOTES Record the temperature of the feedwater at the reactor feed pumps. TB 67' 1 IF FWS-MOV27A, B, or C, FWREG VLV 1A(1B)(1C) INLTwere closed MCR 1 with feedwater temperature at the reactor feed pumps greater than 200F, THEN refer to Section 5.7 for further stroking requirements. WHEN the FWREG Valve is at 0% open, THEN record demanded MCR /TB 67' 1 Not required position in the MCR, position indication in the MCR, and local position for plant indication. shutdown or cooldown Line up RWCU reject to the main condenser per SOP-0090, Reactor MCR 1 Not Feedwater System. required, but the preferred method to control level if shutdown long term. '., RWCU blowdown operations (SOP-0090) . ' 3 i'.. .4 ' ' Request chemistry sample to verify reactor water quality is within the MCR 1 specifications of Technical Requirement 3.4.13. Notify Radiation Protection prior to rejecting water to the Main MCR 1 Condenser or Radwaste. IF rejecting to the Main Condenser, THEN open G33 F046, RWCU MCR 1 DRAIN TO MN COND. IF rejecting during cold shutdown or refueling, THEN open G33 F031, MCR 1 RWCU DRAIN ORIFICE BYP. IF rejecting with the RWCU HXs isolated, THEN perform the following: MCR 1

Open G33-F107, RWCU REGEN HX BYPASS .
Throttle open G33-PVF033, RWCU REJECT FLOW VALVE to establish reject flow as indicated on G33-R602, RWCU REJECT FLOW.
IF necessary to establish adequate reject flow, THEN close G33-F040, RWCU INBD RETURN VALVE. To establish the reject and maintain RWCU flowrate on G33 R609, MCR 1 RWCU INLET FLOW nearly constant, simultaneously throttle the following:
G33-PVF033, RWCU REJECT FLOW VALVE open using G33 R606, RWCU REJECT FLOW CONTROLLER
G33 F042, RWCU REGEN HX OUTLET closed Observe blowdown flow on G33 R602, RWCU REJECT FLOW. MCR 1 Monitor reactor water level while blowdown is in progress.

MCR 1 ' Page 262 of 278 u -:::.:=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP / SOP ACTIONS LOCATION IF desired, THEN shutdown Reactor Recirculation HPU A(B) per SOP-MCR 0003 to prevent unnecessary Flow Control Valve movement. Initiate a Manual Scram per AOP-0001, Reactor Scram. MCR

Verify the Hydrogen Water Chemistry (HWC) System shuts down on scram si nal. WHEN it is desired to bypass the Feedwater Pump Level 8 Trip, THEN MCR perform Attachment 5, Feedwater Pump Level 8 Trip Jumper Installation/Restoration Step 1. WHEN it is desired to bypass the MSO Level 8 Trip, THEN perform MCR Attachment 8, MSO Level 8 BYPASS Switch Step 1. Monitor Bottom Head Drain Temperature on B21-R643 Point 4 or MCR B33NA002 and take the following actions, as necessary, in a timely controlled manner to prevent an excessive temperature change. (STP-050-0700, RCS Pressure/Temperature Limits Verification).
Reset the Scram .
Reset any FCV runback per ARP-680-04 . Within one hour after THERMAL POWER < 10% RTP in MODE 1, MCR complete the following steps:
Verify/ensure that the RCIS data mode is selected to "CHAN 1 .and CHAN 2".
Select and attempt to withdraw an out-of-sequence control rod .
Verify no rod motion occurs .
Verify Annunciator, P680-07A-C01, CONTROL ROD WITHDRAWAL BLOCK is actuated.
Verify WITHDRAWAL BLOCK Status Light is ON and not flashing.

SR3.3.2.1.4) Place the APRM FCTR Cards to the Normal trip setpoints by MCR depressing the Normal/Setup pushbutton and verifying the normal/setup LED indication is green After the Main Turbine is tripped, open the Feedwater Heater Vents per MCR SOP-0010, MSR & FW Heaters Extraction Steam and Drains.

Establish MSR Steam Blanketing per SOP-0010, MSR & FW MCR Heaters Extraction Steam and Drains. Establishing Steam Blanketing (SOP-0010)

IF Aux. Steam is available, THEN perform the following: MCR

Throttle ASR-MOV104, MSR STM BLANKET SHUTOFF Page 263 of 278 1 1/3 3 3 3 3 3 3 3 3 Not required for plant shutdown or cooldowri Not required for plant shutdown or cooldown

-===* Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP I SOP ACTIONS LOCATION MODE NOTES open until both red and green indication is received.

WHEN Aux steam to MSR steam blanketing line is warm as indicated on Computer point ASRTA01, THEN fully open ASR-MOV104.

Verify the following are closed:

MSS-MOV111, MSR 1 STM SPL Y SHUTOFF
MSS-MOV112, MSR 2 STM SPL Y SHUTOFF
MSS-PVRSHLV1, MSR 1 HIGH LOAD VALVE
MSS-PVRSHLV2, MSR 2 HIGH LOAD VALVE
MSS-PVRSLLV1, MSR 1 LOW LOAD VALVE
MSS-PVRSLLV2, MSR 2 LOW LOAD VALVE
DSR-MOV107, SCAV STM TO 1ST PT HTR A
DSR-MOV109, SCAV STM TO 1ST PT HTR B
DSR-MOV108, SCAV STM TO CONDA
10) DSR-MOV110, SCAV STM TO COND B
DTM-MOV54A, MSL TO MSR 1 COND DR
DTM-MOV54B, MSL TO MSR 2 COND DR IF Aux Steam is available, THEN open the following:
ASR-MOVBSFV1, MSR 1 STM BLANKET SPL Y
ASR-MOVBSFV2, MSR 2 STM BLANKET SPL Y WHEN several minutes have elapsed after opening ASR-MOVBSFV1 and ASR-MOVBSFV2, THEN place the following control switches to CLOSE:
MSS-MOV111, MSR 1 STM SPL Y SHUTOFF
MSS-MOV112, MSR 2 STM SPLY SHUTOFF IF notching is observed during the depressurization and magnitude is MCR 3 Not required less than six inches, THEN: for plant
Make all possible attempts to maintain reactor pressure . shutdown or
Have l&C backfill the reference leg in which notching was cooldown observed, even if reference leg was overfilled prior to this event. IF notching is observed *during the depressurization and magnitude is greater than six inches, THEN declare the trip channels associated with that signal inoperable and comply with Technical Specification requirements.

Reduce the number of running condensate pumps to one per MCR 3 SOP-0007, Condensate System. Shutdown of CNM-P1A(B)(C) CONDENSATE PUMPS (SOP-0007) Request Aux Control Room remove unnecessary Condensate Filters ACR 1 Not required from service per SOP-0124, Condensate Filtration System. for plant shutdown or cooldown Page 264 of 278 -::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP/ SOP ACTIONS LOCATION MODE NOTES Request Aux Control Room remove unnecessary Condensate Demins ACR 1 Not required from service per SOP-0093, Condensate Demineralizer System. for plant shutdown or cooldown IF shutting down CNM-P1 C, CNDS PUMP 1 C, THEN secure Oxygen MCR /TB 95' 1 Not required injection per SOP-0123, Hydrogen Water Chemistry H2 and 02 for plant System. shutdown or cooldown IF securing the Condensate System, THEN perform the following: TB 123' 1 Not required

Close all CNM-V3105 A, B, C, D, and E, CNM-FLT1 A, B, C, for plant D, and E BACKWASH AIR SUPPL Yvalves. shutdown or cooldown
IF desired to isolate and depressurize CNM-TK100, AIR RECEIVING TK, THEN perform the following:

0 CLOSE CNM-V3110, SVCE AIR ISOL VLV INLET SERV. AIR ISOL VLV. 0 Uncap and install hose on CNM-V3112, CNM-TK100 DRAIN ISOLATION VALVE. 0 Open CNM-V3112. Depress the CLOSE pushbutton for CNM-MOV3A(B)(C), CNDS PUMP MCR 1 1A(18)(1C) DISCH. WHEN pump motor current lowers below 100 amps, THEN stop CNM-MCR 1 P1A(B)(C), CNDS PUMP 1A(18)(1 C). WHEN CNM-MOV3A(B)(C), CNDS PUMP 1A(1B)(1C) DISCH is full MCR 1 closed, THEN depress the STOP pushbutton.* Verify associated CCS-MOV67A(B)(C), CNDS PMP 1A(1B)(1C) MOT MCR 1 CLR close for pump stopped. Verify associated CCS-MOV68A(B)(C), CNDS PMP 1A(18)(1 C) BRG MCR 1 CLR close for pump stopped. Locally verify breaker relay trip flags are reset for Condensate Pump NSW 98' Not required stopped in Step 6.1.6 for plant shutdown or cooldown WHEN the Steam Jet Air Ejectors (SJAEs) and Gland Seal and MCR 1 Exhaust System are removed from service, THEN adjust CNM-H/A114 to 10% or to a setpoint determined by the CRS/OSM. * {{: ii TI *.

i) g. ::: 7l*tl WHEN less than or equal to 145 MWT, THEN perform the following:

MCR 1

Secure SJAE per SOP-0092, Offgas System .
Start a mechanical vacuum pump per SOP-0025, Condenser TB 123' & 95' Not required Air Removal System. for plant shutdown or L --Page 265 of 278

~------------------------------------------ Q *::::=-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP / SOP ACTIONS LOCATION MODE NOTES cooldown WHEN all rods have been fully inserted, THEN initiate RPV cooldown MCR 3 at less than or equal to 100°F/hr by one of the following methods:

Initiate an automatic cooldown of the RPVfrom HMI Screen 5532, Pressure Control by performing the following:

0 Depress AUTO RATE in the Cooldown/Heatup Rate section and verify the pushbutton turns cornsilk and disabled. 0 Entered the desired cooldown rate and depress ENTER. 0 Enter the desired Target Pressure in the Throttle Pressure Control section and verify the Target pressure reflects the value entered. 0 Depress the GO pushbutton and verify:

Pressure regulator setpoint is changing automatically.
Bypass valves modulate to control pressure .
RPV temperature lowers in accordance with the selected cooldown rate. '
Initiate a cooldown by slowing and periodically reducing turbine pressure regulator setpoint from HMI Screen 5532, Pressure Control as follows: 0 Enter the desired Target pressure into the HMI display and depress ENTER. 0 Depress the GO pushbutton and verify Press Set changes to reflect the Target value.
ON HMI Screed 5537, BPV Jack, control the Bypass valves as follows to establish the required steam flow to the main condenser for cool down: 0 Enter the BPV Jack Rate and depress ENTER. 0 Enter the BPV Target and depress ENTER. 0 Verify the BPVs open to the desired Target value to establish steam flow to the main condenser.

Prior to reaching 600 psig, initiate monitoring of ERIS Narrow Range MCR 3 level for "notching" of cme or more level indications (IF ERIS trending is not available, THEN contact l&C to arrange for al.ternate Narrow Range trending). / Down range IRMs to maintain indication between downscale alarm and MCR 3 upscale alarm. Insert SRMs to maintain SRM counts between 1 x 103 and 1 x 105 cps. MCR 3 Fully insert SRMs before the IRMs are on range 3. Verify overlap between SRM and IRM. (All SRMs reading < 1x105 cps MCR 3 Page 266 of 278 Q --:::=-Entergy River Bend Station EAL Basis Document Rev*ision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP / SOP ACTIONS LOCATION MODE NOTES prior to IRMs reading< 5 on range 1.) WHEN all control rods are fully inserted, THEN perform one of the MCR 3 following:

NORMAL SHUTDOWN 0 Place the REACTOR SYSTEM MODE SWITCH to SHUTDOWN.

0 WHEN at least 10 seconds have elapsed, THEN reset the Reactor Scram.

SOFT SHUTDOWN 0 Bypass the REACTOR MODE SWITCH POSITION SCRAM per SOP-0079, Reactor Protection System Attachment

7. 0 Place the REACTOR SYSTEM MODE SWITCH to SHUTDOWN.

0 WHEN at least 10 seconds have elapsed, THEN restore the REACTOR MODE SWITCH POSITION SCRAM per SOP-0079, Reactor Protection System Attachment

7. WHEN the reactor is shutdown AND at the direction of the Operations ow ,3 Not required Shift Manager, THEN perform a drywell inspection per Attachment 3, 141'/131'/118 for plant Drywell Inspection Checklist.

'/107'/95'/82' shutdown or cooldown Maintain hot shutdown condition with RPV pressure between 250 psig MCR 3 and 1055 psig. \ IF RHR Pump warmup and flushing is required, THE~ perform MCR 3 warmup/flushing per SOP-0031, Residual Heat Removal. Shutdown Cooling Flush, Warmup, and Startup (SOP-0031) f !* * .* ;'. ~t ,, ::, *** j . IF RPV pressure is less than 135 psig, THEN have Electrical MCR 3 Maintenance implement the PM Task to re-land the thermal overload/loss of power annunciator leads for E12-F009, RHR SHUTDOWN COOLING INBD ISOL VALVE. Verify the following breakers are ON: MCR 3

EHS-MCC2E BKR 5C, C002A DISCH MIN FLOW VALVE
EHS-MCC2F BKR 7B, C002B DISCH MIN FLOW Shutdown Cooling Flush MCR 3 Not required
Request Chemistry to verify Suppression Pool is within best for plant practice limits of CSP-0006, Chemistry Surveillance and shutdown or Scheduling System. cooldown
IF suppression pool conductivity is NOT within best practice limits of CSP-0006, THEN perform a complete flush.
IF desired and suppression pool conductivity is within best practice limits of CSP-0006, THEN perform the following:

0 Place RHR A(B) in suppression pool cooling. 0 Monitor E12-R61 OA(B), HX A(B) OUTLET CONDUCTIVITY OR Chemistry sample from SST-PNL80, and continue flush until conductivity is less than 2 umho/cm. Page 267 of 278 Q -::::=* Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP/ SOP ACTIONS o Vent RHR A(B) HX as follows:

Open E12-F074A(B), RHR A(B) HX UP STREAM VENT VALVE.
Open E12-F073A(B), RHR A(B) HX ON STREAM VENT VALVE.
WHEN at least 1 minute has elapsed, THEN close the following:
E12-F073A(B)
E12-F074A(B)
WHEN conductivity is less than 2 umho/cm, THEN perform the following:

o Secure RHR A(B) from suppression pool cooling. o Close E12-VF085A(B), LPCS FILL PUMP STOP CHECK TO RHR A DISCH (DISCH FILL PUMP STOP CHECK TO RHR B DISCH). o Continue with pump warm-up as desired. LOCATION During cooldown, review Attachment 7, High Critical Non-Safety MCR Related MOVs That Are Susceptible to Thermal Binding and carryout actions to prevent thermal binding or actions to unbind the valves if they *are-closed when the valve temperature is greater than 200°F. IF MSIVs have been closed for pressure control, THEN the following MCR systems may be utilized as necessary to continue a cool down at less than or equal to 100°F/hr:

RWCU system per SOP-0090, Reactor Water Cleanup System (i.e. RWCU Slowdown Operation)
Main Steam Line Drains MODE NOTES 3 3 IFNVHEN reactor pressure is less than 400 psig, THEN any running MCR / TB 67' 3 Not required for plant shutdown or cooldown reactor feedwater pumps may be shutdown per SOP-0009, Reactor Feedwater System. Prior to reaching 135 psig, initiate monitoring of the following MCR parameters:
RHR Room sump levels (monitor for possible reactor vessel inventory loss from shutdown cooling leakage). (DFR-Ll135 and DFR-LI 138).
Suppression Pool for unexpected level rise (monitor for reactor vessel inventory loss from RHR to the Suppression Pool). WHEN RPV pressure has been 16wered to below 135 psig, THEN MCR place one loop of RHR in Shutdown Cooling per SOP-0031, Residual Heat Removal. RHR Pump A Warmup (SOP-0031)

Verify closed E12-F064A, RHR PUMP A MIN FLOW TO SUP PL. Page 268 of 278 3 3 ~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP/ SOP ACTIONS LOCATION MODE NOTES Verify open E12-F047A, RHR A HX INLET VALVE. MCR 3 Verify closed E12-F004A, RHR PUMP A SUP PL SUCTION VALVE. MCR 3 Verify open E12-F006A, RHR PUMP A SOC SUCTION VALVE. MCR 3 On H13-P870, verify closed SPC-AOV16, SPC HX SW DISCH VLV. MCR 3 On H13-P870, throttle open E12-F068A, RHR HX A SVCE WTR RTN MCR 3 to establish less than or equal to 5800 gpm flow as indicated on E12-R602A, RHR HX A SVCE WTR FLOW. Close E12-F048A, RHR A HX BYPASS VALVE. MCR 3 Verify open E12-F003A, RHR A HX OUTLET VALVE. MCR 3 Verify open E12-F010, RHR SOC MAN ISOL VLV., MCR 3 In the Div 1 RSS Room at C61-PNL001, verify E12-MOVFOOS CB 95' Div 1 3 REQUIRED ENABLE/DISABLE Switch is in ENABLE. RSS Room Perform the following: MCR 3

Depress B21H-S33, INBD ISOLATION SEAL-IN RESET Pushbutton.
Depress B21H-S32, OUTBD ISOLATION SEAL-IN RESET Pushbutton.

At H13-P601, check RHR ISOLATION Status Lights are ON for E12-MCR 3 FOOS and F009. Note current indicated CNS flow at LWS-PNL 187 on CNS-Fl 116. ACR 3 Not required for plant shutdown or cooldown Slowly open E12-VF020, SHUTDOWN COOLING SUCTION FILL. AB 95' 3 Not required RHR C Pump for plant Room shutdown or cooldown WHEN CNS flow into the shutdown cooling header stops as indicated AB 95' 3 Not required by a lack of flow noise or flow indication of approximately the same RHR C Pump for plant value as previously noted on CNS-Fl116, THEN close E12-VF020. Room shutdown or cooldown Open E12-F008, RHR SHUTDOWN COOLING OUTBD ISOL VALVE. MCR 3 Open E12-F009, RHR SHUTDOWN COOLING INBD ISOL VALVE. MCR 3 Perform STP-204-0204, RHR Shutdown Cooling Piping Fill Verification. Steam 3 Not required Tunnel 114' for plant shutdown or cooldown Notify Radwaste of reactor water flush to the Waste Collector Tanks. MCR 3 Page 269 of 278 --::::::-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP / SOP ACTIONS LOCATION MODE NOTES Open E12-F049, RHRA TO RADWASTE UP STREAM ISOL VALVE. MCR 3 Throttle open E12-F040, RHR A TO RADWASTE DN STREAM ISOL MCR 3 VALVE. Monitor E12-R601, RHR TEMPERATURES, Point 1, RHR INLET TO MCR 3 HX1 A-1 (E12-N004A) for temperature rise and throttle E12-F040 to maintain less than or equal to 1 OO'F/hr heatup. Continue the warm-up until E12-R601 Point 1 is within 1 OO'F of reactor MCR 3 water temperature. Close E12-F049, RHR A TO RADWASTE UP STREAM ISOL VALVE. MCR 3 Close E12-F040, RHR A TO RADWASTE DN STREAM ISOL VALVE. MCR 3 Open E12-F048A, RHR HXA BYPASS VALVE. MCR 3 RHR Pump B Warm-up (SOP-0031) j t ,f : ,, ! ~i **::: *.: . Verify closed E12-F064B, RHR PUMP B MIN FLOW TO SUP PL. MCR 3 Verify open E12-F047B, RHR B HX INLET VALVE. MCR 3 Verify closed E12-F004B, RHR PUMP B SUP PL SUCTION VALVE. MCR 3 Verify open E12-F006B, RHR PUMP B SOC SUCTION VALVE. MCR 3 IF Standby Service Water is supplying service water loads, THEN on MCR 3 H13-P870, verify closed SPC-AOV16, SPC HX SW DISCH VLV. On H13-P870, throttle open E12-F068B, RHR HX B SVCE WTR RTN MCR 3 to establish less than or equal to 5800 gpm flow as indicated on E12-R602B, RHR HX B SVCE WTR FLOW. Verify open E12-F010, RHR SOC MAN ISOL VLV. MCR 3 Verify closed E12-F049, RHR A TO RADWASTE UP STREAM ISOL MCR 3 VALVE. Verify closed E12-F040, RHR A TO RADWASTE DN STREAM ISOL MCR 3 VALVE. In the Div 1 RSS Room at C61-PNL001, verify E12-MOVF008 CB 95' Div 1 3 REQUIRED ENABLE/DISABLE Switch is in ENABLE. RSS Room Perform the following: MCR 3

Depress B21H-S32, OUTBD ISOLATION SEAL-IN RESET Pushbutton.
Depress B21 H-S33, INBD ISOLATION SEAL-IN RESET Pushbutton.

At H13-P601, check RHR ISOLATION Status Lights are ON for E12-MCR 3 FOOS and F009. Note current indicated CNS flow at LWS-PNL 187 on CNS-Fl 116. ACR 3 Not required for plant shutdown or Page 270 of 278 Q ~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP / SOP ACTIONS LOCATION MODE NOTES cooldown Slowly open E12-VF020, SHUTDOWN COOLING SUCTION FILL. AB 95' RHR 3 Not required C Pump for plant Room shutdown or cooldown WHEN CNS flow into the shutdown cooling header stops as indicated AB 95' RHR 3 Not required by a lack of flow noise or flow indication of approximately the same C Pump for plant value as previously noted on CNS-Fl116, THEN close E12-VF020. Room shutdown or cooldown Open E12-F008, RHR SHUTDOWN COOLING OUTBD !SOL VALVE. MCR 3 Open E12-F009, RHR SHUTDOWN COOLING INBD !SOL VALVE. MCR 3 Perform STP-204-0204, RHR Shutdown Cooling Piping Fill Verification. Steam 3 Not required Tunnel 114' for plant shutdown or cooldown Notify Radwaste of reactor water flush to the Waste Collector Tanks. MCR 3 Unlock and open E12-VF072B, RHR B DISCH LINE FLUSH. AB 70' RHR 3 REQUIRED B Pump Room Unlock and throttle open E12-VF070, RHR DR TO RADWASTE. AB 80' RHR 3 REQUIRED A Pump Room Monitor E12-R601, RHR TEMPERATURES, Point 11, RHR DISCH TO MCR 3 RADWASTE (E12-N024) and throttle E12~VF070 to maintain less than or equal to 100°F/hr heatup. Continue the warm-up until E12-R601 Point 11 is within 100°F of MCR 3 reactor water temperature. Close and lock E12-VFO~O, RHR DR TO RADWASTE. AB 80' RHR 3 REQUIRED A Pump Room Close and lock E12-VF072B, RHR B DISCH LINE FLUSH. AB 70' RHR 3 REQUIRED B Pump Room Startup of Shutdown Cooling (SOP-0031) j(lt i:; if r !]:1> ./Ju :1J ;,':' ; \,t; ":~, ,co :I;,~;;:" IF any of the following RHR Shutdown Cooling interlocks are to be MCR 3 bypassed, THEN obtain senior plant management review and approval and verify contingency methods are in place to supply sufficient makeup water if a draining event occurs while the SDC interlocks are bypassed:

Low reactor water level isolation of E12-F008, RHR SHUTDOWN COOLING OUTBD ISOL VALVE and E12-F009, RHR SHUTDOWN COOLING !NBD ISOL VALVE. Page 271 of 278

~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP/ SOP ACTIONS LOCATION MODE NOTES

Interlocks between E12-F004, RHR PUMP SUP PL SUCTION VALVE and E12-F006, RHR PUMP SDC SUCTION. On H13-P601, verify less than 135 psig Reactor Pressure as indicated MCR 3 on B21-R623A(B), RX LEVEUPRESSURE RECORDER A(B). Verify closed the following:

MCR 3

E12-F004A(B), RHR PUMP A(B) SUP PL SUCTION VALVE
E12-F064A(B), RHR PUMP A(B) MIN FLOW TO SUP PL
E12-F024A(B), RHR PUMP A(B) TEST RTN TO SUP PL
E12-F037A(B), RHR A(B) TO UPPER POOL FPC ASSIST
E12-F048A(B), RHR A(B) HX BYPASS VALVE .
E12-F011A(B), RHR A(B) HX CNDS FLUSH TO SUP PL. Place in OFF and initiate administrative controls for EHS-MCC2E(2F)

AB 114' West 3 REQUIRED BKR 5C(78), C002A(B) DISCH MIN FLOW VALVE. IF Standby Service Water is supplying service water loads, THEN on MCR 3 H13-P870, verify closed SPC-AOV16, SPC HX SW DISCH VLV. On H13-P870, throttle open E12-F068A(B), RHR HX A(B) SVCE WTR MCR 3 RTN to establish less than or equal to 5800 gpm flow as indicated on H13-P601, E12-R602A(B), RHR HX A(B) SVCE WTR FLOW. Verify Step 4.4.2 has been performed. MCR 3 At H13-P601, depress 821H-S32, OUTBD ISOLATION SEAL-IN MCR 3 RESET Pushbutton. At H13-P601, depress 821 H-S33, INBD ISOLATION SEAL-IN RESET MCR 3 Pushbutton. At H13-P601, check RHR ISOLATION Status Lights are ON for E12-MCR 3 FOOS and E12-F009. In the Div 1 RSS Room at C61-PNL001, verify E12-MOVF008 CB 95' Div 1 3 REQUIRED ENABLE/DISABLE Switch is in ENABLE. RSS Room Verify open the following: MCR 3

E12-F010, RHR SDC MAN ISOL VLV
E12-F009, RHR SHUTDOWN COOLING INBD ISOL VALVE
E12-F008, RHR SHUTDOWN COOLING OUTBD ISOL VALVE
E12-F006A(B), RHR PUMP A(B) SDC SUCTION VALVE
E12-F047A(B), RHR A(B) HX INLET VALVE Verify open one of the following:

MCR 3

E12-F053A(B), RHR PUMP A(B) SDC INJECTION VALVE
E12-F037A(B), RHR A(B) TO UPPER POOL FPC ASSIST Close E12-F003A(B), RHR A(B) HX OUTLET VALVE. MCR 3 Start E12-C002A(B), RHR PUMP A(B) and IMMEDIATELY throttle MCR 3 open E12-F048A(B), RHR A(B) HX BYPASS VALVE to obtain greater Page 272 of 278 Q ~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation

& Shutdown Areas Tables A-3 & H-2 Bases GOP / SOP ACTIONS LOCATION MODE NOTES than or equal to 2000 gpm and less than or equal to 3000 gpm. Establish a stable flow of greater than or equal to 4000 gpm and less MCR 3 than or equal to 5000 gpm by throttling E12-F048A{B), RHR A(B) HX BYPASS VALVE. Throttle open E12-F003A(B), RHR A(B) HX OUTLET VALVE to MCR 3 approximately 10 PERCENT as indicated on E12-R611 A(B), HX A(B) OUTLET VLV POS. Establish a cool down rate of less than 1 OO'F/hr as follows: MCR 3

Slowly throttle open E12-F003A(B)

RHR A(B) HX OUTLET VALVE and monitor the cool down rate.

Throttle E12-F003A(B), RHR A(B) HX OUTLET VALVE and E12-F048A(B), RHR A(B) HX BYPASS VALVE to obtain the desired cooldown rate or maintain the desired coolant temperature while maintaining a constant RHR loop flow.
IF shifting divisions of Shutdown Cooling per Section 5.6, THEN in the other RHR loop, throttle E12-F003B(A), RHR B(A) HX OUTLET VALVE and E12-F048B(A), RHR B(A) HX BYPASS VALVE to maintain the desired cooldown rate or coolant temperature while maintaining a constant RHR loop flow.
Close FWS-MOV?A(B), A(B) FW OUTBD !SOL. IF it is desired to establish RHR Shutdown Cooling Protection, THEN MCR/ AB 95' Not required Go To Section 4.5. & 115' for plant shutdown or cooldown ' c\ ,,, ' ,, WHEN RHR Shutdown Cooling is established and adequate RPV MCR 3 makeup 'is assured via CRD or Feedwater, THEN close FWS-MOV?A(B), A(B) FW OUTBD !SOL valve on the Feedwater Header supporting RHR Shutdown Cooling. WHEN RPV cool down is being conducted using RHR Shutdown MCR 3 Cooling, THEN stop discharging steam to the main condenser, break condenser vacuum and continue to shutdown the turbine plant as follows:
Place CONDENSER LOW VACUUM BYPASS Switches to BYPASS.
Close all turbine bypass valves and steam drain valves .
Open CNM-AOWB, CNDS VAC BRKR .
WHEN condenser vacuum reaches approximately O" Hg, THEN shutdown steam seals per SOP-0015, Gland Seal System and Exhaust System. WHEN condenser vacuum reaches O" Hg AND mechanical vacuum TB 67' 3 Not required pump operation is no longer required, THEN align Alternate Hotwell for plant Level tygon tubing per SOP-0008, Condensate Storage, Makeup and shutdown or Transfer.

cooldown Page 273 of 278 Q ~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP/ SOP ACTIONS LOCATION MODE NOTES At less than 190°F, perform the following: MCR 3

Open B21-MOVF001, RX DN STREAM HEAD VENT TO DW EQPT DR SUMP.
Open B21-MOVF002, RX UP STREAM HEAD VENT TO DW EQPT DR SUMP.
Close B21-MOVF005, RX HEAD VENT TO MSL A
Establish administrative controls to maintain the established vessel vent path until vessel head piping is disassembled.

IF required/desired, THEN close the MSIVs by placing the following in MCR 3 CLOSE:

B21-F028B, MSL B OUTBD MSIV
B21-F028D, MSL D OUTBD MSIV
B21-F028A, MSL A OUTBD MSIV
B21-F028C, MSL C OUTBD MSIV
B21-F022B, MSL B INBD MSIV
B21-F022D, MSL D INBD MSIV
B21-F022A, MSL A INBD MSIV
B21-F022C, MSL C INBD MSIV '--\ < ,'1: f i At less than 200°F, Mode 4, perform the following:

MCR 3 At H13-P632, place the following switches to BYPASS:

E31A-S1A, RWCU ISOLATION BYPASS DIV 1
E31A-S2A, RCIC ISOLATION BYPASS DIV 1
E31A-S4A, RHR ISOLATION BYPASS DIV 1 At H13-P642, place the following switches to BYPA!;>S:

MCR 3

E31A-S1B, RWCU ISOLATION BYPASS DIV 2
E31A-S2B, RCIC ISOLATION BYPASS DIV 2
E31A-S4B, RHR ISOLATION BYPASS DIV 2 Implement Shutdown Cooling Protection per SOP-0031, Residual Heat MCR / AB 95' 3 Not required Removal. & 115' for plant shutdown or cooldown Bypass RPS trip logic using EOP-0005 Enclosure 12 Bypass Switches MCR 3 per SOP-0079, Reactor Protective System. Bypass ARI logic trips per SOP-0079.

MCR 3 Bypass Backup Scram Valve trips per SOP-0079. MCR 3 At less than 200°F, Mode 4, perform the following to prevent isolating MCR 3 Not required Breathing Air: for plant

Verify open SAS-MOV102, SVCE AIR OUTBD ISOL shutdown or Page 27 4 of 278 Q *====* Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation

& Shutdown Areas Tables A-3 & H-2 Bases GOP/ SOP ACTIONS LOCATION MODE NOTES cooldown

Open EHS-MCC2J Bkr 3C, SAS-MOV102 CONTAINMENT AB 141' West 3 Not required SERVICE AIR ISOLATION.

for plant shutdown or cooldown

Hang the following SAS-MOV102 Caution Tags stating that MCR 3 Not required before closing the breaker or valve, verify that Breathing Air is for plant not in use.: shutdown or 0 EHS-MCC2J Bkr 3C cooldown 0 SAS-MOV102 local handwheel 0 SAS-MOV102 MCR control switch
Hang "Breathing Air in Use" sign in the MCR. MCR 3 Not required for plant shutdown or cooldown
Make an "Open Item" Narrative Log entry by checking the MCR 3 Not required "Open Item" box stating that "Breathing Air Is in Use" to carry for plant over until Breathing Air is no longer in use. shutdown or cooldown At less than 200°F, notify Chemistry to consider securing the Durability AB 114' 3 Not required Monitor. Crescent for plant Area shutdown or cooldown IF required/desired, THEN shutdown Reactor Recirculation System per MCR 3 Not required SOP-0003, Reactor Recirculation System and raise reactor water level for plant to at least 75 inches on shutdown range level instrumentation.

shutdown or cooldown As necessary, reduce the number of operating Turbine Building Chillers TB 67' 3 Not required per SOP-0064 to prevent the chillers from tripping on low load. for plant shutdown or cooldown .GOP-0003 Scram Recovery Verify/establish on-scale neutron monitoring on the SRMs and IRMs MCR 3

VERIFY the SRM Channel Functional Tests are current. If Channel Functional Tests are not current, refer to Tech Spec 3.3.1.2. Maintain RPV pressure to prevent excessive cooldown rates or RPV MCR 3 overpressurization by:
Use of Main Turbine Bypass System .
Use of Normal Plant Steam Loads/Steam Line Drains .
RCIC in CST to CST Mode per SOP-0035 Reactor Core Isolation Cooling System. Page 275 of 278 Q ~Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation

& Shutdown Areas Tables A-3 & H-2 Bases GOP / SOP ACTIONS LOCATION MODE NOTES

Alternate opening of SRVs as needed (only on RPV isolation).

Maintain RPV Level using: MCR 3

Condensate/Feedwater ,
RCIC
HPCS
CRD/RWCU , If tripped, restart Reactor Recirculation Pumps on LFMG per MCR/ LFMG 3 Not required SOP-0003, Reactor Recirculation:

Room for plant

Open flow control valves to the full open position . shutdown or cooldown Verify Main Turbine steam seals are being maintained at approximately MCR 3 4 psig. Start air removal pumps per SOP-0025 Condenser Air Removal MCR /TB 67' 3 Not required System. for plant
Maintain condenser vacuum between 23" Hg and 28" Hg . shutdown or cooldown IF the Steam Jet Air Ejectors have been lost, THEN Secure Offgas MCR /TB 95' 3 Not required System per SOP-0092, to establish purge air flow in order to prevent & 123' for plant system reverse flow. shutdown or cooldown Notify Chemistry Department to operate the Offgas Hydrogen MCR/ TB 3 Not required Analyzers per COP-0227, Operation of the Offgas Hydrogen Analyzers.

123' for plant shutdown or cooldown Record the highest vessel pressure indicated by tracking pointer on RB 114' 3 Not required B21-PIR004A and B21-PIR004B (114' Containment). for plant shutdown or Reset the tracking pointer. cooldown Inspect all CRD HCUs for leakage due to piping cracks. RB 114' 3 Not required

Notify Engineering NOE that visual inspections of HCU for plant charging water piping are required.

shutdown or cooldown WHEN FWREG Valves are removed from service, THEN perform MCR /TB 67' 3 Not required SOP-0009 Attachment for Calibration Check of FWREG Valve. for plant shutdown or cooldown Go To GOP-0002 -Start at the beginning of GOP-0002 and complete MCR 3 Page 276 of 278 @ -~-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation & Shutdown Areas Tables A-3 & H-2 Bases GOP/ SOP ACTIONS LOCATION MODE NOTES all steps required to place the plant in the desired mode after scram. IF the plant tripped while connected to the grid, THEN notify Site MCR 3 Design Engineering to notify the TOP personnel of the event per ENS-DC-201, ENS Transmission Grid Monitoring Attachment 9.3 Step 3.0[1]. Engineering perform a review of post scram cool down data and MCR 3 compare to PT Curves provided in STP-050-0700 Attachment

3. Also verify the cooldown rate is bounded by analyzed thermal cycles. Shift Manager to perform a Post SCRAM crew critique identifying all MCR 3 Human Performance issues and Equipment Malfunctions.

Document each item on separate CRs. Attach Crew Critique to this procedure. AOM/Shift Manager review equipment malfunctions and recommend to MCR 3 OSRC required repairs prior to restart. This should include an evaluation of risk mitigating Non TRM structures, systems, and components. Control Room ventilation systems have adequate engineered safety/design features in place to preclude a Control Room evacuation due to the release of a hazardous gas. Therefore, the Control Room is not included in this assessment or in Table H-2. Page 277 of 278

-Entergy River Bend Station EAL Basis Document Revision XXX Attachment 2 -Safe Operation

& Shutdown Areas Tables A-3 & H-2 Bases Table A-3 & H-2 Results Table A-3 & H-2 Safe Operation & Shutdown Rooms/Areas Room/Area Mode Auxiliary Building 70' RHR B Pump Room 3 Auxiliary Building 80' RHR A Pump Room 3 Auxiliary Building 114' West 3 Control Building 95' Div 1 RSS Room 3 Mode 3 is included above for SOC-related activities because the procedures begin alignment in Mode 3; however, these actions could be delayed until Mode 4, if necessary. In order to ensure adequate guidance to emergency response personnel, the above areas are added to the EAL in order to provide prompt operator guidance for EAL declaration. Control Room ventilation systems have adequate engineered safety/design features in place to preclude a Control Room evacuation due to the re[ease of a hazardous gas. Therefore, the Control Room is not included in this assessment or in Table H-2. Page 278 of 278 ENCLOSURE 4 RBG-47847 NEI 99-01, REV. 6, DEVIATIONS AND DIFFERENCES, RBS RBS NEI 99-01 Revision 6 EAL Comparison Matrix RBS EAL Comparison Matrix Table of Contents Section I n trod ucti on ----------------------------------------------------------------------------------

--------------------------------------------------------------------
-1 Com pa ri son Matrix Form at -------------------------------------------------------------------------------------------------------------------------------------1 EAL Wording -------------------------------------------------------------------------------------------------------------------------------------------------------1 EAL Em p ha sis Tech n i q u es-------------------------------------------------------------------------------------------------------------------------------------1 GI o bal D iff e ren ces ------------------------------------------------------------------------------------------------------------------------------------------------2 D iff eren ces and Deviations

3 Category A -Abnormal Rad Levels/ Rad Effluent-----------------------------------------------------------------------------------------------------16 Category C -Cold Shutdown/

Refueling System Malfunction-------------------------------------------------------------------------------------- 35 Category D -Permanently Defueled Station Malfunction

59 Category E -Independent Spent Fuel Storage Installation ( IS FS I)-----------------------------------------~-------------------------------------- 61 Category F -Fission Product Barrier Degradation----------------------------------------------------------------------------------------------------- 63 Category H -Hazards and Other Conditions Affecting Plant Safety------------------------------------------------------------------------------- 78 Category S -System M alf u n cti on -------------------------------------------------------------------------------------------------------------------------- 98 Tab I e 1 -RBS EAL Categ ori es/Su bca teg ori es------------------------------------------------------------------------------------------------------------5 Table 2 -NE I / RBS EAL Id entifi cation Cross-Reference

.:---------------------------6 Table 3 -Summary of Deviations

11 i of i RBS EAL Comparison Matrix Introduction This document provides a line-by-line comparison of the Initiating Conditions (ICs), Mode Applicability and Emergency Action Levels (EALs) in NEI 99-01 Rev. 6 Final, Development of Emergency Action Levels for Non-Passive Reactors, ADAMS Accession Number ML 12326A805, and River Bend Station (RBS) ICs, Mode Applicability and EALs. This document provides a means of assessing RBS differences and deviations from the NRC endorsed guidance given in NEI 99-01. Discussion of RBS EAL bases and lists of source document references are given in the EAL Technical Bases Document. It is, therefore, advisable to reference the EAL Technical Bases Document for background information while using this document. Comparison Matrix Format The ICs and EALs discussed in this document are grouped according to NEI 99-01 Recognition Categories. Within each Recognition Category, the ICs and EALs are listed in tabular format according to the order in which they are given. in NEI 99-01. Generally, each row of the comparison matrix provides the following information:

NEI EAUIC identifier
. NEI EAUIC wording
RBS EAL/IC identifier
RBS EAL/IC wording
Description of any differences or deviations EAL Wording In Section 4.1, NEI recommends the following: "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 (EALs) within the context of site-specific characteristics -locale, plant design, operating features, terminology, etc. Meeting this goal will result in a shorter and less cumbersome NRC review and approval process, closer alignment with the schemes of other nuclear power plant sites and better positioning to adopt future industry-wide scheme enhancements" 1 of 127 To assist the Emergency Director (ED), the RBS EALs have been written in a clear and concise style (to the extent that the differences from the NEI EAL wording could be reasonably documented and justified). This supports timely and accurate classification in the tense atmosphere of an emergency-event. The EAL differences introduced to reduce reading burden comprise almost all of the differences justified in this document. EAL Emphasis Techniques Due to the width of the table columns and table formatting constraints in this document, line breaks and indentation may differ slightly from the appearance of comparable wording in the source documents. NEI 99-01 is the source document for the NEI EALs; the RBS EAL Technical Bases Document for the RBS EALs. Development of the RBS IC/EAL wording has attempted to minimize inconsistencies and apply sound human factors principles. As a result, differences occur between NEI and RBS ICs/EALs for these reasons alone. When such difference may infer a technical difference in the associated NEI IC/EAL, the difference is identified and a justification provided. The print and paragraph formatting conventions summarized below guide presentation of the RBS EALs in accordance with the EAL writing criteria. Space restrictions in the EAL table of this document sometimes override this criteria in cases when following the criteria would introduce undesirable complications in the EAL layout.

Upper case-bold print is used for the logic terms AND, OR and EITHER.
Bold font is used for certain logic terms, negative terms (not, cannot, etc.), any, all.
Upper case print is reserved for defined terms, acronyms, system abbreviations, logic terms (and, or, etc. when not used as a conjunction), annunciator window engravings.
Three or more items in a list are normally introduced with "Any of the following

... " or "All of the following ... " Items of the list begin with bullets when a priority or sequence is not inferred.

The use of AND/OR logic within the same EAL has*been avoided when possible.

When such logic cannot be avoided, indentation and separation of subordinate contingent phrases is employed. RBS EAL Comparison Matrix Global Differences The differences listed below generally apply throughout the set of EALs and are not repeated in the Justification sections of this document. The global differences do not decrease the effectiveness of the intent of NEI 99-01. 1. The NEI phrase "Notification* of Unusual Event" has been changed to "Unusual Event" or abbreviated "UE" to reduce EAL-user reading burden. 2. NEI 99-01 IC Example EALs are implemented in separate plant EALs to improve clarity and readability. For example, NEI lists all IC HU3 Example EALs under one IC. The corresponding RBS EALs appear as unique EALs (e.g., HU3.1 through HU3.4). 3. Mode applicability identifiers (numbers/letter) modify the NEI 99-01 mode applicability names as follows: 1 -Power Operation, 2 -Startup, 3 -Hot Shutdown, 4 -Cold Shutdown, 5 -Refueling, DEF -Defueled. NEI 99-01 defines Defueled as follows: "Reactor Vessel contains no irradiated fuel (full core off-load during refueling or extended outage)." 4. The Hot Standby mode is applicable only to PWRs and therefore is not used for RBS. 5. NEI 99-01 uses the terms greater than, less than, greater than or equal to, etc. in the wording of some example EALs. For consistency and reduce EAL-user reading burden, RBS has adopted use of Boolean symbols in place of the NEI 99-01 text modifiers within the EAL wording. 6. "min." is the standard abbreviation for "minutes" and is used to reduce EAL user reading burden. 7. The terms "increase" and "decrease" have been replaced with the terms "rise" and "lower". 8. IC/EAL identification:

NEI 99-01 defines the thresholds requiring emergency classification (example EALs) and assigns them to ICs which, in turn, are grouped in "Recognition Categories." RBS endeavors to optimize the NEI EAL organization and identification scheme to 2 of 127 enhance usability of the plant-specific EAL set. To this end, the RBS IC/EAL scheme includes the following features:

a. Division of the NEI EAL set into three groups: o EALs applicable under ill!. plant operating modes -This group would be reviewed by the EAL-user any time emergency classification is considered.

o EALs applicable only under hot operating modes -This group would only be reviewed by the EAL-user when the plant is in Hot Shutdown, Startup or Power Operation mode. o EALs applicable only under cold operating modes -This group would only be reviewed by the EAL-user when the plant is in Cold Shutdown, Refueling or Defueled mode. The purpose of the groups is to avoid review of hot condition EALs when the plant is in a cold condition and avoid review of cold condition EALs when the plant is in a hot condition. This approach significantly minimizes the total number of EALs that must be reviewed by the user for a given plant condition, reduces EAL-user reading burden and, thereby, speeds identification of the EAL that applies to the emergency.

b. Within each of the above three groups, assignment of EALs to categories/subcategories

-Category and subcategory titles are selected to represent conditions that are operationally significant to the EAL-user. Subcategories are used as necessary to further divide the EALs of a category into logical sets of possible emergency classification thresholds. The RBS EAL categories/subcategories and their relationship to NEI Recognition Categories are listed in Table 1. RBS EAL Comparison Matrix c. Unique identification of each EAL -Four characters comprise the EAL identifier as illustrated in Figure 1. Figure 1 -EAL Identifier EAL Identifier xxx.x Category (A. H. E. S. F. C) _J I I L Sequential number within subcategory/classification Emergency classification (G, S, A, U) _J L Subcategory number (1 if no subcategory) The first character is a letter associated with the category in which the EAL is located. The second character is a letter associated with the emergency classification level (G for General Emergency, S for Site Area Emergency, A for Alert, and U for Notification of Unusual Event). The third character is a number associated with one or more subcategories within a given category. Subcategories are sequentially numbered beginning with the number "1". If a category does not have a subcategory, this character is assigned the number "1 ". The fourth character is a number preceded by a period for each EAL within a subcategory. EALs are sequentially numbered within the emergency classification level of a subcategory beginning with the number "1". The EAL identifier is designed to fulfill the following objectives: o Uniqueness -The EAL identifier ensures that there can be no confusion over which EAL is driving the need for emergency classification. o Speed in locating the EAL of concern -When the EALs are displayed in a matrix format, knowledge of the EAL identifier alone can lead the EAL-user to the location of the EAL within the classification matrix. The identifier conveys the category, subcategory and classification level. This assists ERO responders (who may not be in the same facility as the ED) to find the EAL of concern in a timely manner without the need for a word description of the classification threshold. o Possible classification upgrade -The category/subcategory/identifier scheme helps the EAL-user find higher emergency classification EALs that may become active if plant conditions worsen. Table 2 lists the RBS ICs and EALs that correspond to the NEI !Cs/Example EALs when the above EAL/IC organization and identification scheme is implemented. Differences and Deviations In accordance NRC Regulatory Issue Summary (RIS) 2003-18 "Use of Nuclear Energy Institute (NEI) 99-01, Methodology for Development of Emergency Action Levels" Supplements 1 and 2, a difference is an EAL change in which the basis scheme guidance 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 RBS EAL. A deviation is an EAL change in which 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 RBS proposed EAL. Administrative changes that do not actually change the textual content are neither differences nor deviations. Likewise, any format change that does not alter the wording of the IC or EAL is considered neither a difference nor a

deviation.

3 of 127 The following are examples of differences:

Choosing the applicable EAL based upon plant type (i.e., BWR vs. PWR).
Using a numbering scheme other than that provided in NEI 99-01 that does not change the intent of the overall scheme.
Where the NEI 99-01 guidance specifically provides an option to not include an EAL if equipment for the EAL does not exist at RBS (e.g., automatic real-time dose assessment capability).
Pulling information from the bases section up to the actual EAL that does not change the intent of the EAL.

RBS EAL Comparison Matrix

Choosing to state ALL Operating Modes are applicable instead of stating N/A, or listing each mode individually under the Abnormal Rad Level/Radiological Effluent and Hazard and Other Conditions Affecting Plant Safety sections.
Using synonymous wording (e.g., greater than or equal to vs. at or above, less than or equal vs. at or below, greater than or less than vs. above or below, etc.)
Adding RBS equipment/instrument identification and/or noun names to EALs.
Combining like ICs that are exactly the same but have different operating modes as long as the intent of each IC is maintained and the overall progression of the EAL scheme is not affected.
Any change to the IC and/or EAL, and/or basis wording, as stated in NEI 99-01, that does not alter the intent of the IC and/or EAL, i.e., the IC and/or EAL continues to: o Classify at the correct classification level. o Logically integrate with other EALs in the EAL scheme. o Ensure that the resulting EAL scheme is complete (i.e., classifies all potential emergency conditions).

The following are examples of deviations:

Use of altered mode applicability.

j

Altering key words or time limits.
Changing words of physical reference (protected area, safety-related equipment, etc.).
Eliminating an IC. This includes the removal of an IC from the Fission Product Barrier Degradation category as this impacts the logic of Fission Product Barrier I Cs.
Changing a Fission Product Barrier from a Loss to a Potential Loss or vice-versa.
Not using NEI 99-01 definitions as the intent is for all NEI 99-01 users to have a standard set of defined terms as defined in NEI 99-01. Differences due to plant types are permissible (BWR or PWR). Verbatim compliance to the wording in NEI 99-01 is not necessary as long as the intent of the defined word is maintained.

Use of the 4 of 127 wording provided in NEI 99-01 is encouraged since the intent is for all users to have a standard set of defined terms as defined in NEI 99-01.

Any change to the IC and/or EAL, and/or basis wording as stated in NEI 99-01 that does alter the intent of the IC and/or EAL, i.e., the IC and/or EAL: o Does not classify at the classification level consistent with NEI 99-01. o Is not logically integrated with other EALs in the EAL scheme. o Results in an incomplete EAL scheme (i.e., does not classify all potential emergency conditions).

The "Difference/Deviation Justification" columns in the remaining sections of this document identify each difference between the NEI 99-01 IC/EAL wording and the RBS IC/EAL wording. An explanation that justifies the reason for each difference is then provided. If the difference is determined to be a deviation, a statement is made to that affect and explanation is given that states why classification may'be different from the NEI 99-01 IC/EAL and the reason for its acceptability. In all cases, however, the differences and deviations do not decrease the effectiveness of the intent of NEI 99-01. A summary list of RBS EAL deviations from NEI 99-01 is given in Table 3. RBS EAL Comparison Matrix Table 1 -RBS EAL Categories/Subcategories Category 'Group: Any Operating Mode: A -Abnormal Rad Levels/Rad Effluent H -Hazards and Other Conditions Affecting Plant Safety E-ISFSI Group: Hot Conditions: S -System Malfunction F -Fission Product Barrier Group: Cold Conditions: C -Cold Shutdown/Refueling System Malfunction RBS EALs I Subcategory 1 -Radiological Effluent 2 -Irradiated Fuel Event 3 -Area Radiation Levels 1 -Security 2 -Seismic Event 3 -Natural or Technological Hazard 4-Fire 5 -Hazardous Gas 6 -Control Room Evacuation 7 -Emergency Director Judgment 1 -Confinement Boundary 1 -Loss of Emergency AC Power 2 -Loss of Vital DC Power 3 -Loss of Control Room Indications 4 -RCS Activity 5 -RCS Leakage 6 -RPS Failure 7 -Loss of Communications 8 -Hazardous Event Affecting Safety Systems None 1 -RPV Level 2 -Loss of Emergency AC Power 3 -RCS Temperature 4 -Loss of Vital DC Power 5 -Loss of Communications 6 -Hazardous Event Affecting Safety Systems 5 of 127 NEI Recognition Category Abnormal Rad Levels/Radiological Effluent ICs/EALs Hazards and Other Conditions Affecting Plant Safety ICs/EALs ISFSI ICs/EALs System Malfunction ICs/EALs Fission Product Barrier ICs/EALs Cold Shutdown./ Refueling System Malfunction ICs/EALs ,. RBS EAL Comparison Matrix Table 2 -NEI / RBS EAL Identification Cross-Reference NEI RBS IC Example ( EAL Category and Subcategory EAL AU1 1 A -Abnormal Rad Levels/ Rad Effluent, 1 -Radiological Effluent AU1.1 -AU1 2 A -Abnormal Rad Levels/ Rad Effluent, 1 -Radiological Effluent AU1.1 AU1 3 A -Abnormal Rad Levels/ Rad Effluent, 1 -Radiological Effluent AU1.2 AU2 1 A -Abnormal Rad Levels/ Rad Effluent, 2 -Irradiated Fuel Event AU2.1 ' AA1 1 A -Abnormal Rad Levels/ Rad Effluent, 1 -Radiological Effluent AA1.1 ' AA1 2 A -Abnormal Rad Levels/ Rad Effluent, 1 -Radiological Effluent AA1.2 AA1 3 A -Abnormal Rad Levels/ Rad Effluent, 1 -Radiological Effluent AA1.3 AA1 4 A -Abnormal Rad Levels/ Rad Effluent, 1 -Radiological Effluent AA1.4 AA2 1 A -Abnormal Rad Levels/ Rad Effluent, 2 -Irradiated Fuel Event AA2.1 AA2 2 A -Abnormal Rad Levels/ Rad Effluent, 2 -Irradiated Fuel Event AA2.2 AA2 3 A -Abnormal Rad Levels/ Rad Effluent, 2 -Irradiated Fuel Event AA2.3 AA3 1 A -Abnormal Rad Levels / Rad Effluent, 3 -Area Radiation Levels AA3.1 AA3 2 A -Abnormal Rad Levels/ Rad Effluent, 3 -Area Radiation Levels *, AA3.2 AS1 1 A -Abnormal Rad Levels / Rad Effluent, 1 -Radiological Effluent AS1.1 AS1 2 A-Abnormal Rad Levels/ Rad Effluent, 1 -Radiological Effluent AS1.2 AS1 3 A -Abnormal Rad Levels/ Rad Effluent, 1 -Radiological Effluent AS1.3 6 of 127 RBS EAL Comparison Matrix NEI RBS IC Example Category and Subcategory EAL EAL AS2 1 A -Abnormal Rad Levels/ Rad Effluent, 2 -Irradiated Fuel Event AS2.1 AG1 1 A -Abnormal Rad Levels/ Rad Effluent, 1 -Radiological Effluent AG1.1 AG1 2 A -Abnormal Rad Levels/ Rad Effluent, 1 -Radiological Effluent AG1.2 AG1 3 A -Abnormal Rad Levels/ Rad Effluent, 1 -Radiological Effluent AG1.3 AG2 1 A -Abnormal Rad Levels/ Rad Effluent, 2 -Irradiated Fuel Event AG2.1 *<.:"",,! CU1 1 C -Cold SD/ Refueling System Malfunction, 1 -RPV Level CU1.1 CU1 2 C -Cold SD/ Refueling System Malfunction, 1 -RPV Level CU1.2 CU2 1 C -Cold SD/ Refueling System Malfunction, 2 -Loss of Emergency AC Power CU2.1 CU3 1 C -Cold SD/ Refueling System Malfunction, 3 -RCS Temperature CU3.1 CU3 2 C -Cold SD/ Refueling System Malfunction, 3 -RCS Temperature CU3.2 CU4 1 C -Cold SD/ Refueling System Malfunction, 4 -Loss of Vital DC Power CU4.1 CU5 1, 2, 3 C -Cold SD/ Refueling System Malfunction, 5 -Loss of Communications CU5.1 CA1 1 C -Cold SD/ Refueling System Malfunction, 1 -RPV Level CA1.1 CA1 2 C -Cold SD/ Refueling System Malfunction, 1 -RPV Level CA1.2 CA2 1 C -Cold SD/ Refueling System Malfunction, 1 -Loss of Emergency AC Power CA2.1 CA3 1, 2 C -Cold SD/ Refueling System Malfunction, 3 -RCS Temperature CA3.1 CA6 1 C -Cold SD/ Refueling System Malfunction, 6 -Hazardous Event Affecting Safety Systems CA6.1 CS1 1 C -Cold SD/ Refueling System Malfunction, 1 -RPV Level CS1.1 7 of 127 RBS EAL Comparison Matrix NEI RBS IC Example Category and Subcategory EAL EAL CS1 2 C -Cold SD/ Refueling System Malfunction, 1 -RPV Level CS1.2 CS1 3 C -Cold SD/ Refueling System Malfunction, 1 -RPV Level CS1.3 CG1 1 C -Cold SD/ Refueling System Malfunction, 1 -RPV Level CG1.1 CG1 2 C -Cold SD/ Refueling System Malfunction, 1 -RPV Level CG1.2 I E-HU1 1 E -ISFSI EU1.1 FA1 1 F -Fission Product Barrier Degradation FA1.1 FS1 1 F -Fission Product Barrier Degradation FS1.1 FG1 1 F -Fission Product Barrier Degradation FG1.1 HU1 1, 2, 3 H -Hazards and Other Conditions Affecting Plant Safety, 1 -Security HU1.1 HU2 1 H -Hazards and Other Conditions Affecting Plant Safety, 2 -Seismic Event HU2.*1 HU3 1 H -Hazards and Other Conditions Affecting Plant Safety, 3 -Natural or Technology Hazard HU3.1 HU3 2 H -Hazards and Other Conditions Affecting Plant Safety, 3 :__ Natural or Technology Hazard HU3.2 HU3 3 H -Hazards and Other Conditions Affecting Plant Safety, 3 -Natural or Technology Hazard HU3.3 HU3 4 H -Hazards and Other Conditions Affecting Plant Safety, 3 -Natural or Technology Hazard HU3.4 HU3 5 H -Hazards and Other Conditions Affecting Plant Safety, 3 -Natural or Technology Hazard N/A HU4 1 H -Hazards and Other Conditions Affecting Plant Safety, 4 -Fire or Explosion HU4.1 HU4 2 H -Hazards and Other Conditions Affecting Plant Safety, 4 -Fire or Explosion HU4.2 HU4 3 H -Hazards and Other Conditions Affecting Plant Safety, 4 -Fire or Explosion HU4.3 8 of 127 RBS EAL Comparison Matrix NEI RBS IC Example Category and Subcategory EAL EAL HU4 4 H -Hazards and Other Conditions Affecting Plant Safety, 4 -Fire or Explosion HU4.4 HU7 1 H -Hazards and Other Conditions Affecting Plant Safety, 7 -Judgment HU7.1 HA1 1, 2 H -Hazards and Other Conditions Affecting Plant Safety, 1 -Security HA1.1 HAS 1 H -Hazards and Other Conditions Affecting Plant Safety, S -Hazardous Gases HAS.1 HA6 1 H -Hazards and Other Conditions Affecting Plant Safety, 6 -Control Room Evacuation HA6.1 HA7 1 H -Hazards and Other Conditions Affecting Plant Safety, 7 -Judgment HA7.1 HS1 1 H -Hazards and Other Conditions Affecting Plant Safety, 1 -Security HS1.1 HS6 1 H -Hazards and Other Conditions Affecting Plant Safety, 6 -Control Room Evacuation HS6.1 HS7 1 H -Hazards and Other Conditions Affecting Plant Safety, 7 -Judgment HS7.1 HG1 1 N/A N/A HG7 1 H -Hazards and Other Conditions Affecting Plant Safety, 7 -Judgment HG7.1 SU1 1 S -System Malfunction, 1 -Loss of Emergency AC Power SU1.1 SU2 1 S -System Malfunction, 3 -Loss of Control Room Indications SU3.1 SU3 1 S -System Malfunction, 4 -RCS Activity SU4.1 SU3 2 S -System Malfunction, 4 -RCS Activity SU4.2 SU4 1, 2, 3 S -System Malfunction, S -RCS Leakage SUS.1 SUS 1 S -System Malfunction, 6 -RPS Failure SU6.1 SUS 2 S -System Malfuncti~n, 6 -RPS Failure SU6.2 9 of 127 RBS EAL Comparison Matrix NEI RBS IC Example Category and Subcategory EAL EAL SU6 1, 2, 3 S -System Malfunction, 7 -Loss of Communications SU7.1 SU? 1, 2 S -System Malfunction, 8 -Containment Failure N/A SA1 1 S -System Malfunction, 1 -Loss of Emergency AC Power SA1.1 SA2 1 S -System Malfunction, 3 -Loss of Control Room Indications SA3.1 SAS 1 S -System Malfunction, 6 -RPS Failure SA6.1 SA9 1 S -Hazardous Event Affecting Safety Systems SA8.1 SS1 1 S -System Malfunction, 1 -Loss of Emergency AC Power SS1.1 SSS 1 S -System Malfunction, 6 -RPS Failure SS6.1 SS8 1 S -System Malfunction, 2 -Loss of Vital DC Power SS2.1 SG1 1 S -System Malfunction, 1 -Loss of Emergency AC Power SG1.1 SGS 1 S -System Malfunction, 1 -Loss of Emergency AC Power SG1.2 10of127 RBS EAL Comparison Matrix Table 3 -Summary of Deviations NEI RBS EAL Description IC Example EAL FA1, FS1 RCS 3.A RCB4 River Bend uses Technical Specification based isolation temperatures vice Max Potential Loss Normal Operating temperatures for NEI RCS barrier potential loss 3.A.1. A valid indication of area temperature(s) greater than or equal to the system MOV Technical Requirements Manual (TRM) isolation value resulting from a primary system discharging into the Auxiliary Building is indicative of conditions in which significant RCS inventory is being lost. This is therefore considered to be a potential loss of the RCS barrier. The Maximum Normal Operating area radiation values are still being used by River Bend and are consistent with the NEI guidance. The alarms for the high area ambient temperature are associated with the TRM 3.3.6.1 values for primary containment isolation. The use of the TRM value provides a readily identifiable condition of the RCS barrier status with an alarm in the Control Room and an isolation signal that (if the condition persists) indicates unisolable leakage as defined by the EAL threshold. The Technical Specification 3.3.6.1 ambient temperature monitoring Allowable Value setpoints are set low enough to detect a leak equivalent to 25 gpm. Primary Containment Operating Values -Temperatures Parameter Max Normal TRM TS Allowable Max Safe Isolation Value " Value Main Steam Line 144°F 173°F 183°F 200°F Tunnel RHR Equip Area A 110°F 117°F 121:1°F 200°F RHR Equip Area B 110°F 117°F 186.4°F 200°F RCIC Equip Area 144°F 182°F 186.4°F 200°F RWCU Pump 145°F 165°F 169.5°F 200°F Room 11 of 127 RBS EAL Comparison Matrix NEI RBS EAL Description IC Example EAL Tecbnical Specification 3.3.6.1 Bases -Isolation References Area Allowable Value Section Main Steam Line 183°F 1.e Isolation -RHR System Isolation 121.1°F 5.a. RCIC Isolation 186.4°F 3.e. RWCU Isolation 169.5°F 4.d. The use of the isolation alarm and EOP condition provides a method of rapid identification of the EAL threshold without the task of performing additional confirmatory actions. The use of isolation temperatures still allows for a discernable margin prior to reaching the Max Safe Operating temperature . described in NEI Containment barrier loss threshold 3.C. Therefore this is an acceptable deviation from the generic NEI 99-01 Revision 6 guidance. HG1 1 NIA IC HG1 and associated example EAL is not implemented in the RBS scheme. There are several other ICs that are redundant with this IC, and are better suited to ensure timely and effective emergency declarations. In addition, the development of new spent fuel pool level EALs, as a result of NRC Order EA-12-051, clarified the intended emergency classification level for spent fuel pool level events. This deviation is justified because: 1. Hostile Action in the Protected Area is bounded by ICs HS1 and HS?. Hostile Action resulting in a loss of physical control is bound by EAL HG?, as well as any event that may lead to radiological releases to the public in excess of Environmental Protection Agency (EPA) Protective Action Guides (PAGs). a. If, for whatever reason, the Control Room must be evacuated, and control of safety functions (e.g., reactivity control, core cooling, and RCS heat removal) cannot be reestablished, then IC HS6 would apply, as well as IC HS7 if desired by the EAL decision-maker.

b. Also, as stated above, any event (including Hostile Action) that could reasonably be expected to have a release exceeding EPA PAGs would be 12 of 127 RBS EAL Comparison Matrix NEI RBS EAL Description IC Example EAL bound by IC HG?. C. From a Hostile Action perspective, I Cs HS1, HS? and HG? are appropriate, and therefore, make this part of HG1 redundant and unnecessary.

d. From a loss of physical control perspective, ICs HS6, HS? and HG? are appropriate, and therefore, make this part of HG1 redundant and unnecessary.

2. Any event which causes a loss of spent fuel pool level will be bounded by I Cs AA2, AS2 and AG2, regardless of whether it was based upon a Hostile Action or not, thus making this part of HG1 redundant and unnecessary.

a. An event that leads to a radiological release will be bounded by I Cs AU1, AA1, AS1 and AG1. Events that lead to radiological releases in excess of EPA PAGs will be bounded by EALs AG1 and HG?, thus making this part of HG1 redundant and unnecessary.

I Cs AA2, AS2, AG2, AS1, AG1, HS1, HS6, HS? and HG? have been implemented consistent with NEI 99-01 Revision 6 and thus HG1 is adequately bounded as described above. Therefore this is an acceptable deviation from the generic NEI 99-01 Revision 6 guidance and is consistent with NRC-approved EP FAQ 2015-013. HS6 1 HS6.1 Deleted defueled mode applicability. Control of the cited safety functions are not critical for a defueled reactor as there is no energy source in the reactor vessel or RCS. The Mode applicability for the reactivity control safety function has been limited to Modes 1, 2, and 3 (hot operating conditions). In the cold operating modes adequate shutdown margin exists under all conditions. Therefore this is an acceptable deviation from the generic NEI 99-01 Revision 6 guidance and is consistent with NRC-approved EP FAQ 2015-014. 13 of 127

RBS EAL Comparison Matrix NEI RBS EAL Description IC Example EAL CA6 1 CA6.1 The proposed RBS CA6.1 and SA8.1 wording is intended to ensure that an Alert SA9 1 SA8.1 should be declared only when actual or potential performance issues with SAFETY SYSTEMS have occurred as a result of a hazardous event. The occurrence of a hazardous event will result in an Unusual Event classification at a minimum. In order to warrant escalation to the Alert classification, the hazardous event should cause indications of degraded performance to one train of a SAFETY SYSTEM with either indications of degraded performance on the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second SAFETY SYSTEM train, such that the operability or reliability of the second train is a concern. In addition, escalation to the Alert classification should not occur if the damage from the hazardous event is limited to a SAFETY SYSTEM that was inoperable, or out of service, prior to the event occurring. As such, the proposed EALs will reduce the potential of declaring an Alert when events are in progress that do not involve an actual or potential substantial degradation of the level of safety of the plant, i.e., does not cause significant concern with shutting down or cooling down the plant. EALs CA6.1 and SA8.1 do not directly escalate to a Site Area Emergency or a General Emergency due to a hazardous event. The Fission Product Barrier and/or Abnormal Radiation Levels/Radiological Effluent recognition categories would provide an escalation path to a Site Area Emergency or a General Emergency. The EALs and the Basis sections have been revised to ensure potential escalations from an Unusual Event 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. The definition for VISIBLE DAMAGE has been revised to reflect the fact that the EALs are based upon SAFETY SYSTEM trains rather than individual components or structures. Note 9 has been added to CA6.1 and SA8.1 as it meets the intent of the EALs, is consistent with other EALs (e.g., EAL HA5.1 which was previously endorsed by the NRC), and ensures that declared emergencies are based upon unplanned 14of127 i~~-RBS EAL Comparison Matrix NEI RBS EAL Description IC Example EAL events with the potential to pose a radiological risk to the public. Note 10 has been added to CA6.1 and SA8.1 to help reinforce and succinctly capture the more detailed information from the revised basis section related to when conditions would require the declaration of an Alert. CA6.1 and SA8.1 are consistent with approved NRG FAQ 2016-002 addressing degraded performance or visible damage to more than one safety system train caused by the specified events. Based on the above information, this revised wording is an acceptable deviation from the generic NEI 99-01 Revision 6 guidance and is consistent with NRC approved EP FAQ 2016-002. 15 of 127 RBS EAL Comparison Matrix Category A Abnormal Rad Levels / Radiological Effluent 16of127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording and Mode RBS RBS IC Wording and Mode Difference/Deviation Justification Applicability -IC#(s) Applicability AU1 Release of gaseous or liquid AU1 Release of gaseous or liquid The RBS ODCM is the site-specific effluent release radioactivity greater than 2 times radioactivity greater than 2 times the controlling document. the (site-specific effluent release ODCM limits for 60 minutes or longer controlling document) limits for MODE: All 60 minutes or longer. MODE: All NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation JustiJication EAL# EAL# 1 Reading on ANY effluent Reading on any Table A-1 effluent Example EALs #1 and #2 have been combined into a single radiation monitor greater than 2 radiation monitor> column "UE" for EAL to simplify presentation. times the (site-specific effluent ;:;: 60 min. (Notes 1, 2, 3) The NEI phrase " ... effluent radiation monitor greater than 2 release controlling document) times the (site-specific effluent release controlling limits for 60 minutes or longer: document)" and "effluent radiation monitor greater than 2 (site-specific monitor list and times the alarm setpoint established by a current radioactivity threshold values corresponding discharge permit" have been replaced with" ... any Table A-1 to 2 times the controlling AU1.1 effluent radiation monitor> column "UE". document limits) UE thresholds for all RBS continuously monitored gaseous 2 Reading on ANY effluent and liquid release pathways are listed in Table A-1 to radiation monitor greater than 2 consolidate the information in a single location and, thereby, times the alarm setpoint simplify identification of the thresholds by the EAL user. The established by a current values shown in Table A-1 column "UE", consistent with the radioactivity discharge permit for NEI bases, represent two times the ODCM release limits for 60 minutes or longer. gaseous and liquid refeases. 3 Sample analysis for a gaseous or AU1.2 Sample analysis for a gaseous or The RBS ODCM is the site-specific effluent release liquid release indicates a liquid release indicates a concentration controlling document. concentration or release rate or release rate > 2 x ODCM limits for;:;: greater than 2 times the (site-60 min. (Notes 1, 2) specific effluent release controlling document) limits for 17 of 127 RBS EAL Comparison Matrix NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 60 minutes or longer. Notes

The Emergency Director N/A Note 1: The Emergency Director The classification timeliness note has been standardized should declare the Unusual should declare the event across the RBS EAL scheme by referencing the "time limit" Event promptly upon promptly upon determining specified within the EAL wording. determining that 60 minutes that the time limit has been Added "The Emergency Director is not allowed an additional has been exceeded, or will exceeded, or will likely be 15 minutes to declare after the time limit is exceeded" to likely be exceeded.

exceeded. The Emergency reinforce the concept that the EAL timing component runs

If an ongoing release is Director is not allowed an concurrent with the classification timeliness clock. additional 15 minutes to detected and the release declare after the time limit is start time is unknown, exceeded.

assume that the release duration has exceeded 60 Note 2: If an ongoing release is The classification timeliness note has been standardized minutes. detected and the release across the RBS EAL scheme by referencing the "time limit"

If the effluent flow past an start time is unknown, specified within the EAL wording. assume that the release effluent monitor is known to duration has exceeded the have stopped due to actions specified time limit. to isolate the release path, then the effluent monitor Note 3: If the effluent flow past an None reading is no longer valid for effluent monitor is known to classification purposes.

have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer VALID for classification purposes. 18 of 127 RBS EAL Comparison Matrix Table A-1 Effluent Monitor Classification Thresholds I Release Point I Monitor I GE I SAE I Alert I UE I Main Plant Vent -Primary RE125 9.56E+08 µCi/sec 9.56E+07 µCi/sec 9.63E+06 µCi/sec 1.01 E+05 µCi/sec Main Plant Vent -RE126 1.66E-01 µCi/ml 1.74E-03 µCi/ml Secondary

U) Fuel Bldg Vent -Primary RE5A 7.75E+08 µCi/sec 7.75E+07 µCi/sec 7.75E+06 µCi/sec 6.50E+03 µCi/sec , ::i 0 Q) U) Fuel Bldg Vent -Secondary RE5B 1. 72E-01 µCi/ml 1.38E-03 µCi/ml ra --------c., Radwaste Bldg Vent -REGA 8.03E+08 µCi/sec 8.03E+07 µCi/sec 8.03E+06 µCi/sec 6.96E+04 µCi/sec Primary Radwaste Bldg Vent -RE6B --------2.12E-01 µCi/ml 1.71 E-04 µCi/ml Secondary

E --::i Liquid Radwaste RE107 CT ------------2 x Alarm Setpoint :J 19 of 127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording and Mode RBS RBS IC Wording and Mode Difference/Deviation Justification Applicability IC#(s) Applicability AU2 UNPLANNED loss of water level AU2 UNPLANNED loss of water level None above irradiated fuel. above irradiated fuel MODE: All MODE:AII NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 a. UNPLANNED water level AU2.1 UNPLANNED water level drop in the Site-specific level indications incorporated.

drop in the REFUELING REFUELING PATHWAY as indicated PATHWAY as indicated by by level instrumentation, low water Site-specific area radiation monitors incorporated. ANY of the follpwing: level alarm or visual observation (site-specific level AND indications). UNPLANNED rise in corresponding AND area radiation levels as indicated by b. UNPLANNED rise in area any of the following radiation monitors: radiation levels as indicated RMS-RE140 Refueling Floor by ANY of the following

radiation monitors.

Near North Entrance (site-specific list of area

RMS-RE141 Refueling Floor radiation monitors)

Near South Entrance

RMS-RE192 Fuel Building Operating Floor -South
RMS-RE193 Fuel Building Operating Floor -North 20 of 127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) AA1 Release of gaseous or liquid AA1 Release of gaseous or liquid None radioactivity resulting in offsite radioactivity resulting in offsite dose dose greater than 10 mrem TEDE greater than 10 mrem TEDE or or 50 mrem thyroid COE. 50 mrem thyroid COE MODE: All MODE: All NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Reading on ANY of the following AA1.1 Reading on any Table A-1 effluent The RBS radiation monitors that detect radioactivity effluent radiation monitors greater than radiation monitor> column "ALERT" release to the environment are listed in Table A-1. UE, Alert, the reading shown for 15 minutes for 15 min. (Notes 1, 2, 3, 4) SAE and GE thresholds for all RBS continuously monitored or longer: gaseous and liquid release pathways are listed in Table A-1 to (site-specific monitor list and consolidate the information in a single location and, thereby, threshold values) simplify identification of the thresholds by the EAL-user.

2 Dose assessment using actual AA1.2 Dose assessment using actual The site boundary is the site-specific receptor point. meteorology indicates doses meteorology indicates doses > 10 greater than 10 mrem TEDE or mrem TEDE or 50 mrem thyroid COE 50 mrem'thyroid COE at or at or beyond the SITE BOUNDARY beyond (site-specific dose (Note 4) receptor point). 3 Analysis of a liquid effluent AA1.3 Analysis of a liquid effluent sample The site boundary is the site-specific receptor point. sample indicates a concentration indicates a concentration or"release or release rate that would result rate that would result in doses > 10 in doses greater than 10 mrem mrem TEDE or 50 mrem thyroid COE TEDE or 50 mrem thyroid COE at at or beyond the SITE BOUNDARY for or beyond ( site-specific dose 60 min. of exposure (Notes 1, 2) receptor point) for one hour of exposure. 21 of 127 RBS EAL Comparison Matrix *-4 Field survey results indicate AA1A Field survey results indicate EITHER The site boundary is the site-specific receptor point. EITHER of the following at or beyond (site-specific dose of the following at or beyond the SITE receptor point): BOUNDARY:

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

inhalation. (Notes 1, 2) Notes

The Emergency Director N/A Note 1: The Emergency Director The classification timeliness note has been standardized should declare the Alert should declare the event across the RBS EAL scheme by referencing the "time limit" promptly upon determining promptly upon determining specified within the EAL wording. that the applicable time has that the time limit has been Added "The Emergency Director is not allowed an additional been exceeded, or will likely exceeded, or will likely be 15 minutes to declare after the time limit is exceeded" to be exceeded.

exceeded. The Emergency reinforce the concept that the EAL timing component runs

If an ongoing release is Director is not allowed an concurrent with the classification timeliness clock. additional 15 minutes to detected and the release declare after the time limit is start time is unknown, exceeded.

assume that the release duration has exceeded 15 Note 2: If an ongoing release is The classification timeliness note has been standardized minutes. detected and the release across the RBS EAL scheme by referencing the "time limit"

If the effluent flow past an start time is unknown, specified within the EAL wording. assume that the release effluent monitor is known to duration has exceeded the have stopped due to actions specified time limit. to isolate the release path, then the effluent monitor Note 3: If the effluent flow past an reading is no longer valid for effluent monitor is known to None classification purposes.

have stopped due to actions to isolate the release path,

The pre-calculated effluent then the effluent monitor reading is no longer valid for monitor values presented in classification purposes.

EAL #1 should be used for emergency classification Note 4 The pre-calculated effluent ,Incorporated site-specific EAL numbers associated with 22 of 127 RBS EAL Comparison Matrix assessments until the results monitor values presented in generic EAL#1. from a dose assessment EA Ls AA 1 .1, AS 1 . 1 and using actual mete0rology are AG1 .1 should be used for available. emergency classification (' assessments until the results from a dose assessment using actual meteorology are available. 23of127 --1 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) AA2 Significant lowering of water AA2 Significant lowering of water level None level above, or damage to, above, or damage to, irradiated fuel irradiated fuel. MODE: All MODE: All NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Uncovery of irradiated fuel in the AA2.1 IMMINENT uncovery of irradiated fuel Added the defined term "IMMINENT." Determination of REFUELING PATHWAY. in the REFUELING PATHWAY irradiated fuel uncovery in the refueling pathway will always be an anticipatory determination as no direct indication is available to determine when the irradiated fuel has become uncovered. 2 Damage to irradiated fuel AA2.2 Damage to irradiated fuel resulting in a Site-specific list of radiation monitors are incorporated in Table resulting in a release of release of radioactivity A-2. radioactivity from the fuel as indicated by ANY of the AND Radiation monitor high alarms are specified. following radiation monitors: High alarm on any Table A-2 radiation (site-specific listing of radiation monitor monitors, and the associated -readings, setpoints and/or alarms) 3 Lowering of spent fuel pool level AA2.3 Lowering of spent fuel pool level to Post-Fukushima order EA-12-051 required the installation of to (site-specific Level 2 value). 108.0 ft. (Level 2) on SFC-Ll29NB [See Developer Notes] reliable SFP level indication (SFC-Ll29NB) capable of identifying normal level (Level 1 ), SFP level approximately 23 ft. above the top of the fuel racks, (Level 2) 107 ft. 10 5/16 in. (rounded to 108.0 ft. for readability) which is that level adequate to provide substantial radiation shielding for a person standing on the SFP operating deck, and SFP level at the top of the fuel racks (Level 3) 85 ft. 10 5/16 in. (rounded to 86.0 ft. for readability). RBS uses a Level 3 of approximately one foot above the highest point of any fuel rack providing 24 of 127 RBS EAL Comparison Matrix I added margin. Table A-2 Fuel Damage Radiation Monitors

RMS-RE140 Refueling Floor Near North Entrance
RMS-RE141 Refueling Floor Near South Entrance
RMS-RE192 Fuel Building Operating Floor -South
RMS-RE193 Fuel Building Operating Floor -North
RMS-RESA(B)

Fuel Building Ventilation Exhaust 25of127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) AA3 Radiation levels that impede AA3 Radiation levels that IMPEDE access EAL AA3.2 mode applicability has been limited to the access to equipment necessary to equipment necessary for normal applicable mode of Table A-3. for normal plant operations, plant operations, cooldown or cooldown or shutdown shutdown MODE: All MODE: All (AA3.2 Mode 3 only) NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Dose rate greater than 15 mR/hr AA3.1 Dose rate> 15 mR/hr in EITHER of No other site-specific areas requiring continuous occupancy in ANY of the following areas:

Control Room the following areas: exist at RBS.
Central Alarm Station
Control Room (RMS-RE170)

The Control Room is monitored for excessive radiation by * ( other site-specific RMS-RE-170. There are no permanently installed area areas/rooms)

Central Alarm Station (by radiation monitors in CAS that may be used to assess this survey) EAL threshold.

Therefore, this threshold is evaluated using local radiation suNey for this area. 2 An UNPLANNED event results AA3.2 An UNPLANNED event results in The site-specific list of plant rooms or areas with entry-related in radiation levels that prohibit or radiation levels that prohibit or mode applicability are tabularized in Table A-3. impede access to any of the IMPEDE access to any Table A-3 The bulleted bases item "the action for which room/area entry following plant rooms or areas: room or area (Note 5) is required is of an admi,nistrative or record keeping nature (site-specific list of plant rooms (e.g., normal rounds or routine inspections)" was removed or areas with entry-related mode from the list of exceptions to classification in the basis applicability identified) information. These actions are a consideration when the site-specific list was developed. Rooms requiring entry for these types of actions are already excluded from the list when it was -developed. Note If the equipment in the listed N/A Note 5 If the equipment in the listed None room or area was already room or area was already inoperable or out-of-seNice inoperable or out-of-seNice before the event occurred, then before the event occurred, 26of127 RBS EAL Comparison Matrix no emergency classification is then no emergency warranted. classification is warranted. TableA-3 Safe Operation & Shutdown Rooms/Areas Room/Area Mode Auxiliary Building 70' RHR B Pump Room 3 Auxiliary Building 80' RHR A Pump Room 3 Auxiliary Building 114' West 3 Control Building 95' Div 1 RSS Room 3 ' 27of127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) AS1 Release of gaseous radioactivity AS1 Release of gaseous radioactivity None resulting in offsite dose greater resulting in offsite dose greater than than 100 mrem TEDE or 500 100 mrem TEDE or 500 mrem thyroid mrem thyroid COE COE MODE: All MODE:AII NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Reading on ANY of the following AS1.1 Reading on any Table A-1 effluent The RBS radiation monitors that detect radioactivity effluent radiation monitors greater than radiation monitor> column "SAE" for ;:: release to the environment are listed in Table A-1. UE, Alert, the reading shown for 15 15 min. SAE and GE thresholds for all RBS continuously monitored minutes or longer: (Notes 1, 2, 3, 4) gaseous and liquid release pathways are listed in Table A-1 to (site-specific monitor list and consolidate the information in a single location and, thereby, threshold values) simplify identification of the thresholds by the EAL-user. -2 Dose assessment using actual AS1.2 Dose assessment using actual The site boundary is the site-specific receptor point. meteorology indicates doses meteorology indicates doses > 100 greater than 100 mrem TEDE or mrem TEDE or 500 mrem thyroid COE 500 mrem thyroid COE at or at or beyond the SITE BOUNDARY beyond (site-specific dose (Note 4) receptor point) 3 Field survey results indicate AS1.3 Field survey results indicate EITHER The site boundary is the site-specific receptor point. EITHER of the following at or of the following at or beyond the SITE beyond (site-specific dose BOUNDARY: receptor point):

Closed window dose rates >
Closed window dose rates 100 mR/hr expected to continue greater than 100 mR/hr for;:: 60 min. expected to continue for 60
Analyses of field survey minutes or longer. samples indicate thyroid COE >
Analyses of field survey 500 mrem for 60 min. of samples indicate thyroid inhalation.

COE greater than 500 (Notes 1, 2) 28 of 127 RBS EAL Comparison Matrix mrem for one hour of inhalation. Notes

The Emergency Director Note 1: The Emergency Director The classification timeliness note has been standardized should declare the Site Area should declare the event across the RBS EAL scheme by referencing the "time limit" Emergency promptly upon promptly upon determining specified within the EAL wording. determining that the that the time limit has been Added "The Emergency Director is not allowed an additional applicable time has been exceeded, or will likely be 15 minutes to declare after the time limit is exceeded" to exceeded, or will likely be exceeded.

The Emergency reinforce the concept that the EAL timing component runs exceeded. Director is not allowed an concurrent with the classification timeliness clock.

If an ongoing release is additional 15 minutes to detected and the release start declare after the time limit is time is unknown, assume that exceeded.

the release duration has Note 2: If an ongoing release is The classification timeliness note has been standardized exceeded 15 minutes.

If the effluent flow past an detected and the release across the RBS EAL scheme by referencing the "time limit" effluent monitor is known to start time is unknown, specified within the EAL wording. have stopped due to actions assume that the release duration has exceeded the to isolate the release path, specified time limit. then the effluent monitor -reading is no longer valid for Note 3: If the effluent flow past an classification purposes.

effluent monitor is known to None

The pre-calculated effluent have stopped due to actions monitor values presented in to isolate the release path, EAL #1 should be used for then the effluent monitor emergency classification reading is no longer VALID assessments until the results for classification purposes.

from a dose assessment Note 4 The pre-calculated effluent using actual meteorology are available. monitor values presented in Incorporated site-specific EAL numbers associated with EALs AA 1 .1 , AS 1 . 1 and generic EAL#1 . AG 1 .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. 29of127 RBS EAL Comparison Matrix ( NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) AS2 Spent fuel pool level at (site-AS2 Spent fuel pool level at the top of the Top of the fuel racks is the site-specific Level 3 description. specific Level 3 description) ' fuel racks MODE: All NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Lowering of spent fuel pool level AS2.1 Lowering of spent fuel pool level to For RBS, Level 3, which corresponds to approximately one to (site-specific Level 3 value) 86.0 ft. (Level 3) on SFC-Ll29A/B foot above the highest point of any fuel rack, providing added margin, is an indicated level of: 85 ft. 10 5/16 in. (rounded to 86.0 ft. for readability) on the specified instrument. 30of127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) AG1 Release of gaseous radioactivity AG1 Release of gaseous radioactivity None resulting in offsite dose greater resulting in offsite dose greater than 1,000 mrem TEDE or than 1,000 mrem TEDE or 5,000 5,000 mrem thyroid COE. mrem thyroid COE MODE: All MODE: All NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Reading on ANY of the following AG1.1 Reading on any Table A-1 effluent The RBS radiation monitors that detect radioactivity effluent release radiation monitors greater than radiation monitor> column "GE" to the environment are listed in Table A-1. the reading shown for 15 for 15 min. UE, Alert, SAE and GE thresholds for all RBS continuously minutes or longer: (Notes 1, 2, 3, 4) monitored gaseous and liquid release pathways are listed in Table (site-specific monitor list and A-1 to consolidate the information in a single location and, thereby, threshold values) simplify identification of the thresholds by the EAL-user. 2 Dose assessment using actual AG1.2 Dose assessment using actual The site boundary is the site-specific receptor point. meteorology indicates doses meteorology indicates doses > greater than 1,000 mrem TEDE 1000 mrem TEDE or or 5,000 mrem thyroid COE at 5000 mrem thyroid COE at or or beyond (site-specific dose beyond the SITE BOUNDARY receptor point). (Note 4) 3 Field survey results indicate AG1.3 Field survey results indicate The site boundary is the site-specific receptor point. EITHER of the following at or EITHER of the following at or beyond (site-specific dose beyond the SITE BOUNDARY: receptor point):

Closed window dose rates >
Closed window dose rates 1000 mR/hr expected to greater than 1,000 mR/hr continue for~ 60 min. expected to continue for 60
Analyses of field survey minutes or longer.
Analyses of field survey samples indicate thyroid COE samples indicate thyroid COE > 5000 mrem for 60 min. of greater than 5,000 mrem for inhalation.

31 of 127 RBS EAL Comparison Matrix one hour of inhalation. (Notes 1, 2) Notes

The Emergency Director Note 1: The Emergency Director The classification timeliness note has been standardized across the should declare the Site Area should declare the event RBS EAL scheme by referencing the "time limit" specified within the Emergency promptly upon promptly upon EAL wording. determining that the determining that the time Added "The Emergency Director is not allowed an additional 15 applicable time has been limit has been minutes to declare after the time limit is exceeded" to reinforce the exceeded, or will likely be exceeded, or will likely concept that the EAL timing component runs concurrent with the exceeded.

be exceeded. The Emergency Director is classification timeliness clock.

If an ongoing release is not allowed an additional detected and the release 15 minutes to declare start time is unknown, after the time limit is assume that the release exceeded.

duration has exceeded 15 minutes. Note 2: If an ongoing release is The classification timeliness note has been standardized across the

If the effluent flow past an detected and the release RBS EAL scheme by referencing the "time limit" specified within the effluent monitor is known to start time is unknown, EAL wording. have stopped due to actions assume that the release duration has exceeded to isolate the release path, the specified time limit. then the effluent monitor reading is no longer valid for Note 3: If the effluent flow past classification purposes.

an effluent monitor is

The pre-calculated effluent known to have stopped due to actions to isolate None monitor values presented in the release path, then EAL #1 should be used for the effluent monitor emergency classification reading is no longer assessments until the results VALID for classification from a dose assessment using actual meteorology are purposes.

available. Note 4 The pre-calculated effluent monitor values r Incorporated site-specific EAL numbers associated with generic presented in EALs EAL#1. AA 1 . 1 , AS 1 . 1 and AG1 .1 should be used for emergency classification assessments until the results from a dose assessment using actual 32 of 127 RBS EAL Comparison Matrix meteorology are available. 33 of 127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) --AG2 Spent fuel pool level cannot be AG2 Spent fuel pool level cannot be Top of the fuel racks is the site-specific Level 3 description. restored to at least (site-specific restored to at least the top of the fuel Level 3 description) for 60 racks for 60 minutes or longer minutes or longer MODE: All MODE: All NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Spent fuel pool level cannot be AG2.1 Spent fuel pool level cannot be For RBS, Level 3, which corresponds to approximately one restored to at least (site-specific restored to at least 86.0 ft. (Level 3) foot above the highest point of any fuel rack, providing added Level 3 value) for 60 minutes or on SFC-Ll29A/B for;;: 60 min. (Note 1) margin, is an indicated level of: 85 ft. 10 5/16 in. (rounded to longer 86.0 ft. for readability) on the specified instrument. Note The Emergency Director should N/A Note 1: The Emergency Director The classification timeliness note has been standardized declare the General Emergency should declare the event across the R!3S EAL scheme by referencing the "time limit" promptly upon determining that promptly upon determining specified within the EAL wording. 60 minutes has been exceeded, that the time limit has been Added "The Emergency Director is not allowed an additional or will likely be exceeded. exceeded, or will likely be 15 minutes to declare after the time limit is exceeded" to exceeded. The Emergency reinforce the concept that the EAL timing component runs Director is not allowed an additional 15 minutes to concurrent with the classification timeliness clock. declare after the time limit is exceeded. 34 of 127 RBS EAL Comparison Matrix Category C Cold Shutdown / Refueling System Malfunction 35of127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification ' IC#(s) CU1 UNPLANNED loss of (reactor CU1 UNPLANNED loss of RPV Deleted the words " ... for 15 minutes or longer" as the 15 minute vessel/RCS [PWR] or RPV inventory criteria only applies to EAL #1 [BWR]) inventory for 15 minutes MODE: 4 -Cold Shutdown, 5 -or longer. -Refueling MODE: Cold Shutdown, Refueling NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 UNPLANNED loss of reactor CU1.1 UNPLANNED loss of primary None coolant results in (reactor coolant results in RPV water vessel/RCS [PWR] or RPV level less than a required lower [BWR]) level less than a limit for e=: 15 min. (Note 1) required lower limit for 15 minutes or longer. 2 a. (Reactor vessel/RCS [PWR] CU1.2 RPV water level cannot be Added the words " ... due to loss of RPV inventory" to be consistent or RPV [BWR]) level cannot monitored with the IC wording. be monitored. AND EITHER Replaced the term "increase" with the word "rise" consistent with AND

UNPLANNED rise in any allowed usage. Table C-1 sump or pool level Site-specific applicable sumps and tanks are listed in Table C-1 to b. UNPLANNED increase in due to a loss of RPV improve the readability of the EAL. The word Tank has been replaced (site-specific sump and/or inventory with Pool. There are no tanks applicable to RBS but have included tank) levels.
Visual observation of the Suppression Pool as a volume where RCS leakage may relocate.

UNISOLABLE RCS leakage Added bulleted criteria "Visual observation ... " to include direct observation of significant unisolable RCS leakage. Note The Emergency Director should N/A Note 1: The Emergency The classification timeliness note has been standardized across the declare the Unusual Event Director should RBS EAL scheme by referencing the "time limit" specified within the promptly upon determining that

declare the event EAL wording. 15 minutes has been exceeded, promptly upon Added "The Emergency Dir,ector is not allowed an additional 15 or will likely be exceeded.

determining that the time limit has been minutes to declare after the time limit is exceeded" to reinforce the exceeded, or will likely concept that the EAL timing component runs concurrent with the 36 of 127 RBS EAL Comparison Matrix be exceeded. The classification timeliness clock. Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. Table C-1 Sumps/Pool

Drywell equipmerit drain sump
Drywell floor drain sump
Pedestal floor drain sump
CTMT equipment drain sump
CTMT floor drain sump
Suppression Pool
RHR A, B, C, HPCS, LPCS, RCIC room sumps
Auxiliary Building floor drain sump 37of127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) CU2 Loss of all but one AC power CU2 Loss of all but one AC power "ENS buses" is the RBS-specific terminology for "emergency buses". source to emergency buses for source to ENS buses for 15 15 minutes or longer. minutes or longer. MODE: Cold Shutdown, MODE: 4 -Cold Shutdown, 5 -Refueling, Defueled Refueling, DEF -Defueled NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 a. AC power capability to (site-CU2.1 AC power capability, Table C-3, DIV I and DIV II 4.16 KV ENS buses are the site-specific emergency specific emergency buses) is to DIV I and DIV II 4.16 KV ENS buses. reduced to a single power buses reduced to a single power Site-specific AC power sources are tabularized in Table C-3. source for 15 minutes or source for~ 15 min. (Note 1) longer. AND AND Any additional single power b. Any additional single power source failure will result in loss of source failure will result in all AC power to SAFETY loss of all AC power to SYSTEMS SAFETY SYSTEMS. Note The Emergency Director should N/A Note 1: The Emergency The classification timeliness note has been standardized across the declare the Unusual Event Director should declare RBS EAL scheme by referencing the "time limit" specified within the promptly upon determining that the event promptly EAL wording. 15 minutes has been exceeded, upon determining that Added "The Emergency Director is not allowed an additional 15 or will likely be exceeded.

the time limit has been exceeded, or will likely minutes to declare after the time limit is exceeded." to reinforce the be exceeded. The concept that the EAL timing component runs concurrent with the Emergency Director is classification timeliness clock. not allowed an additional 15 minutes to declare after the time limit is exceeded. 38of127 RBS EAL Comparison Matrix Table C-3 AC Power Sources I-Offsite

1RTX-XSR1C
1 RTX-XSR1 D Onsite
EGS-EG1A
EGS-EG1B NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) CU3 UNPLANNED increase in RCS CU3 UNPLANNED rise in RCS Replaced the term "increase" with the word "rise" consistent with temperature temperature allowed usage. MODE: Cold Shutdown, MODE: 4 -Cold Shutdown, 5 -Refueling Refueling NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 UNPLANNED increase in RCS CU3.1 UNPLANNED rise in RCS Replaced the term "increase" with the word "rise" consistent with temperature to greater than (site-temperature to> 200°F due to allowed usage. -specific Technical Specification loss of decay heat removal 200°F is the site-specific Tech. Spec. cold shutdown temperature cold shutdown temperature limit) capability limit. Added "due to loss of decay heat removal capability" to reinforce the generic bases that states "EAL #1 involves a loss of decay heat removal capability." 2 Loss of ALL RCS temperature CU3.2 Loss of all RCS temperature and None and (reactor vessel/RCS

[PWR] RPV water level indication for or RPV [BWR]) level indication 15 min. (Note 1) for 15 minutes or longer. Note The Emergency Director should NIA Note 1: The Emergency Director The classification timeliness note has been standardized across the declare the Unusual Event should declare the event RBS EAL scheme by referencing the "time limit" specified within the 39 of 127 RBS EAL Comparison Matrix promptly upon determining that promptly upon EAL wording. 15 minutes has been exceeded, determining that the time Added "The Emergency Director is not allowed an additional 15 or will likely be exceeded limit has been exceeded, minutes to declare after the time limit is exceeded" to reinforce the or will likely be concept that the EAL timing component runs concurrent with the exceeded. The classification timeliness clock. Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. 40 of 127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) CU4 Loss of Vital DC power for 15 CU4 Loss of Vital DC power for 15 None minutes or longer. minutes or longer. MODE: Cold Shutdown, MODE 4 -Cold Shutdown, 5 -Refueling Refueling NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Indicated voltage is less than CU4.1 Indicated voltage is < 105 VDC on 105 VDC is the site-specific minimum vital DC bus voltage. (site-specific bus voltage value) required Safety Related DIV I and Safety-related DC bus operability requirements are specified in on required Vital DC buses for 15 DIV II 125 VDC buses for~ 15 Technical Specifications. minutes or longer. min. (Note 1) Note The Emergency Director should NIA Note 1: The Emergency Director The classification timeliness note has been standardized across the declare the Unusual Event should declare the event RBS EAL scheme by referencing the "time limit" specified within the promptly upon determining that promptly upon EAL wording. 15 minutes has been exceeded, determining that the time Added "The Emergency Director is not allowed an additional 15 or will likely be exceeded. limit has been exceeded, minutes to declare after the time limit is exceeded" to reinforce the or will likely be exceeded. concept that the EAL timing component runs concurrent with the The Emergency Director classification timeliness clock. is not allowed an additional 15 minutes to declare after the time limit is exceeded. 41 of 127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) CU5 Loss of all onsite or offsite CU5 Loss of all onsite or offsite None -communications capabilities. communications capabilities. MODE: Cold Shutdown, MODE: 4 -Cold Shutdown, 5 -Refueling, Defuele.d Refueling, DEF -Defueled NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Loss of ALL of the following CU5.1 Loss of all Table C-5 onsite Example EALs #1, 2 and 3 have been combined into a single communication methods onsite communication methods: OR EAL for simplification of presentation. (site specific list of Loss of all Table C-5 State and Replaced "ORO" with "State and local agency" for clarification. communications methods) local agency communication Table C-5 provides a site-specific list of onsite, State and Local methods 2 Loss of ALL of the following ORO OR (ORO) and NRC communications methods. communications methods: Loss of all Table C-5 NRC (site specific list of communication methods communications methods) 3 Loss of ALL of the following NRC communications methods: (site specific list of communications methods) 42of127 RBS EAL Comparison Matrix Table C-5 Communication Methods System Onsite State/L NRC ocal Plant radio system X Plant Paging System X -Sound powered phones X In-plant telephones X Emergency Notification System (ENS) X '* Commercial Telephone System X X Satellite Phones ')( X State of Louisiana Radio X '* State and Local Hotline radio X INFORM Notification System X 43 of 127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) CA1 Loss of (reactor vessel/RCS CA1 Significant loss of RPV Added the word "Significant. .. " to differentiate the Alert loss of RPV [PWR] or RPV [BWR]) inventory inventory inventory IC from the Unusual Event IC which is "Unplanned loss of MODE: Cold Shutdown, MODE: 4 -Cold Shutdown, 5 -RPV inventory." Refueling Refueling NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Loss of (reactor vessel/RCS CA1.1 Loss of RPV inventory as -43 in. is the Level 2 actuation setpoint for HPCS and RCIC. [PWR] or RPV [BWR]) inventory indicated by RPV water level < as indicated by level less than -43 in. (Level 2) (site-specific level). 2 a. (Reactor vessel/RCS [PWR] CA1.2 RPV water level cannot be Site-specific applicable sumps and tanks are listed in Table C-1 to or RPV [BWR]) level cannot monitored for ;:: 15 min. (Note 1) improve the readability of the EAL. The word Tank has been be monitored for 15 minutes AND EITHER replaced with Pool. There are no tanks applicable to RBS but have or longer

UNPLANNED rise in any included the Suppression Pool as a volume where RCS leakage may AND Table C-1 sump or pool relocate.

levels due to a loss of RPV Replaced the term "increase" with the word "rise" consistent with b. UNPLANNED increase in inventory allowed usage. (site-specific sump and/or tank) levels due to a loss of

Visual observation of Added bulleted criteria "Visual observation" to include direct (reactor vessel/RCS

[PWR] UNISOLABLE RCS leakage observation of significant RCS leakage. or RPV [BWR]) inventory. Note The Emergency Director should N/A Note 1: The Emergency The classification timeliness note has been standardized across the declare the Alert promptly upon Director should declare RBS EAL scheme by referencing the "time limit" specified within the determining that 15 minutes has the event promptly upon EAL wording. been exceeded, or will likely be determining that the Added "The Emergency Director is not allowed an additional 15 exceeded time limit has been minutes to declare after the time limit is exceeded" to reinforce the exceeded, or will likely concept that the EAL timing component runs concurrent with the be exceeded. The classification timeliness clock. Emergency Director is 44 of 127 RBS EAL Comparison Matrix not allowed an additional 15 minutes to declare after the time limit is exceeded. 45of127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) CA2 Loss of all offsite and all onsite CA2 Loss of all offsite and all onsite "ENS buses" is the RBS-specific terminology for "emergency buses". AC power to emergency buses AC power to ENS buses for 15 for 15 minutes or longer minutes or longer. MODE: Cold Shutdown, MODE: 4 -Cold Shutdown, 5 -Refueling, Defueled Refueling, DEF -Defueled NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Loss of ALL offsite and ALL CA2.1 Loss of all offsite and all onsite DIV I and DIV II 4.16 KV ENS buses are the site-specific emergency onsite AC Power to (site-specific AC power capability to DIV I and buses. emergency buses) for 15 DIV II 4.16 KV ENS buses for minutes or longer. 15 min. (Note 1) Note The Emergency Director should N/A Note 1: The Emergency The classification timeliness note has been standardized across the declare the Unusual Event Director should declare RBS EAL scheme by referencing the "time limit" specified within the promptly upon determining that the event promptly EAL wording. 15 minutes has been exceeded, upon determining that Added "The Emergency Director is not allowed an additional 15 or will likely be exceeded. the time limit has been exceeded, or will likely minutes to declare after the time limit is exceeded" to reinforce the be exceeded. The concept that the EAL timing component runs concurrent with the Emergency Director is classification timeliness clock. not allowed an additional 15 minutes to declare after the time limit is exceeded. 46 of 127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) CA3 Inability to maintain the plant in CA3 Inability to maintain plant in cold None cold shutdown. shutdown. L MODE: Cold Shutdown, MODE: 4 -Cold Shutdown, 5 -Refueling Refueling NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 UNPLANNED increase in RCS UNPLANNED rise in RCS Example EALs #1 and #2 have been combined into a single EAL temperature to greater than temperature to> 200°F for as EAL #2 is the alternative threshold based on a loss of RCS (site-specific Technical > Table C-4 duration temperature indication. Specification cold shutdown (Note 1) Replaced the term "increase" with the word "rise" consistent with temperature limit) for greater OR allowed usage. than the duration specified in the following table. CA3.1 UNPLANNED RPV pressure 200°F is the site-specific Tech. Spec. cold shutdown temperature rise > 10 psig limit. 2 UNPLANNED RCS pressure Table C-4 *is the site-specific implementation of the gener.ic RCS increase greater than (site-Reheat Duration Threshold table. specific pressure reading). (This EAL does not apply during 10 psig is the site-specific pressure increase readable by Control water-solid plant conditions. Room indications. [PWR]) Note The Emergency Director should NIA Note 1: The Emergency The classification timeliness note has been standardized across the declare the Unusual Event Director should RBS EAL scheme by referencing the "time limit" specified within the promptly upon determining that declare the event EAL wording. 15 minutes has been exceeded, promptly upon Added "The Emergency Director is not allowed an additional 15 or will likely be exceeded. determining that the time limit has been minutes to declare after the time limit is exceeded" to reinforce the exceeded, or will likely concept that the EAL timing component runs concurrent with the be exceeded. The classification timeliness clock. Emergency Director is not allowed an additional 15 minutes to declare after the 47 of 127 RBS EAL Comparison Matrix time limit is exceeded. , Table: RCS Heat-up Duration Thresholds RCS Status Containment Closure Status Heat-up Duration Intact (but not at reduced Not applicable 60 minutes* inventory [PWR]) Not intact ( or at reduced Established 20 minutes* inventory [PWR]) Not Established 0 minutes

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

Table C-4 RCS Heat-up Duration Thresholds RCS Status CONTAINMENT CLOSURE Heat-up Duration Status Intact N/A 60 min.* Established 20 min.* Not intact Not established 0 min.

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

48 of 127 _ j RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) CA6 Hazardous event affecting a CA6 Hazardous event affecting Pluralized safety systems to be consistent with NRC EP FAQ 2016-SAFETY SYSTEM needed for SAFETY SYSTEMS needed for 002 that specifies degraded performance or visible damage in more the current operating mode. the current operating mode. than one safety system train. MODE: Cold Shutdown, MODE: 4 -Cold Shutdown, 5 -Refueling Refueling 49of127 RBS EAL Comparison Matrix NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 a. The occurrence of ANY of CA6.1 The occurrence of any Table The hazardous events have been tabularized in Table C-6. the following hazardous evenJs: C-6 hazardous event CA6.1 reflects NRC FAQ 2016-002 requiring degraded

Seismic event AND performance or visible damage to more than one train of a safety (earthquake) system caused by the specified events.
Internal or external Event damage has caused This wording is a deviation from NEI 99-01 Revision 6 CA6 flooding event indications of degraded performance on one train of a generic wording and bases but is deemed acceptable in order
High winds or tornado SAFETY SYSTEM needed for to ensure that an Alert is declared only when a hazardous strike the current operating mode event causes actual or potential performance issues with
FIRE safety systems. This is consistent with NRC-approved EP
EXPLOSION AND EITHER: FAQ 2016-002.
(site-specific hazards)
Event damage has The word "a" is replaced with "the" in the FAQ wording to provide
Other events with similar hazard characteristics as caused indications of agreement with the FAQ basis information indicating that the determined by the Shift degraded performance criteria is applicable to another train of the same safety system. Manager to the second train of AND the SAFETY SYSTEM b. EITHER of the following:

needed for the current 1. Event damage has caused operating mode indications of degraded

Event damage has performance in at least resulted in VISIBLE one train of a SAFETY DAMAGE to the SYSTEM needed for the second train of the current operating mode. SAFETY SYSTEM OR needed for the current 2. The event has caused operating mode VISIBLE DAMAGE to a (Notes 9, 10) SAFETY SYSTEM component or structure needed for the current operating mode. 50of127 RBS EAL Comparison Matrix N/A N/A N/A Note 9: If the affected SAFETY Added Note 9 consistent with the recommendation of NRC EP FAQ SYSTEM train was already 2016-002.

inoperable or out of service before the hazardous event occurred, then emergency classification is not warranted. N/A N/A N/A Note 1 O: If the hazardous event Added Note 10 consistent with the recommendation of NRC EP only resulted in VISIBLE FAQ 2016-002. DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted. 51 of 127 RBS EAL Comparison Matrix Table C-6 Hazardous Events

Seismic event (earthquake)
Internal or external FLOODING event
High winds or tornado strike )
FIRE
EXPLOSION
Other events with similar hazard characteristics as determined by the Shift Manager 52 of 127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) CS1 Loss of (reactor vessel/RCS CS1 Loss of RPV inventory affecting None [PWR] or RPV [BWR]) inventory core decay heat removal affecting core decay heat capability removal capability.

MODE: 4 -Cold Shutdown, 5 -MODE: Cold Shutdown, Refueling Refueling NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 a. CONTAINMENT CLOSURE CS1.1 CONTAINMENT CLOSURE not -143 in. is the low-low-low ECCS actuation setpoint (Level 1). not established. established , AND AND b. (Reactor vessel/RCS [PWR] RPV water level< -143 in. or RPV [BWR]) level less than (Level 1) (site-specific level). 2 a. CONTAINMENT CLOSURE CS1.2 CONTAINMENT CLOSURE -162 in. is the indicated RPV water level corresponding to the top of established. established active fuel. AND AND b.(Reactor vessel/RCS [PWR] RPV water level< -162 in. or RPV [BWR]) level less than (TAF) (site-specific level). -3 a. (Reactor vessel/RCS [PWR] CS1.3 RPV level cannot be monitored Replaced the term "increase" with the word "rise" consistent with or RPV [BWR]) level cannot for;;:: 30 min. (Note 1) allowed usage. be monitored for 30 minutes AND Site-specific applicable sumps and tanks are listed in Table C-1 to or longer. Core uncovery is* indicated by any of the following: improve the readability of the EAL. The word Tank has been AND replaced with Pool. There are no tanks applicable to RBS but have b. Core uncovery is indicated by

UNPLANNED rise in any included the Suppression Pool as a volume where RCS leakage may ANY of the following:

Table C-1 sump or pool relocate. levels of sufficient 53of127 RBS EAL Comparison Matrix * (Site-specific radiation magnitude to indicate core Added bulleted criteria "Visual observation ... " to include direct monitor) reading greater uncovery observation of significant unisolable RCS leakage than (site-specific value) Visual observation of The dose rate due to core shine when the top of the core becomes *

Erratic source range UNISOLABLE RCS leakage uncovered should result in the indicated value. monitor indication

[PWR] of sufficient magnitude to

UNPLANNED increase in indicate core uncovery (site-specific sump and/or
RMS-RE140 Refueling tank) levels of sufficient Floor Near North Entrance, magnitude to indicate RMS-RE141 Refueling core uncovery Floor Near South Entrance * (Other site-specific or RMS-RE16 A/B Primary indications) containment

-PAM A/B reading > 9 R/hr Note The Emergency Director should N/A Note 1: The Emergency Director The classification timeliness note has been standardized across the declare the Site Area should declare the event RBS EAL scheme by referencing the "time limit" specified within the Emergency promptly upon promptly upon EAL wording. determining that 30 minutes has determining that the Added "The Emergency Director is not allowed an additional 15 been exceeded, or will likely be time limit has been minutes to declare after the time limit is exceeded" to reinforce the exceeded exceeded, or will likely concept that the EAL timing component runs concurrent with the be exceeded. The classification timeliness clock. Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded. 54of127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) -* CG1 Loss of (reactor vessel/RCS CG1 Loss of RPV inventory affecting None [PWR] or RPV [BWR]) inventory fuel clad integrity with affecting fuel clad integrity with Containment challenged containment challenged MODE: 4 -Cola Shutdown, 5 -MODE: Cold Shutdown, Refueling Refueling NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 a. (Reactor vessel/RCS [PWR] CG1.1 RPV level< -162 in. (TAF) for;:;:: -162 in. is the indicated RPV water level corresponding to the top or RPV [BWR]) level less than 30 min. (Note 1) of active fuel. (site-specific level) for 30 minutes or longer. AND 4% hydrogen concentration in the presence of oxygen is the AND minimum necessary to support a hydrogen burn (4%). The Igniter b. ANY indication from the Any Containment Challenge System is designed to prevent hydrogen accumulation by locally Containment Challenge Table indication, Table C-2 burning hydrogen in a controlled manner as soon as the (see below). hydrogen enters the containment atmosphere and reaches the ) ., igniters. For high rates of hydrogen production, ignition occurs at ( the lowest concentration that can support ignition. Following __ ignition, hydrogen is consumed through formation of diffusion ( flames where the gas enters the containment, thus controlling hydrogen concentration at approximately 4%. The MAX SAFE Operating Radiation Levels are the highest value of these parameters at which neither: (1) equipment necessary for the safe shutdown of the plant will fail, nor (2) personnel access necessary for the safe shutdown of the plant will be precluded. These are the site-specific secondary containment radiation monitor readings and.are listed in EOP-3 Table SC-2. Table C-2 lists the values for equipment that is in service in Cold Shutdown. 2 a. (Reactor vessel/RCS [PWR] CG1.2 RCS level cannot be monitored Site-specific applicable sumps and tanks are listed in Table C-1 or RPV [BWR]) level cannot ' for ;:;:: 30 min. (Note 1) to improve the readability of the EAL. The word Tank has been be monitored for 30 minutes replaced with Pool. There are no tanks applicable to RBS but 55 of 127 RBS EAL Comparison Matrix or longer. AND have included the Suppression Pool as a volume where RCS AND Core uncovery is indicated by leakage may relocate.

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

Although "Visual Observation" is neither a sump nor tank, it is ANY of the following:

UNPLANNED rise in any included in order to implement the intent of the NEI basis which Table C-1 sump or pool states:" ... operators may determine that an inventory loss is * (Site-specific radiation occurring by observing changes ... " levels of sufficient magnitude monitor) reading greater to indicate core uncovery The dose rate due to core shine when the top of the core than (site-specific value) becomes uncovered should result in the indicated value.
Erratic source range
Visual observation of monitor indication

[PWR] UNISOLABLE RCS leakage 4% hydrogen concentration in the presence of oxygen is the of sufficient magnitude to mini_mum necessary to support a hydrogen burn (4%). The Igniter ** UNPLANNED increase in indicate core uncovery System is designed to prevent hydrogen accumulation by locally (site-specific sump and/or

RMS~RE140 Refueling Floor burning hydrogen in a controlled manner as soon as the tank) levels of sufficient hydrogen enters the containment atmosphere and reaches the magnitude to indicate core Near North Entrance, RMS-igniters.

For high rates of hydrogen production, ignition occurs at uncovery RE141 Refueling Floor Near the lowest concentration that can support ignition. Following South Entrance or RMS-ignition, hydrogen is consumed through formation of diffusion

(Other site-specific RE16 A/B Primary flames where the gas enters the containment, thus controlling indications) containment

-PAM A/B hydrogen concentration at approximately 4%. AND reading The MAX SAFE Operating Radiation Levels are the highest value > 9 R/hr c. ANY indication from the AND of these parameters at which neither: (1) equipment necessary Containment Challenge Table for the safe shutdown of the plant will fail, nor (2) personnel (see below). Any Containment Challenge access necessary for the safe shutdown of the plant will be indication, Table C-2 precluded. These are the site-specific secondary containment radiation monitor readings and are listed in EOP-3 Table SC-2. Note The Emergency Director should N/A Note 1: The Emergency Director The classification timeliness note has been standardized across declare the General Emergency should declare the event the RBS EAL scheme by referencing the "time limit" specified promptly upon determining that promptly upon within the EAL wording. 30 minutes has been exceeded, determining that the time Added "The Emergency Director is not allowed an additional 15 or will likely be exceeded. limit has been exceeded, minutes to declare after the time limit is exceeded." to reinforce or will likely be the concept that the EAL timing component runs concurrent with exceeded. The the classification timeliness clock. Emergency Director is not allowed an additional N/A 15 minutes to declare after the time limit is exceeded. 56of127 RBS EAL Comparison Matrix Note 6: If CONTAINMENT Note 6 implements the asterisked note associated with the CLOSURE is re-Containment Closure requirement. established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required. Containment Challenge Table

CONTAINMENT CLOSURE not established*
(Explosive mixture) exists inside containment
UNPLANNED increase in containment pressure
Secondary containment radiation monitor reading above (site-specific value) [BWR]
If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.

57of127 RBS EAL Comparison Matrix Table C-2 Containment Challenge Indications

CONTAINMENT CLOSURE not established (Note 6)
Drywell or containment hydrogen concentration

> 4%

UNPLANNED rise in containment pressure
Exceeding one or more Secondary Containment Control MAX SAFE area radiation levels: Area DRMS Grid 2 Max. Safe Operating Value RHR Equip Rm A 1213 9.5E+03 mR/hr RHR Equip Rm B 1214 9.5E+03 mR/hr RHR Equip Rm C 1215 9.5E+03 mR/hr 58of127 RBS EAL Comparison Matrix Category D Permanently Defueled Station Malfunction 59 of 127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) PD-AU1 Recognition Category D NIA NIA NEI Recognition Category PD I Cs and EALs are applicable only to PD-AU2 Permanently Defueled Station permanently defueled stations.

RBS is not a defueled station. PD-SU1 -PD-HU1 -PD-HU2 PD-HU3 PD-AA1 PD-AA2 PD-HA1 PD-HA3 60 of 127 RBS EAL Comparison Matrix Category E Independent Spent Fuel Storage Installation (ISFSI) 61 of127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording ' Difference/Deviation Justification IC#(s) E-HU1 Damage to a loaded cask EU1 Damage to a loaded cask None CONFINEMENT BOUNDARY CONFINEMENT BOUNDARY MODE: All MODE: All NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Damage to a loaded cask EU1 .1 Damage to a loaded cask The specified dose rates represent 2 times the site-specific cask CONFINEMENT BOUNDARY as CONFINEMENT BOUNDARY technical specification allowable levels per the ISFSI Technical indicated by an on-contact as indicated by an on-contact Specifications. radiation reading greater than (2 radiation reading on the surface times the site-specific cask of a loaded spent fuel cask {HI-specific technical specification STORM overpack) allowable radiation level) on the > EITHER of the following: surface of the spent fuel cask. 60 mRem/hr (T + ri) on the

top of the overpack
600 mRem/hr (T + ri) on side of the overpack (excluding inlet and outlet. ducts) 62 of 127 RBS EAL Comparison Matrix Category F Fission Product Barrier Degradation 63of127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) FA1 Any Loss or any Potential Loss of FA1 Any loss or any potential loss of None either the Fuel Clad or RCS either Fuel Clad or RCS barrier. MODE: 1 -Power Operation, 2 -MODE: Power Operation, Hot Startup, 3 -Hot Shutdown Standby, Startup, Hot Shutdown NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Any Loss or any Potential Loss of FA1.1 Any loss or any potential loss of Table F-1 provides the fission product barrier loss and potential loss either the Fuel Clad or RCS either Fuel Clad or RCS barrier thresholds.

barrier. (Table F-1) 64of127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) FS1 Loss or Potential Loss of any two FS1 Loss or potential loss of any two None barriers barriers MODE: Power Operation, Hot MODE: 1 -Power Operation, 2 -Standby, Startup, Hot Shutdown Startup, 3 -Hot Shutdown NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Loss or Potential Loss of any two FS1.1 Loss or potential loss of any two Table F-1 provides the fission product barrier loss and potential loss barriers barriers {Table F-1) thresholds. 65of127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) FG1 Loss of any two barriers and FG1 Loss of any two barriers and loss

None Loss or Potential Loss of third or potential loss of the third barrier barrier MODE: Power Operation, Hot MODE: 1 -Power Operation, 2 -Standby, Startup, Hot Shutdown Startup, 3 -Hot Shutdown NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Loss of any two barriers and FG1.1 Loss of any two barriers Table F-1 pmvides the fission product barrier loss and potential loss Loss or Potential Loss of third thresholds.

'* barrier AND ' ', Loss or potential loss of the third barrier (Table F-1) 66 of 127 RBS EAL Comparison Matrix BWR Fuel Clad Fission Product Barrier Degradation Thresholds NEI NEI Threshold Wording RBS RBS FPB Wording Difference/Deviation Justification FPB# FPB #(s) FC Loss RCS Activity FC Loss Coolant activity> 300 µCi/gm 1-300 µCi/gm DEl-131 is the site-specific indication for this reactor 1 A. (Site-specific indications that FCB4 131 Dose Equivalent coolant activity. reactor coolant activity is greater than 300 µCi/gm dose equivalent 1-131 ). FC Loss RPV Water Level FC Loss SAP entry is required Revised to read "SAP entry is required." Requirements for Primary 2 A. Primary containment flooding FCB1 Containment Flopding correspond to entry into the Severe Accident Guidelines (SAG's) and are established in EOP RPV Control, EOP required. RPV Control, ATWS and EOP RPV Flooding. Per the developer's guide "the phrase, "Primary containment flooding required," should be modified to agree with the site-specific EOP phrase indicating exit from all EOPs and entry to the SAGs (e.g., drywell flooding required, etc.)." Implements EP FAQ 2015-004. FC Loss Not Applicable N/A N/A N/A 3 Not Applicable FC Loss Primary Containment FC Loss Containment radiation (RMS-A 3,000 R/hr reading in the containment is used to indicate a loss of 4 Radiation FCB3 RE16) > 3,000 R/hr the Fuel Clad barrier and a release of reactor coolant, with elevated activity (300 µCi/gm dose equivalent 1-131) indicative of fuel A. Primary containment damage, into the drywell or containment. radiation monitor reading greater than (site-specific value). FC Loss Other Indications N/A N/A No other site-specific Fuel Clad Loss indication has been identified 5 A. (site-specific as applicable) for RBS. 67of127 RBS EAL Comparison Matrix NEI NEI Threshold Wording RBS RBS FPB Wording Difference/Deviation Justification FPB# FPB #(s) -. FC Loss Emergency Director FC Loss Any condition in the opinion of None 6 Judgment FCB5 the Emergency Director that A. ANY condition in the opinion indicates loss of the Fuel Clad of the Emergency Director that Barrier indicates Loss of the Fuel Clad Barrier. ,_ FC RCS Activity N/A N/A N/A P-Loss Not Applicable 1 FC RPV Water Level FC RPV water level cannot be -162 in. is the site-specific indicated RPV water level corresponding P-Loss A. RPV water level cannot be P-Loss restored and maintained > 162 to the top of active fuel. 2 restored and maintained above FCB2 in. (TAF) or cannot be (site-specific RPV water level determined. corresponding to the top of active fuel) or cannot be determined. -FC Not Applicable N/A N/A N/A P-Loss Not Applicable 3 FC Primary Containment N/A N/A N/A P-Loss Radiation 4 Not Applicable " FC Other Indications N/A N/A No other site-specific Fuel Clad Potential Loss indication has been P-Loss A. (site-specific as applicable) identified for RBS. 5 68 of 127 RBS EAL Comparison Matrix NEI NEI Threshold Wording RBS RBS FPS Wording Difference/Deviation Justification FPS# FPS #(s) FC Emergency Director FC Any condition in the opinion of None P-Loss Judgment P-Loss the Emergency Director that 6 A. Any condition in the opinion FCB6 indicates potential loss of the of the Emergency Director that Fuel Clad Barrier indicates Potential Loss of the Fuel Clad Barrier. 69 of 127 RBS EAL Comparison Matrix BWR RCS Fission Product Barrier Degradation Thresholds NEI NEI IC Wording RBS FPB RBS FPB Wording Difference/Deviation Justification FPB# #(s) RCS Primary Containment RCS Loss Drywell pressure > 1.68 psid 1.68 psid is the site-specific primary containment pressure Loss Pressure RCB5 due to RCS leakage corresponding to the drywell high pressure scram and isolation 1 A. Primary containment pressure setpoint. greater than (site-specific value) due to RCS leakage. RCS RPV Water Level RCS Loss RPV water level cannot be -162 in. is the site-specific indicated RPV water level Loss A. RPV water level cannot be RCB1 restored and maintained corresponding to the top of active fuel. 2 restored and maintained above > -162 in. (TAF) or cannot be (site-specific RPV water level determined. corresponding to the top of active fuel) or cannot be determined. RCS RCS Leak Rate RCS Loss UNISOLABLE break in any of Main Steam Line, RCIC Steam Line, RWCU, and Feedwater are Loss A. UNISOLABLE break in ANY RCB2 the following: the site-specific systems with potential for high energy line breaks. 3 of the following: (site-specific . Main Steam Line Added "Failure to isolate the leak (from the Control Room or systems with potential for high-. RCIC Steam Line locally), within 15 minutes or if known that the leak cannot be energy line breaks) isolated within 15 minutes, from the start of the leak requires OR . RWCU immediate classification" to the basis. This provides agreement . Feedwater with the definition of "Unisolable" and ensures isolation attempts, B. Emergency RPV both locally and remotely, are achieved in a timely manner. Depressurization. RCS Loss Emergency Depressurization The term "RPV" has been deleted to agree with the use of this RCB3 is required phrase in RBS EOP-1 RPV Control "Emergency Depressurizatio_n". 70of127 RBS EAL Comparison Matrix NEI NEI IC Wording RBS FPB RBS FPB Wording Difference/Deviation Justification FPB# #(s) RCS Primary Containment RCS Loss Drywell radiation (RMS-RE20) The drywell radiation monitor reading (38 R/h rounded to 30 R/hr Loss Radiation RCB6 > 30 R/hr for readability) corresponds to an instantaneous release of all 4 A. Primary containment reactor coolant mass into the primary containment, assuming that radiation monitor reading greater reactor coolant activity equals Technical Specification allowable than (site-specific value). limits. Thermally Induced Currents (TIC) discussion is added to the NEI basis information to prevent potential incorrect classification during the time period post-LOCA when the radiation monitor may be impacted by this effect. RCS Other Indications NIA NIA No other site-specific RCS Loss indication has been identified for Loss A. (site-specific as applicable) RBS. 5 RCS Emergency Director Judgment RCS Loss Any condition in the opinion None -Loss A. ANY condition in the opinion RCB7 of the Emergency Director that indicates loss of the RCS 6 of the Emergency Director that Barrier indicates Loss of the RCS Barrier. RCS Primary Containment NIA NIA NIA P-Loss 1 Pressure Not Applicable RCS RPV Water Level NIA NIA NIA P-Loss 2 Not Applicable 71 of 127 RBS EAL Comparison Matrix NEI NEI IC Wording RBS FPS RBS FPS Wording Difference/Deviation Justification FPS# #(s) RCS RCS Leak Rate RCS UNISOLABLE primary system Reference to EOP-3 has been added for clarification. P-Loss 3 A. UNISOLABLE primary P-Loss leakage that results in Table F-2 was added to provide the Isolation Temperature alarm system leakage that results in RCB4 exceeding EITHER: setpoints and Max Normal Operating Values from EOP-3. exceeding EITHER of the

One or more EOP-3 Max The alarms for the high area ambient temperature are associated following:

Normal area radiation with the TRM 3.3.6.1 allowable values for primary containment

1. Max Normal Operating operating value isolation.

The use of the TRM value provides a readily identifiable Temperature {Table F-2) condition of the RCS barrier status with an alarm in the Control

One or more Isolation Room and an isolation signal that (if the condition persists)

OR Temperature alarms indicates unisolable leakage as defined in the EAL threshold. The 2. Max Normal Operating Area {Table F-2) use of isolation temperatures still allows for a discernable margin Radiation Level. prior to reaching the Max Safe Operating temperature This is an acceptable deviation from the NEI 99-01 Revision 6 guidance. Added "Failure to isolate the leak (from the Control Room or locally), within.15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification" to the basis. This provides agreement with the definition of "Unisolable" and ensures isolation attempts, both locally and remotely, are achieved in a timely manner. RCS Pr1mary Containment N/A N/A N/A P-Loss 4 Radiation Not Applicable RCS Other Indications N/A N/A No other site-specific RCS Potential Loss indication has been P-Loss 5 A. (site-specific as applicable) identified for RBS. RCS Emergency Director Judgment RCS Any condition in the opinion of None P-Loss 6 A. ANY condition in the opinion P-Loss the Emergency Director that indicates potential loss of the of the Emergency Director that RC88 RCS Barrier indicates Potential Loss of the RCS Barrier. 72 of 127 RBS EAL Comparison Matrix Table F-2 Secondary Containment Operating Values -Area Temperatures Parameter Isolation Temperature Max Safe Main Steam Line Tunnel 173°F (P601-19A-200°F A1/A3/B1/B3) RHR Equipment Area 1 (A) 117°F (P601-20A-B4) 200°F RHR Equipment Area 2 (B) 117°F ( P601-20A-B4) 200°F RCIC Equipment Area 182°F (P601-21 A-86) 200°F RWCU Pump Room 1 (A)/ 2 (B) 165°F (P680-1 A-A2/B2) 200°F / Area Radiation Levels Parameter Max Normal Max Safe HPCS Area (1212) Grid 2 8.20E+01 mR/hr 9.5E+03 mR/hr RHR Equipment Room A (1213) Grid 2 8.20E+01 mR/hr 9.5E+03 mR/hr RHR Equipment Room B (1214) Grid 2 8.20E+01 mR/hr 9.5E+03 mR/hr RHR Equipment Room C (1215) Grid 2 8.20E+01 mR/hr 9.5E+03 mR/hr LPCS Equipment Room (1216) Grid 2 8.20E+01 mR/hr 9.5E+03 mR/hr HPCS Penetration Area (1217) Grid 2 8.20E+01 mR/hr 9.5E+03 mR/hr LPCS Penetration Area (1218) Grid 2 8.20E+01 mR/hr 9.5E+03 mR/hr RCIC Equipment Room (1219) Grid 2 1.20E+02 mR/hr 9.5E+03 mR/hr 73 of 127 RBS EAL Comparison Matrix BWR Containment Fission Product Barrier Degradation Thresholds NEI NEI IC Wording RBS RBS FPB Wording Difference/Deviation Justification FPB# FPB #(s) PC Primary Containment Conditions PC Loss UNPLANNED rapid drop in The NEI phrase "primary containment" has been Loss A. UNPLANNED rapid drop in primary CNB3 containment pressure following changed to "containment" to use terminology specific 1 containment pressure following containment pressure rise to the Mark Ill containment design. primary containment pressure rise OR Containment pressure response not The NEI phrase "Primary containment" has been PC Loss B. Primary containment pressure CNB4 consistent with LOCA conditions changed to "Containment" to use terminology specific response not consistent with LOCA to the Mark Ill containment design. conditions. PC RPV Water Level N/A N/A N/A Loss Not Applicable 2 PC Primary Containment Isolation PC Loss UNISOLABLE direct downstream The NEI phrase "Primary Containment" has been Loss Failure CNB9 pathway to the environment exists after changed to "Containment" to use terminology-specific 3 A. UNISOLABLE direct downstream Containment isolation signal to the Mark Ill containment design. pathway to the environment exists Added "Failure to isolate the leak (from the Control after primary containment isolation Room or locally), within 15 minutes or if known that signal the leak cannot be isolated within 15 minutes, from OR the start of the leak requires immediate classification" to the basis. This provides agreement with the B. Intentional primary containment definition of "Unisolable" and ensures isolation venting per EOPs attempts, both locally and remotely, are achieved in a timely manner. 74 of 127 RBS EAL Comparison Matrix NEI NEI IC Wording RBS RBS FPB Wording Difference/Deviation Justification FPB# FPB #(s) OR PC Loss Intentional Containment venting per The NEI phrase "Primary Containment" has been C. UNISOLABLE primary system CNB10 EOPs changed to "Containment" to use terminology specific leakage that results in exceeding to the Mark Ill containment design. EITHER of the following:

1. Max Safe Operating Temperature.

PC Loss UNISOLABLE primary system leakage Reference to EOP-3 has been added for clarification. OR that results in exceeding EITHER: CNB2 Table F-2 was added to provide the Max Safe 2. Max Safe Operating Area

One or more EOP-3 Max Safe Operating Values from EOP-3. Radiation Level. area radiation operating value Added "Failure to isolate the leak (from the Control (Table F-2) Room or locally), within 15 minutes or if known that
One or more EOP-3 Max Safe the leak cannot be isolated within 15 minutes, from area temperature operating value the start of the leak requires immediate classification" \ (Table F-2) to the basis. This provides agreement with the definition of "Unisolable" and ensures isolation attempts, both locally and remotely, are achieved in a timely manner. PC Primary Containment Radiation NIA NIA NIA Loss Not Applicable 4 PC Other Indications N/A N/A No other site-specific Containment Loss indication Loss A. (site-specific as applicable) has been identified for RBS. 5 PC Emergency Director Judgment PC Any condition in the opinion of the None Loss ANY condition in the opinion of the Loss Emergency Director that indicates loss 6 Emergency Director that indicates CNB11 of the Containment Barrier Loss of the Containment Barrier. 75 of 127 RBS EAL Comparison Matrix NEI NEI IC Wording RBS RBS FPS Wording Difference/Deviation Justification FPS# FPS #(s) PC Primary Containment Conditions PC Containment pressure > 15 psig 15 psig is the RBS containment design pressure.

P-Loss A. Primary containment pressure P-Loss 1 greater than (site-specific value) CNB5 OR PC Drywell or containment hydrogen 4% hydrogen concentration is the minimum B. (site-specific explosive mixture) P-Loss concentration > 4% necessary to support a hydrogen burn. The Igniter exists inside primary containment CNB6 System is designed to prevent hydrogen OR accumulation by locally burning hydrogen in a controlled manner as soon as the hydrogen enters C. HCTL exceeded. the containment atmosphere and reaches the igniters. For high rates of hydrogen production, ignition occurs at the lowest concentration that can support ignition. Following ignition, hydrogen is consumed through formation of diffusion flames where the gas enters the containment, thus controlling hydrogen concentration at approximately 4%. / PC Parameters cannot be restored and The NEI phrase "HCTL exceeded" has been changed P-Loss maintained within the safe zone of the to "Parameters cannot be restored and maintained CNB7 HCTL curve (EOP Figure 2) within the safe zone of the HCTL curve (EOP Figure 2)" to use terminology consistent with the EP use of the HCTL. EOP Figure 2 is the RBS HCTL curve. PC RPV Water Level PC SAP entry is required Revised to read "SAP entry is required." P-Loss A. Primary containment flooding P-Loss Requirements for Primary Containment Flooding correspond to entry into the Severe Accident 2 required. CNB1 Guidelines (SAGs) and are established in EOP RPV Control, EOP RPV Control -ATWS and EOP RPV Flooding. Per the developer's guide the phrase, "Primary containment flooding required," was / modified to agree with the site-specific EOP phrase indicating exit from all EOPs and entry to the SAGs This difference implements EP FAQ 2015-004. 76of127 RBS EAL Comparison Matrix NEI NEI IC Wording RBS RBS FPB Wording Difference/Deviation Justification FPB# FPB #(s) PC Primary Containment Isolation N/A N/A N/A . P-Loss Failure 3 Not Applicable ,, Containment radiation (Rfv!S-RE16) A 12,000 R/hr reading in the containment is used to PC Primary Containment Radiation PC P-Loss P-Loss > 12,000 R/hr indicate a potential loss of the containment barrier A Primary containment radiation and a release of reactor coolant, with significant 4 monitor reading greater than (site-CNB8 activity indicative of 20% fuel damage, into the drywall specific value). or containment. This value assumes an instantaneous_ release and dispersal of the reactor coolant noble gas ) and iodine inventory associated with a concentration associat~d with 20% clad damage into the drywall or containment atmosphere. PC Other Indications N/A N/A No other site-specific Containment Potential Loss P-Loss A (site-specific as applicable) indication has been identified for RBS. J 5 PC Emergency Director Judgment PC Any condition in the opinion of the None P-Loss A ANY condition in the opinion of the P-Loss Emergency Director that indicates potential loss of the Containmenf -6 Emergency Director that indicates CNB12 Barrier Potential Loss of the Containment Barrier. -77of127 RBS EAL Comparison Matrix \ Category H Hazards and Other Conditions Affecting Plant Safety 78 of 127 RBS EAL 'comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) HU1 Confirmed SECURITY HU1 Confirmed SECURITY None CONDITION or threat CONDITION or threat. MODE: All MODE:AII NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 A SECURITY CONDITION that HU1.1 A SECURITY CONDITION that Example EALs #1, 2 and 3 have been combined into a single EAL does not involve a HOSTILE does not involve a HOSTILE for ease of presentation and use. ACTION as reported by the (site-ACTION as reported by RBS specific security shift supervision). Security Shift Supervision 2 Notification of a credible security OR threat directed at the site. Notification of a credible security threat directed at the site 3 A validated notification from the OR NRC providing information of an aircraft threat. A validated notification from the NRC providing information of an aircraft threat . 79of127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) HU2 Seismic event greater than OBE HU2 Seismic event greater than QBE None levels levels MODE: All MODE: All NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Ju stifi cation EAL# EAL# 1 Seismic event greater than HU2.1 Seismic event > QBE as Annunciator P680-02A-C06, SEISMIC EVENT HIGH is actuated by Operating Basis Earthquake indicated by EITHER of the ERS-NBS4B, Reactor Mat Seismic Switch (AB 70' EL) with a (OBE) as indicated by: following: setpoint of 0.083 g Vertical Axis or 0.082 g Horizontal Axis. (site-specific indication that a

Annunciator P680-02A-Annunciator P680-02A-B06, SEISMIC EVENT HIGH/HIGH is seismic event mE)t or exceeded C06, SEISMIC EVENT actuated by Reactor Mat Response Spectrum Recorder ERS-OBE limits) HIGH NBR2D (AB 70' EL) ((Any of 16 Mechanical Accelerometers) with
Annunciator P680-02A-setpoint ranges of .09 -1.83g. B06, SEISMIC EVENT The amber lights on H13-P869 ERS-NBl101 indicate 100% of QBE HIGH/HIGH and amber (Operating Basis Event) acceleration limits have been exceeded for lights illuminated on the affected frequency.

H13-P869 ERS-NBl101. 80 of 127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) HU3 Hazardous event. HU3 Hazardous event None MODE: All MODE:AII NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 A tornado strike within the HU3.1 A tornado strike within the None PROTECTED AREA. PROTECTED AREA 2 Internal room or area flooding of a HU3.2 Internal room or area FLOODING Clarified the word "needed" with "required by Technical magnitude sufficient to require of a magnitude sufficient to Specifications" consistent with the generic bases. manual or automatic electrical require manual or automatic isolation of a SAFETY SYSTEM electrical isolation of a SAFETY component needed for the current SYSTEM component required by operating mode. Technical Specifications for the current operating mode 3 Movement of personnel within the HU3.3 Movement of personnel within the Changed the term "offsite" to "external to the PROTECTED AREA" PROTECTED AREA is impeded PROTECTED AREA is to address events located external to the PROTECTED AREA but due to an offsite event involving IMPEDED due to an event not considered offsite. hazardous materials (e.g., an external to the PROTECTED offsite chemical spill or toxic gas AREA involving hazardous release). materials (e.g., an offsite < chemical spill or toxic gas release) 4 A hazardous event that results in HU3.4 A hazardous event that results in Added reference to Note 7. on-site conditions sufficient to on-site conditions sufficient to prohibit the plant staff from prohibit the plant staff from accessing the site via personal accessing the site via personal vehicles. vehicles (Note 7) 5 (Site-specific list of natural or N/A N/A No other site-specific hazard has been identified for RBS. technological hazard events) 81 of 127 ---RBS EAL Comparison Matrix No te EAL #3 does not apply to routine N/A Note 7: This EAL does not This note, designated Note #7, is intended to apply to generic traffic impediments such as fog, apply to routine traffic example EAL #4, not #3 as specified in the generic guidance. snow, ice, or vehicle breakdowns impediments such as or accidents. fog, snow, ice, or vehicle breakdowns or accidents. C 82of127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) HU4 FIRE potentially degrading the HU4 FIRE potentially degrading the None level of safety of the plant. level of safety of the plant MODE: All MODE: All NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 a. A FIRE is NOT extinguished HU4.1 A FIRE is not extinguished within Table H-1 provides a list of site-specific fire areas. within 15-minutes of ANY of the 15 min. of any of the following following FIRE detection FIRE detection indications (Note indications: 1 ):

Report from the field (i.e.,
Report from the field (i.e., visual observation) visual observation)
Receipt of multiple (more
Receipt of multiple (more than 1) fire alarms or than 1) fire alarms or indications indications
Field verification of a single
Field verification of a single fire alarm fire alarm AND AND b. The FIRE is located within , The FIRE is located within any ANY of the following plant rooms Table H-1 area or areas: (site-specific list of plant rooms or areas) 2 a. Receipt of a single fire alarm HU4.2 Receipt of a single fire alarm . Table H-1 provides a list of site-specific fire areas. (i.e., no other indications of a (i.e., no other indications of a FIRE). FIRE) AND AND b. The FIRE is located within The fire alarm is indicating a 83 of 127 RBS EAL Comparison Matrix ANY of the following plant rooms FIRE within any Table H-1 area or areas: AND (site-specific list of plant rooms or The existence of a FIRE is not areas) verified within 30 min. of alarm AND receipt (Note 1) C. The existence of a FIRE is not verified within 30-minutes of alarm receipt. 3 A FIRE within the plant or ISFSI HU4.3 A FIRE within the PROTECTED RBS has an ISFSI located inside the plant Protected Area. [for plants with an ISFSI outside AREA not extinguished within 60 the plant Protected Area] min. of the initial report, alarm or PROTECTED AREA not indication (Note 1) extinguished within 60-minutes of the initial report, alarm or indication.

4 A FIRE within the plant or ISFSI HU4.4 A FIRE within the PROTECTED RBS has an ISFSI located inside the plant Protected Area. [for plants with an ISFSI outside AREA that requires firefighting the plant Protected Area] support by an offsite fire PROTECTED AREA that requires response agency to extinguish firefighting support by an offsite fire response agency to extinguish. Note Note: The Emergency Director N/A Note 1: The Emergency Director The classification timeliness note has been standardized across the should declare the Unusual Event should declare the event RBS EAL scheme by referencing the "time limit" specified within the promptly upon determining that promptly upon EAL wording. the applicable time has been determining that the time Added "The Emergency Director is not allowed an additional 15 exceeded, or will likely be limit has been exceeded, exceeded. minutes to declare after the time limit is exceeded" to reinforce the or will likely be exceeded. concept that the EAL timing component runs concurrent with the ~he Emergency Director classification timeliness clock. is not allowed an additional 15 minutes to declare after the time limit is exceeded. -84of127 RBS EAL Comparison Matrix Table H-1 Fire Areas

Reactor Building
Auxiliary Building
Fuel Building
Control Building
Standby Cooling Tower
Diesel Generator Building
Tunnels (8, D,E, F, G) 85of127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) HU? Other conditions exist which in the HU? Other conditions exist that in the None judgment of the Emergency judgment of the Emergency Director warrant declaration of a Director warrant declaration of a (NO)UE UE MODE: All MODE: All , NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Other conditions exist which in the HU7.1 Other conditions exist which, in None judgment of the Emergency the judgment of the Emergency Director indicate that events are in Director, indicate that events are progress or have occurred which in progress or have occurred indicate a potential degradation of which indicate a potential the level of safety of the plant or degradation of the level of safety indicate a security threat to facility of the plant or indicate a security protection has been initiated.

No threat to facility protection has releases of radioactive material been initiated. No releases of requiring offsite response or radioactive material requiring monitoring are expected unless offsite response or monitoring further degradation of safety are expected unless further systems occurs. degradation of SAFETY SYSTEMS occurs. 86 of 127 L RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) HA1 HOSTILE ACTION within the HA1 HOSTILE ACTION within the None OWNER CONTROLLED AREA OWNER CONTROLLED AREA or or airborne attack threat within 30 airborne attack threat within 30 minutes. minutes MODE: All MODE: All NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 A HOSTILE ACTION is occurring or HA1.1 A HOSTILE ACTION is Example EALs #1 and #2 have been combined into a single EAL has occurred within the OWNER occurring or has occurred within for ease of use. CONTROLLED AREA as reported the OWNER CONTROLLED ' by the (site-specific security shift AREA as reported by RBS supervision). Security Shift Supervision 2 A validated notification from NRC of OR an aircraft attack threat within 30 A validated notification from minutes of the site. NRC of an aircraft attack threat

within 30 min. of the site 87 of 127 RBS EAL Comparison Matrix NEI IC# HAS NEI Ex. EAL# 1 NEI IC Wording* Gaseous release impeding 9 ccess _to equipment necessary for normal plant operations, cooldown or shutdown.

MODE: All NEI Example EAL Wording a. Release of a toxic, corrosive, asphyxiant or flammable gas into any of the following plant rooms or areas: (site-specific list of plant rooms or areas with entry-related mode applicability identified) AND b. Entry into the room or area is prohibited or impeded. Note Note: If the equipment in the listed room or area was already inoperable or out-of-service lbefore the event occurred, then no emergency classification is warranted. RBS IC#(s) HAS RBS EAL# HA5.1 N/A RBS IC Wording Gaseous release IMPEDING access to equipment necessary for normal plant operations, cooldown or shutdown MODE: 3 -Hot Shutdown RBS EAL Wording Release of a toxic, corrosive, asphyxiant or flammable gas into any Table H-2 room or area AND Entry into the room or area is prohibited or IMPEDED (Note 5) Difference/Deviation Justification Mode applicability has been limited to the mode restrictions of Table H-2, Mode 3 only. ' Difference/Deviation Justification The site-specific list of plant rooms or areas with entry-related mode applicability are tabulariz~d in Table H-2. The bulleted bases item "the action for which room/area-entry is required is of an administrative or record keeping nature (e.g., normal rounds or routine inspections)" was removed from the list of exceptions to classification in the basis information. These actions are a consideration when the site-specific list was developed. Rooms requiring entry for these types of actions are already excluded from the list when it was developed. Note 5: If the equipment in the None listed room or area was already inoperable or of-service before the event occurred, then.no emergency classification i~ warranted. 88of127 / / RBS EAL Comparison Matrix Table H-2 Safe Operation & Shutdown Rooms/Areas Room/Area Mode Auxiliary Building 70' RHR B Pump Room 3 Auxiliary Building 80' RHR A Pump Room 3 Auxiliary Building 114' West 3 Control Building 95' Div 1 RSS Room 3 89 of 127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) HA6 Control Room evacuation HA6 Control Room evacuation None resulting in transfer of plant resulting in transfer of plant control to alternate locations. control to alternate locations MODE: All MODE: All NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 An event has resulted in plant HA6.1 An event has resulted in plant None control being transferred from the control being transferred from the Control Room to (site-specific Control Room to the Remote remote shutdown panels and Shutdown Panels local control stations). 90of127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) HA? Other conditions exist which in the HA? Other conditions exist that in the None judgment of the Emergency Director judgment of the Emergency Director warrant declaration of an Alert. warrant declaration of an ALERT MODE: All MODE:AII NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Other conditions exist which, in the HA?.1 Other conditions exist which, in the None judgment of the Emergency Director, judgment of the Emergency Director, indicate that events are in progress or indicate that events are in progress or have occurred which involve an actual or have occurred which involve an actual or potential substantial degradation of the potential substantial degradation of the level of safety of the plant or a security level of safety of the plant or a security event that involves probable life event that involves probable life threatening risk to site personnel or threatening risk to site personnel or damage to site equipment because of damage to site equipment because of HOSTILE ACTION. Any releases are HOSTILE ACTION. Any releases are expected to be limited to small fractions expected to be limited to small fractions of the EPA Protective Action Guideline of the EPA Protective Action Guideline exposure levels. exposure levels. \ 91 of 127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) HS1 HOSTILE ACTION within the HS1 HOSTILE ACTION within the None PROTECTED AREA PROTECTED AREA MODE: All MODE: All NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 A HOSTILE ACTION is occurring HS1.1 A HOSTILE ACTION is occurring or has None or has occurred within the occurred within the PROTECTED AREA PROTECTED AREA as reported as reported by RBS Security Shift by the (site-specific security shift Supervision supervision). 92of127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) HS6 Inability to control a key safety HS6 Inability to control a key safety function Deleted defueled mode applicability. Control of the cited function from outside the Control from outside the Control Room safety functions is not critical for a defueled reactor as there Room. is no energy source in the RPV or RCS. MODE: 1 -Power Operation, 2 -Startup, MODE: All 3 -Hot Shutdown, 4 -Cold Shutdown, 5 -This is an acceptable deviation from the generic NEI 99-Refueling 01 Revision 6 guidance. NEI Ex. RBS ( EAL# NEI Example EAL Wording EAL# RBS EAL Wording Difference/Deviation Justification 1 a. An event has resulted in HS6.1 An event has resulted in plant control The Mode applicability for the reactivity control safety plant control being transferred being transferred from the Control Room function has been limited to Modes 1 and 2 (hot operating from the Control Room to (site-to the Remote Shutdown Panels conditions). In the hot shutdown and cold operating modes specific remote shutdown panels AND adequate shutdown margin exists under all conditions. and local control stations). EP FAQ 2015-014. AND Control of any of the following key safety functions is not re-established This is an acceptable deviation from the generic NEI 99-b. Control of ANY of the within 15 min. (Note 1 ): 01 Revision 6 guidance. following key safety functions is ) not reestablished within (site-* Reactivity (Modes 1 and 2 only) specific number of minutes).

RPV water level
Reactivity control
RCS heat removal
Core cooling [PWR] I RPV water level [BWR]
RCS heat removal 93 of 127 RBS EAL Comparison Matrix RBS \ ; NEI IC# NEI IC Wording IC#(s) RBS IC Wording Difference/Deviation Justification HS? Other conditions exist which in HS? Other conditions exist that in the None the judgment of the Emergency judgment of the Emergency Director Director warrant declaration of a warrant declaration of a SITE AREA Site Area Emergency.

EMERGENCY MODE: All MODE: All NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Differe'nce/Deviation Justification EAL# EAL# 1 Other conditions exist which in HS7.1 Other conditions exist which, in the None the judgment of the Emergency judgment of the Emergency Director, r Director indicate that events are indicate that events are in progress or in progress or have occurred have occurred which involve actual or which involve actual or likely likely major failures of plant functions major failures of plant functions needed for protection of the public or needed for protection of the HOSTILE ACTION that results in public or HOSTILE ACTION that intentional damage or malicious acts, (1) results in intentional damage or toward site personnel or equipment that malicious acts, (1) toward site could lead to the likely failure of or, (2) that personnel or equipment that prevent effective access to equipment ,_ could lead to the likely failure of needed for the protection of the public. or, (2) that prevent effective Any releases are not expected to result access to equipment needed for in exposure levels which exceed EPA the protection of the public. Any Protective Action Guideline exposure releases are not expected to levels beyond the SITE BOUNDARY. result in exposure levels which exceed EPA Protective Action ' Guideline exposure levels beyond the site boundary. 94 of 127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) HG1 HOSTILE ACTION resulting in N/A N/A IC HG1 and associated example EAL are not implemented in loss of physical control of the the RBS scheme. facility. There are several other ICs that are redundant with this IC, MODE: All and are better suited to ensure timely and effective emergency declarations. In addition, the development of new spent fuel pool level EALs, as a result of NRC Order EA-12-051, clarified the intended emergency classification level for spent fuel pool level events. This is an acceptable deviation from the generic NEI 99-01 Revision 6 guidance. NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 a. A HOSTILE ACTION is N/A N/A IC HG1 and associated example EAL are not implemented occurring or has occurred within in the RBS scheme. the PROTECTED AREA as There are several other ICs that are redundant with this IC, reported by the (site-specific and are better suited to ensure timely and effective security shift supervision). emergency declarations. In addition, the development of new AND spent fuel pool level EALs, as a result of NRC Order EA-12-b. EITHER of the following has 051, clarified the intended emergency classification level for spent fuel pool level events. This deviation is justified occurred: because: ' , 1. ANY of the following safety 1. Hostile Action in the Protected Area is bounded by ICs functions cannot be HS1 and HS?. Hostile Action resulting in a loss of controlled or maintained. physical control is bound by EAL HG?, as well as any

Reactivity control event that may lead to radiological releases to the public
Core cooling in excess of Environmental Protection Agency (EPA) [PWR]/RPV water Protective Action Guides (PAGs). level [BWR] a. If, for whatever reason, the Control Room must be
RCS heat removal evacuated, and control of safety functions (e.g., reactivity control, RPV water level, and RCS heat 95of127 RBS EAL Comparison Matrix OR removal) cannot be reestablished, then IC HS6 would 2. Damage to spent fuel has apply, as well as IC HS7 if desired by the EAL occurred or is IMMINENT.

decision-maker.

b. Also, as stated above, any event (including Hostile Action) that could reasonably be expected to have a release exceeding EPA PAGs would be-bound by IC HG?. C. From a Hostile Action perspective, !Cs HS1, HS7 and HG? are appropriate, and therefore, make this part of HG1 redundant and unnecessary.

d. From a loss of physical control perspective, !Cs HS6, HS7 and HG? are appropriate, and therefore, make this part of HG1 redundant and unnecessary.

2. Any event which causes a loss of spent fuel pool level will be bounded by !Cs AA2, AS2 and AG2, regardless of I whether it was based upon a Hostile Action or not, thus making this part of HG1 redundant and unnecessary.

a. An event that leads to a radiological release will be bounded by !Cs AU1, AA1, AS1 and AG1. Events that lead to radiological releases in excess of EPA PAGs will be bounded by EALs AG1 and HG?, thus making this part of HG1 redundant and unnecessary.

!Cs AA2, AS2, AG2, AS1, AG1, HS1, HS6, HS7 and HG? have been implemented consistent with NEI 99-01 Revision 6 and thus HG1 is adequately bounded as described above. EP FAQ 2015-013 This is an acceptable deviation from the generic NEI 99-01 Revision 6 guidance. 96 of 127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) HG? Other conditions exist which in HG? Other conditions exist that in the None the judgment of the Emergency judgment of the Emergency Director Director warrant declaration of a warrant declaration of a GENERAL General Emergency EMERGENCY MODE: All MODE: All NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 other conditions exist which in HG7.1 Other conditions exist which, in the None the judgment of the Emergency judgment of the Emergency Director, Director indicate that events are indicate that events are in progress or in progress or have occurred have occurred which involve actual or which involve actual or IMMINENT substantial core degradation IMMINENT substantial core or melting with potential for loss of degradat!on or melting with containment integrity or HOSTILE _/ potential for loss of containment ACTION that results in an actual loss of integrity or HOSTILE ACTION physical control of the facility. Releases that results in an actual loss of can be reasonably expected to exceed physical control of the facility. EPA Protective Action Guideline Releases can be reasonably exposure levels offsite for more than the expected to exceed EPA immediate site area. Protective Action Guideline exposure levels offsite for more than the immediate site area. 97 of 127 RBS EAL Comparison Matrix Category S System Malfunction 98 of 127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) -SU1 Loss of all offsite AC power SU1 Loss of all offsite AC power "ENS buses" is the RBS-specific terminology for "emergency buses". capability to emergency buses for capability to ENS buses for 15 15 minutes or longer. minutes or longer MODE: Power Operation, Startup, MODE: 1 -Power Operation, 2 -Hot Standby, Hot Shutdown Startup, 3 -Hot Shutdown -NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Loss of ALL offsite AC power SU1.1 Loss of all offsite AC power DIV I and DIV II 4.16 KV ENS buses are the site-specific emergency capability to (site-specific capability, Table S-1, to DIV I and buses. emergency buses) for 15 minutes DIV II 4.16 KV ENS buses for ;;:: Site-specific AC power sources are listed in Table S-1. or longer. 15 min. (Note 1) Note The Emergency Director should N/A Note 1: The Emergency Director The classification timeliness note has been standardized across the declare the Unusual Event should declare the event RBS EAL scheme by referencing the "time limit" specified within. the promptly upon determining that 15 promptly upon EAL wording. minutes has been exceeded, or determining that the time Added "The Emergency Director is not allowed an additional 15 will likely be exceeded. limit has been exceeded, or will likely be minutes to declare after the time limit is exceeded" to reinforce the -exceeded. The concept that the EAL timing component runs concurrent with the Emergency Director is classification timeliness clock. not allowed an additional 15 minutes to declare after the time limit is exceeded. 99of127 RBS EAL Comparison Matrix Table 5-1 AC Power Sources Offsite

1 RTX-XSR1 C
1 RTX-XSR1 D Onsite
EGS-EG1A
EGS-EG1B 100 of 127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) SU2 UNPLANNED loss of Control SU3 UNPLANNED loss of Control None Room indications for 15 minutes Room indications for 15 minutes or longer. or longer. MODE: Power Operation, MODE: 1 -Power Operation, 2 -Startup, Hot Standby, Hot Startup, 3 -Hot Shutdown Shutdown . NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 An UNPLANNED event results in SU3.1 An UNPLANNED event results in The site-specific Safety System Parameter list is tabulated in Table the inability to monitor one or the inability to monitor one or S-2. more of the following parameters more Table S-2 parameters from from within the Control Room for within the Control Room for~ 15 15 minutes or longer. ' min. (Note 1) Note The Emergency Director should NIA Note 1: The Emergency Director The classification timeliness note has been standardized across the declare the Unusual Event should declare the event RBS EAL scheme by referencing the "time limit" specified within the promptly upon determining that promptly upon EAL wording. 15 minutes has been exceeded,
determining that the time Added "The Emergency Director is not allowed an additional 15 or will likely be exceeded.

limit has been exceeded, or will likely be minutes to declare after the time limit is exceeded" to reinforce the exceeded. The concept that the EAL timing component runs concurrent with the Emergency Director is classification timeliness clock. not allowed an additional 15 minutes to declare after the time limit is exceeded. 101 of 127 RBS EAL Comparison Matrix [BWR parameter list] [PWR parameter list] Reactor Power Reactor Power RPV Water Level RCS Level RPV Pressure RCS Pressure Primary Containment Pressure In-Core/Core Exit Temperature Suppression Pool Level Levels in at least (site-specific number) steam generators Suppression Pool Temperature Steam Generator Auxiliary or Emergency Feed Water Flow Table 5-2 Safety System Parameters

Reactor power
RPV water level
RPV pressure
Containment pressure
Suppression Pocil water level
Suppression Pool temperature 102 of 127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) SU3 Reactor coolant activity greater SU4 RCS activity greater than Replaced "Reactor coolant" with "RCS" consistent with site specific than Technical Specification Technical Specification allowable usage. allowable limits. limits MODE: Power Operation, Startup, MODE: Hot Standby, Hot Shutdown 1 -Power Operation, 2 -Startup, 3 -Hot Shutdown NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 (Site-specific radiation monitor) SU4.1 Offgas Pretreatment radiation The High alarm indicates that the radioactivity present at the reading greater than (site-specific monitor high alarm (P601-22A-recombiner effluent discharge is approaching the Technical value). F03, OFF GAS PRE-TREAT Specification

3. 7.4 limit. The nominal setpoint of 1.5 times the full HIGH RADIATION) power process background radiation level ensures that the activity will not exceed a value corresponding to the Technical Specification LCO 3. 7.4 allowable release rate. 2 Sample analysis indicates that a SU4.2 Coolant activity>

0.2 µCi/gm These limits ensure the source term assumed in the safety analysis reactor coolant activity value is dose equivalent 1-131 for> 48 for the Main Steam Line Break (MSLB) outside containment is not greater than an allowable limit hours exceeded, so any release of radioactivity to the environment during specified in Technical OR an MSLB is less than a small fraction of the guidelines of 10 CFR Specifications. 50.67. Coolant activity > 4.0 µCi/gm Consistent with Regulatory Guide 1.183, two cases are evaluated: dose equivalent 1-131 (1) an equilibrium iodine case with an iodine concentration in the instantaneous reactor coolant of 0.2 µCi/gm dose equivalent 1-131, and (2) an iodine spiking case with an iodine concentration in the reactor coolant of 4.0 µCi/gm dose equivalent 1-131. 103 of 127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS IC#(s) SU4 RCS leakage for 15 minutes or SU5 longer. MODE: Power Operation, Startup, Hot Standby, Hot Shutdown NEI Ex. NEI Example EAL Wording RBS EAL# EAL# 1 RCS unidentified or pressure SU5.1 boundary leakage greater than (site-specific value) for 15 minutes or longer. 2 RCS identified leakage greater than (site-specific value) for 15 minutes or longer. 3 Leakage from the RCS to a location outside containment greater than 25 gpm for 15 minutes or longer. Note The Emergency Director should N/A declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded. RBS IC Wording RCS leakage for 15 minutes or longer MODE: 1 -Power Operation, 2 -Startup, 3 -Hot Shutdown RBS EAL Wording RCS unidentified or pressure boundary leakage > 10 gpm fore". 15 min. (Note 1) OR RCS identified leakage > 25 gpm fore". 15 min. (Note 1) OR Leakage from the RCS to a location outside Containment> 25 gpm fore". 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is 104 of 127 Difference/Deviation Justification None Difference/Deviation Justification Example EALs #1, 2 and 3 have been combined into a single EAL for usability. Added "Failure to isolate the leak (from the Control Room or locally), within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification" to the basis. This provides agreement with the definition of "Unisolable" and ensures isolation attempts, both locally and remotely, are achieved in a timely manner -The classification timeliness note has been standardized across the RBS EAL scheme by referencing the "time limit" specified within the EAL wording. Added "The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded" to reinforce the concept that the EAL timing component runs concurrent with the classification timeliness clock. I _J RBS EAL Comparison Matrix exceeded. 105 of 127 RBS EAL Comparison Matrix RBS ' NEI IC# NEI IC Wording IC#(s) RBS IC Wording Difference/Deviation Justification r ' SUS Automatic or manual (trip SU6 Automatic or manual scram fails Mode 2 -Startup has been included. For BWRs, including RBS, the [PWR] / scram [BWR]) fails to to shut down the reactor plant operating mode is defined by the position of the Reactor Mode shutdown the reactor. MODE: 1 -Power Operation, 2 -Switch. During a normal plant startup the Reactor Mode Switch is MODE: Power Operation Startup placed in the Startup p9sition (Startup Mode 2) as reactor power is increased. Typically reactor power is increased to 8% before the Reactor Mode Switch is placed in Run (Power Operations Mode 1 ). 5% reactor power (APRM downscale) is the site-specific indication of .-a successful reactor scram. Therefore it is appropriate to include Startup Mode 2 to failure to scram I Cs. NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 a. An automatic (trip [PWR] / SU6.1 An automatic scram did not shut As specified in the generic developers guidance "Developers may scram [BWR]) did not shutdown down the reactor as indicated by include site-specific EOP criteria indicative of a successful reactor the reactor. reactor power > 5% after any shutdown in an EAL statement, the Basis or both (e.g., a reactor AND RPS_ setpoint is exceeded power level)." Reactor powers 5% is the site-specific indication of a AND successful reactor scram. ' b. A subsequent manual action Added the words" ... as indicated by reactor power> 5% after any taken at the reactor control A subsequent automatic scram consoles is successful in or manual scram action taken at RPS setpoint is exceeded" to clari_fy that it is a failure of the shutting down the reactor. the reactor control console automatic trip when a valid trip signal has been exceed. (Mode Switch, Manual PBs, ARI) Mode Switch, Manual PBs, and initiation of ARI are the manual is successful in shutting down the actions taken to shut down the reactor. reactor as indicated by reactor power::; 5% (APRM downscale) -2 a. A manual trip ([PWR] / -SU6.2 A manual scram did not shut As specified in the generic developers guidance "Dev:elopers may scram [BWR]) did not shutdown down the reactor as indicated by include site~specific EOP criteria indicative of a/successful reactor the reactor. reactor power > 5% after any shutdown in an EAL statement, the Basis or both (e.g., a reactor AND manual trip action was initiated power level)." Reactor power::; 5% is the site-specific indication of a AND successful reactor scram. b. EITHER of the following: Added the words" ... as indicated by reactor power> 5% after any A subsequent automatic scram 1. A subsequent manual or manual scram action taken at manual trip action was initiated" to clarify that it is a failure of any action taken at the reactor 106of127 RBS EAL Comparison Matrix control consoles is the reactor control console manual trip when an actual manual trip signal has been inserted. successful in shutting (Mode Switch, Manual PBs, ARI) Combined conditions b.1 and b.2 into a single statement to simplify down the reactor. is successful in shutting down the reactor as indicated by reactor the presentation. OR power :s; 5% (APRM downscale) Mode Switch, Manual PBs, and initiation of ARI are the manual 2 A subsequent automatic (Note 8) actions taken to shut down the reactor. (trip [PWR] / scram [BWR]) is successful in shutting down the reactor. Notes Note: A manual action is any N/A Note 8: A manual action is any None operator action, or set of actions, which causes the control rods to operator action, or set of be rapidly inserted into the core, actions, which causes the control rods to be and does not include manually rapidly inserted into the driving in control rods or implementation of boron core, and does not injection strategies. include manually driving in control rods or implementation of boron injection strategies. 107of127 RBS EAL Comparison Matrix ' NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Ju stifi cation IC#(s) SU6 Loss of all onsite or offsite SU? Loss of all onsite or offsite None communications capabilities. communications capabilities. MODE: Power Operation, MODE: 1 -Power Operation, 2 -Startup, Hot Standby, Hot Startup, 3 -Hot Shutdown Shutdown NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Loss of ALL of the following SU7.1 Loss of all Table S-4 onsite Example EALs #1, 2 and 3 have been combined into a single EAL onsite communication methods: communication methods for simplification of presentation. (site-specific list of OR Replaced "ORO" with "State and local agency" for clarification communications methods) Loss of all Table S-4 State and Table S-4 provides a site-specific list of onsite, State and Local local agency communication (ORO) and NRC communications methods. 2 Loss of ALL of the following methods ORO communications methods: OR (site-specific list of Loss of all Table S-4 NRC communications methods) communication methods 3 Loss of ALL of the following NRC communications methods: (site-specific list of communications methods) 108 of 127 RBS EAL Comparison Matrix Table 5-4 Communication Methods System Onsite State/L NRC ocal Plant radio system X Plant Paging System X Sound powered phones X In-plant telephones \ X Emergency Notification System (ENS) X Commercial Telephone System X X Satellite Phones X X State of Louisiana Radio X State and Local Hotline radio X INFORM Notification System X \ 109 of 127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) SU? Failure to isolate containment or NIA NIA This IC and its associated example EALs are applicable to PWRs loss of containment pressure only and therefore not included. control. [PWR] MODE: Power Operation, Startup, Hot Standby, Hot Shutdown NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 a. Failure of containment to NIA NIA This IC and its associated example EALs are applicable to PWRs isolate when required by an only and therefore not included. actuation signal. AND b. ALL required penetrations are not closed within 15 minutes of the actuation signal. '> 2 a. Containment pressure greater than (site-specific pressure). AND ' b. Less than one full train of (site-specific system or equipment) is operating per design for 15 minutes or longer. NIA NIA NIA NIA This IC and its associated example EALs are applicable to PWRs only and therefore not included. 110 of 127 L RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification -IC#(s) SA1 Loss of all but one AC power SA1 Loss of all but one AC power "ENS buses" is the RBS-specific terminology for "emergency buses". source to emergency buses for source to ENS buses for 15 15 minutes or longer. . minutes or longer. MODE: Power Operation, / .r MODE: 1 -Power Operation, 2 -I Startup, Hot Standby, Hot -Startup, 3 -Hot Shutdown Shutdown NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 a. AC power capability to (site-SA1.1 AC power capability, Table S-1, DIV I and DIV II 4.16 KV ENS buses are the site-specific emergency specific emergency buses) is to DIV I and DIV II 4.16 KV ENS buses. reduced to a single power source buses reduced to a single power Site-specific AC power sources are listed in Table S-1. --for 15 minutes or longer. source for 15 min. (Note 1) AND AND b. Any additional single power Any additional single power source failure will result in a loss source failure will result in loss of of all AC power to SAFETY all AC power to SAFETY SYSTEMS. SYSTEMS Note The Emergency Director should N/A Note 1: The Emergency Director The classification timeliness note has been standardized across the declare the Alert promptly upon should declare the RBS EAL scheme by referencing the "time limit" specified within the determining that 15 minutes has event promptly upon EAL wording. been exceeded, or will likely be determining that the Added "The Emergency Director is not allowed an additional 15 exceeded. time limit has been exceeded, or will likely minutes to declare after the time limit is exceeded" to reinforce the be exceeded. The concept that the EAL timing component runs concurrent with the Emergency Director is classification timeliness clock. not allowed an additional 15 minutes to declare after the time limit is exceeded. 111 of 127 RBS EAL Comparison Matrix Table 5-1 AC Power Sources Offsite

1RTX-XSR1C
1RTX-XSR1D Onsite
EGS-EG1A
EGS-EG1B 112of127 RBS EAL Comparison Matrix ' NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) SA2 UNPLANNED loss of Control SA3. UNPLANNED loss of Control None Room indications for 15 minutes Room indications for 15 minutes or longer with a significant or longer with a significant transient in progress.

transient in progress. MODE: Power Operation, MODE: 1 -Power Operation, 2 -Startup, Hot Standby, Hot Startup, 3 -Hot Shutdown Shutdown NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# . 1 An UNPLANNED event results in SA3.1 An UNPLANNED event results in The site-specific Safety System Parameter list is in Table S-2. the inability to monitor one or the inability to monitor one or more of the following parameters more Table S-2 parameters from The NEI phrase "Primary Containment" has been changed to from within the Control Room for within the Control Room for~ 15 "Containment" to use terminology specific to the Mark Ill 15 minutes or longer. min. (Note 1) containment design. AND The significant transient list has been tabularized in Table S-3 for AND Any significant transient is in ease of use. ANY of the following transient progress, Table S-3 events in progress.

Automatic or manual runback greater than 25% thermal reactor power
Electrical load rejection greater than 25% full electrical load
Reactor scram [BWR] I trip [PWR] .
ECCS (SI) actuation
Thermal power oscillations

' greater than 10% [BWR] -113 of 127 RBS EAL Comparison Matrix Note The Emergency Director should NIA Note 1: The Emergency Director The classification timeliness note has been standardized across the declare the Unusual Event should declare the event RBS EAL scheme by referencing the "time limit" specified within the promptly upon determining that promptly upon EAL wording. 15 minutes has been exceeded, determining that the time Added The Emergency Director is not allowed an additional 15 or will likely be exceeded. limit has been exceeded, or will likely be minutes to declare after the time limit is exceeded" to reinforce the exceeded. The concept that the EAL timing component runs concurrent with the Emergency Director is classification timeliness clock. not allowed an additional 15 minutes to declare after the time limit is exceeded. [BWR parameter list] [PWR parameter list] Reactor Power Reactor Power RPV Water Level RCS Level RPV Pressure RCS Pressure Primary Containment Pressure In-Core/Core Exit Temperature Suppression Pool Level Levels in at least (site-specific number) steam generators Suppression Pool Temperature Steam Generator Auxiliary or Emergency Feed Water Flow 114 of 127 RBS EAL Comparison Matrix Table S-2 Safety System Parameters

Reactor power
RPV water level
RPV pressure
Containment pressure
Suppression Pool water level
Suppression Pool temperature Table S-3 Significant Transients
Reactor scram
Runback > 25% thermal reactor power
Electrical load rejection

> 25% full electrical load

ECCS injection
Thermal power oscillations>

10% 115of127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) SA5 Automatic or manual (trip [PWR] SA6 Automatic or manual scram fails Mode 2 -Startup has been included. For BWRs, including RBS, the I scram [BWR]) fails to shutdown to shut down the reactor and plant operating mode is defined by the position of the Reactor Mode the reactor, and subsequent subsequent manual actions Switch. During a normal plant startup the Reactor Mode Switch is manual actions taken at the taken at the reactor control placed in the Startup position (Startup Mode 2) as reactor power is reactor control consoles are not consoles are not successful in increased. Typically reactor power is increased to 8% before the successful in shutting down the shutting down the reactor Reactor Mode Switch is placed in Run (Power Operations Mode 1 ). reactor. MODE: 1 -Power Operation, 2 -5% reactor power (APRM downscale) is the site-specific indication of MODE: Power Operation Startup a successful reactor scram. Therefore it is appropriate to include Startup Mode 2 to failure to scram I Cs. NEI Ex. RBS ( EAL# NEI Example EAL Wording EAL# RBS EAL Wording Difference/Deviation Justification 1 a. An automatic or manual (trip SA6.1 An automatic or manual scram As specified in the generic developers guidance "Developers may [PWR] / scram [BWR]) did not fails to shut down the reactor as include site-specific EOP criteria indicative of a successful reactor shutdown the reactor. indicated by reactor power shutdown in an EAL statement, the Basis or both (e.g., a reactor AND >5% power level)." Reactor power 5% is the site-specific indication of a AND successful reactor scram. b. Manual actions taken at the Mode Switch, Manual PBs, and initiation of ARI are the manual reactor control consoles are not Manual scram actions taken at successful in shutting down the the reactor control console actions taken to shut down the reactor. reactor. (Mode Switch, Manual PBs, ARI) are not successful in shutting down the reactor as indicated by reactor power > 5% (Note 8) Notes Note: A manual action is any N/A Note 8: A manual action is any None operator action, or set of actions, which causes the control rods to operator action, or set of be rapidly inserted into the core, actions, which causes the control rods to be and does not include manually rapidly inserted into the driving in control rods or core, and does not implementation of boron injection include manually driving strategies. in control rods or 116of127 RBS EAL Comparison Matrix implementation of boron injection strategies 117 of 127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) SA9 Hazardous event affecting a SA8 Hazardous event affecting Pluralized safety systems to be consistent with NRC EP FAQ 2016-SAFETY SYSTEM needed for SAFETY SYSTEMS needed for 002 that specifies degraded performance or visible damage in more the current operating mode. the current operating mode than one safety system train. MODE: Power Operation, MODE: 1 -Power Operation, 2 -) Startup, Hot Standby, Hot Startup, 3 -Hot Shutdown Shutdown / 118of127 RBS EAL Comparison Matrix NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 a. The occurrence of ANY of SA8.1 The occurrence of any Table S-The hazardous events have been tabularized in Table S-5. the following hazardous events:

Seismic event (earthquake) 5 hazardous event SA8.1 reflects NRC FAQ 2016-002 requiring degraded
Internal or external flooding AND performance or visible damage to more than one train of a safety event system caused by the specified events.
High winds or tornado strike Event damage has caused This wording is a deviation from NEI 99-01 Revision 6 SA9
FIRE indications of degraded generic wording and bases but is deemed acceptable in order
EXPLOSION performance on one train of a SAFETY SYSTEM needed for to ensure that an Alert is declared only when a hazardous
( site-specific hazards) event causes actual or potential performance issues with
Other events with similar the current operating mode safety systems. This is consistent with NRG-approved EP hazard characteristics as AND EITHER: FAQ 2016-002.

determined by the Shift

Event damage has The word "a" is replaced with "the" in the FAQ wording to provide Manager AND caused indications of agreement with the FAQ basis information indicating that the b. EITHER of the following:

degraded performance criteria is applicable to another train of the same safety system. 1. Event damage has to the second train of caused indications of the SAFETY SYSTEM degraded performance in needed for the current at least one train of a operating mode SAFETY SYSTEM

Event damage has needed for the current resulted in VISIBLE operating mode. DAMAGE to the OR second train of the 2. The event has caused SAFETY SYSTEM VISIBLE DAMAGE to a needed for the current SAFETY SYSTEM operating mode component or structure (Notes 9, 1 O) needed for the current operating mode. 119 of 127 RBS EAL Comparison Matrix NIA NIA NIA Note 9: If the affected SAFETY Added Note 9 consistent with the recommendation of NRC EP FAQ SYSTEM train was already 2016-002.

inoperable or out of service before the hazardous event occurred, then emergency classification is not warranted NIA NIA NIA Note 10: If the hazardous event Added Note 10 consistent with the recommendation of NRC EP only resulted in VISIBLE FAQ 2016-002. DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted. 120 of 127 RBS EAL Comparison Matrix Table 5-5 Hazardous Events

Seismic event (earthquake)
Internal or external FLOODING event
High winds or tornado strike
FIRE
EXPLOSION
Other events with similar hazard characteristics as determined by the Shift Manager 121 of 127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) SS1 Loss of all offsite and all onsite SS1 Loss of all offsite and all onsite "ENS buses" is the RBS-specific terminology for "emergency buses". AC power to emergency buses AC power to ENS buses for 15 for 15 minutes or longer. minutes or longer MODE: Power Operation, MODE: 1 -Power Operation, 2 -Startup, Hot Standby, Hot Startup, 3 -Hot Shutdown Shutdown NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Loss of ALL offsite and ALL SS1.1 Loss of all offsite and all onsite DIV I and DIV II 4.16 KV ENS buses are the site-specific emergency onsite AC power to (site-specific AC power capability to DIV I buses. emergency buses) for 15 minutes and DIV II 4.16 KV ENS buses or longer. for~ 15 min. (Note 1) -Note The Emergency Director should N/A Note 1: The Emergency The classification timeliness note has been standardized across the declare the Unusual Event Director should declare RBS EAL scheme by referencing the "time limit" specified within the promptly upon determining that the event promptly EAL wording. 15 minutes has been exceeded, upon determining that Added "The Emergency Director is not allowed an additional 15 or will likely be exceeded.

the time limit has been exceeded, or will likely minutes to declare after the time limit is exceeded" to reinforce the be exceeded. The concept that the EAL timing component runs concurrent with the Emergency Director is classification timeliness clock. not allowed an additional 15 minutes to declare after the time limit is exceeded. 122 of 127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) SS5 Inability to shutdown the reactor SS6 Inability to shut down the Mode 2 -Startup has been included. For BWRs, including RBS, the causing a challenge to ( core reactor causing a challenge to plant operating mode is defined by the position of the Reactor Mode cooling [PWR] I RPV water level RPV water level or RCS heat Switch. During a normal plant startup the Reactor Mode Switch is [BWR]) or RCS heat removal. removal placed in the Startup position (Startup Mode 2) as reactor power is MODE: Power Operation MODE: 1 -Power Operation, 2 -increased. Typically reactor power is increased to 8% before the Startup Reactor Mode Switch is placed in Run (Power Operations Mode 1 ). 5% reactor power (APRM downscale) is the site-specific indication of a successful reactor scram. Therefore it is appropriate to include Startup Mode 2 to failure to scram !Cs. NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 a. An automatic or manual (trip SS6.1 An automatic or manual scram As specified in the generic developers guidance "Developers may [PWR] / scram [BWR]) did not fails to shut down the reactor as include site-specific EOP criteria indicative of a successful reactor shutdown the reactor. indicated by reactor power shutdown in an EAL statement, the Basis or both (e.g., a reactor AND power level)." Reactor power::; 5% is the site-specific indication of a b. All manual actions to >5% successful reactor scram. shutdown the reactor have AND Deleted the term "manual actions" from the second condition. For been unsuccessful. AND All actions to shut down the generic IC SS5, all actions to shut down the reactor can be credited, c. EITHER of the following reactor are not successful as including emergency boration which is not considered a "manual" conditions exist: indicated by reactor power > 5% scram action. * (Site-specific indication of AND EITHER: Indication of an inability to adequately remove heat from the core an inability to adequately RPV water level cannot occurs when RPV water level cannot be restored and maintained remove heat from the core) be restored and above -187 in., which is the EOP RPV water level indicative of a * (Site-specific indication of maintained > -187 in. loss of adequate core cooling. an inability to adequately --Indication of an inability to adequately remove heat from the RCS remove heat from the RCS) OR occurs when parameters cannot be restored and maintained within Heat Capacity the safe region of the HCTL curve. Temperature Limit -(HCTL) exceeded (EOP Figure 2) 123of127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) SSB Loss of all Vital DC power for 15 SS2 Loss of all vital DC power for 15 None minutes or longer. minutes or longer. MODE: Power Operation, MODE: 1 -Power Operation, 2 -Startup, Hot Standby, Hot Startup, 3 -Hot Shutdown Shutdown NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Indicated voltage is less than SS2.1 Indicated voltage is < 105 voe 105 VDC is the site-specific minimum vital DC bus voltage. (site-specific bus voltage value) on Safety Related DIV I and DIV Safety Related DIV I and DIV II 125 VDC buses are the site-specific on ALL (site-specific Vital DC 11 125 voe buses for credited vital DC buses. busses) for 15 minutes or longer. 15 min. (Note 1) Note The Emergency Director should NIA Note 1: The Emergency The classification timeliness note has been standardized across the declare the Unusual Event Director should declare the RBS EAL scheme by referencing the "time limit" specified within the promptly upon determining that event promptly upon EAL wording. 15 minutes has been exceeded, determining that the time limit Added "The Emergency Director is not allowed an additional 15 or will likely be exceeded. has been exceeded, or will likely minutes to declare after the time limit is exceeded" to reinforce the be exceeded. The Emergency concept that the EAL timing component runs concurrent with the Director is not allowed an classification timeliness clock. additional 15 minutes to declare after the time limit is exceeded. 124of127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) SG1 Prolonged loss of all offsite and SG1a Prolonged loss of all offsite and "ENS buses" is the RBS-specific terminology for "emergency buses". all onsite AC power to all onsite AC power to ENS emergency buses. buses MODE: Power Operation, MODE: 1 -Power Operation, Startup, Hot Standby, Hot 2 -Startup, 3 -Hot Shutdown Shutdown NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 a. Loss of ALL offsite and ALL SG1.1 Loss of all offsite and all onsite DIV I and DIV II 4.16 KV ENS buses are the site-specific onsite AC power to (site-specific AC power capability to DIV I and emergency buses. emergency buses). DIV II 4.16 KV ENS buses 4 hours is the site-specific SBO coping analysis time. AND AND EITHER: Indication of an inability to adequately remove heat from the core b. EITHER of the following:

Restoration of at least one occurs when RPV water level cannot be restored and maintained 4.16 KV ENS bus in< 4 above -187 in., which is the EOP RPV water level indicative of a
Restoration of at least hours is not likely (Note 1) loss of adequate core cooling . one AC emergency bus in less than (site-specific
RPV water level cannot hours) is not likely. be restored and maintained>

-187 in. * (Site-specific indication of an inability to adequately remove heat from the ' core) -Note The Emergency Director should N/A Note 1: The Emergency The classification timeliness note has been standardized across the declare the General Emergency Director should declare RBS EAL scheme by referencing the "time limit" specified within the promptly upon determining that the event promptly EAL wording. (site-specific hours) has been upon determining that Added "The Emergency Director is not allowed an additional 15 exceeded, or will likely be the time limit has been exceeded. exceeded, or will likely minutes to declare after the time limit is exceeded" to reinforce the be exceeded. The concept that the EAL timing component runs concurrent with the Emergency Director is classification timeliness clock. not allowed an 125 of 127 RBS EAL Comparison Matrix additional 15 minutes to declare after the time limit is exceeded. 126 of 127 RBS EAL Comparison Matrix NEI IC# NEI IC Wording RBS RBS IC Wording Difference/Deviation Justification IC#(s) SGB Loss of all AC and Vital DC SG1b Loss of all ENS AC and vital DC "ENS AC" is the RBS-specific terminology for "emergency AC". power sources for 15 minutes or power sources for 15 minutes or longer. longer MODE: Power Operation, MODE: 1 -Power Operation, 2 -Startup, Hot Standby, Hot Startup, 3 -Hot Shutdown Shutdown NEI Ex. NEI Example EAL Wording RBS RBS EAL Wording Difference/Deviation Justification EAL# EAL# 1 a. Loss of ALL offsite and ALL SG1.2 Loss of all offsite and all onsite DIV I and DIV II 4.16 KV ENS buses are the site-specific emergency onsite AC power to (site-specific AC power capability to DIV I buses. emergency buses) for 15 minutes and DIV II 4.16 KV ENS buses 105 VDC is the site-specific minimum vital DC bus voltage. or longer. for~ 15 min. (Note 1) AND AND Safety Related DIV I and DIV II 125 VDC buses are the safety-related DC buses that are the credited.

b. Indicated voltage is less than Indicated voltage is< 105 voe (site-specific bus voltage value) on Safety Related DIV I and on ALL (site-specific Vital DC DIV II 125 VDC buses for busses) for 15 minutes or longer. 15 min. (Note 1) Note The Emergency Director should N/A Note 1: The Emergency The classification timeliness note has been standardized across the declare the Unusual Event Director should declare the RBS EAL scheme by referencing the "time limit" specified within the promptly upon determining that 15 event promptly upon EAL wording. minutes has been exceeded, or determining that the time limit Added "The Emergency Director is not allowed an additional 15 will likely be exceeded.

has been exceeded, or will minutes to declare after the time limit is exceeded" to reinforce the likely be exceeded. The concept that the EAL timing component runs concurrent with the Emergency Director is not classification timeliness clock. allowed an additional 15 minutes to declare after the time limit is exceeded. 127of127}}

ML18128A052

Text

SUMMARY

6.0 REFERENCES

1.0 INTRODUCTION

4.0 REFERENCES

1.0 INTRODUCTION

4.0 REFERENCES

4.0 REFERENCES

1.0 INTRODUCTION

4.0 REFERENCES

Navigation menu

Search