ML15296A060

Enclosure 3, Revised (Clean) Copy of the SSES EAL Basis Document Which Includes the Changes Made in Enclosure 2

ML15296A060
Person / Time
Site:	Susquehanna
Issue date:	10/16/2015
From:	Franke J Susquehanna, Talen Energy
To:	Office of Nuclear Reactor Regulation
Shared Package
ML15296A057	List: ML15296A059 ML15296A060 PLA-7393, Submittal of Response to NRC Request for Additional Information Regarding License Amendment Request to Adopt NEI 99-01, Revision 6, Emergency Action Level Scheme Change
References
PLA-7393, TAC MF6057, TAC MF6058 EP-RM-004, Rev. X
Download: ML15296A060 (282)

Text

Enclosure 3 to PLA-7393 Revised (Clean) Copy of the SSES EAL Basis Document which Includes the Changes Made in Enclosure 2

PROCEDURE COVER SHEET SUSQUEHANNA, LLC PROCEDURE 02/13/2015 EAL CLASSIFICATION BASES (CLEAN COPY) EP-RM-004 Revision [X]

Page 1 of 281 ADHERENCE LEVEL: INFORMATION USE QUALITY CLASSIFICATION: APPROVAL CLASSIFICATION:

(X) QA Program ( ) Non-QA Program (X) Plant ( ) Non-Plant

( ) Instruction EFFECTIVE DATE: ________

PERIODIC REVIEW FREQUENCY: 2 Year PERIODIC REVIEW DUE DATE: REWL X0200 RECOMMENDED REVIEWS: All Procedure Owner: Emergency Planning Responsible Supervisor: Manager-EP Responsible FUM: Manager-EP Responsible Approver: Manager-NRA

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Page 2 of 281 TABLE OF CONTENTS SECTION PAGE 1.0 PURPOSE 3 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

13 4.1 Developmental 13 4.2 Implementing 13 5.0 DEINITIONS, ACRONYMS & ABBREVIATIONS 14 6.0 Susquehanna TO NEI 99-01 Rev. 6 EAL CROSS-REFERENCE 23 7.0 ATTACHMENTS 27 1 EAL Bases 28 Categqory R Abnormal Rad Release /Rad Effluent 28 Catecqory C Cold Shutdown / Refueling System Malfunction 75 Categqory H Hazards 122 Categqory S System Malfunction 159 Catecqory E ISFSI 206 Categqory F Fission Product Barrier Degradation 208 2 Fission Product Barrier Loss / Potential Loss "

Matrix and Bases 213 3 Safe Shutdown Room/Areas Tables R-2 & H-2 Bases 273 4 Table R - Abnormal Rad Levels / Rad Effluents (Form EP-RM-004-R) 276 5 Table C - Cold Shutdown/Refueling System Malfunctions (Form EP-RM-004-C) 277 6 Table H - Hazards (Form EP-RM-004-H) 278 7 Table S - System Malfunctions (Form EP-RM-004-S) 279 8 Table E -ISFSI (Form EP-RM-004-E) 280 9 Table F - Fission Product Barrier Degradation (Form EP-RM-004-F) 281

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Page 3 of 281 CAUTION This Reference Manual shall follow the process described in NDAP-QA-0004, Procedure Change Process, for subsequent revisions AND NOT the change process described under EP-112, Emergency Plan Reference Manual Program. It must be maintained in accordance with 10 CFR50.54(q).

1.0 PURPOSE This document provides an explanation and rationale for each Emergency Action Level (EAL) included in the EAL Upgrade Project for Susquehanna, LLC. It should be used to facilitate review of the Susquehanna EALs and provide historical documentation for future reference.

Decision-makers responsible for implementation of EP-PS-1 00, Emergency Director Control Room (ref. 4.2.1), EP-PS-101, TSC Emergency Director (ref. 4.2.2) and EP-PS-200, Recovery Manager (ref. 4.2.3), may use this document as a technical reference in support of EAL interpretation. This information may assist the Emergency Director/Recovery Manager in making classifications, particularly those involving judgment or multiple events. The bases 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.

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 Susquehanna 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 (ISFSIs).

• 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 (ADAMS Accession Number ML12326A805) (ref.

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Page 4 of 281 4.1.1), Susquehanna conducted an EAL implementation upgrade project that produced the EALs discussed herein.

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 an increased probability of barrier loss and decreased certainty of maintaining the barrier.

The primary fission product barriers are:

A. Fuel Clad (FC): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets.

B. Reactor Coolant System (RCS): The RCS Barrier is the reactor coolant system pressure boundary and includes the Reactor Pressure Vessel (RPV) and all reactor coolant system piping up to and including the isolation valves.

C. Primary Containment (PC): The Primary Containment Barrier includes the drywell, the wetwell, their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves. Primary Containment Barrier thresholds are used as criteria for escalation of the Emergency Classification Level (ECL) 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 Emergqency:

Loss or potential loss of any two barriers General Emergency:

Loss of any two barriersand loss or potentialloss of the third barrier

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Page 5 of 281 2.4 EAL Organization The Susquehanna EAL scheme includes the following features:

• Division of the EAL set into three broad groups:

o EALs applicable under all plant operating modes - This group would be reviewed by the EAL-user any time emergency classification is considered.

o EALs applicable only under ho._t 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 Susquehanna EAL categories are aligned to and represent the NEI 99-01"Recognition Categories." Subcategories are used in the Susquehanna scheme as necessary to further divide the EALs of a category into logical sets of possible emergency classification thresholds. The Susquehanna EAL categories and subcategories are listed in Table 2.4-1.

The primary tool for determining the emergency classification level is the EAL Classification Matrix. The bases provide the EAL user with the background and justification behind the EAL threshold values identified using the guidance set forth in NEI 99-01 Revision 6. If there is any doubt with regard to the applicability of any EAL, the technical basis should be reviewed. The user should consult Section 3.0 and Attachments 1 (EAL technical bases) & 2 (fission product barrier technical bases) of this document for such information.

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Page 6 of 281 Table 2.4-1 EAL Groups, Categories and Subcategories All Operatingi Mode:

R - Abnormal Rad Levels / Rad 1 - Radiological Effluent 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 - ED/RM Judgment E - Independent Spent Fuel Storage 1 - Confinement Boundary Installation (ISFSI)

Hot Conditions:

S - System Malfunction 1 - Loss of Essential 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 None Degradation Cold Conditions:

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

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Page 7 of 281 2.5 Technical Bases Information EAL technical bases are provided in Attachment 1 for each EAL according to EAL group (All, Hot, Cold), EAL category (R, C, H, 5, F and E) 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:

Categqory Letter & Title Subcateqory Number & Title Initiatinq Condition (IC)

Site-specific description of the generic IC given in NEI 99-01 Rev. 6.

EAL Identifier (enclosed in rectangqle)

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 (R, C, H, S, F or E)

2. Second character (letter): The emergency classification (G, 5, 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 rectangqle):

Unusual Event (U), Alert (A), Site Area Emergency (5) or General Emergency (G)

EAL (enclosed in rectanqle)

Exact wording of the EAL as it appears in the EAL Classification Matrix Mode Applicability One or more of the following plant operating conditions comprise the mode to which each EAL is applicable: 1 - Power Operations, 2 - Startup, 3 - Hot Shutdown, 4 - Cold Shutdown, 5 - Refueling, D - 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. Defined terms used in Initiating Condition and Emergency Action Level statements are set in all capital letters (e.g., ALL CAPS).

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Page 8 of 281 Basis:

A basis section that provides Susquehanna-relevant information concerning the EAL. This section incorporates the basis that provides a description of the rationale for the EAL as provided in NEI 99-01 Rev. 6.

Susquehanna Basis Reference(s):

Site-specific source documentation from which the EAL is derived 2.6 Operating Mode Applicability (ref. 4.1.2 except Defueled) 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 (with one or more reactor vessel head closure bolts less than fully tensioned) and average reactor coolant temperature is >200°F 4 Cold Shutdown The mode switch is in SHUTDOWN (with one or more reactor vessel head closure bolts less than fully tensioned) and average reactor coolant temperature is < 200°F 5 Refuelincq The mode switch is in REFUEL or SHUTDOWN with one or more reactor vessel head closure bolts are less than fully tensioned D Defueled.

All fuel removed from the reactor vessel (i.e., full core offload during refueling or extended outage) (ref. 4.1.1)

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.

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Page 9 of 281 3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS 3.1 General Considerations When making an emergency classification, the Emergency Director/Recovery Manager must consider all information having a bearing on the proper assessment of an Initiating Condition (IC). This includes the Emergency Action Level (EAL) plus the associated Operating Mode Applicability, Notes, and the informing basis information. In the Recognition Category F matrices, EALs are based on loss or potential loss of Fission Product Barrier Thresholds.

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-lSG-01, "Interim Staff Guidance, Emergency Planning for Nuclear Power Plants" (ref. 4.1.3).

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 indicator operability, the condition existence, or the report 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/Recovery Manager 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.

3.1.4 Planned vs. Unplanned Events A planned work activity that results in an expected event or condition which meets or exceeds an EAL does not warrant an emergency declaration provided that: 1) the activity proceeds as planned, and 2) the plant remains within the limits imposed by the operating license. Such activities include planned work to test, manipulate, repair, maintain or modify a system or component. In these cases, the controls associated with the planning, preparation and execution of the work will ensure that compliance is maintained with all aspects of the operating license provided that the activity proceeds and concludes as expected. Events or conditions of this type may be subject to the reporting requirements of 10 § CFR 50.72 (ref. 4.1.4).

/

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Page 10 of 281 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/Recovery Manager 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/Recovery Manager 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/Recovery Manager 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., 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 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. 4.1.3).

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 exam pie:

• If an Alert EAL and a Site Area Emergency EAL are met, whether at one unit or at both 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 both units, an Alert should be declared.

Related guidance concerning classification of rapidly escalating events or conditions is provided in Regulatory Issue Summary (RIS) 2007-02, Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events (ref. 4.1.5).

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Page 11 of 281 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/Recovery Manager 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/Recovery Manager, 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 Downgrading An ECL may be downgraded when the event or condition that meets the highest IC and EAL no longer exists, and other site-specific downgrading requirements are met. If downgrading the ECL is deemed appropriate, the new ECL would then be based on a lower applicable IC(s) and EAL(s). The ECL may also simply be terminated.

As noted above, guidance concerning classification of rapidly escalating events or conditions is provided in RIS 2007-02 (ref. 4.1.5).

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 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 ICs 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.

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Page 12 of 281 EAL momentarily met duringq expected plant response - In instances where an EAL is briefly met during an expected (normal) plant response, an emergency declaration is not warranted provided that associated systems and components are operating as expected, and operator actions are performed in accordance with procedures.

EAL momentarily met but the condition is corrected prior to an emergency declaration - If an operator takes prompt manual action to address a condition, and the action is successful in correcting the condition prior to the emergency declaration, then the applicable EAL is not considered met and the associated emergency declaration is not required. For illustrative purposes, consider the following example:

An ATWS occurs and the high pressure ECCS systems fail to automatically start. RPV level rapidly decreases and 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 ATWS 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/Recovery Manager 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.

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-1 022 (ref. 4.1.6) 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. 4.1.6).

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4.0 REFERENCES

4.1 Developmental 4.1.1 NEI 99-0 1 Revision 6, Methodology for the Development of Emergency Action Levels for Non-Passive Reactors, ADAMS Accession Number ML12326A805 4.1.2 Technical Specifications Table 1.1-1 Modes 4.1.3 NSIR/DPR-ISG-01 Interim Staff Guidance, Emergency Planning for Nuclear Power Plants 4.1.4 10 § CFR 50.72 Immediate Notification Requirements for Operating Nuclear Power Reactors 4.1.5 RIS 2007-02 Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events, February 2, 2007 4.1.6 NUREG-1 022 Event Reporting Guidelines: 10CFR50.72 and 50.73 4.1.7 NUH-003 Updated Final Safety Analysis Report for the Standardized NUHOMS Horizontal Modular Storage System for Irradiated Nuclear Fuel, Section 1.3.1 4.1.8 NDAP-QA-0309 Primary Containment Access and Control 4.1.9 NDAP-QA-0321 Secondary Containment Integrity Control 4.1.10 Susquehanna LLC, Susquehanna Steam Electric Station Emergency Plan, Section 1.0, Definitions 4.1.11 10 § CFR 50.73 License Event Report System 4.1.12 ON-FPC-101(201) Loss of Fuel Pool Cooling 4.1 .13 EO-000-1 04 Secondary Containment Control 4.1.14 FSAR Section 3.7a Seismic Design 4.2 Implementing 4.2.1 EP-PS-100 Emergency Director Control Room 4.2.2 EP-PS-101 TSC Emergency Director 4.2.3 EP-PS-200 Recovery Manager 4.2.4 NEI 99-01 Rev. 6 to Susquehanna EAL Comparison Matrix

EP-RM-004 Revision [XI Page 14 of 281 5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS 5.1 Definitions (ref. 4.1.1 except as noted)

Selected defined 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 limited to small fractions of the EPA PAG exposure levels.

CAN/CANNOT BE MAINTAINED ABOVE/BELOW The value of an identified parameter is/is not able to be held within the specified limit. The determination requires an evaluation of system performance and availability in relation to parameter values and trends. An instruction prescribing action when a parameter cannot be maintained above or below a specified limit neither requires nor prohibits anticipatory action-depending upon plant conditions, the action may be taken as soon as it is determined that the limit will ultimately be exceeded, or delayed until the limit is actually reached. Once the parameter does exceed the limit, however, the action must be performed; it may not be delayed while attempts are made to restore the parameter to within the desired control band.

CAN/CANNOT BE RESTORED ABOVE/BELOW The value of an identified parameter is/is not able to be brought within the specified limit. The determination requires an evaluation of system performance and availability in relation to parameter values and trends. An instruction prescribing action when a value CANNOT BE RESTORED AND MAINTAINED above or below a specified limit does not require immediate action simply because the current values is outside the range, but does not permit extended operation beyond the limit; the action must be taken as soon as it is apparent that the specified range cannot be attained.

CONFINEMENT BOUNDARY The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the Susquehanna ISFSI, Confinement Boundary is defined as the Dry Shielded Canister (DSC) (Ref. 4.1.7).

CONTAINMENT CLOSURE The procedurally defined conditions or actions taken to secure Primary or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.

As applied to Susquehanna, CONTAINMENT CLOSURE is established per NDAP-QA-0309 (ref 4.1.8) for Primary Containment OR is established per NDAP-QA-0321 (ref 4.1.9) for Secondary Containment.EMERGENCY PLAN BOUNDARY (EPB) (ref. 4.1.10)

Same as the Exclusion Area Boundary, i.e., that area around SSES within a radius of 1800 feet determined in accordance with 10 CFR 100.11.

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Page 15 of 281 EMERGENCY ACTION LEVEL (EAL)

A pre-determined, site-specific, observable threshold for an Initiating Condition that, when met or exceeded, places the plant in a given emergency classification level.

EMERGENCY CLASSIFICATION LEVEL (ECL)

One of a set of names or titles established by the US Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions according to (1) potential or actual effects or consequences, and (2) resulting onsite and offsite response actions. The emergency classification levels, in ascending order of severity, are:

• Notification of Unusual Event (NOUE) - synonymous with Unusual Event (UE)

• Alert

• Site Area Emergency (SAE)

• General Emergency (GE)

EPA PAGs Environment Protection Agency Protective Action Guidelines. The EPA PAGs are expressed in terms of dose commitment: 1 Rem TEDE or 5 Rem ODE Thyroid. Actual or projected offsite exposures in excess of the EPA PAGs require Susquehanna to recommend protective actions for the general public to offsite planning agencies.

The dose program complies with the "Manual of Protective Action Guides and Protective Actions for Nuclear Incidents," (EPA-400), adopting the dose calculation methodology I ICRP

1. 26/30. The accident dose assessments are based on the adult physiology per EPA 400, except for one case - that is, child thyroid dose conversion factors are used in calculating thyroid ODE. However adult physiology is used in calculating thyroid ODE for purposes of evaluating the need for a sheltering only PAR and evaluating controlled venting of containment.

Calculations of TEDE are made using the (adult) dose factors provided in EPA-400.

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.

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Page 16 of 281 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 An act toward Susquehanna 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 Susquehanna. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA).

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

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

INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI)

A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage.

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 measures such as temporary shielding, SCBAs or dose extensions beyond Emergency Plan RWP that are not routinely employed to access the room/area).

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.

INTRUSION The act of entering without authorization. Discovery of a bomb in a specified area is indication of intrusion into that area by a hostile force.

MAINTAIN Take appropriate action to hold the value of an identified parameter within specified limits.

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Page 17 of 281 NORMAL LEVELS As applied to radiological IC/EALs, the highest reading in the past twenty-four hours excluding the current peak value.

NOTIFICATION OF UNUSUAL EVENT - synonymous with Unusual Event (UE)

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.

OPERATING BASIS EARTHQUAKE (OBE)

An earthquake which, considering the regional and local geology, and seismology and specific characteristics of local subsurface material, could reasonably be expected to affect the plant site during the operating life of the plant. It is that earthquake which produces the vibratory ground motion for which these features of the nuclear power plant necessary for continued operation without undue risk to the health and safety of the public are designed to remain functional (ref.

4.1.14).

OWNER CONTROLLED AREA (ref. 4.1.10)

Includes the area within the expanded security perimeter, i.e., the areas that are bordered by the Vehicle Barriers System. The OWNER CONTROLLED AREA also includes the Monitored OWNER CONTROLLED AREA (MOCA) as defined in Security Procedures.

PROJECTILE An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.

PROTECTED AREA (ref. 4.1 .10)

Area within the station inner security fence (PROTECTED AREA Barrier) designated to implement the requirements of 10 CFR 73.

RCS INTACT The RCS should be considered intact when the RCS pressure boundary is in its normal condition for the cold shutdown mode of operation (e.g., no freeze seals). REFUELING PATHWAY The reactor refueling cavity, spent fuel pool and fuel transfer canal comprise the refueling pathway (ref. 4.1.12).

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 safety-related (as defined in 10CFR50.2):

Those structures, systems and components that are relied upon to remain functional during a'nd following design basis events to assure:

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Page 18 of 281 (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.

SAFE SHUTDOWN EARTHQUAKE (SSE)

An earthquake which is based upon an evaluation of the maximum earthquake potential considering the regional and local geology, and seismology and specific characteristics of local subsurface material. It is that earthquake which produces the maximum vibratory ground motion for which Seismic Category I systems and components are designed to remain functional (ref 4.1.14).

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 BOUNDARY The line beyond which the land is not owned, leased or otherwise controlled by the licensee (Susquehanna drawing C243786, Sh 1, "Site Facilities and Boundary Map.") (ref. 4.1.10).

UNISOLABLE An open or breached system line that cannot be isolated, remotely or locally.

• The term UNISOLABLE also includes any decision by plant staff or procedure direction to not isolate a primary system.

• Normal leakage past a closed isolation valve is not considered UNISOLABLE leakage.

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.

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.

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Page 19 of 281 VISIBLE DAMAGE Damage to a component or structure that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure.

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Page 20 of 281 5.2 Abbreviations/Acronyms

°F .................................................................................... Degrees Fahrenheit

°.............................................................................................. Degrees 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 CS ............................................................................................ Core Spray DBA .............................................................................. Design Basis Accident DC ......................................................................................... Direct Current EAL............................................................................. Emergency Action Level ECCS................................................................ Emergency Core Cooling System ECL .................................................................... Emergency Classification Level ED ................................................................................... Emergency Director EDST ............................................................... Evaporator Distillate Sample Tank EOF..................................................................... Emergency Operations Facility EOP.................................................................. Emergency Operating Procedure EPA.................................................................. Environmental Protection Agency EPB ......................................................................... Emergency Plan Boundary EPG................................................................... Emergency Procedure Guideline EPIP ....................................................... Emergency Plan Implementing Procedure ESFE......................................................................... Engineered Safety Feature ESS ................................................................... Engineered Safeguards System FAA .................................................................... Federal Aviation Administration FBI ..................................................................... Federal Bureau of Investigation FEMA ...................................................... Federal Emergency Management Agency FPRR.................................................................... Fire Protection Review Report FSAR...................................................................... Final Safety Analysis Report GE .................................................................................. General Emergency HCTL ................................................................ Heat Capacity Temperature Limit HPCI................................................................... High Pressure Coolant Injection IC ...................................................................................... Initiating Condition iPEEE...................... Individual Plant Examination of External Events (Generic Letter 88-20)

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

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Page 21 of 281 LER.............................................................................. Licensee Event Report LOCA.......................................................................... Loss of Coolant Accident LRW .................................................................................... Liquid Radwaste LRW................................................................................ Light Water Reactor MPC .................................. Maximum Permissible Concentration/Multi-Purpose Canister MPH ...................................................................................... Miles Per Hour MSIV........................................................................ Main Steam Isolation Valve MSL .................................................................................... Main Steam Line mR, mRem, mrem, mREM ........................................... milli-Roentgen Equivalent Man MW ............................................................................................. Megawatt NEI ............................................................................. Nuclear Energy Institute NESP.......................................................... National Environmental Studies Project NPP ................................................................................ Nuclear Power Plant NRC................................................................... Nuclear Regulatory Commission NSSS ................................................................... Nuclear Steam Supply System NORAD .............................................. North American Aerospace Defense Command (NO)UE................................................................... Notification of Unusual Event OBE........................................................................ Operating Basis Earthquake OCA ............................................................................. Owner Controlled Area ODCM/O DAM....................................... Off-site Dose Calculation (Assessment) Manual O RO .................................................................... Offsite Response Organization PA ........................................................................................ Protected Area PRA/PSA ......................... Probabilistic Risk Assessment / Probabilistic Safety Assessment PWR........................................................................ Pressurized Water Reactor PSIG................................................................... Pounds per Square Inch Gauge R................................................................................................. Roentgen RB ...................................................................................... Reactor Building RCIC.................................................................... Reactor Core Isolation Cooling RCS............................................................................ Reactor Coolant System Rem, rem, REM ............................................................ Roentgen Equivalent Man RETS................................................... Radiological Effluent Technical Specifications RHR............................................................................. Residual Heat Removal RM ................................................................................... Recovery Manager RPS ......................................................................... Reactor Protection System RPV ........................................................................... Reactor Pressure Vessel RRC............................................................................... Reactor Recirculation RWCU .......................................................................... Reactor Water Cleanup SAR.............................................................................. Safety Analysis Report SBO ..................................................................................... Station Blackout SCBA ............................................................ Self-Contained Breathing Apparatus

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Page 22 of 281 SDHR............................................................. Supplemental Decay Heat Removal SGTS ................................................................ Stand-By Gas Treatment System SPDS................................................................ Safety Parameter Display System SRO ........................................................................... Senior Reactor Operator SSE........................................................................ Safe Shutdown Earthquake SSES ............................................................ Susquehanna Steam Electric Station SW ......................................................................................... Service Water TEDE .................................................................. Total Effective Dose Equivalent TAF ................................................................................... Top of Active Fuel TRM................................................................... Technical Requirements Manual TSC........................................................................... Technical Support Center

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Page 23 of 281 6.0 Susquehanna-TO-NEI 99-01 Rev. 6 EAL CROSS-REFERENCE This cross~reference is provided to facilitate association and location of a Susquehanna EAL within the NEI 99-01 IC/EAL identification scheme. Further information regarding the development of the Susquehanna EALs based on the NEI guidance can be found in the EAL Comparison Matrix.

NEI 99-01 Rev. 6 Susquehanna EAL ICExample IC EAL RUI.1 AU1 1, 2 RU1.2 AU1 3 RU2.1 AU2 1 RA1.1 AA1 1 RA1.2 AA1 2 RA1.3 AA1 3 RA1.4 AA1 4 RA2.1 AA2 1 RA2.2 AA2 2 RA2.3 AA2 3 RA3.1 AA3 1 RA3.2 AA3 2 RSl.1 AS1 1 RS1.2 AS1 2 RSl.3 AS1 3 RS2.1 AS2 1 RG1.1 AG1 1 RG1.2 AG1 2 RG1.3 AG1 3 RG2.1 AG2 1 CU1.1 CU1 1

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Page 24 of 281 NEI 99-01 Rev. 6 Susquehanna EAL ICExample IC EAL CU1.2 CU1 2 CU2.1 CU2 1 CU3.1 CU4 1 CU4.1 CU3 1 CU4.2 CU3 2 CU4.1 CU4 1 CU5.1 CU5 1, 2, 3 CA1. 1 CA1l CA1.2 CA1 2 CA2.1 CA2 1 CA4.1 CA3 1, 2 CA6.1 CA6 1 CS1.1 CS1 1 CS1.2 CS1 2 0S1.3 0S1 3 CG1.1 CG1 1 CG1.2 CG1 2 FAI.1 FA1 1 FS1.1 FS1 1 FGI.1 FG1 1 HU1.1 HU1 1,2 3 HU2.1 HU2 1 HU3.1 HU3 1 HU3.2 HU3 2 HU3.3 HU3 3

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Page 25 of 281 NEI 99-01 Rev. 6 Susquehanna EAL ICExample EAL HU3.4 HU3 4 N/A HU3 5 HU4.1 HU4 1 HU4.2 HU4 2 HU4.3 HU4 3 HU4.4 HU4 4 HU7.1 HU7 1 HA1.1 HA1 1, 2 HA5.1 HA5 1 HA6.1 HA6 1 HA7.1 HA7 1 HS1.1 HS1 1 HS6.1 HS6 1 HS7.1 HS7 1 HG1.1 HG1 1 HG7.1 HG7 1 SU1.1 SU1 1 SU3.1 SU2 1 SU4.1 SU3 1 SU4.2 SU3 2 SU5.1 SU4 1, 2, 3 SU6.1 SU5 1 SU6.2 SU5 2 SU7.1 SU6 1, 2, 3 SA1.1 SAl 1

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Page 26 of 281 NEI 99-01 Rev. 6 Susquehanna EAL ICExample IC EAL SA3.1 SA2 1 SA6.1 SA5 1 SA8.1 SA9 1 SSl.1 SSl1 SS2.1 SS8 1 SS6.1 SS5 1 SG1.1 SGI 1 SG1.2 SG8 1 EU1.1 E-HU1 1

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Page 27 of 281 7.0 ATTACHMENTS 7.1 Attachment 1, Emergency Action Level Technical Bases 7.2 Attachment 2, Fission Product Barrier Matrix and Bases 7.3 Attachment 3, Safe Shutdown Room/Areas Tables R-2 & H-2 Bases 7.4 Attachment 4, Table R - Abnormal Rad Levels / Rad Effluents (Form EP-RM-004-R) 7.5 Attachment 5, Table E - ISFSI (Form EP-RM-004-E) 7.6 Attachment 6, Table H - Hazards (Form EP-RM-004-H) 7.7 Attachment 7, Table S - System Malfunctions (Form EP-RM-004-S) 7.8 Attachment 8, Table F - Fission Product Barrier Degradation (Form EP-RM-004-F) 7.9 Attachment 9, Table C - Cold Shutdown/Refueling System Malfunctions (Form EP-RM-004-C)

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Page 28 of 281 ATTACHMENT 1 Emergency Action Level Technical Bases Category R - Abnormal Rad Release I 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 the 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. Radioloq~ical 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 1 of 186

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Page 29 of 281 Category: R - 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 child thyroid CDE

• EAL:

RGtI1 General Emergency Gaseous effluent > Table R-1 column "GE" for >15 min. (Notes 1, 2, 3, 4)

Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be 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, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.

Note 4: The pre-calculated effluent monitor values presented in EALs RAI .1, RSI .1 and RG1 .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Table R-1 Effluent Monitor Classification Thresholds (Note 4)-______

Release Point [ Monitor ] GE j SAE ] Alert ] UE oPlant Vent 00630 1 .9E+09 1 .9E+08 1 .9E+07 4.0E+06 nbegs)067jCi/min p IpCi/min pCi/min pCi/min (olga)067(site total) (site total) (site total) (site total)

LRW RR-06433 ......... 2 x hi alarm

'51(2) RHRSW A/B D.... ..... 2 xhialarm o" 1 (2)R606 1(2) SW/SDHR 1RR-D12-(2)R606 ........... 2 xhi alarm Mode Applicability:

All Definition(s):

None Page 2 of 185

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Page 30 of 281 Basis:

This EAL address gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to EMERGENCY PLAN BOUNDARY doses that exceed either (ref. 1):

• 1000 mrem TEDE

• 5000 mrem ODE Child Thyroid The SITE BOUNDARY is irregularly shaped and therefore would result in dose projections and protective action recommendations that can vary significantly depending on plume direction and affected sector. Using dose projections calculated using the EPB (i.e. EXCLUSION AREA) provides a more consistent approach to Public Protective Action Recommendations since the EPB is more consistently defined in all directions. The EPB is at or within the SITE BOUNDARY in all compass sectors.

The EPA PAGs are expressed in terms of the projected sum of the effective dose equivalent (EDE) from external radiation and the committed effective dose equivalent (CEDE) incurred from inhalation of radioactive materials, or as the committed dose equivalent (CDE) to the thyroid. For the purpose of these IC/EALs, the projected dose quantity Total Effective Dose Equivalent (TEDE), as defined in 10 CFR 20 is used in lieu of "the projected sum of...EDE...and CEDE...", with CEDE considering significant dose from inhaled radionuclides during the early phase of the event. The EPA protective action guidance provides for the use of adult thyroid dose conversion factors. However, the Commonwealth of Pennsylvania requires the use of child thyroid CDE for purposes of comparison of projected thyroid CDE to the PAG for thyroid CDE.

The monitor reading threshold for RGI.1 was determined as described in ref. 1.

The column "GE" gaseous effluent release values in Table R-1 correspond to calculated doses of 100% of the EPA Protective Action Guidelines (TEDE or CDE child Thyroid). For multi-release point gaseous releases, classification should be based on dose assessment that considers the total site release rate.

The SPING monitors the radioactive effluent from the Units 1 and 2 Turbine Building and Reactor Building Ventilation Stacks and the Standby Gas Treatment System Exhaust Vent. All five collectively are the Plant Vent on Table R-1. The SPING system is normally aligned to be operated from Brother Control Terminals 0C630 in the Control Room, using 0C677 as a backup in the TSC. Three Post-Accident Vent Stack Sampling Systems (PAVSSS) have been installed as backup to the SPING Units. They are used following an accident involving fuel degradation if the SPING monitoring capabilities are lost. Control Terminal CT-i with System Operator Console CT-i1B interrogates each of the SPING and PAVSSS Radiation Monitors for particulate, iodine, and noble gases and informs the operator of changes in operational status, alarm condition, or system parameters within seconds of their occurrence. (ref. 2)

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 Page 3 of 185

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Page 31 of 281 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 ODE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid ODE.

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.

Basis Reference(s):

1. NEP Technical Basis02-005 Rev. #2 Noble Gas Release Rate Limits for EALs

2. ESAR 18.1.30 Accident-Monitoring instrumentation

3. NEI 99-01 AG1 Page 4 of 185

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Page 32 of 281 Category: R - 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 child thyroid ODE EAL:

RGI.2 General Emergency Dose assessment using actual meteorology indicates doses > 1,000 mrem TEDE or 5,000 mrem child thyroid ODE at or beyond the EMERGENCY PLAN BOUNDARY (Note 4)

Note 4: The pre-calculated effluent monitor values presented in EALs RAI .1, RSI .1 and RGI .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):

EMERGENCY PLAN BOUNDARY (EPB) - Same as the Exclusion Area Boundary, i.e., that area around SSES within a radius of 1800 feet determined in accordance with 10 CER 100.11 Basis:

Dose assessments are performed by computer-based methods (ref. 2).

The EMERGENCY PLAN BOUNDARY (EPB) is used in assessing dose effects to the public rather than the SITE BOUNDARY. The EPB is at or within the SITE BOUNDARY in all compass sectors. The SSES dose projection model (MIDAS) utilizes the EPB when performing dose calculations (ref. 5).

The SITE BOUNDARY is irregularly shaped and therefore would result in dose projections and protective action recommendations that can vary significantly depending on plume direction and affected sector. Using dose projections calculated using the EPB (i.e. Exclusion Area) provides a more consistent approach to Public Protective Action Recommendations since the EPB is more consistently defined in all directions. The EPB is at or within the SITE BOUNDARY in all compass sectors.

The EPA PAGs are expressed in terms of the projected sum of the effective dose equivalent (EDE) from external radiation and the committed effective dose equivalent (CEDE) incurred from inhalation of radioactive materials, or as the committed dose equivalent (CDE) to the thyroid. For the purpose of these IC/EALs, the projected dose quantity Total Effective Dose Equivalent (TEDE), as defined in 10 CFR 20 is used in lieu of "the projected sum of...EDE...and CEDE.. .", with CEDE considering significant dose from inhaled radionuclides during the early phase of the event. The EPA protective action guidance provides for the use of adult thyroid dose conversion factors. However, the Commonwealth of Pennsylvania requires the use of Page 5 of 185

Attachment 1 EP.-RM-004 Revision [X]

Page 33 of 281 child thyroid CDE for purposes of comparison of projected thyroid CDE to the PAG for thyroid ODE.

Since dose assessment is based on actual meteorology, whereas the monitor reading RG1 .1 is not, the results from these assessments may indicate that the classification is not warranted. For this reason, emergency implementing procedures call for the timely performance of dose assessments using actual meteorology and release information. If the results of these dose assessments are available when the classification is made (e.g., initiated at a lower classification level), the dose assessment results override the monitor reading RG1 .1. Classification should not be delayed pending the results of these dose assessments.

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 ODE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid ODE.

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.

Basis Reference(s):

1. Susquehanna LLC, Susquehanna Steam Electric Stations Emergency Plan, Section 7.1.1, Off Site Dose Calculations

2. EP-RM-005 SSES MIDAS-N U User Manual

3. NEI 99-01 AG1 Page 6 of 185

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Page 34 of 281 Category: R - 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 child thyroid ODE EAL:

RGI.3 General Emergency Field survey results indicate EITHER of the following at or beyond the EMERGENCY PLAN BOUNDARY:

• Closed window dose rates > 1,000 mR/hr expected to continue for >_60 mai.

• Analyses of field survey samples indicate child thyroid ODE > 5,000 mrem for 60 mai. of inhalation.

(Notes 1, 2)

Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be 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):

EMERGENCY PLAN BOUNDARY (EPB) - Same as the Exclusion Area Boundary, i.e., that area around SSES within a radius of 1800 feet determined in accordance with 10 CER 100.11 Basis:

The EMERGENCY PLAN BOUNDARY (EPB) is used in assessing dose effects to the public rather than the SITE BOUNDARY. The EPB is at or within the SITE BOUNDARY in all compass sectors. The SSES dose projection model (MIDAS) utilizes the EPB when performing dose calculations (ref. 1).

The SITE BOUNDARY is irregularly shaped and therefore would result in dose projections and protective action recommendations that can vary significantly depending on plume direction and affected sector. Using dose projections calculated using the EPB (i.e. EXCLUSION AREA) provides a more consistent approach to Public Protective Action Recommendations since the EPB is more consistently defined in all directions. The EPB is at or within the SITE BOUNDARY in all compass sectors.

The EPA PAGs are expressed in terms of the projected sum of the effective dose equivalent (EDE) from external radiation and the committed effective dose equivalent (CEDE) incurred from inhalation of radioactive materials, or as the committed dose equivalent (ODE) to the thyroid. For the purpose of these IC/EALs, the projected dose quantity Total Effective Dose Page 7 of 1 86

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Page 35 of 281 Equivalent (TEDE), as defined in 10 CFR 20 is used in lieu of "the projected sum of. ..EDE...and CEDE.. .", with CEDE considering significant dose from inhaled radionuclides during the early phase of the event. The EPA protective action guidance provides for the use of adult thyroid dose conversion factors. However, the Commonwealth of Pennsylvania requires the use of child thyroid ODE for purposes of comparison of projected thyroid CDE to the PAG for thyroid CDE.

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 CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

Basis Reference(s):

1. EP-RM-005 SSES MIDAS-N U User Manual

2. NEI 99-01 AG1 Page 8 of 185

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Page 36 of 281 Category: R - 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 child thyroid CDE EAL:

RStI. Site Area Emergency Gaseous effluent > Table R-1 column "SAE" for _>15 min. (Notes 1, 2, 3, 4)

Note 1" The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be 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, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.

Note 4: The pre-calculated effluent monitor values presented in EALs RAI.1, RSI.1 and RG1 .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Table R-1 Effluent Monitor Classification Thresholds (Note 4) -_____

Release Point [ Monitor ] GE SAE ] Alert ] UE oPlant Vent 00630 1 .9E+09 1 .9E+08 1 .9E+07 4.0E+06 S(olga)067pCi/min pCi/mmn pCi/min ICi/min (olga)067(site total) (site total) (site total) (site total)

LRW RR-06433 ......... 2 x hi alarm "o RR-D12-

"5 1(2) RHRSW A/B 1....R....... 2 x hi alarm 1(2) SW/SDHR 1RR-D12-(2)R606 ........ 2 xhi alarm Mode Applicability:

All Definition(s):

None Page 9 of 185

Attachment I EP-RM-004 Revision [X]

Page 37 of 281 Basis:

This EAL address gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to EMERGENCY PLAN BOUNDARY (EPB) doses that exceed either (ref. 1):

• 100 mrem TEDE

• 500 mrem ODE Child Thyroid The SITE BOUNDARY is irregularly shaped and therefore would result in dose projections and protective action recommendations that can vary significantly depending on plume direction and affected sector. Using dose projections calculated using the EPB (i.e. EXCLUSION AREA) provides a more consistent approach to Public Protective Action Recommendations since the EPB is more consistently defined in all directions. The EPB is at or within the SITE BOUNDARY in all compass sectors.

The EPA PAGs are expressed in terms of the projected sum of the effective dose equivalent (EDE) from external radiation and the committed effective dose equivalent (CEDE) incurred from inhalation of radioactive materials, or as the committed dose equivalent (CDE) to the thyroid. For the purpose of these IC/EALs, the projected dose quantity Total Effective Dose Equivalent (TEDE), as defined in 10 CFR 20 is used in lieu of "the projected sum of.. .EDE. .. and CEDE...", with CEDE considering significant dose from inhaled radionuclides during the early phase of the event. The EPA protective actiOn guidance provides for the use of adult thyroid dose conversion factors. However, the Commonwealth of Pennsylvania requires the use of child thyroid CDE for purposes of comparison of projected thyroid ODE to the PAG for thyroid ODE.

The monitor reading threshold for RSI .1 was determined as described in ref. 1.

The column "SAE" gaseous effluent release values in Table R-1 correspond to calculated doses of 10% of the EPA Protective Action Guidelines (TEDE or ODE child Thyroid). For multi-release point gaseous releases, classification should be based on dose assessment that considers the total site release rate.

The SPING monitors the radioactive effluent from the Units I and 2 Turbine Building and Reactor Building Ventilation Stacks and the Standby Gas Treatment System Exhaust Vent. All five collectively are the Plant Vent on Table R-1. The SPING system is normally aligned to be operated from Brother Control Terminals 00630 in the Control Room, using 00677 as a backup in the TSC. Three Post-Accident Vent Stack Sampling Systems (PAVSSS) have been installed as backup to the SPING Units. They are used following an accident involving fuel degradation if the SPING monitoring capabilities are lost. Control Terminal CT-I with System Operator Console CT-I1B interrogates each of the SPING and PAVSSS Radiation Monitors for particulate, iodine, and noble gases and informs the operator of changes in operational status, alarm condition, or system parameters within seconds of their occurrence. (ref. 2)

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 Page 10 of 185

Attachment 1 EP-RM-004 Revision IX]

Page 38 of 281 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 ODE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid ODE.

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 RGI.

Basis Reference(s):

1. NEP Technical Basis02-005 Rev. #2 Noble Gas Release Rate Limits for EALs

2. FSAR 18.1.30 Accident-Monitoring Instrumentation

3. NEI 99-01 AS1 Page 11 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 39 of 281 Category: R - 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 child thyroid CDE EAL:

RSl.2 Site Area Emergency Dose assessment using actual meteorology indicates doses > 100 mrem TEDE or 500 mrem child thyroid ODE at or beyond the EMERGENCY PLAN BOUNDARY (Note 4)

Note 4: The pre-calculated effluent monitor values presented in EALs RAI .1, RS1 .1 and RG1 .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):

EMERGENCY PLAN BOUNDARY (EPB) - Same as the Exclusion Area Boundary, i.e., that area around SSES within a radius of 1800 feet determined in accordance with 10 CER 100.11 Basis:

Dose assessments are performed by computer-based methods (ref. 2).

The EMERGENCY PLAN BOUNDARY (EPB) is used in assessing dose effects to the public rather than the SITE BOUNDARY. The EPB is at or within the SITE BOUNDARY in all compass sectors. The SSES dose projection model (MIDAS) utilizes the EPB when performing dose calculations (ref. 5).

The SITE BOUNDARY is irregularly shaped and therefore would result in dose projections and protective action recommendations that can vary significantly depending on plume direction and affected sector. Using dose projections calculated using the EPB (i.e. Exclusion Area) provides a more consistent approach to Public Protective Action Recommendations since the EPB is more consistently defined in all directions. The EPB is at or within the SITE BOUNDARY in all compass sectors.

The EPA PAGs are expressed in terms of the projected sum of the effective dose equivalent (EDE) from external radiation and the committed effective dose equivalent (CEDE) incurred from inhalation of radioactive materials, or as the committed dose equivalent (ODE) to the thyroid. For the purpose of these IC/EALs, the projected dose quantity Total Effective Dose Equivalent (TEDE), as defined in 10 CFR 20 is used in lieu of "the projected sum of...EDE...and CEDE.. .", with CEDE considering significant dose from inhaled radionuclides during the early phase of the event. The EPA protective action guidance provides for the use of adult thyroid dose conversion factors. However, the Commonwealth of Pennsylvania requires the use of Page 12 of 185

Attachment 1 EP-RM-004 Revision IX]

Page 40 of 281 child thyroid ODE for purposes of comparison of projected thyroid ODE to the PAG for thyroid ODE.

Since dose assessment is based on actual meteorology, whereas the monitor reading RS1 .1 is not, the results from these assessments may indicate that the classification is not warranted. For this reason, emergency implementing procedures call for the timely performance of dose assessments using actual meteorology and release information. If the results of these dose assessments are available when the classification is made (e.g., initiated at a lower classification level), the dose assessment results override the monitor reading RS1 .1. Classification should not be delayed pending the results of these dose assessments.

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 ODE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid ODE.

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 RGI.

Basis Reference(s):

1. Susquehanna LLO, Susquehanna Steam Electric Stations Emergency Plan, Section 7.1.1, Off Site Dose Calculations

2. EP-RM-005 SSES MIDAS-N U User Manual

3. NEI 99-01 AS1 Page 13 of 185

Attachment 1 EP-RM-004 Revision IX]

Page 41 of 281 Category: R - 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 child thyroid ODE EAL:

R81 .3 Site Area Emergency Field survey results indicate EITHER of the following at or beyond the EMERGENCY PLAN BOUNDARY:

• Closed window dose rates > 100 mR/hr expected to continue for _>60 mmn.

• Analyses of field survey samples indicate child thyroid CDE > 500 mrem for 60 min. of inhalation.

(Notes 1, 2)

Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be 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):

EMERGENCY PLAN BOUNDARY (EPB) - Same as the Exclusion Area Boundary, i.e., that area around SSES within a radius of 1800 feet determined in accordance with 10 CER 100.11.

Basis:

The EMERGENCY PLAN BOUNDARY (EPB) is used in assessing dose effects to the public rather than the SITE BOUNDARY. The EPB is at or within the SITE BOUNDARY in all compass sectors. The SSES dose projection model (MIDAS) utilizes the EPB when performing dose calculations (ref. 1).

The SITE BOUNDARY is irregularly shaped and therefore would result in dose projections and protective action recommendations that can vary significantly depending on plume direction and affected sector. Using dose projections calculated using the EPB (i.e. EXCLUSION AREA) provides a more consistent approach to Public Protective Action Recommendations since the EPB is more consistently defined in all directions. The EPB is at or within the SITE BOUNDARY in all compass sectors The EPA PAGs are expressed in terms of the projected sum of the effective dose equivalent (EDE) from external radiation and the committed effective dose equivalent (CEDE) incurred from inhalation of radioactive materials, or as the committed dose equivalent (CDE) to the thyroid. For the purpose of these IC/EALs, the projected dose quantity Total Effective Dose Equivalent (TEDE), as defined in 10 CER 20 is used in lieu of "the projected sum of...EDE...and Page 14 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 42 of 281 CEDE.. .", with CEDE considering significant dose from inhaled radionuclides during the early phase of the event. The EPA protective action guidance provides for the use of adult thyroid dose conversion factors. However, the Commonwealth of Pennsylvania requires the use of child thyroid CDE for purposes of comparison of projected thyroid CDE to the PAG for thyroid CDE.

Since dose assessment is based on actual meteorology, whereas the monitor reading RS1 .1 is not, the results from these assessments may indicate that the classification is not warranted. For this reason, emergency implementing procedures call for the timely performance of dose assessments using actual meteorology and release information. If the results of these dose assessments are available when the classification is made (e.g., initiated at a lower classification level), the dose assessment results override the monitor reading RS1I.1. Classification should not be delayed pending the results of these dose assessments.

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 CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

Escalation of the emergency classification level would be via IC RGI.

Basis Reference(s):

1. EP-RM-005 SSES MIDAS-N U User Manual

2. NEI 99-01 AS1 Page 15 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 43 of 281 Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dlose greater than 10 mrem TEDE or 50 mrem child thyroid ODE EAL:

RA1.1 Alert Gaseous effluent > Table R-1 column "Alert" for _>15 min. (Notes 1, 2, 3, 4)

Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be 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, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.

Note 4 The pre-calculated effluent monitor values presented in EALs RAI .1, RS1I.1 and RG1 .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Release-1 Effluent Monitor Classification Thresholds (Note 4) U RlaePoint Monitor GE ] SAE f Alert U Pl et0601 .9E+09 1 .9E+08 I1.9E+07 4.0E+06 o ln et060pCi/min pCi/min pCi/min pCi/min S(noble gas) 0C677 (site total) (site total) (site total) (site total)

LRW RR-06433 ........ 2 x hi alarm 0"=o1(2) RHRSW A/B RD1(2)R0 ...... 2 xhi alarm 1(2) SW/SDHR 1RR-D12-(2)R604 ....... 2 xhi alarm Mode Applicability:

All Definition(s):

None Page 16 of 185

Attachment I EP-RM-004 Revision [X]

Page 44 of 281 Basis:

This EAL address gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to EMERGENCY PLAN BOUNDARY (EPB) doses that exceed either (ref. 1):

• 10 mremTEDE

• 50 mrem ODE Child Thyroid The SITE BOUNDARY is irregularly shaped and therefore would result in dose projections and protective action recommendations that can vary significantly depending on plume direction and affected sector. Using dose projections calculated using the EPB (i.e. Exclusion Area) provides a more consistent approach to Public Protective Action Recommendations since the EPB is more consistently defined in all directions. The EPB is at or within the SITE BOUNDARY in all compass sectors.

The EPA PAGs are expressed in terms of the projected sum of the effective dose equivalent (EDE) from external radiation and the committed effective dose equivalent (CEDE) incurred from inhalation of radioactive materials, or as the committed dose equivalent (ODE) to the thyroid. For the purpose of these IC/EALs, the projected dose quantity Total Effective Dose Equivalent (TEDE), as defined in 10 CER 20 is used in lieu of "the projected sum of... EDE. .. and CEDE...", with CEDE considering significant dose from inhaled radionuclides during the early phase of the event. The EPA protective action guidance provides for the use of adult thyroid dose conversion factors. However, the Commonwealth of Pennsylvania requires the use of child thyroid ODE for purposes of comparison of projected thyroid ODE to the PAG for thyroid ODE.

The monitor reading threshold for RAI .1 was determined as described in ref. 1.

The column "ALERT" gaseous effluent release value in Table R-1 correspond to calculated doses of 1% (10% of the SAE thresholds) of the EPA Protective Action Guidelines (TEDE or ODE Thyroid). For multi-release point gaseous releases, classification should be based on dose assessment that considers the total site release rate.

The SPING monitors the radioactive effluent from the Units I and 2 Turbine Building and Reactor Building Ventilation Stacks and the Standby Gas Treatment System Exhaust Vent. All five collectively are the Plant Vent on Table R-1. The SPING system is normally aligned to be operated from Brother Control Terminals 00630 in the Control Room, using 00677 as a backup in the TSC. Three Post-Accident Vent Stack Sampling Systems (PAVSSS) have been installed as backup to the SPING Units. They are used following an accident involving fuel degradation if the SPING monitoring capabilities are lost. Control Terminal CT-i with System Operator Console CT-I1B interrogates each of the SPING and PAVSSS Radiation Monitors for particulate, iodine, and noble gases and informs the operator of changes in operational status, alarm condition, or system parameters within seconds of their occurrence. (ref. 2)

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).

Page 17 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 45 of 281 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 ODE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid ODE.

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 RSI.

Basis Reference(s):

1. NEP Technical Basis02-005 Rev. #2 Noble Gas Release Rate Limits for EALs

2. FSAR 18.1.30 Accident-Monitoring Instrumentation

4. NEI 99-01 AA1 Page 18 of 185

Attachment 1 EP-RM-004 Revision IX]

Page 46 of 281 Category: R - 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 child thyroid ODE EAL:

RAI.2 Alert Dose assessment using actual meteorology indicates doses > 10 mrem TEDE or 50 mrem child thyroid ODE at or beyond the EMERGENCY PLAN BOUNDARY (Note 4)

Note 4: The pre-calculated effluent monitor values presented in EALs RAlI.1, RSI .1 and RG1 .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):

EMERGENCY PLAN BOUNDARY (EPB) - Same as the Exclusion Area Boundary, i.e., that area around SSES within a radius of 1800 feet determined in accordance with 10 CER 100.11.

Basis:

Dose assessments are performed by computer-based methods (ref. 2).

The EMERGENCY PLAN BOUNDARY (EPB) is used in assessing dose effects to the public rather than the SITE BOUNDARY. The EPB is at or within the SITE BOUNDARY in all compass sectors. The SSES dose projection model (MIDAS) utilizes the EPB when performing dose calculations (ref. 5).

Page 19 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 47 of 281 The SITE BOUNDARY is irregularly shaped and therefore would result in dose projections and protective action recommendations that can vary significantly depending on plume direction and affected sector. Using dose projections calculated using the EPB (i.e. Exclusion Area) provides a more consistent approach to Public Protective Action Recommendations since the EPB is more consistently defined in all directions. The EPB is at or within the SITE BOUNDARY in all compass sectors.

The EPA PAGs are expressed in terms of the projected sum of the effective dose equivalent (EDE) from external radiation and the committed effective dose equivalent (CEDE) incurred from inhalation of radioactive materials, or as the committed dose equivalent (ODE) to the thyroid. For the purpose of these IC/EALs, the projected dose quantity Total Effective Dose Equivalent (TEDE), as defined in 10 CFR 20 is used in lieu of "the projected sum of...EDE...and CEDE...", with CEDE considering significant dose from inhaled radionuclides during the early phase of the event. The EPA protective action guidance provides for the use of adult thyroid dose conversion factors. However, the Commonwealth of Pennsylvania requires the use of child thyroid CDE for purposes of comparison of projected thyroid CDE to the PAG for thyroid CDE.

Since dose assessment is based on actual meteorology, whereas the monitor reading RAI.1 is not, the results from these assessments may indicate that the classification is not warranted. For this reason, emergency implementing procedures call for the timely performance of dose assessments using actual meteorology and release information. If the results of these dose assessments are available when the classification is made (e.g., initiated at a lower classification level), the dose assessment results override the monitor reading RA1 .1. Classification should not be delayed pending the results of these dose assessments.

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 CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.

Escalation of the emergency classification level would be via IC RS1.

Basis Reference(s):

1. Susquehanna LLC, Susquehanna Steam Electric Stations Emergency Plan, Section 7.1.1, Off Site Dose Calculations Page 20 of 185

Attachment 1 EP-RM-004 Revision IX]

Page 48 of 281

2. EP-RM-005 SSES MIDAS-N U User Manual

3. NE! 99-01 AA1 Page 21 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 49 of 281 Category: R - 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 child thyroid ODE EAL:

RAI.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 child thyroid ODE at or beyond the EMERGENCY PLAN BOUNDARY for 60 min. of exposure (Notes 1, 2)

Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be 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):

EMERGENCY PLAN BOUNDARY (EPB) - Same as the Exclusion Area Boundary, i.e., that area around SSES within a radius of 1800 feet determined in accordance with 10 CER 100.11.

Basis:

RA1 .3 addresses uncontrolled releases that are detected by sample analyses, particularly on unmonitored pathways, e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.

For a radiological liquid release, the calculated effluent concentration from a chemistry sample is compared to the emergency action level. Shift Management utilizes emergency response procedures to notify risk counties and to obtain river water samples.

The Susquehanna station incorporates features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, or control and monitor intentional releases. These controls are located in the Technical Requirements Manual (TRM). The occurrence of extended, uncontrolled radioactive releases to the environment is indicative of degradation in these features and/or controls.

EAL RA1 .3 addresses uncontrolled releases that are detected by sample analyses, particularly on unmonitored pathways, e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc. This EAL reflects the concern that releases in excess of the referenced offsite dose values represent an uncontrolled situation and hence a potential degradation in the level of safety. Although the calculated dose is very low, it is the degradation in plant control as indicated by the failure to terminate the release that is of primary concern.

Page 22 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 50 of 281 The EMERGENCY PLAN BOUNDARY is referenced in EAL RA1 .3 because this EAL is based upon liquid release limits from the plant. The release limits are contained in the TRM and are in turn based upon calculation methodology specified in the ODOM. The ODCM utilizes the EMERGENCY PLAN BOUNDARY to establish plant release limits.

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 ODE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid ODE.

Escalation of the emergency classification level would be via IC RS1.

Basis Reference(s):

1. ODCM-QA-003 Effluent Monitor Setpoints

2. NEI 99-01 AA1 Page 23 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 51 of 281 Category: R - 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 child thyroid CDE EAL:

RAI.4 Alert Field survey results indicate EITHER of the following at or beyond the EMERGENCY PLAN BOUNDARY:

• Closed window dose rates > 10 mR/hr expected to continue for >_60 min.

• Analyses of field survey samples indicate child thyroid CDE > 50 mrem for 60 min. of inhalation.

(Notes 1, 2)

Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be 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):

EMERGENCY PLAN BOUNDARY (EPB) - Same as the Exclusion Area Boundary, i.e., that area around SSES within a radius of 1800 feet determined in accordance with 10 CER 100.11 Basis:

The EMERGENCY PLAN BOUNDARY (EPB) is used in assessing dose effects to the public rather than the SITE BOUNDARY. The EPB is at or within the SITE BOUNDARY in all compass sectors. The SSES dose projection model (MIDAS) utilizes the EPB when performing dose calculations (ref. 1).

The SITE BOUNDARY is irregularly shaped and therefore would result in dose projections and protective action recommendations that can vary significantly depending on plume direction and affected sector. Using dose projections calculated using the EPB (i.e. Exclusion Area) provides a more consistent approach to Public Protective Action Recommendations since the EPB is more consistently defined in all directions. The EPB is at or within the SITE BOUNDARY in all compass sectors.

The EPA PAGs are expressed in terms of the projected sum of the effective dose equivalent (EDE) from external radiation and the committed effective dose equivalent (CEDE) incurred from inhalation of radioactive materials, or as the committed dose equivalent (CDE) to the thyroid. For the purpose of these IC/EALs, the projected dose quantity Total Effective Dose Page 24 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 52 of 281 Equivalent (TEDE), as defined in 10 CFR 20 is used in lieu of "the projected sum of...EDE...and CEDE.. .", with CEDE considering significant dose from inhaled radionuclides during the early phase of the event. The EPA protective action guidance provides for the use of adult thyroid dose conversion factors. However, the Commonwealth of Pennsylvania requires the use of child thyroid CDE for purposes of comparison of projected thyroid CDE to the PAG for thyroid CDE.

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 CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.

Escalation of the emergency classification level would be via IC RSI.

Basis Reference(s):

1. EP-RM-005 SSES MIDAS-N U User Manual

2. NEI 99-01 AA1 Page 25 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 53 of 281 Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the TRM limits for 60 minutes or longer EAL:

RUI.1 Unusual Event Gaseous or liquid effluent > Table R-1 column "UE" for _>60 min. (Notes 1, 2, 3)

Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be 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, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.

Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor ] GE SAE Alert UE oPlant Vent 00630 1 .9E+09 1 .9E+08 1.9E+07 4.0E+06 S(olga)067pCi/min pCi/min pCi/min pCi/min (olga)067(site total) (site total) (site total) (site total)

LRW RR-06433 ... 2 x hi alarm

,1 RHRSW (2) A/B RR.D.. 2 xhi alarm o" 1 (2)R606 RR-D12-1 (2) SW/SDHR 1 (2)R604 ... 2 xhi alarm Mode Applicability:

All Definition(s):

None Basis:

The EMERGENCY PLAN BOUNDARY is used in RU1 .1 because it is based upon release limits from the plant. The release limits are based upon the Plant Technical Requirements Manual and in turn based upon calculation methodology specified in the ODCM. The ODCM utilizes the Page 26 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 54 of 281 EMERGENCY PLAN BOUNDARY to establish plant release limits. EAL RUI .1 is an indication of degradation in the level of safety of the plant.

This EAL represents radioactivity releases, that for whatever reason, cause liquid effluent radiation monitor readings to exceed two times the Technical Requirements Manual limit and releases are not terminated within 60 minutes. This alarm setpoint may be associated with a planned batch release, or a continuous release path. In either case, the setpoint is established by the ODCM to warn of a release that is not in compliance with the applicable TRM release limit. Indexing the EAL threshold to the 00CM setpoints in this manner ensures that the EAL threshold will never be less than the setpoint established by a specific discharge permit.

For a radiological liquid release, the calculated effluent concentration from a chemistry sample is compared to the emergency action level. Shift Management utilizes emergency response procedures to notify risk counties and to obtain river water samples.

Gaseous Releases The column "UE" gaseous release value in Table R-1 represents two times the appropriate TRM release rate limits (ref. 1-3).

The SPING monitors the radioactive effluent from the Units 1 and 2 Turbine Building and Reactor Building Ventilation Stacks and the Standby Gas Treatment System Exhaust Vent. All five collectively are the Plant Vent on Table R-I. The SPING system is normally aligned to be operated from Brother Control Terminals 0C630 in the Control Room, using 0C677 as a backup in the TSC. Three Post-Accident Vent Stack Sampling Systems (PAVSSS) have been installed as backup to the SPING Units. They are used following an accident involving fuel degradation if the SPING monitoring capabilities are lost. Control Terminal CT-i with System Operator Console CT-i1B interrogates each of the SPING and PAVSSS Radiation Monitors for particulate, iodine, and noble gases and informs the operator of changes in operational status, alarm condition, or system parameters within seconds of their occurrence. (ref. 4)

Liquid Releases The column "UE" liquid release values in Table R-1 represent two times the appropriate TRM release rate limits associated with the specified release point (ref. 5-9).

• LRW A sample pump takes a portion of the LRW effluent line flow and passes it through a scintillation detector (RE-06433). The contents of EDST, the LRW sample tanks, or the laundry drain sample tank can be lined up for release. Release flow is through two isolation valves, HV-06432A1 and A2 to the cooling t~wer blowdown line.

The isolation valves will close on:

• Low Sample Flow (< 0.5 gpm)

• Low Blowdown Flow (< 5500 gpm)

• High radiation (calculated for each release based on sample)

• Rad Monitor downscale (calculated)

• Rad Monitor inoperable Page 27 of 185

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Page 55 of 281

• RHRSW The Residual Heat Removal (RHR) Service Water RMS detects primary coolant leakage into the RHR Service Water during RHR Heat Exchanger operation. Local indication of RHR Service Water Loop 'A' radiation is provided by RITS-11216A and RHR Service Water Loop 'B' by RITS-1 121 6B. Output is also sent to Control Room alarms and Radiation Recorder RR-D12-1 R606 on Panel 1C600. The following RHR Service Water RMS high radiation annunciators are located on Control Room Panel 10601: RHR SW A HI RADIATION (AR-i109-F01) is actuated at a setpoint determined by Chemistry.

• RHR SW B HI RADIATION (AR-I1I3-F01) is actuated at a setpoint determined by Chemistry.

• SW/SDHR Service Water/Supplemental Decay Heat Removal RMS detects radioactive material in-leakage to the service water system from the spent fuel pool heat exchangers. During unit outages when SDHR is placed in service, the SDHR Detector is connected to the Radiation Monitoring Unit in place of the Service Water Detector.

The Service Water/Supplemental Decay Heat Removal RMS annunciator SERVICE WATER EFFLUENT HI RADIATION (AR-123-D04) setpoint is variable as determined by Chemistry Group.

This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.

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 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.

This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways. EAL Bases Escalation of the emergency classification level would be via IC RAI.

Basis Reference(s):

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Page 56 of 281

1. TRM 3.3.4 TRM Post-Accident Monitoring Instrumentation

2. TRM 3.11.2 Gaseous Effluents

3. Susquehanna Calculation EC ENVR 1041 Airborne Effluent Limiting Site Release Rate &

Plant Vent Effluent Monitor Setpoints

4. FSAR 18.1.30 Accident-Monitoring Instrumentation

5. OP-1 79-002 Process Radiation Monitoring System

6. ESAR 9.2 Water Systems

7. TRM 3.11.1.4 Liquid Radwaste Effluent Monitoring Instrumentation

8. TRM 3.11.1.5 Radioactive Liquid Process Monitoring Instrumentation

9. ON-069-001 Abnormal Rad Release Liquid

10. SSES Offsite Dose Calculation Manual

11. NEI 99-01 AU1 Page 29 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 57 of 281 Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the TRM limits for 60 minutes or longer.

EAL:

Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be 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:

For a radiological liquid release, the calculated effluent concentration from a chemistry sample is compared to the emergency action level. Shift Management utilizes emergency response procedures to notify risk counties and to obtain river water samples.

Limits associated with liquid and gaseous radioactive effluents are contained in the Technical Requirements Manual (TRM). The methodology for calculation of offsite dose or release rates to ensure compliance with the applicable TRM limits is outlined in the ODCM.

EAL RU1 .2 addresses radioactivity releases, that for whatever reason, cause effluent radiation monitor readings to exceed two times the applicable TRM limit and are not terminated within 60 minutes. The effluent monitor alarm setpoints are established by the ODCM to warn of a release that is not in compliance with the applicable TRM limit (ref.1). The ODCM establishes a methodology for determining effluent radiation monitor setpoints. The ODCM specifies default source terms and, for gaseous releases, prescribes the use of pre-determined annual average meteorology in the most limiting downwind sector for showing compliance with the regulatory commitments. The fundamental basis of this IC is NOT a dose or dose rate, but rather the degradation in the level of safety of the plant implied by the uncontrolled release EAL RU1 .2 includes any release for which a radioactivity discharge permit was not prepared or not applicable, or a release that exceeds the conditions (e.g., minimum dilution flow, maximum discharge flow, alarm setpoints, etc.) on the applicable permit. Although the calculated dose is very low, it is the degradation in plant control as indicated by the failure to terminate the release that is of primary concern.

Page 30 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 58 of 281 This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e~g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.

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. 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 RAI.

Basis Reference(s):

1. ODCM-QA-003 Effluent Monitor Setpoints

2. NEI 99-01 AU1 Page 31 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 59 of 281 Category: R - 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 spent fuel racks for 60 minutes or longer EAL:

RG2.1 General Emergency Spent fuel pool level CANNOT BE RESTORED to at least 0.5 ft. above the top of the spent fuel racks for >60 min. (Note 1)

Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability:

All Definition(s):

None Basis:

Each fuel storage pool is designed to maintain the water level in the pool above the top of active fuel providing cooling for the fuel rods. Technical Specifications require greater than or equal to 22 ft. of water be maintained over the top of irradiated fuel assemblies seated in the spent fuel storage racks at all times (ref. 1). In the event of loss of fuel pool inventory, Operations will assess multiple indications in accordance with ON-FPC-101 (201) and AOP-081-001 (ref. 3, 4).

Post-Fukushima order EA-12-051 required the installation of reliable SEP level indication capable of identifying normal SPE level 22.75 feet above the top of the fuel racks (Level I or el.

81 7'-1 "), SFP level 10 ft. above the top of the fuel racks (Level 2 or el. 804'-4") and SFP level at the top of the fuel racks (Level 3 or el. 794'-1 0" which for SSES is 0.5 feet on instrument above the top of the fuel racks). Each spent fuel pool is equipped with primary and backup guided wave radar probes to measure pool level. The range is continuous from the high pool level elevation (23.25 feet on instrument or el. 817'-7") to the top of the spent fuel racks (0.0 feet on instrument or el. 794'-4") (ref. 2).

Each new SEP LI system provides two alarms to the associated Unit's Control Room Benchboard. The first alarm identifies a [ow water level condition at 20 ft. (instrument) above the spent fuel rack (elevation 814'-4"). The second alarm identifies a low low water level condition at 10 ft. (instrument) above the spent fuel rack (Level 2 or elevation 804'-4"). These alarms are intended to alert the control room operators of the loss of SEP inventory so actions are taken to provide make-up as soon as possible (ref. 2).

Page 32 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 60 of 281 This IC addresses a significant loss of spent fuel pool inventory control and makeup capability leading to a prolonged uncovery of spent fuel. This condition will lead to fuel damage and a radiological release to the environment.

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

Basis Reference(s):

1. Technical Specifications 3.7.7 Spent Fuel Storage Pool Water Level

2. PLA-6980 Enclosure 1 Susquehanna Units 1 & 2 Overall Integrated Plan with Regard to Reliable Spent Fuel Pool Instrumentation

3. ON-FPC-101(201) Loss of Fuel Pool Cooling

4. AOP-081-001 Fuel Handling Abnormal Operating Procedure

5. NEI 99-01 AG2 Page 33 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 61 of 281 Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Spent fuel pool level at the top of the fuel racks EAL:

RS2.1 Site Area Emergency Lowering of spent fuel pool level to <.0.5 ft. above the top of the spent fuel racks Mode Applicability:

All Definition(s):

None Basis:

Each fuel storage pool is designed to maintain the water level in the pool above the top of active fuel providing cooling for the fuel rods. Technical Specifications require greater than or equal to 22 ft. of water be maintained over the top of irradiated fuel assemblies seated in the spent fuel storage racks at all times (ref. 1). In the event of loss of fuel pool inventory, Operations will assess multiple indications in accordance with ON-FPC-1 01 (201) and AOP-081 -001 (ref. 3, 4).

Post-Fukushima order EA-12-051 required the installation of reliable SEP level indication capable of identifying normal SPF level 22.75 feet above the top of the fuel racks (Level 1 or el.

81 7'-1 "), SFP level 10 ft. above the top of the fuel racks (Level 2 or el. 804'-4") and SFP level at the top of the fuel racks (Level 3 or el. 794'-10" which for SSES is 0.5 feet on instrument above the top of the fuel racks). Each spent fuel pool is equipped with primary and backup guided wave radar probes to measure pool level. The range is continuous from the high pool level elevation (23.25 feet on instrument or el. 81 7'-7") to the top of the spent fuel racks (0.0 feet on instrument or el. 794'-4") (ref. 2).

Each new SFP LI system provides two alarms to the associated Unit's Control Room Benchboard. The first alarm identifies a low water level condition at 20 ft. (instrument) above the spent fuel rack (elevation 814'-4"). The second alarm identifies a low low water level condition at 10 ft. (instrument) above the spent fuel rack (Level 2 or elevation 804'-4"). These alarms are intended to alert the control room operators of the loss of SEP inventory so actions are taken to provide make-up as soon as possible (ref. 2).

This IC addresses a significant loss of spent fuel pool inventory control and makeup capability leading to IMMINENT fuel damage. This condition entails major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.

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

Escalation of the emergency classification level would be via IC RG1 or RG2.

Basis Reference(s):

Page 34 of 185

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Page 62 of 281

1. Technical Specifications 3.7.7 Spent Fuel Storage Pool Water Level

2. PLA-6980 Enclosure 1 Susquehanna Units 1 & 2 Overall Integrated Plan with Regard to Reliable Spent Fuel Pool Instrumentation

3. ON-FPC-101(201) Loss of Fuel Pool Cooling

4. AOP-081-001 Fuel Handling Abnormal Operating Procedure

5. NEI 99-01 AA2 Page 35 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 63 of 281 Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL:

RA2.1 Alert Uncovery of irradiated fuel in the REFUELING PATHWAY Mode Applicability:

All Definition(s):

REFUELING PATHWAY-. The reactor refueling cavity, spent fuel pool and fuel transfer canal comprise the refueling pathway.

Basis:

This EAL applies to all instances of irradiated fuel handling, including those that are not directly in support of a reactor refueling outage. (e.g. unit-to-unit fuel shuffles, dry fuel storage canister loading, etc.).

Each fuel storage pool is designed to maintain the water level in the pool above the top of active fuel providing cooling for the fuel rods. Technical Specifications require greater than or equal to 22 feet of water be maintained over the top of irradiated fuel assemblies seated in the spent fuel storage racks at all times (ref. 1). In the event of loss of fuel pool inventory, Operations will assess multiple indications in accordance with ON-FPC-1O1 (201) and AOP-081-OO1 (ref. 3, 4).

This IC3 addresses events that have caused imminent or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant.

This EAL escalates from RU2.1 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 boil-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 an increase in a dose rate due to a lowering of water level in some portion of the REFUELING PATHWAY, the reading may not be a reliable indication of whether or not the fuel is actually uncovered. To the degree possible, readings should be considered in combination with other available indications of inventory loss.

Page 36 of 185

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Page 64 of 281 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 RSI.

Basis Reference(s):

1. Technical Specifications 3.7.7 Spent Fuel Storage Pool Water Level

2. PLA-6980 Enclosure 1 Susquehanna Units 1 & 2 Overall Integrated Plan with Regard to Reliable Spent Fuel Pool Instrumentation

3. ON-FPC-101(201) Loss of Fuel Pool Cooling

4. AOP-081-001 Fuel Handling Abnormal Operating Procedure

5. NEI 99-01 AA2 Page 37 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 65 of 281 Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL:

RA2.2 Alert Damage to irradiated fuel resulting in a release of radioactivity AND Any of the following radiation monitor indications:

• Refuel Floor High Exhaust (> 18 mR/hr)

• Refuel Floor Wall Exhaust (> 21 mR/hr)

• Channel 14 Spent Fuel Pool Area Criticality Monitor (> 100 mR/hr)

• Channel 15 Refueling Floor Area (> 80 mR/hr)

° Channel 42 Refueling Floor Area (> 80 mR/hr)

,, Channel 47 (U1) /44 (U2) Spent Fuel Pool Area Criticality Monitor (> 100 mR/hr)

• Channel 49 Refueling Floor High Range Monitor (on scale)

Mode Applicability:

All Definition(s):

None Basis:

The Reactor Building ventilation process monitoring system isolates Zone 3 HVAC on high exhaust radiation. Zone 3 exhaust can be monitored at 1C600 (2C600) (ref. 4):

• RR D1 2 1 R605 (2R605), Refuel Floor Wall Exhaust Radiation Monitor

• RR 012 1 R607 (2R607), Refuel Floor High Exhaust Radiation Monitor The listed radiation monitors and specified alarm setpoints/ indications are those associated with a fuel handling accident or damaged spent fuel (ref. 1, 3, 5).

This IC addresses events that have caused imminent or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant.

Escalation of the emergency would be based on either Recognition Category R or C ICs.

This EAL addresses a release of radioactive material caused by mechanical damage to Page 38 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 66 of 281 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).

Escalation of the emergency classification level would be via IC RSI.

Basis Reference(s):

1. OP-i179(279)-O01 Area Radiation Monitoring System

2. PLA-6980 Enclosure 1 Susquehanna Units 1 & 2 Overall Integrated Plan with Regard to Reliable Spent Fuel Pool Instrumentation

3. AOP-081-O01 Fuel Handling Abnormal Operating Procedure

4. EO-000-104 Secondary Containment Control

5. AR-101(201)-001

6. NEI 99-01 AA2 Page 39 of 185

Attachment 1 EP-RM-004 Revision IX]

Page 67 of 281 Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL:

RA2.3 Alert Lowering of spent fuel pool level to < 10 ft. above the top of the spent fuel racks Mode Applicability:

All Definition(s):

None Basis:

Each fuel storage pool is designed to maintain the water level in the pool above the top of active fuel providing cooling for the fuel rods. Technical Specifications require greater than or equal to 22 ft. of water be maintained over the top of irradiated fuel assemblies seated in the spent fuel storage racks at all times (ref. 1). In the event of loss of fuel pool inventory, Operations will assess multiple indications in accordance with ON-FPC-101(201 ) and AOP-081-001 (ref. 3, 4).

Post-Fukushima order EA-12-051 required the installation of reliable SEP level indication capable of identifying normal SPE level 22.75 feet above the top of the fuel racks (Level 1 or el.

81 7'-1 "), SEP level 10 ft. above the top of the fuel racks (Level 2 or el. 804'-4") and SEP level at the top of the fuel racks (Level 3 or el. 794'-10" which for SSES is 0.5 feet on instrument above the top of the fuel racks). Each spent fuel pool is equipped with primary and backup guided wave radar probes to measure pool level. The range is continuous from the high pool level elevation (23.25 feet on instrument or el. 81 7'-7") to the top of the spent fuel racks (0.0 feet on instrument or el. 794'-4") (ref. 2).

Each new SEP LI system provides two alarms to the associated Unit's Control Room Benchboard. The first alarm identifies a low water level condition at 20 ft. (instrument) above the spent fuel rack (elevation 814'-4"). The second alarm identifies a low low water level condition at l0 ft. (instrument) above the spent fuel rack (Level 2 or elevation 804'-4"). These alarms are intended to alert the control room operators of the loss of SEP inventory so actions are taken to provide make-up as soon as possible (ref. 2).

This IC addresses events that have caused imminent or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant.

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.

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 Page 40 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 68 of 281 operations in the vicinity of the spent fuel pooi. 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 RS1.

Basis Reference(s):

1. Technical Specifications 3.7.7 Spent Fuel Storage Pool Water Level

2. PLA-6980 Enclosure 1 Susquehanna Units 1 & 2 Overall Integrated Plan with Regard to Reliable Spent Fuel Pool Instrumentation

3. ON-FPC-101(201) Loss of Fuel Pool Cooling

4. AOP-081-001 Fuel Handling Abnormal Operating Procedure

5. NEI 99-01 AA2 Page 41 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 69 of 281 Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Unplanned loss of water level above irradiated fuel EAL:

RU2.1 Unusual Event UNPLANNED water level drop in the REFUELING PATHWAY as indicated by any of the following on EITHER unit:

• Fuel Pool Water Low Level alarm

• Skimmer Surge Tank Low Level alarm

• Visual observation of a water level drop below a fuel pool skimmer surge tank inlet

• Observation of water draining down the outside wall of primary containment AND UNPLANNED rise in area radiation levels as indicated by any of the following radiation monitors:

• Channel 14 Spent Fuel Pool Area Criticality Monitor

• Channel 15 Refueling Floor Area

• Channel 42 Refueling Floor Area 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-. The reactor refueling cavity, spent fuel pool and fuel transfer canal comprise the refueling pathway.

Basis:

Each fuel storage pool is designed to maintain the water level in the pool above the top of active fuel providing cooling for the fuel rods. Technical Specifications require greater than or equal to 22 feet of water be maintained over the top of irradiated fuel assemblies seated in the spent fuel storage racks at all times (ref. 1). In the event of loss of fuel pool inventory, Operations will assess multiple indications in accordance with ON-FPC-101 (201) and AOP-081-001 (ref. 4, 6).

When the spent fuel pool and reactor cavity are connected, there could exist the possibility of uncovering irradiated fuel. Therefore, this EAL is applicable for conditions in which irradiated fuel is being transferred to and from the RPV and spent fuel pool.

Page 42 of 185

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Page 70 of 281 Control room alarms associated with elevated refuel floor area radiation levels include (ref. 4, 5):

• REFUELING FLOOR AREA HI RADIATION (AR-101-D05)

• SPENT FUEL POOL AREA HI RADIATION (AR-101-E05)

The listed ARMs are the normal range monitors that detect increasing area radiation due to a lack of shielding in the REFUELING PATHWAY (ref. 6).

This IC addresses a decrease in water level above irradiated fuel sufficient to cause elevated radiation levels. This condition could be a precursor to a more serious event and is also indicative of a minor loss in the ability to control radiation levels within the plant. It is therefore a potential degradation in the level of safety of the plant.

A water level decrease will be primarily determined by indications from available level instrumentation. Other sources of level indications may include reports from plant personnel (e.g., from a refueling crew) or video camera observations (if available). A significant drop in the water level may also cause an increase in the radiation levels of adjacent areas that can be detected by monitors in those locations.

The effects of planned evolutions should be considered. For example, a refueling bridge area radiation monitor reading may increase due to planned evolutions such as lifting of the reactor vessel head or movement of a fuel assembly. Note that this EAL is applicable only in cases where the elevated reading is due to an unplanned loss of water level.

A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes.

Escalation of the emergency classification level would be via IC RA2.

Basis Reference(s):

1. Technical Specifications 3.7.7 Spent Fuel Storage Pool Water Level

2. OP-179(279)-001 Area Radiation Monitoring System

3. AOP-081-001 Fuel Handling Abnormal Operating Procedure

4. EO-000-104 Secondary Containment Control

5. SSES-FSAR Table 12.3-7 Area Radiation Monitoring System

6. ON-FPC-101(201) Loss of Fuel Pool Cooling

7. NEI 99-01 AU2 Page 43 of 185

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Page 71 of 281 Category: R - 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:

RA3.1 Alert Dose rates > 15 mR/hr in any of the following areas:

• Main Control Room

• Radwaste Control Room

• Both the Central Alarm Station (CAS) and Secondary Alarm Station (SAS)

Mode Applicability:

All Definition(s):

None Basis:

Central Alarm Station (CAS) and Secondary Alarm Station (SAS) are included in this EAL because of their importance in permitting access to areas required to assure safe plant operations (ref. 1). Both are included in this EAL because either security station can effectively permit access to areas required to assure safe plant operations. It is not the intent of this EAL that there be continuous radiation monitoring in the CAS or SAS. However, if a radiological release is in progress and there are indications that the release may affect the CAS and SAS and dose rates in both areas are determined by manual radiological survey to be greater than 15 mR/hr, an Alert shall be declared.

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/Recovery Manager should consider the cause of the increased radiation levels and determine if another IC may be applicable.

Escalation of the emergency classification level would be via Recognition Category R, C or F ICs.

Basis Reference(s):

1. AR-1650683

2. NEI 99-01 AA3 Page 44 of 185

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Page 72 of 281 Category: R - 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:

SRA3.2 Alert

• An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to any Table R-2 rooms or areas (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 R-2 Safe Operation & Shutdown Areas Elevation Unit I Area(s) ** Unit 2 Area(s) ** Mode(s) 670' RB 27 RB 32 3,4,5 683' RB 27, 28, 29 RB 32, 33, 34 3,4,5 703' RB 28, 29 RB 33, 34 3,4,5 719' RB 25, 29 RB 30, 34 3,4,5 749' RB 25, 29 RB 32, 33 3,4,5

• • See Chart I for location of plant areas Chart 1- Plant Area Key Plan 70 LOWWASTE RAD LEVEL WATER TREATMENT

]SPRAY POND 48ACIDAND I I I= VALVE VAULT I]CHLORINE 532510 LIBUILDING [~

PUMPHOUSE

)

ESSW PUMPHOUSE L BINE BLDG. RAD WASTE STURB. 116115 1471341 31 2 1 38 I37 18 I17 201 194. ---4 817 6 5

-I-* I J11 22 21 12 10 9 42 1411 I I

& RF 32 6LN 301276'N 25 36 35 ___COND STORAG

~ T33129T 28

' l I 44 43 SDIESEL GEPNERATOR

_ RIVER E #2 E#1 DIESEL -*

IIINTAKE REACTOR REACTOR STRUCTURE 8*1 E ENERATOR GE Page 45 of 185

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Page 73 of 281 Mode Applicability:

3 - Hot Shutdown 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 measures such as temporary shielding, SCBAs or beyond Emergency Plan RWP dose extensions that are not routinely employed to access the room/area).

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:

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 areas in Table R-2 specify those areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. 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. See Chart I for the specific locations of areas listed in Table R-2. See Attachment 3 for more details of how the Table R-2 was developed (ref. 1).

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/Recovery Manager should consider the cause of the increased radiation levels and determine if another IC may be applicable.

For RA3.2, an Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect at the time of the elevated radiation levels. The emergency classification is not contingent upon whether entry is actually necessary at the time of the increased radiation levels. Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., installing temporary shielding, requiring use of non-routine protective equipment, requesting an extension in dose limits beyond normal administrative limits).

An emergency declaration is not warranted if any of the following conditions apply:

• The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the elevated radiation levels). For example, the plant is in Mode 1 when the radiation Page 46 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 74 of 281 increase occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4.

• The increased radiation levels are a result of a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., radiography, spent filter or resin transfer, etc.).

• The action for which room/area entry is required is of an administrative or record keeping nature (e.g., normal rounds or routine inspections).

• The access control measures are of a conservative or precautionary nature, and would not actually prevent or impede a required action.

• 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.

Escalation of the emergency classification level would be via Recognition Category R, C or F l~s.

Basis Reference(s):

1. Attachment 3 Safe Operation & Shutdown Areas Tables R-2 & H-2 Bases

2. NEI 99-01 AA3 Page 47 of 185

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Page 75 of 281 Categiory 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 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, D- Defueled).

The events of this category pertain to the following subcategories:

1. RPV Level Reactor Pressure Vessel water level is directly related to the status of adequate core cooling and, therefore, fuel clad integrity.

2. Loss of Essential AC Power Loss of essential 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 ESS buses.

3. 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 125 VDC vital buses.

4. RCS Temperature Uncontrolled or inadvertent temperature or pressure increases are indicative of a potential loss of safety functions.

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.

6. Hazardous Event Affectinci Safety Systems Certain hazardous natural and technological events may result in visible damage to or degraded performance of safety systems warranting classification.

Page 48 of 185

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Page 76 of 281 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:

CGI.1 General Emergency RPV level < -161 in. (TAF) for _>30 min. (Note 1)

AND Any Containment Challenge indication, Table C-2 Note 1: The EDIRM should declare the event promptly upon determining that time limit has been exceeded, or will likely be 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)

• PC hydrogen concentration > 6%

• UNPLANNED rise in PC pressure

• Exceeding one or more Secondary Containment Control Max Safe Radiation Levels (EO-000-104 Table 9) that can be read in the Control Room (Table C-6)

Page 49 of 185

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Page 77 of 281 Table C-6 Max Safe Reactor Building Radiation Limits Max Safe Rad RB Area ARM ChannelLit Elevation (f RMNube Description (IR 818 49 Refuel Fioor Area 10 749 52 RWCU Recirc PP Access 10 54 Fuel Pool PP Area 719 50 CR0 North 10 51 CR0 South 670 53 Remote Shutdown Room 10 645 48 HPCI PP & Turbine Room 10 57 ROIC PP & Turbine Room 55 RHR A CPPRoom 56 RHR B DPP Room Mode Applicability:

4 - Cold Shutdown, 5 - Refueling Definition(s):

CONTAINMENT CLOSURE - The procedurally defined conditions or actions taken to secure Primary or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.

As applied to Susquehanna, CONTAINMENT CLOSURE is established per NDAP-QA-0309 (ref. 4) for Primary Containment OR is established per NDAP-QA-0321 (ref. 5) for Secondary Containment.

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 -161 in., core uncovery starts to occur (ref. 1).

Four conditions are associated with a challenge to Primary Containment (PC) integrity:

• CONTAINMENT CLOSURE is not established.

Page 50 of 185

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Page 78 of 281 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 Primary Containment. However, Primary Containment monitoring and/or sampling should be performed to verify this assumption and a General Emergency declared if it is determined that an explosive mixture exists. An explosive mixture can be formed when hydrogen gas concentration in the Primary Containment atmosphere is greater than 6%

by volume in the presence of oxygen (>5%) (ref. 2). In Cold Shutdown and Refueling modes it is assumed that the Primary Containment is de-inerted.

• Any unplanned increase in PC pressure in the Cold Shutdown or Refueling mode indicates a potential loss of CONTAINMENT CLOSURE capability. Unplanned Primary Containment pressure increases indicates CONTAINMENT CLOSURE cannot be assured and the Primary Containment cannot be relied upon as a barrier to fission product release.

• Secondary Containment radiation monitors that can be read in the Control Room should provide indication of increased 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 EO-000-1 04, Secondary Containment Control, (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 re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.

The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a challenge to containment integrity.

In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment. If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged.

Page 51 of 185

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Page 79 of 281 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.

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 Operationat Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.

Basis Reference(s):

1. EO-000-1 01 RPV Control

2. EP-DS-001 Containment Combustible Gas Control

3. EO-000-104 Secondary Containment Control

4. NDAP-QA-0309 Primary Containment Access and Control

5. NDAP-QA-0321 Secondary Containment Integrity Control

6. NEI 99-01 CG1 Page 52 of 185

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Page 80 of 281 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 level cannot be monitored for _>30 min. (Note 1)

AND UNPLANNED increase in any Table C-i sump or tank level due to a loss of RPV inventory of sufficient magnitude to indicate core uncovery AND Any Containment Challenge indication, Table C-2 Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Note 6:lf CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required.

Table C-I Sumps & Tanks

• Drywell equipment drain tank

• Drywell sumps

• Reactor Building sump

• LRW collection tanks

• Main condenser hotwell

• Suppression pool

• Visual observation Table C-2 Containment Challenge Indications]

• CONTAINMENT CLOSURE not established (Note 6)

• PC hydrogen concentration > 6%

• UNPLANNED rise in PC pressure

• Exceeding one or more Secondary Containment Control Max Safe Radiation Levels (EO-000-104 Table 9) that can be read in the Control Room (Table C-6)

Page 53 of 185

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Page 81 of 281 Table C-6 Max Safe Reactor Building Radiation Limits Max Safe Rad RB Area ARM ChannelLit Elevation (ft) AR ubrDescription (IR 818 49 Refuel Floor Area 10 749 52 RWCU Recirc PP Access 10 54 Fuel Pool PP Area 719 50 CRD North 10 51 CRD South 670 53 Remote Shutdown Room 10 645 48 HPCI PP & Turbine Room 10 57 RCIC PP & Turbine Room 55 RHR A CPP Room 56 RHR B DPP Room 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.

CONTAINMENT CLOSURE - The procedurally defined conditions or actions taken to secure Primary or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.

As applied to Susquehanna, CONTAINMENT CLOSURE is established per NDAP-QA-0309 (ref. 7) for Primary Containment OR is established per NDAP-QA-0321 (ref. 8) for Secondary Containment.

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 or temporary RPV shutdown level transmitter (ref. 1).

If RPV level monitoring capability is unavailable, the RPV inventory loss must be detected by the leakage indications listed in Table C-I. Level increases must be evaluated against other Page 54 of 185

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Page 82 of 281 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 tank level and drywell floor sump level is the normal method of monitoring and calculating leakage from the RPV (ref. 2). A Reactor Building sump level rise may also be indicative of RCS inventory losses external to the Primary 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 (ref. 3, 4). 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.

Four conditions are associated with a challenge to Primary Containment (PC) 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 Primary Containment. However, Primary Containment monitoring and/or sampling should be performed to verify this assumption and a General Emergency declared if it is determined that an explosive mixture exists. An explosive mixture can be formed when hydrogen gas concentration in the Primary Containment atmosphere is greater than 6%

by volume in the presence of oxygen (>5%) (ref. 5). In Cold Shutdown and Refueling modes it is assumed that the Primary Containment is de-inerted.

• Any unplanned increase in PC pressure in the Cold Shutdown or Refueling mode indicates a potential loss of CONTAINMENT CLOSURE capability. Unplanned Primary Containment pressure increases indicates CONTAINMENT CLOSURE cannot be assured and the Primary Containment cannot be relied upon as a barrier to fission product release.

• Secondary Containment radiation monitors should provide indication of increased 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 EO-000-1 04, Secondary Containment Control, (ref. 6).

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 re-established prior to exceeding the 30-minute time limit, then declaration of a General Page 55 of 185

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Page 83 of 281 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.

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 RCS.

This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 9 1-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear PowerPlants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.

Basis Reference(s):

1. IC 180 005 Installation and Removal of Temporary Unit 1 RPV Shutdown Level Transmitter

2. ON-100(200)-005 Excess Drywell Leakage Identification

3. OP-149(249)-002 RHR Operation in Shutdown Cooling Mode

4. ON-149(249)-001 Loss of RHR Shutdown Cooling Mode

5. EP-DS-001 Containment Combustible Gas Control

6. EO-000-104 Secondary Containment Control

7. NDAP-QA-0309 Primary Containment Access and Control

8. NDAP-QA-0321 Secondary Containment Integrity Control

9. NEI 99-01 CG1 Page 56 of 185

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Page 84 of 281 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 CLOSU RE not established AND RPV level < -129 in. (Level 1)

Mode Applicability:

4 - Cold Shutdown, 5 - Refueling Definition(s):

CONTAINMENT CLOSURE - The procedurally defined conditions or actions taken to secure Primary or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.

As applied to Susquehanna, CONTAINMENT CLOSURE is established per NDAP-QA-0309 (ref. 3) for Primary Containment OR is established per NDAP-QA-0321 (ref. 4) for Secondary Containment.

Basis:

RPV level is normally monitored using the instruments in Figure C-i (ref. 1, 2).

When RPV level decreases to -129 in., RPV water level is below the low-low-low ECCS actuation setpoint (Level 1) (ref. 1).

The magnitude of this loss of water indicates that makeup systems have not been effective and may not be capable of preventing further RPV water level decrease 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.

This IC addresses a significant and prolonged loss of ROS level 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.

Page 57 of 185

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Page 85 of 281 The difference in the specified RPV levels of CS 1.1 and CS 1.2 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 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 RGI.

Basis Reference(s):

1, M-142 P&ID Nuclear Boiler Vessel Instrumentation, Sheets 1, 2

2. ON-145(245)-004 RPV Water Level Anomaly

3. NDAP-QA-0309 Primary Containment Access and Control

4. NDAP-QA-0321 Secondary Containment Integrity Control

5. NEI 99-01 CS1 Page 58 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 86 of 281 Figure C-I RPV Levels (ref. 1, 2) 5O00 SHUTDOWN RANGE 0"

H;-INDICAIlON (INCHES)

+54 7 566.5 +39 6 562.5 +35 5 562.5 +35 4 557.5 +30 3 540.5 +13 2 489.5 -38 1 398.5 -129 0 324.5 -203 Page 59 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 87 of 281 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RPV Level Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability EAL:

CSI.2 Site Area Emergency CONTAI NMENT CLOSU RE established AND RPV level < -161 in. (TAF)

Mode Applicability:

4 - Cold Shutdown, 5 - Refueling Definition(s):

CONTAINMENT CLOSURE - The procedurally defined conditions or actions taken to secure Primary or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.

As applied to Susquehanna, CONTAINMENT CLOSURE is established per NDAP-QA-0309 (ref. 2) for Primary Containment OR is established per NDAP-QA-0321 (ref. 3) for Secondary Containment.

Basis:

When RPV level drops to the top of active fuel (TAF) (an indicated RPV level of -161 in.), core uncovery starts to occur (ref. 1).

This IC addresses a significant and prolonged loss of RPV level 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 vessel levels of 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 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.

Page 60 of 185

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Page 88 of 281 Escalation of the emergency classification level would be via IC CG1 or RG1.

Basis Reference(s):

1. EO-O00-1 02 RPV Control

2. NDAP-QA-0309 Primary Containment Access and Control

3. NDAP-QA-0321 Secondary Containment Integrity Control

4. NEI 99-01 CS1 Page 61 of 185

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Page 89 of 281 C - Cold Shutdown / Refueling System Malfunction Category:

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 UNPLANNED increase in any Table C-i sump or tank level due to a loss of RPV inventory of sufficient magnitude to indicate core uncovery Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Table C-I Sumps & Tanks

• Drywell equipment drain tank

• Drywell sumps

• Reactor Building sump

• LRW collection tanks

• Main condenser hotwell

• Suppression pool

• Visual observation 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:

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 or temporary RPV shutdown level transmitter (ref. 1).

If RPV level monitoring capability is unavailable, the RPV inventory loss must be detected by the leakage indications listed in Table C-i. Level increases 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 tank level and drywell floor sump level is the normal method of monitoring and calculating leakage from the RPV (ref. 2). A Reactor Building sump level rise may also be indicative of RCS inventory losses external to the Primary Containment from Page 62 of 185

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Page 90 of 281 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 (ref. 3, 4). 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 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.

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 RCS.

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 RG1 Basis Reference(s):

1. IC 180 005 Installation and Removal of Temporary Unit 1 RPV Shutdown Level Transmitter

2. ON-100(200)-005 Excess Drywell Leakage Identification

3. OP-149(249)-002 RHR Operation in Shutdown Cooling Mode

4. ON-149(249)-001 Loss of RHR Shutdown Cooling Mode

5. NEI 99-01 CS1 Page 63 of 185

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Page 91 of 281 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RPV Level Initiating Condition: Loss of RPV inventory EAL:

CA1.1 Alert Loss of RPV inventory as indicated by RPV level < -38 in. (Level 2)

Mode Applicability:

4 - Cold Shutdown, 5 - Refueling Definition(s):

None Basis:

The threshold RPV level of -38 in. is the low-low ECOS actuation setpoint (ref. 1). RPV level is normally monitored using the instruments in Figure C-i (ref. 1, 2).

When reactor vessel water level drops to -38 in. high pressure steam-driven injection sources HPCI (ECCS) and RCIC receive an initiation signal (ref. 1, 2). Although these systems cannot restore RCS inventory in the cold condition, the Level 2 actuation setpoint is operationally significant and is indicative of a loss of RCS inventory significantly below the low level scram setpoint specified in CUlI..

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 water level below - 38 in. indicates that operator actions have not been successful in restoring and maintaining RPV water level. The heat-up rate of the coolant will increase as the available water inventory is reduced. A continuing decrease in water level will lead to core uncovery.

Although related, this EAL is concerned with the loss of RCS inventory and not the potential concurrent effects on systems needed for decay heat removal (e.g., loss of a Residual Heat Removal suction point). An increase in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA4.

If RPV water level continues to lower, then escalation to Site Area Emergency would be via IC CS1.

Basis Reference(s):

1. M-142 P&ID Nuclear Boiler Vessel Instrumentation, Sheets 1, 2

2. ON-145(245)-004 RPV Water Level Anomaly

3. NEI 99-01 CA1 Page 64 of 185

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Page 92 of 281 Figure C-I RPV Levels (ref. 1, 2) 1 180" -180" 658.50" LMAIN U

STEAM UP EXTENED SHUTDOWN UNARO WIDANGEN IRANGE

• 0" i 1 50"r i

-11" -150" ZONF LEEL HEIGHT ABOVE INDICATION RANGEVESSEL ZERO (INCHES)

!t RANGE(INCHES) 630 562.5 +35 5 562.5 +35 INTUET4 557.5 +30 TAP 3 540.5 +13 2 489.5 -38

• ,1 398.5 -129 0 324.5 -203 Page 65 of 185

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Page 93 of 281 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RPV Level Initiating Condition: Loss of RPV inventory EAL:

CA1.2 Alert RPV level cannot be monitored for _>15 min. (Note 1)

AND UNPLANNED increase in any Table C-i sump or tank levels due to a loss of RPV inventory Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Table C-I Sumps & Tanks

• Drywell equipment drain tank

• Drywell sumps

• Reactor Building sump

• LRW collection tanks

• Main condenser hotwell

• Suppression pool

• Visual observation 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:

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 or temporary RPV shutdown level transmitter (ref. 1).

In this EAL, all water level indication is unavailable for greater than 15 minutes and the RPV inventory loss must be detected by the leakage indications listed in Table C-i. Level increases 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 tank level and drywell floor sump level is the normal method of monitoring and calculating leakage from the RPV (ref. 2). A Reactor Building sump level rise may also be indicative of RCS inventory losses external to the Primary Containment from systems connected to the RPV. With Page 66 of 185

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Page 94 of 281 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 (ref. 3, 4). 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 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, 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 CSI.

Basis Reference(s):

1. IC 180 005 Installation and Removal of Temporary Unit 1 RPV Shutdown Level Transmitter

2. ON-100(200)-005 Excess Drywell Leakage Identification

3. OP-149(249)-002 RHR Operation in Shutdown Cooling Mode

4. ON-149(249)-001 Loss of RHR Shutdown Cooling Mode

5. NEI 99-01 CA1 Page 67 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 95 of 281 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RPV Level Initiating Condition: UNPLANNED loss of RPV inventory for 15 minutes or longer EAL:

CUI.1 Unusual Event UNPLANNED loss of reactor coolant results in RPV level less than a required lower limit for _>15 min. (Note 1)

Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be 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:

Figure C3-i illustrates the elevations of the RPV level instrument ranges (ref. 1).

With the plant in Cold Shutdown, RPV water level is normally maintained above the RPV low level scram setpoint of +13 in. (ref. 1, 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 RO3S 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 3.9.6 requires at least 22 ft of water above the top of the reactor vessel flange in the refueling cavity during refueling operations). The RPV flange is at an indicated level of 217.5 in. as indicated on the Shutdown Range RPV water level instrument (ref. 3).

EAL RU2.1 may also be applicable based on increasing radiation levels due to loss of inventory in the REFULING PATHWAY.

This IC3 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 decrease RC3S water inventory are carefully planned and controlled.

An UNPLANNED event that results in water level decreasing below a procedurally required limit warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the core covered.

Page 68 of 185

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Page 96 of 281 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.

Basis Reference(s):

1. M-t42 P&ID Nuclear Boiler Vessel Instrumentation, Sheets 1, 2

2. ON-145(245)-004 RPV Water Level Anomaly

3. GO-100(200)-006 Cold Shutdown, Refueling and Defueled

5. NEI 99-01 CU1 Page 69 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 97 of 281 Figure C-I RPV Levels (ref. 1, 2) 500" SHUTDOWN RANGE 0"

HEIGHT ABOVE INDICATION LVL VESSEL ZERO (INCHES)

(INCHES) ____

8 581.5 +54 7 566.5 +39 6 502.5 +35 5 562.5 +35 4 557.5 +30 3 540.5 +13 2 489.5 -38 1 398.5 -129 0 324.5 -203 Page 70 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 98 of 281 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RPV Level Initiating Condition: UNPLANNED loss of RPV inventory for 15 minutes or longer EAL:

CU1.2 Unusual Event RPV water level cannot be monitored AND UNPLANNED increase in any Table C-i sump or tank levels due to a loss of RPV inventory (Notes 1,9)

Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Note 9: If the conditions specified in this EAL last longer than 15 minutes, CA1 .2 applies.

Table C-I Sumps & Tanks ]

• Drywell equipment drain tank

• Drywell sumps

• Reactor Building sump

• LRW collection tanks

• Main condenser hotwell

• Suppression pool

• Visual observation 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:

If the conditions specified in this EAL last longer than 15 minutes, CAl1.2 applies.

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 or temporary RPV shutdown level transmitter (ref. 1).

Page 71 of 185

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Page 99 of 281 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-i. Level increases 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 tank level and drywell floor sump level is the normal method of monitoring and calculating leakage from the RPV (ref. 2). A Reactor Building sump level rise may also be indicative of RCS inventory losses external to the Primary 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 (ref. 3, 4). 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 decrease RCS water inventory are carefully planned and controlled.

An UNPLANNED event that results in water level decreasing below a procedurally required limit warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the core covered.

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. 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 CA1 or CA4.

Basis Reference(s):

1. IC 180 005 Installation and Removal of Temporary Unit 1 RPV Shutdown Level Transmitter

2. ON-i100(200)-005 Excess Drywell Leakage Identification

3. OP-149(249)-002 RHR Operation in Shutdown Cooling Mode

4. ON-149(249)-001 Loss of RHR Shutdown Cooling Mode

5. NEI 99-01 CUi Page 72 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 100 of 281 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 essential buses for 15 minutes or longer EAL:

CA2.1 Alert Loss of ALL offsite and ALL onsite AC power capability to ALL 4.16 kV ESS buses on EITHER unit for

Ž>15 min. (Note 1)

Note 1: The EDIRM should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability:

4 - Cold Shutdown, 5 - Refueling, D - Defueled Basis:

The Class IlE 4.16 kV system supplies all the Engineered Safety Feature (ESF) loads and other loads that are needed for a safe and orderly plant shutdown, and for keeping the plant in a safe shutdown condition. See Figure C-2 (ref. 1, 2) The eight Class IlE 4.16 kV ESS Buses 1(2)A through I (2)D receive power from either the four ESS 13.8/4.16 kV transformers or the diesel generators (A, B, C, D and additional diesel generator E). Buses IA-I D supply Unit 1 and common loads and Buses 2A-2D supply Unit 2 loads. This configuration prevents a loss of all ESS Buses for one unit in the event one of the ESS Transformers is lost.

During normal plant operation, ESS Transformer 101 supplies preferred power to ESS Bus IA and 2A and is an alternate power supply to ESS bus 1ID and 2D. ESS Transformer 111 supplies preferred power to ESS Bus 1IC and 2C, and is an alternate power supply to ESS bus 1IB and 2B. ESS Transformer 201 supplies preferred power to ESS Bus 1ID and 2D, and is an alternate power supply to ESS bus IA and 2A. ESS Transformer 211 supplies preferred power to ESS Bus 1IB and 2B, and is an alternate power supply to ESS bus 1IC and 2C.

On a loss of a preferred power source, the bus rapidly transfers to the alternate power source to maintain component power. If both the preferred and alternate power sources are lost, the associated standby diesel generator connects to the ESS bus. (ref. 2-6)

This cold condition EAL is equivalent to the hot condition loss of all offsite AC power EAL SSI .I.

This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink.

When in the cold shutdown, refueling, or defueled mode, this condition is not classified as a Site Area Emergency because of the increased time available to restore an emergency bus to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this Page 73 of 185

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Page 101 of 281 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 RSI.

Basis Reference(s):

1. FSAR Section 8.20Offsite Power System

2. ESAR Section 8.30Onsite Power System

3. Technical Specifications 3.8.2 AC Sources - Shutdown

4. Technical Specifications 3.8.8 Distribution System - Shutdown

5. ON-i104 (204)-001 Units 1(2) Response to Loss of All Offsite Power

6. EO-100 (200)-030 UNIT 1(2) Response to Station Blackout

7. NEI 99-01 CA2 Page 74 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 102 of 281 Figure C-2 ESS 13.8/4.16 kV Transformers and Distribution (ref. 1)

MONTOU IF

( H*

TI 0I o I-r*

T-211 ESS211 0X(213 Page 75 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 103 of 281 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 essential buses for 15 minutes or longer EAL:

0U2.1 Unusual Event AC power capability to ALL 4.16 kV ESS buses on EITHER unit (Table C-7) 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 EDIRM should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

[ Table C-7 AC 4.16 kV ESS buses Unit 1:

• 1A201

• 1A202

• 1A203

• 1A204 Unit 2:

• 2A201

• 2A202

• 2A203

• 2A204 Mode Applicability:

4 - Cold Shutdown, 5 - Refueling, D - 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 76 of 185

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Page 104 of 281 (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

Basis:

The Class 1E 4.16 kV system supplies all the Engineered Safety Feature (ESF) loads and other loads that are needed for a safe and orderly plant shutdown, and for keeping the plant in a safe shutdown condition. See Figure C-2 (ref. 1, 2) The eight Class 1E 4.16 kV ESS Buses 1(2)A through 1 (2)D receive power from either the four ESS 13.8/4.16 kV transformers or the diesel generators (A, B, C, D and additional diesel generator E). Buses 1A-1 D supply Unit 1 and common loads and Buses 2A-2D supply Unit 2 loads. This configuration prevents a loss of all ESS Buses for one unit in the event one of the ESS Transformers is lost.

During normal plant operation, ESS Transformer 101 supplies preferred power to ESS Bus 1A and 2A and is an alternate power supply to ESS bus 1 D and 2D. ESS Transformer 111 supplies preferred power to ESS Bus 1C and 2C, and is an alternate power supply to ESS bus 1B and 2B. ESS Transformer 201 supplies preferred power to ESS Bus 1 D and 2D, and is an alternate power supply to ESS bus 1A and 2A. ESS Transformer 211 supplies preferred power to ESS Bus 1 B and 2B, and is an alternate power supply to ESS bus 1C and 2C.

On a loss of a preferred power source, the bus rapidly transfers to the alternate power source to maintain component power. If both the preferred and alternate power sources are lost, the associated standby diesel generator connects to the ESS bus. (ref. 2-6)

This cold condition EAL is equivalent to the hot condition EAL SAl1.1.

This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment.

When in the cold shutdown, refueling, or defueled mode, this condition is not classified as an Alert because of the increased time available to restore another power source to service.

Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition is considered to be a potential degradation of the level of safety of the plant.

An "AC power source" is a source recognized in AOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.

• A loss of all offsite power with a concurrent failure of one division of emergency power sources (e.g., onsite diesel generators).

• A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from the unit main generator.

• A loss of emergency power sources (e.g., onsite diesel generators) with a single division of emergency buses being back-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.

Page 77 of 185

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Page 105 of 281 Basis Reference(s):

1. FSAR Section 8.20ffsite Power System

2. FSAR Section 8.3 Onsite Power System

3. Technical Specifications 3.8.2 AC Sources - Shutdown

4. Technical Specifications 3.8.8 Distribution System - Shutdown

5. ON-i104 (204)-001 Units 1(2) Response to Loss of All Offsite Power

6. EO-100 (200)-030 UNIT 1(2) Response to Station Blackout

7. NEI 99-01 CU2 Page 78 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 106 of 281 Figure C-2 ESS 13.814.16 kV Transformers and Distribution (ref. 1)

T-211 ESS2t1 0X213 Page 79 of 185

Attachment 1 EP-RM-004 Revision XIX Page 107 of 281 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - Loss of Vital DC Power Initiating Condition: Loss of vital DC power for 15 minutes or longer EAL:

0U3.1 Unusual Event

< 105 VDC bus voltage indications on Technical Specification required 125 VDC buses for _>15 min. (Note 1)

Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability:

4 - Cold Shutdown, 5 - Refueling Definition(s):

None Basis:

The Class IE Battery Banks are 1(2)D610 (Channel A), 1 (2)D620 (Channel B), 1(2)0630 (Channel C), and 1 (2)D640 (Channel D). Each bank consists of 60 cells connected in series.

Each cell produces a nominal voltage of 2.06 VDC resulting in a total battery bank terminal voltage of 123.6 VDC. All battery banks are designed to supply power to its load center for four hours in the event of a loss of power from its battery charger (ref. 1-3).

105 VDC is the minimum design voltage limit (ref. 4).

Indicated voltage for the vital 125 VDC main distribution buses is local only. Local voltage indication is available for each bus based on dispatching a field operator in accordance with Control Room alarm response procedure AR-1(2)06-001 (A12,B12,C12,D12). The 15 minute classification clock begins upon receipt of the 125V DC Panel System Trouble alarm in the Control Room. If battery voltage cannot be verified to be greater than or equal to 105 VDC within the 15 minutes, emergency classification must be made under this EAL.

This EAL is the cold condition equivalent of the hot condition loss of DC power EAL SS2.1.

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 increase the time available to restore a vital DC bus to service. Thus, this condition is considered to be a potential degradation of the level of safety of the plant.

As used in this EAL, "required" means the vital DC buses necessary to support operation of the in-service, or operable, train or trains of SAFETY SYSTEM equipment. For example, if Division I is out-of-service (inoperable) for scheduled outage maintenance work and Division II is in-service (operable), then a loss of Vital DC power affecting Division II would require the Page 80 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 108 of 281 declaration of an Unusual Event. A loss of Vital DC power to Division I 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 CA4, or an IC in Recognition Category R.

Basis Reference(s):

1. ESAR Section 8.3.2 DC Power Systems

2. Susquehanna Drawing No. E107159, Sheet 1, "Single Line Meter & Relay Diagram 125 VDC, 250 VDC & 120 VAC Systems"

3. Technical Specifications 3.8.5 DC Sources - Shutdown

4. ON-102(202)-610, -620, -630, -640 Loss of 125V DC

5. AR-1(2)06-001 Main Turbine/Generator, Computer HVAC, Instrument AC, 24V DC, 125V DC, 250V DC Panel 2C651

6. NEI 99-01 CU4 Page 81 of 185

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Page 109 of 281 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 4 - RCS Temperature Initiating Condition: Inability to maintain the plant in cold shutdown EAL:

CA4.1 Alert UNPLANNED increase in RCS temperature to > 200°F for > Table 0-3 duration (Notes 1, 10)

OR UNPLANNED RPV pressure increase > 10 psig due to loss of decay heat removal capability Note 1: The ED/RM should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

Note 10: In the absence of reliable RCS temperature indication caused by the loss of decay heat removal capability, classification should be based on the RPV pressure increase 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.

Table C-3: RCS Heat-up Duration Thresholds RCS SatusCONTAINMENT CLOSURE Ha-pDrto RCS Status ~~Status Ha-pDrto INTACT N/A 60 min.*

established 20 min.*

Not INTACT 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 procedurally defined conditions or actions taken to secure Primary or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.

As applied to Susquehanna, CONTAINMENT CLOSURE is established per NDAP-QA-0309 (ref. 3) for Primary Containment OR is established per NDAP-QA-0321 (ref. 4) for Secondary Containment.

RCS/INTACT - The RCS should be considered intact when the RCS pressure boundary is in its normal condition for the cold shutdown mode of operation (e.g., no freeze seals).

Page 82 of 185

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Page 110 of 281 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 the absence of reliable RCS temperature indication caused by the loss of decay heat removal capability, classification should be based on the RPV pressure increase 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.

Available methods of determining RCS temperature can be found in Operations surveillance procedures (ref. 1, 2).

10 psig is the lowest pressure increase increment that can be reasonably read in the control room (ref. 2):

This IC addresses conditions involving a loss of decay heat removal capability or an addition of heat to the RCS in excess of that which can currently be removed. Either condition represents an actual or potential substantial degradation of the level of safety of the plant.

A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification.

The RCS Heat-up Duration Thresholds table addresses an increase in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact.. The 20-minute criterion was included to allow time for operator action to address the temperature increase.

The RCS Heat-up Duration Thresholds table also addresses an increase in RCS temperature with the RCS intact. The status of CONTAINMENT CLOSURE is not crucial in this condition since the intact RCS is providing a high pressure barrier to a fission product release. The 60-minute time frame should allow sufficient time to address the temperature increase without a substantial degradation in plant safety.

Finally, in the case where there is an increase in RCS temperature, the RCS is not intact, and CONTAINMENT CLOSURE is not established, no heat-up duration is allowed (i.e., 0 minutes).

This is because 1) the evaporated reactor coolant may be released directly into the Primary Containment or Reactor Building atmosphere and subsequently to the environment, and 2) there is reduced reactor coolant inventory above the top of irradiated fuel.

The RCS pressure increase threshold 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 RSI.

Basis Reference(s):

1. SO-100(200)-0l1 Reactor Vessel Temperature and Pressure Recording

2. SO-100(200)-006 Shiftly Surveillance Operating Log

3. NEI 99-01 CA3 Page 83 of 185

Attachment 1 EP-RM-004 Revision IX]

Page 111 of 281 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 4 - RCS Temperature Initiating Condition: UNPLANNED increase in RCS temperature EAL:

CU4.1 Unusual Event UNPLANNED increase in RCS temperature to > 200°F due to loss of decay heat removal capability 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:

In the absence of reliable ROS temperature indication caused by a loss of decay heat removal capability, classification should be based on EAL CU4.2 should RPV level indication be subsequently lost.

Available methods of determining RCS temperature can be found in Operations surveillance procedures (ref. 2, 3).

This IC addresses an UNPLANNED increase 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/Recovery Manager should also refer to IC CA4.

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 inventory may result in a rapid increase in reactor coolant temperature depending on the time after shutdown.

Page 84 of 185

Attachment 1 EP-RM-004 Revision IX]

Page 112 of 281 Escalation to Alert would be via IC CAl based on an inventory loss or IC CA4 based on exceeding plant configuration-specific time criteria.

Basis Reference(s):

1. Technical Specifications Table 1.1-1

2. SO-100(200)-Ol11 Reactor Vessel Temperature and Pressure Recording

3. SO-100(200)-006 Shiftly Surveillance Operating Log

4. NEI 99-01 CU3 Page 85 of 185

Attachment 1 EP-RM-004 Revision IX]

Page 113 of 281 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 4 - RCS Temperature Initiating Condition: UNPLANNED increase in ROS temperature EAL:

CU4.2 Unusual Event Loss of ALL RCS temperature and RPV level indication for _>15 min. (Note 1)

Note 1: The ED/RM should declare the event promptly upon determining that'time limit has been exceeded, or will likely be exceeded.

Mode Applicability:

4 - Cold Shutdown, 5 - Refueling Definition(s):

None Basis:

Available methods of determining RCS temperature can be found in Operations surveillance procedures (ref. 4, 5).

RPV level is normally monitored using the instruments in Figure 0-2 (ref. 1, 2).

This EAL addresses the inability to determine RCS temperature and RPV 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/Recovery Manager should also refer to IC CA4.

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.

Escalation to Alert would be via IC CAl based on an inventory loss or IC CA4 based on exceeding plant configuration-specific time criteria.

Basis Reference(s):

1. M-142 P&ID Nuclear Boiler Vessel Instrumentation, Sheets 1, 2

2. ON-145(245)-004 RPV Water Level Anomaly

3. Technical Specifications Table 1.1-1

4. SO-100(200)-01l Reactor Vessel Temperature and Pressure Recording Page 86 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 114 of 281

5. SO-100(200)-006 Shiftly Surveillance Operating Log

6. NE! 99-01 CU3 Page 87 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 115 of 281 Figure C-2 RPV Levels (ref. 1, 2) 500" SHUTDOWN RANGE 0"

- I -*R I

- I- I -

INDICATION (INCHES)

II *I*..B 171Gt.*

+54

-t I

+39

-t I

+35

+35

+30

+13 2 489.5 -38 1 398.5 -129 0 324.5 -203 Page 88 of 185

Attachment 1 EP-RM-004 Revision IX]

Page 116 of 281 C - Cold Shutdown / Refueling System Malfunction Category:

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-4 onsite communication methods OR Loss of ALL Table C-40ORO communication methods OR Loss of ALL Table C-4 NRC communication methods Table C-4 Communication Methods System Onsite ORO NRC UHF Radio X Plant PA System X Dedicated Conference Lines X Commercial Telephone Systems X X X Cellular Telephone X X FTS-2001 (ENS) X X Mode Applicability:

4 - Cold Shutdown, 5 - Refueling, D - Defueled Definition(s):

None Basis:

Onsite/offsite communications include one or more of the systems listed in Table C-4 (ref. 1, 2, 3).

UHF Radio Onsite portable radio communication systems are described in the Susquehanna SES Physical Security Plan and in the Susquehanna SES Emergency Plan. Four UHF channels, each Page 89 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 117 of 281 consisting of two frequencies for duplex operation through one of five in-plant repeaters, provide onsite portable radio communications. Operations is assigned two channels; one channel is assigned to Unit 1 and one to Unit 2. Operators in the plant on rounds and on specific assignments are equipped with handheld two-way radios.

Plant PA System The plant PA system is an intra-plant public address providing the following functions:

• A 5-channel page-talk handset intercom system for on-site communications between plant locations.

• Broadcast accountability and fire alarms designed to warn personnel of emergency conditions.

The system consists of telephone handsets, amplifiers and loudspeakers located at various selected areas throughout the plant.

Dedicated Conference Lines (Centrex Three (3) digqit dialingq)

The Dedicated Conference Lines are those normally used to communicate with several offsite agencies at one time (e.g., 191 conference line).

Commercial Telephone Systems Two independent telecommunications networks exist to provide primary and backup telephone communications between ERFs and offsite agencies.

Plant Cellular Telephone Cell phones can be utilized to perform both ORO and NRC communications.

ETS 2001 (ENS)

This system is for NRC offsite communications but may also be used to perform ORO notifications.

This EAL is the cold condition equivalent of the hot condition EAL SU7.1.

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 OROs and the NRC.

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

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 OROs of an emergency declaration. The OROs referred to here are the Commonwealth of Pennsylvania, Luzerne and Columbia County EOCs The third EAL addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.

Page 90 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 118 of 281 Basis Reference(s):

1. EP-RM-007 Emergency Telephone Instructions and Directory

2. SSES Emergency Plan Section 8

3. ESAR Section 9.5.2

4. NEI 99-01 CU5 Page 91 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 119 of 281 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 6 - Hazardous Event Affecting Safety Systems Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode EAL:

CA6.1 Alert The occurrence of any Table 0-5 hazardous event AND EITHER:

• Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode

• The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode Table C-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 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 post-event inspection to determine if the attributes of an explosion are present.

Page 92 of 185

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Page 120 of 281 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 100FR50.2):

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 component or structure that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure.

Basis:

• The significance of a seismic event is discussed under EAL HU2.1 (ref. 1, 2).

• Internal FLOODING may be caused by events such as component failures, equipment misalignment, or outage activity mishaps (ref. 3, 4, 5).

• Seismic Category I structures are analyzed to withstand a sustained, design wind velocity of at least 80 mph. (ref. 6, 7).

• Areas containing functions and systems required for safe shutdown of the plant are identified by Fire Zone in the fire response procedure (ref. 8, 9).

• An EXPLOSION that degrades the performance of a SAFETY SYSTEM train or visibly damages a SAFETY SYSTEM component or structure would be classified under this EAL.

This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure containing SAFETY SYSTEM components, needed for the current operating mode.

This condition significantly reduces the margin to a loss or potential loss of a fission product barrier, and therefore represents an actual or potential substantial degradation of the level of safety of the plant.

The first conditional 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.

The second conditional addresses damage to a SAFETY SYSTEM component that is not in service/operation or readily apparent through indications alone, or to a structure containing SAFETY SYSTEM components. Operators will make this determination based on the totality of Page 93 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 121 of 281 available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage.

Escalation of the emergency classification level would be via IC CS1 or RSI.

Basis Reference(s):

1. ON-000-002 Severe Weather / Natural Phenomena

2. FSAR Section 3.7 Seismic Design

3. ON-169(269)-001 Flooding in Turbine Building

4. ON-169(269)-002 Flooding in Reactor Building

5. ESAR Section 3.4 Water Level (Flood) Design

6. FSAR Section 3.3 Wind and Tornado Loadings

7. FSAR Section 3.5 Missile Projection

8. SSES-FPRR Section 6.2 Fire Area Description

9. ON-013-001 Response to Fire

9. NEI 99-01 CA6 Page 94 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 122 of 281 Category H - Hazards and Other Conditions Affecting Plant Safety EAL Group: ANY (EALs in this category are applicable to any piant 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 Technologqy 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 site 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 If the Control Room must be evacuated, additional support for monitoring and controlling plant functions is necessary through the emergency response facilities.

7. ED/RM 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/Recovery Manager the latitude to classify emergency conditions consistent with the established classification criteria based upon Emergency Director/Recovery Manager judgment.

Page 95 of 185

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Page 123 of 281 Category: H - Hazards Subcategory: 1 - Security Initiating Condition: HOSTILE ACTION resulting in loss of physical control of the facility EAL:

HGI.1 General Emergency A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the Security Shift Supervision AND EITHER of the following has occurred:

Any of the following safety functions cannot be controlled or maintained

• Reactivity

• RPV water level

• RCS heat removal OR Damage to spent fuel has occurred or is IMMINENT Mode Applicability:

All Definition(s):

HOSTILE ACTION- An act toward Susquehanna 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 Susquehanna. 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).

PROTECTED AREA - Area within the station inner security fence (PROTECTED AREA Barrier) designated to implement the requirements of 10 CFR 73.

Basis:

Security Shift Supervision are the designated on-site personnel qualified and trained to confirm that a security event is occurring or has occurred. Training on security event classification confirmation is closely controlled due to the strict secrecy controls placed on the SSES Physical Security Plan (Safeguards) information (ref. 1).

This IC does not apply to a HOSTILE ACTION directed at an ISESI PROTECTED AREA; such an attack should be assessed using IC HAl.

Page 96 of 185

Attachment 1 EP-RM-004 Revision IX]

Page 124 of 281 If the plant equipment necessary to maintain the safety functions can be controlled from another location, then the EAL is not met.

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

Timely and accurate communications between 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.

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 Susquehanna Physical Security Plan (ref.1).

Basis Reference(s):

1. SSES Physical Security Plan

2. ON-000-010 Security Event

3. NEI 99-01 HG1 Page 97 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 125 of 281 Category: H - Hazards Subcategory: 1 - Security Initiating Condition: HOSTILE ACTION within the PROTECTED AREA EAL:

HSI.1 Site Area Emergency A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the Security Shift Supervision Mode Applicability:

All Definition(s):

HOSTILE ACTION- An act toward Susquehanna 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 Susquehanna. 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).

PROTECTED AREA - Area within the station inner security fence (PROTECTED AREA Barrier) designated to implement the requirements of 10 CFR 73.

Basis:

Security Shift Supervision are the designated on-site personnel qualified and trained to confirm that a security event is occurring or has occurred. Training on security event classification confirmation is closely controlled due to the strict secrecy controls placed on the SSES Physical Security Plan (Safeguards) information (ref. 1).

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.

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and QualificationPlan, Safeguards Contingency Plan.

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.

Page 98 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 126 of 281 This IC does apply to a HOSTILE ACTION directed at the ISESI since it is within the PROTECTED AREA.. It 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 Susquehanna Physical Security Plan (ref. 1).

Escalation of the emergency classification level would be via IC HGI.

Basis Reference(s):

1. SSES Physical Security Plan

2. ON-000-010 Security Event

3. NEI 99-01 HS1 Page 99 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 127 of 281 Category: H - Hazards Subcategory: 1 - Security Initiating Condition: HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes EAL:

HAI.1 Alert A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the 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):

HOSTILE ACTION - An act toward Susquehanna 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 Susquehanna. 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 - Includes the area within the expanded security perimeter, i.e.,

the areas that are bordered by the Vehicle Barriers System. The OWNER CONTROLLED AREA also includes the Monitored OWNER CONTROLLED AREA (MOCA) as defined in Security Procedures.

Basis:

Security Shift Supervision are the designated on-site personnel qualified and trained to confirm that a security event is occurring or has occurred. Training on security event classification confirmation is closely controlled due to the strict secrecy controls placed on the SSES Physical Security Plan (Safeguards) information (ref. 1).

This IC addresses the occurrence of a HOSTILE ACTION within the OWNER CONTROLLED AREA or notification of an aircraft attack threat. This event will require rapid response and assistance due to the possibility of the attack progressing to the PROTECTED AREA, or the need to prepare the plant and staff for a potential aircraft impact.

Timely and accurate communications between the 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.

Page 100 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 128 of 281 As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protectiye 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 IC does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc.

Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72.

The first threshold is applicable for any HOSTILE ACTION occurring, or that has occurred, in the OWNER CONTROLLED AREA. This includes any action directed against an ISFSI that is located outside the plant PROTECTED 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 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 site-specific security procedures.

The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may be provided by NORAD through the NRC.

In some cases, it may not be readily apparent if an aircraft impact within the OWNER CONTROLLED AREA was intentional (i.e., a HOSTILE ACTION). It is expected, although not certain, that notification by an appropriate Federal agency to the site would clarify this point. In this case, the appropriate federal agency is intended to be NORAD, FBI, FAA or NRC. The emergency declaration, including one based on other ICs/EALs, should not be unduly delayed while awaiting notification by a Federal agency.

Emergency plans and implementing procedures are public documents; therefore, EALs 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 Susquehanna Physical Security Plan (ref. 1).

Basis Reference(s):

1. SSES Physical Security Plan

2. ON-000-010 Security Event

3. NEI 99-01 HA1 Page 101 of 185

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Page 129 of 281 Category: H - Hazards Subcategory: 1 - Security Initiating Condition: Confirmed SECURITY CONDITION or threat EAL:

HUI.1 Unusual Event A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the 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):

SECURITY CONDITION - Any security event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A security condition does not involve a HOSTILE ACTION.

HOSTILE ACTION - An act toward Susquehanna 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 Susquehanna. 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).

Basis:

This EAL is based on the SSES Physical Security Plan (ref. 1).

Security Shift Supervision are the designated on-site personnel qualified and trained to confirm that a security event is occurring or has occurred. Training on security event classification confirmation is closely controlled due to the strict secrecy controls placed on the SSES Physical Security Plan (Safeguards) information (ref. 1).

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, HS1 and HG1.

Page 102 of 185

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Page 130 of 281 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.

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 Susquehanna Physical Security Plan (ref. 1).

The third threshold 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 the Susquehanna Physical Security Plan (ref. 1).

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 Susquehanna Physical Security Plan (ref. 1).

Escalation of the emergency classification level would be via IC HAl.

Basis Reference(s):

1. SSES Physical Security Plan

2. ON-000-010 Security Event

3. NEI 99-01 HU1 Page 103 of 185

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Page 131 of 281 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 greater than OPERATING BASIS EARTHQUAKE (OBE) as indicated by seismic instrumentation in the Control Room recording level greater than an OBE Mode Applicability:

All Definition(s):

None Basis:

Ground motion acceleration of 0.05g is the OPERATING BASIS EARTHQUAKE (OBE) for Susquehanna (ref. 1). Earthquake Monitoring Panel 0C696 provides indication of strong motion, OBE, or SAFE SHUTDOWN EARTHQUAKE (SSE) events for the Unit 1 Containment Foundation, Unit 2 Containment Foundation, and the ESW Pumphouse (as well as tape printout). Input from all six channels are recorded when a trigger initiates the system. A seismic event generally starts with an indication in the Control Room, Annunciator SEISMIC MON SYSTEM TRIGGERED (AR-01 6-001 window G06) on 0C653. The OBE is signaled by an LED illuminated green on the upper panel adjacent to the label, OPERATING BASIS EARTHQUAKE (ref. 2-4).

If above seismic monitoring is inoperable, ensure on-shift Operations and Security crews are aware of the need to report vibratory ground motion (earthquakes) to the control room, and the need for control room staff to contact outside agencies to confirm level of earthquake per OP-099-002 and ON-000-002.

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) can confirm that an earthquake has occurred in the area of the plant. Such confirmation should not, however, preclude a timely emergency declaration. The NEIC can be contacted by calling (303) 273-8500. Select option #1 and inform the analyst you wish to confirm recent seismic activity in the vicinity of Susquehanna. Provide the analyst with the following Susquehanna coordinates: 41O5'2O north latitude, 76o8'56

west longitude (ref. 4). Alternatively, near real-time seismic activity can be accessed via the NElIC website: http://earthquake,usgs.gov/eqcenter/

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 walk-downs and Page 104 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 132 of 281 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 excess of 0.05g). The Shift Manager or Emergency Director/Recovery Manager may seek external verification if deemed appropriate (e.g., a call to the USGS, check internet news sources, etc.); however, the verification action must not preclude a timely emergency declaration.

Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA8.

Basis Reference(s):

1. TRM 3.3.2 Seismic Monitoring Instrumentation

2. ESAR Section 3.7a Seismic Design

3. ON-000-002 Natural Phenomena

4. OP-099-002 Seismic Monitoring System

5. NEI 99-01 HU2 Page 105 of 185

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Page 133 of 281 Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technology Hazard Initiating Condition: Hazardous event EAL:

HU3.1 Unusual Event A tornado strike within the PROTECTED AREA Mode Applicability:

All Definition(s):

PROTECTED AREA - Area within the station inner security fence (PROTECTED AREA Barrier) designated to implement the requirements of 10 CFR 73.

Basis:

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.

This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.

EAL HU3.1 addresses a tornado striking (touching down) within the PROTECTED AREA.

Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, S or C.

Basis Reference(s):

1. NEI 99-01 HU3 Page 106 of 185

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Page 134 of 281 Category: H - Hazards and Other Conditions Affecting Plant Safety 8ubcategory: 3 - Natural or Technology 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 needed 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 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:

If the SAFETY SYSTEM component was operating at the time of isolation, EAL SA8.1 or CA6.1 may be applicable based on degraded SAFETY SYSTEM performance.

This EAL addresses water entering a room or area faster than installed equipment is capable of removing but is not applicable to water spraying on equipment of a magnitude that does not meet the definition of FLOODING.

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 caused by the FLOODING. Classification is also required if the water level or related wetting caused by the FLOODING 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 ICs in Recognition Categories R, F, S or C.

Page 107 of 185

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Page 135 of 281 Basis Reference(s):

1. NPE 91 001 SSES IPE

2. NEI 99-01 HU3 Page 108 of 185

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Page 136 of 281 Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technology Hazard Initiating Condition: Hazardous event EAL:

HU3.3 Unusual Event Movement of personnel within the PROTECTED AREA is IMPEDED due to an offsite event involving hazardous materials (e.g., an offsite chemical spill or toxic gas release)

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 measures such as temporary shielding, SCBAs or beyond Emergency Plan RWP dose extensions that are not routinely employed to access the room/area).

PROTECTED AREA - Area within the station inner security fence (PROTECTED AREA Barrier) designated to implement the requirements of 10 CFR 73.

Basis:

As used here, the term "offsite" is meant to be areas external to the PROTECTED AREA.

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 an offsite location and of sufficient magnitude to impede the movement of personnel within the PROTECTED AREA.

Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, S or C.

Basis Reference(s):

1. NEI 99-01 HU3 Page 109 of 185

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Page 137 of 281 Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technology 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):

None Basis:

Events to consider are river flooding, hurricane, wind storms that block all of the multiple routes to get to the site.

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 around Ft. Calhoun Station in 2011.

Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, S or C.

Basis Reference(s):

1. NEI 99-01 HU3 Page 110 of 185

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Page 138 of 281 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-i area Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Table H-I Fire AreasI

• Control Structure

• Diesel Generator Buildings

• ESSW Pump House

• Reactor Buildings

• Turbine Buildings 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.

Basis: '

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. A single 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. If a fire is verified to be occurring by field report, the 15 minute time limit is from the original receipt of multiple fire detection alarms/indications or field confirmation of the single fire detection alarm.

Page 111 of 185

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Page 139 of 281 A valid fire detection alarm and a valid fire suppression alarm in the same area are considered receipt of multiple fire alarms or indications. Independent fire detection or suppression alarms in Table H-i areas that are in close proximity are treated as multiple indications of a fire.

Table H-i Fire Areas are based on the SSES Fire Protection Review Report (FPRR). Table H-i Fire Areas include those structures containing functions and systems required for safe shutdown of the plant (SAFETY SYSTEMS). (ref. 1)

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.

For EAL HU4.1 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 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 knowledge that a fires exists starts at the time of a report from the field, receipt of multiple fire detection alarms or indications, or field confirmation of a single alarm or indication. The fire duration clock starts at the time that knowledge of a fire exists.

Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA8.

Basis Reference(s):

1. SSES FPRR Section 6.2 Fire Area Description

2. NEI 99-01 HU4 Page 112 of 185

Attachment 1 EP-RM-004 Revision [XJ Page 140 of 281 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-I area AND The existence of a FIRE is not verified (i.e., proved or disproved) within 30 min. of alarm receipt (Note 1)

Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Table H-I Fire AreasI

• Control Structure

• Diesel Generator Buildings

• ESSW Pump House

• Reactor Buildings

• Turbine Buildings 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.

Basis:

The 30 minute requirement begins upon receipt of a single valid fire detection or fire suppression system 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. 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 within 30 minutes, classification shall be made based on EAL HU4.1.

Page 113 of 185

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Page 141 of 281 Table H-i Fire Areas are based on the SSES Fire Protection Review Report (FPRR). Table H-i Fire Areas include those structures containing functions and systems required for safe shutdown of the plant (SAFETY SYSTEMS). (ref. 1)

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.

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 HU4.2, 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 IC CA6 or SA8.

Page 114 of 185

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Page 142 of 281 Basis Reference(s):

1. SSES FPRR Section 6.2 Fire Area Description

2. NE! 99-01 HU4 Page 115 of 185

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Page 143 of 281 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 plant PROTECTED AREA not extinguished within 60 min. of the initial report, alarm or indication (Note 1)

Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be 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 - Area within the station inner security fence (PROTECTED AREA Barrier) designated to implement the requirements of 10 CFR 73.

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 emergency classification level would be via IC CA6 or SA8.

The ISFSI is located within the PROTECTED AREA.

Basis Reference(s):

1. NEI 99-01 HU4 Page 116 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 144 of 281 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 plant 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 - Area within the station inner security fence (PROTECTED AREA Barrier) designated to implement the requirements of 10 CFR 73.

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 IC CA6 or SA8.

The ISFSI is located within the PROTECTED AREA.

Basis Reference(s):

1. NEI 99-01 HU4 Page 117 of 185

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Page 145 of 281 Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 5 - Hazardous Gases Initiating Condition: Gaseous release IMPEDING access to equipment necessary for normal plant operations, cooldown or shutdown EAL:

HA5.1 Alert Release of a toxic, corrosive, asphyxiant or flammable gas into any Table H-2 rooms or areas 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 Areas Elevation Unit 1 Area(s) ** Unit 2 Area(s)** Mode(s) 670' RB 27 RB 32 3,4,5 683' RB 27, 28, 29 RB 32, 33, 34 3,4,5 703' RB 28, 29 RB 33, 34 3,4,5 719' RB 25, 29 RB 30, 34 3,4,5 749' RB 25, 29 RB 32, 33 3,4,5 729' CS 12, 21 CS 12, 21 1, 2,3, 4,5, D

• • See Chart 1 for location of plant areas Chart 1- Plant Area Key Plan RAD WASTE

'*LOW LEVEL WATER TREATMENT r*SPRAY POND 141ACIDOAND IICHLORINE rI S* I - I---I

~~~S VALVEVAULT L+/-JBUILDING 2S2T0E PUMPHOUSE ESSW[1 StT&29 2* PUMPHNO USE RAD WASTE I1 Ii 14 I13 413 2 1 38 37 t TU{K --

I- 20 I19 I18 I17 817 6 5 flU * -I- -I--- 4 -)----- - I -

22 21 12 10 9 42 1411 27 251~

COND.STRG 36 -I5----COND STORAG EF 31 26 4 DIESEL 34 33 29 28

.LJ GENERTO E #2 E #i SRIVER INTAKE REACTOR REACTOR 81]B DIESEL STRUCTURE GENERATOR Page 118 of 185

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Page 146 of 281 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 measures such as temporary shielding, SCBAs or beyond Emergency Plan RWP dose extensions that are not routinely employed to access the room/area).

Basis:

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 areas in Table H-2 specify those areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. 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. See Chart 1 for the specific locations of areas listed in Table R-2. See Attachment 3 for more details of how the Table R-2 was developed (ref. 1).

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 emergency classification is not contingent upon whether entry is actually necessary at the time of the release.

Evaluation of the IC and EAL do not require atmospheric sampling; it only requires the Emergency Director's/Recover Manager'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, Page 119 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 147 of 281 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).

• The action for which room/area entry is required is of an administrative or record keeping nature (e.g., normal rounds or routine inspections).

• The access control measures are of a conservative or precautionary nature, and would not actually prevent or impede a required action.

• 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.

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 IDLH level of around 19.5%, which can lead to breathing difficulties, unconsciousness or even death.

This EAL does not apply to firefighting activities that automatically or manually activate a fire suppression system in an area, or to intentional inerting of containment.

Escalation of the emergency classification level would be via Recognition Category R, C or F l~s.

oasis Reference(s):

1. Attachment 3 Safe Operation & Shutdown Areas Tables R-2 & H-2 Bases

2. NEI 99-01 HA5 Page 120 of 185

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Page 148 of 281 H - Hazards and Other Conditions Affecting Plant Safety Category:

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 reestablished within 15 min. (Note 1):

• Reactivity

• RPV water level

• RCS heat removal Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability:

All Definition(s):

None Basis:

The Shift Manager determines if the Control Room is uninhabitable and requires evacuation.

Control Room inhabitability may be caused by fire, dense smoke, noxious fumes, bomb threat in or adjacent to the Control Room, or other life threatening conditions (Ref. 1, 2).

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) is based on Emergency Director/Recovery Manager judgment. The Emergency Director/Recovery Manager 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).

Escalation of the emergency classification level would be via IC FG1 or CG1 Basis Reference(s):

Page 121 of 185

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Page 149 of 281

1. ON-i100(200)-009 Control Room Evacuation

2. NEI 99-01 HS6 Page 122 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 150 of 281 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:

The Shift Manager (SM) determines if the Control Room is uninhabitable and requires evacuation. Control Room inhabitability may be caused by fire, dense smoke, noxious fumes, bomb threat in or adjacent to the Control Room, or other life threatening conditions (Ref. 1, 2).

Inability to establish plant control from outside the Control Room escalates this event to a Site Area Emergency per EAL HS6.1.

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.

Escalation of the emergency classification level would be via IC HS6.

Basis Reference(s):

1. ON-100(200)-009 Control Room Evacuation

2. NEI 99-01 HA6 Page 123 of 185

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Page 151 of 281 Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - ED/RM Judgment Initiating Condition: Other conditions exist which in the judgment of the Emergency Director/Recovery Manager warrant declaration of a General Emergency EAL:

HG7.1 General Emergency Other conditions exist which in the judgment of the Emergency Director/Recovery Manager 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):

HOSTILE ACTION - An act toward Susquehanna 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 Susquehanna. 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.

As applied to Susquehanna, CONTAINMENT CLOSURE is established per NDAP-QA-0309 (ref. 2) for Primary Containment OR is established per NDAP-QA-0321 (ref. 3) for Secondary Containment.

Basis:

The Emergency Director/Recovery Manager is the designated onsite individual having the responsibility and authority for implementing the Emergency Response Plan. The Shift Manager(SM) initially acts in the capacity of the Emergency Director/Recovery Manager and takes actions as outlined in the Emergency Plan implementing procedures. If required by the emergency classification or if deemed appropriate by the Emergency Director/Recovery Manager, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in Page 124 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 152 of 281 anticipation of the possible wide-ranging responsibilities associated with managing a major emergency (ref. 1).

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/Recovery Manager to fall under the emergency classification level description for a General Emergency.

Basis Reference(s):

1. Susquehanna LLC, Susquehanna Steam Electric Station Emergency Plan, Section 6.0 Organizational Control of Emergencies

2. NDAP-QA-0309 Primary Containment Access and Control

3. NDAP-QA-0321 Secondary Containment Integrity Control

4. NE! 99-01 HG7 Page 125 of 185

Attachment 1 EP-RM-004 Revision IX]

Page 153 of 281 Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7- ED/RM Judgment Initiating Condition: Other conditions existing that in the judgment of the Emergency Director/Recovery Manager warrant declaration of a Site Area Emergency EAL:

HS7.1 Site Area Emergency Other conditions exist which in the judgment of the Emergency Director/Recovery Manager indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts, (1) toward site personnel or equipment that could lead to the likely failure of or, (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the EMERGENCY PLAN BOUNDARY Mode Applicability:

All Definition(s):

HOSTILE ACTION - An act toward Susquehanna 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 Susquehanna. 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).

EMERGENCY PLAN BOUNDARY (EPB) -Same as the Exclusion Area Boundary, i.e., that area around SSES within a radius of 1800 feet determined in accordance with 10 CFR 100.11.

Basis:

The Emergency Director/Recovery Manager is the designated onsite individual having the responsibility and authority for implementing the Emergency Response Plan. The Shift Manager (SM) initially acts in the capacity of the Emergency Director/Recovery Manager and takes actions as outlined in the Emergency Plan implementing procedures. If required by the emergency classification or if deemed appropriate by the Emergency Director/Recovery Manager, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency (ref. 1).

Page 126 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 154 of 281 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/Recovery Manager to fall under the emergency classification level description for a Site Area Emergency.

Basis Reference(s):

1. Susquehanna LLC, Susquehanna Steam Electric Station Emergency Plan, Section 6.0 Organizational Control of Emergencies

2. NEI 99-01 HS7 2.,

Page 127 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 155 of 281 Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - ED/RM Judgment Initiating Condition: Other conditions exist that in the judgment of the Emergency Director/Recovery Manager warrant declaration of an Alert EAL:

HA7.1 Alert Other conditions exist which, in the judgment of the Emergency Director/Recovery Manager, 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):

HOSTILE ACTION - An act toward Susquehanna 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 Susquehanna. 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).

Basis:

The Emergency Director/Recovery Manager is the designated onsite individual having the responsibility and authority for implementing the Emergency Response Plan. The Shift Manager (SM) initially acts in the capacity of the Emergency Director/Recovery Manager and takes actions as outlined in the Emergency Plan implementing procedures. If required by the emergency classification or if deemed appropriate by the Emergency Director/Recovery Manager, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency (ref.1).

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/Recovery Manager to fall under the emergency classification level description for an Alert.

Page 128 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 156 of 281 Basis Reference(s):

I. Susquehanna LLC, Susquehanna Steam Electric Station Emergency Plan, Section 6.0 Organizational Control of Emergencies

2. NEI 99-01 HA7 Page 129 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 157 of 281 Category: H -. Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - ED/RM Judgment Initiating Condition: Other conditions existing that in the judgment of the Emergency Director/Recovery Manager warrant declaration of a UE EAL:

HU7.1 Unusual Event Other conditions exist which in the judgment of the Emergency Director/Recovery Manager 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 100FR50.2):

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:

The Emergency Director/Recovery Manager is the designated onsite individual having the responsibility and authority for implementing the Emergency Response Plan. The Shift Manager (SM) initially acts in the capacity of the Emergency Director/Recovery Manager and takes actions as outlined in the Emergency Plan implementing procedures. If required by the emergency classification or if deemed appropriate by the Emergency Director/Recovery Manager, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency (ref. 1).

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/Recovery Manager to fall under the emergency classification level description for an Unusual Event.

Page 130 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 158 of 281 Basis Reference(s):

1. Susquehanna LLC, Susquehanna Steam Electric Station Emergency Plan, Section 6.0 Organizational Control of Emergencies

2. NEI 99-01 HU7 Page 131 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 159 of 281 Category S - System Malfunction EAL Group: Hot Conditions (RCS temperature > 2000 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 Emergqency AC Power Loss of emergency 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 sources for 4.16 kV ESS buses.

2. Loss of Vital DC Power Loss of emergency 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 125 VDC 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 increase 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 coolant and can be detected by coolant sampling.

5. RCS Leakagqe 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 Primary Containment integrity.

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 Page 132 of 185

Attachment 1 EP-RM-004 Revision XIX Page 160 of 281 as Anticipated Transient Without Scram (ATWS) events. For EAL classification, however, ATWS is intended to mean any trip 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, ROS and Primary 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 Affectin~q Safety Systems Certain hazardous natural and technological events may result in visible damage to or degraded performance of safety systems warranting classification.

Page 133 of 186

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Page 161 of 281 Category: S -System Malfunction Subcategory: 1 - Loss of Essential AC Power Initiating Condition: Prolonged loss of all offsite and all onsite AC power to essential buses EAL:

SGI.1 General Emergency Loss of ALL offsite and ALL onsite AC power capability to ALL 4.16 kV ESS buses on EITHER unit AND EITHER:

Restoration of at least one 4.16 kV ESS bus in < 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not likely (Note 1)

OR RPV water level CANNOT BE RESTORED AND MAINTAINED > -179 in.

Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability:

1 - Power Operations, 2 - Startup, 3 - Hot Shutdown Definition(s):

None Basis:

The Class I1E 4.16 kV system supplies all the Engineered Safety Feature (ESE) loads and other loads that are needed for a safe and orderly plant shutdown, and for keeping the plant in a safe shutdown condition. The eight Class 1E 4.16 kV ESS Buses 1 (2)A through 1 (2)D receive power from either the four ESS 13.8/4.16 kV transformers or the diesel generators (A, B, C, D and additional diesel generator E). Buses 1A-1D supply Unit 1 and common loads and Buses 2A-2D supply Unit 2 loads. This configuration prevents a loss of all ESS Buses for one unit in the event one of the ESS Transformers is lost.

During normal plant operation, ESS Transformer 101 supplies preferred power to ESS Bus IA and 2A and is an alternate power supply to ESS bus I1D and 2D. ESS Transformer 111 supplies preferred power to ESS Bus 1C and 2C, and is an alternate power supply to ESS bus I1B and 2B. ESS Transformer 201 supplies preferred power to ESS Bus 1D and 2D, and is an alternate power supply to ESS bus 1A and 2A. ESS Transformer 211 supplies preferred power to ESS Bus 1 B and 2B, and is an alternate power supply to ESS bus IC and 2C.

On a loss of a preferred power source, the bus rapidly transfers to the alternate power source to maintain component power. If both the preferred and alternate power sources are lost, the associated standby diesel generator connects to the ESS bus. (ref. 2-6)

Four hours is the station blackout coping time (ref. 7).

Page 134 of 185

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Page 162 of 281 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 (-179 in.)

(ref. 8). 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 emergency buses. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A prolonged loss of these buses will lead to a loss of one or more fission product barriers. In addition, fission product barrier monitoring capabilities may be degraded under these conditions.

The EAL should require declaration of a General Emergency prior to meeting the thresholds for IC FGI. This will allow additional time for implementation of offsite protective actions.

Escalation of the emergency classification from Site Area Emergency will occur if it is projected that power cannot be restored to at least one AC emergency 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 increased likelihood of challenges to multiple fission product barriers.

The estimate for restoring at least one 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.

Basis Reference(s):

1. FSAR Section 8.20Offsite Power System

2. ESAR Section 8.30Onsite Power System

3. Technical Specifications 3.8.1 AC Sources - Operating

4. Technical Specifications 3.8.7 Distribution System - Operating

5. ON-i104 (204)-001 Units 1(2) Response to Loss of All Offsite Power

6. EO-100 (200)-030 UNIT 1(2) Response to Station Blackout

7. ESAR Section 15.9 STATION BLACKOUT (SBO)

8. EO-000-102 RPV Control

9. NEI 99-01 SG1 Page 135 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 163 of 281 Category: S -System Malfunction Subcategory: 1 - Loss of Essential AC Power Initiating Condition: Loss of all essential 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 ALL 4.16 kV ESS buses on EITHER unit for _>15 min.

AND Indicated voltage is < 105 VDC on ALL of the following vital 125 VDC main distribution buses on the affected unit for _>15 min. (Note 1):

• 1D612 (2D612)

• 1 D622 (2D622)

• 10632 (2D632)

• 10642 (2D642)

Note 1: The EDIRM should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability:

1 - Power Operations, 2 - Startup, 3 - Hot Shutdown Definition(s):

None Basis:

This EAL addresses operating experience from the March 2011 accident at Fukushima Daiichi.

The Class 1 E 4.16 kV system supplies all the Engineered Safety Feature (ESE) loads and other loads that are needed for a safe and orderly plant shutdown, and for keeping the plant in a safe shutdown condition. The eight Class 1 E 4.16 kV ESS Buses 1 (2)A through 1 (2)0 receive power from either the four ESS 13.8/4.16 kV transformers or the diesel generators (A, B, C, 0 and additional diesel generator E). Buses IA-iD supply Unit 1 and common loads and Buses 2A-2D supply Unit 2 loads. This configuration prevents a loss of all ESS Buses for one unit in the event one of the ESS Transformers is lost.

During normal plant operation, ESS Transformer 101 supplies preferred power to ESS Bus IA and 2A and is an alternate power supply to ESS bus ID and 20. ESS Transformer 111 supplies preferred power to ESS Bus IC and 2C, and is an alternate power supply to ESS bus I1B and 2B. ESS Transformer 201 supplies preferred power to ESS Bus 10 and 20, and is an alternate power supply to ESS bus IA and 2A. ESS Transformer 211 supplies preferred power to ESS Bus 1 B and 2B, and is an alternate power supply to ESS bus 1C and 2C.

Page 136 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 164 of 281 On a loss of a preferred power source, the bus rapidly transfers to the alternate power source to maintain component power. If both the preferred and alternate power sources are lost, the associated standby diesel generator connects to the ESS bus. (ref. 1-6)

The Class I1E Battery Banks are 1(2)D61 0 (Channel A), I1(2)D620 (Channel B), I (2)D630 (Channel C), and I1(2)D640 (Channel 0). Each bank consists of 60 cells connected in series.

Each cell produces a nominal voltage of 2.06 VDC resulting in a total battery bank terminal voltage of 123.6 VDC. All battery banks are designed to supply power to its load center for four hours in the event of a loss of power from its battery charger (ref. 7-9).

105 VDC is the minimum design voltage limit (ref. 10).

Indicated voltage for the vital 125 VDC main distribution buses is local only. Local voltage indication is .available for each bus based on dispatching a field operator in accordance with Control Room alarm response procedure AR-I1(2)06-001 (A12,B12,C12,D12). The 15 minute classification clock begins upon receipt of the 125V DC Panel System Trouble alarm in the Control Room. If battery voltage cannot be verified to be greater than or equal to 105 VDC within the 15 minutes, emergency classification must be made under this EAL.

This IC addresses a concurrent and prolonged loss of both emergency AC and Vital DC power.

A loss of all emergency AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A loss of vital DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both emergency 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.

Basis Reference(s):

1. FSAR Section 8.20Offsite Power System

2. FSAR Section 8.30Onsite Power System

3. Technical Specifications 3.8.1 AC Sources - Operating

4. Technical Specifications 3.8.7 Distribution System - Operating

5. ON-I104 (204)-001 Units 1(2) Response to Loss of All Offsite Power

6. EO-100 (200)-030 UNIT 1(2) Response to Station Blackout

7. FSAR Section 8.3.2 DC Power Systems

8. Susquehanna Drawing No. E107159, Sheet 1, "Single Line Meter & Relay Diagram 125 VDC, 250 VDC & 120 VAC Systems"

9. Technical Specifications 3.8.5 DC Sources - Shutdown

10. ON-102(202)-610, -620, -630, -640 Loss of 125V DC

11. AR-I1(2)06-001 Main Turbine/Generator, Computer HVAC, Instrument AC, 24V DC, 125V DC, 250V DC Panel 2C651

12. NEI 99-01 SG8 Page 137 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 165 of 281 Category: S - System Malfunction Subcategory: 1 - Loss of Essential AC Power Initiating Condition: Loss of all offsite power and all onsite AC power to essential buses for 15 minutes or longer EAL:

SS1.1 Site Area Emergency Loss of ALL offsite and ALL onsite AC power capability to ALL 4.16 kV ESS buses on EITHER unit for _>15 min. (Note 1)

Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability:

1 - Power Operations, 2 - Startup, 3 - Hot Shutdown Definition(s):

None Basis:

The Class 1E 4.16 kV system supplies all the Engineered Safety Feature (ESE) loads and other loads that are needed for a safe and orderly plant shutdown, and for keeping the plant in a safe shutdown condition. See Figure S-1 (ref. 1, 2) The eight Class 1E 4.16 kV ESS Buses 1(2)A through 1 (2)0 receive power from either the four ESS 13.8/4.16 kV transformers or the diesel generators (A, B, C, 0 and additional diesel generator E). Buses 1A-1D supply Unit 1 and common loads and Buses 2A-2D supply Unit 2 loads. This configuration prevents a loss of all ESS Buses for one unit in the event one of the ESS Transformers is lost.

During normal plant operation, ESS Transformer 101 supplies preferred power to ESS Bus 1A and 2A and is an alternate power supply to ESS bus 10 and 20. ESS Transformer 111 supplies preferred power to ESS Bus 1lC and 2C, and is an alternate power supply to ESS bus 1 B and 2B. ESS Transformer 201 supplies preferred power to ESS Bus ID and 20, and is an alternate power supply to ESS bus 1A and 2A. ESS Transformer 211 supplies preferred power to ESS Bus 1B and 2B, and is an alternate power supply to ESS bus 1C and 2C.

On a loss of a preferred power source, the bus rapidly transfers to the alternate power source to maintain component power. If both the preferred and alternate power sources are lost, the associated standby diesel generator connects to the ESS bus. (ref. 2-6)

The 15-minute interval was selected as a threshold to exclude transient or momentary power losses. The interval begins when both offsite and onsite AC power are lost.

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.

Page 138 of 185

Attachment 1 EP-RM-004 Revision IX]

Page 166 of 281 In addition, fission product barrier monitoring capabilities may be degraded under these conditions. This IC represents a condition that involves actual or likely major failures of plant functions needed for the protection of the public.

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

Escalation of the emergency classification level would be via l~s RG1, FG1 or SG1.

Basis Reference(s):

1. ESAR Section 8.20Offsite Power System

2. ESAR Section 8.30Onsite Power System

3. Technical Specifications 3.8.1 AC Sources - Operating

4. Technical Specifications 3.8.7 Distribution System - Operating

5. ON-I104 (204)-001 Units 1(2) Response to Loss of All Offsite Power

6. EO-100 (200)-030 UNIT 1(2) Response to Station Blackout

7. NEI 99-01 SS1 Page 139 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 167 of 281 Figure S-I ESS 13.8/4.16 kV Transformers and Distribution (ref. 1) 5~ kV/230kVTiELINE iRlOI

)NOUR*

230kV s/u TRANS 1-20 T-211 ESS2'I1 0X213 Page 140 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 168 of 281 Category: S - System Malfunction Subcategory: 1 - Loss of Essential AC Power Initiating Condition: Loss of all but one AC power source to essential buses for 15 minutes or longer EAL:

SAI.1 Alert AC power capability to ALL 4.16 kV ESS buses on EITHER unit (Table S-5) 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 ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

[Table S-5 AC 4.16 kV ESS buses Unit 1:

• 1A201

• 1A202

• 1A203

• 1A204 Unit 2:

• 2A201

• 2A202

• 2A203

• 2A204 Mode Applicability:

1 - Power Operations, 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 10OCFR50.2):

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 141 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 169 of 281 (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

Basis:

Basis:

The Class 1 E 4.16 kV system supplies all the Engineered Safety Feature (ESF) loads and other loads that are needed for a safe and orderly plant shutdown, and for keeping the plant in a safe shutdown condition. See Figure S-1 (ref. 1, 2) The eight Class 1E 4.16 kV ESS Buses 1(2)A through 1(2)D receive power from either the four ESS 13.8/4.16 kV transformers or the diesel generators (A, B, C, D and additional diesel generator E). Buses 1A-1D supply Unit 1 and common loads and Buses 2A-2D supply Unit 2 loads. This configuration prevents a loss of all ESS Buses for one unit in the event one of the ESS Transformers is lost.

During normal plant operation, ESS Transformer 101 supplies preferred power to ESS Bus 1A and 2A and is an alternate power supply to ESS bus 1D and 2D. ESS Transformer 111 supplies preferred power to ESS Bus 1C and 2C, and is an alternate power supply to ESS bus 1 B and 2B. ESS Transformer 201 supplies preferred power to ESS Bus 1 D and 2D, and is an alternate power supply to ESS bus 1A and 2A. ESS Transformer 211 supplies preferred power to ESS Bus 1B and 2B, and is an alternate power supply to ESS bus 1 C and 2C.

On a loss of a preferred power source, the bus rapidly transfers to the alternate power source to maintain component power. If both the preferred and alternate power sources are lost, the associated standby diesel generator connects to the ESS bus. (ref. 2-6)

The 15-minute interval was selected as a threshold to exclude transient or momentary power losses. If the capability of a second source of emergency bus power is not restored within 15 minutes, an Alert is declared under this EAL.

This hot condition EAL is equivalent to the cold condition EAL CU2.1.

This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment. This IC provides an escalation path from IC SUI.

An "AC power source" is a source recognized in AOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.

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

• A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from the unit main generator.

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

Basis Reference(s):

Page 142 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 170 of 281

1. FSAR Section 8.20Offsite Power System

2. ESAR Section 8.30Onsite Power System

3. Technical Specifications 3.8.1 AC Sources - Operating

4. Technical Specifications 3.8.7 Distribution System - Operating

5. ON-i104 (204)-001 Units 1(2) Response to Loss of All Ofisite Power

6. EO-100 (200)-030 UNIT 1(2) Response to Station Blackout

7. NEI 99-01 SAt Page 143 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 171 of 281 Figure s-I ESS 13.8/4.16 kV Transformers and Distribution (ref. 1)

SUNBUPY#2 ONTOUR OUNTA*I*

T-211 ESS2I1 0X213 CKTBKR CLOSED L OPEN T CKT BKR

"'°'°"°'"°'"°'"'"'°""°'""'" 0G501E TWO UNIT OPERATIONAL LINEUPSHOWN J°,..°,°°...,°.°,,°=,°=°,,,°,..,°,°.

Page 144 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 172 of 281 Category: S - System Malfunction Subcategory: 1 - Loss of Essential AC Power Initiating Condition: Loss of all offsite AC power capability to essential buses for 15 minutes or longer EAL:

SUI.1 Unusual Event Loss of ALL offsite AC power capability to ALL 4.16 kV ESS buses on EITHER unit (Table S-5) for

_>15 min. (Note 1)

Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Table S-5 AC 4.16 kV ESS buses Unit 1:

• 1A201

• 1A202

• 1A203

• 1A204 Unit 2:

• 2A201

• 2A202

• 2A203

• 2A204 Mode Applicability:

1 - Power Operations, 2 - Startup, 3 - Hot Shutdown Definition(s):

None Basis:

Basis:

The Class 1 E 4.16 kV system supplies all the Engineered Safety Feature (ESE) loads and other loads that are needed for a safe and orderly plant shutdown, and for keeping the plant in a safe shutdown condition. See Figure S-1 (ref. 1, 2) The eight Class 1E 4.16 kV ESS Buses 1(2)A through 1 (2)D receive power from either the four ESS 13.8/4.16 kV transformers or the diesel generators (A, B, C, D and additional diesel generator E). Buses 1A-1 D supply Unit 1 and common loads and Buses 2A-2D supply Unit 2 loads. This configuration prevents a loss of all ESS Buses for one unit in the event one of the ESS Transformers is lost.

Page 145 of 185

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Page 173 of 281 During normal plant operation, ESS Transformer 101 supplies preferred power to ESS Bus 1A and 2A and is an alternate power supply to ESS bus 1 D and 2D. ESS Transformer 111 supplies preferred power to ESS Bus 1C and 20, and is an alternate power supply to ESS bus 1 B and 2B. ESS Transformer 201 supplies preferred power to ESS Bus 1 D and 2D, and is an alternate power supply to ESS bus 1A and 2A. ESS Transformer 211 supplies preferred power to ESS Bus 1 B and 2B, and is an alternate power supply to ESS bus 1 C and 20.

On a loss of a preferred power source, the bus rapidly transfers to the alternate power source to maintain component power. If both the preferred and alternate power sources are lost, the associated standby diesel generator connects to the ESS bus. (ref. 2-6)

The 15-minute interval was selected as a threshold to exclude transient or momentary power losses.

This IC addresses a prolonged loss of offsite power. The loss of offsite power sources renders the plant more vulnerable to a complete loss of power to AC emergency 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 emergency buses, whether or not the buses are powered from it.

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

Escalation of the emergency classification level would be via IC SAl.

Basis Reference(s):

1. FSAR Section 8.20Offsite Power System

2. FSAR Section 8.30Onsite Power System

3. Technical Specifications 3.8.1 AC Sources - Operating

4. Technical Specifications 3.8.7 Distribution System - Operating

5. ON-i104 (204)-001 Units 1(2) Response to Loss of All Offsite Power

6. EO-100 (200)-030 UNIT 1(2) Response to Station Blackout

7. NEI 99-01 SU1 Page 146 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 174 of 281 Figure S-I ESS 13.8/4.16 kV Transformers and Distribution (ref. 1)

,)J TRANS T-20 OX104 1-201 ESS201 0X203 j1"-211 ESS211

' 0X213 Page 147 of 185

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Page 175 of 281 Category: S - System Malfunction Subcategory: 2 - Loss of Vita! 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 ALL of the following vita! 125 VDC main distribution buses on the affected unit for _Ž15 min. (Note 1):

• 1D612 (2D612)

• 10622 (2D622)

• 10632 (2D632)

• 1 D642 (2D642)

Note 1: The EDIRM should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability:

1 - Power Operations, 2 - Startup, 3 - Hot Shutdown Definition(s):

None Basis:

The Class 1E Battery Banks are 1 (2)0610 (Channel A), 1 (2)0620 (Channel B), 1 (2)D630 (Channel C), and 1(2)D640 (Channel 0). Each bank consists of 60 cells connected in series.

Each cell produces a nominal voltage of 2.06 VDC resulting in a total battery bank terminal voltage of 123.6 VDC. All battery banks are designed to supply power to its load center for four hours in the event of a loss of power from its battery charger (ref. 1-3).

105 VDC is the minimum design voltage limit (ref. 4).

Indicated voltage for the vital 125 VDC main distribution buses is local only. Local voltage indication is available for each bus based on dispatching a field operator in accordance with Control Room alarm response procedure AR-1 (2)06-001 (A12, B12,C12,0D12). The 15 minute classification clock begins upon receipt of the 125V DC Panel System Trouble alarm in the Control Room. If battery voltage cannot be verified to be greater than or equal to 105 VDC within the 15 minutes, emergency classification must be made under this EAL.

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 RG1, FG1 or SGI.

Basis Reference(s):

Page 148 of 185

Attachment 1 EP-RM-004 Revision XIX Page 176 of 281

1. FSAR Section 8.3.2 DC Power Systems

2. Susquehanna Drawing No. E107159, Sheet 1, "Single Line Meter & Relay Diagram 125 VDC, 250 VDC & 120 VAC Systems"

3. Technical Specifications 3.8.5 DC Sources - Shutdown

4. ON-102(202)-610, -620, -630, -640 Loss of 125V DC

5. AR-i1(2)06-00 1 Main Turbine/Generator, Computer HVAC, Instrument AC, 24V DC, 125V DC, 250V DC Panel 2C651

6. NEI 99-01 SS8 Page 149 of 185

Attachment 1 EP-RM-004 Revision IX]

Page 177 of 281 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-1 parameters from within the Control Room for >_15 min. (Note 1)

AND Any significant transient is in progress, Table S-2 Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Table S-I Safety System Parameters

• Reactor power

• RPV water level

• RPV pressure

• Primary Containment pressure

• Suppression Pool water level

• Suppression Pool temperature Table S-2 Significant Transients

• Reactor scram

• Runback > 25% reactor power

• RRC pump trip while > 25% reactor power

• ECCS injection

• Thermal power oscillations > 10%

Mode Applicability:

1 - Power Operations, 2 - Startup, 3 - Hot Shutdown 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.

Page 150 of 185

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Page 178 of 281 Basis:

SAFETY SYSTEM parameters listed in Table S-I are monitored in the Control Room through a combination of hard control panel indicators as well as computer based information systems.

The Plant Process Computer (PPC) and SPDS are redundant compensatory indication which may be utilized in lieu of normal Control Room indicators (ref. 1, 2).

Significant transients are listed in Table S-2 and include response to automatic or manually initiated functions such as scrams, runbacks involving greater than 25% thermal power change, RRC pump trip while > 25% reactor power, ECCS injections, or thermal power oscillations of 10% or greater.

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-1 022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.

This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, RPV level and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for RPV water level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well.

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

Escalation of the emergency classification level would be via ICs FS1 or IC RS1 Basis Reference(s):

1. FSAR Section 18.1.17 Plant Safety Parameter Display System

2. OP-1 31(231 )-002 Plant Computer Systems

3. EO-000-102 RPV Control Page 151 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 179 of 281

4. EO-000-103 Primary Containment Control

5. NEI 99-01 SA2 Page 152 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 180 of 281 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-I parameters from within the Control Room for >15 min. (Note 1)

Note 1: The ED/RM should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Table S-I Safety System Parameters

• Reactor power

• RPV water level

• RPV pressure

• Primary Containment pressure

• Suppression Pool water level

• Suppression Pool temperature Mode Applicability:

1 - Power Operations, 2 - Startup, 3 - Hot Shutdown 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:

SAFETY SYSTEM parameters listed in Table S-1 are monitored in the Control Room through a combination of hard control panel indicators as well as computer based information systems.

The Plant Process Computer (PPC) and SPDS are redundant compensatory indication which may be utilized in lieu of normal Control Room indicators (ref. 1, 2).

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 Page 153 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 181 of 281 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-1 022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.

This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, RPV level and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for RPV water level cannot be determined from the indications and recorders on a main control board, the SPOS or the plant computer, the availability of other parameter values may be compromised as well.

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

Escalation of the emergency classification level would be via IC SA3.

Basis Reference(s):

1. FSAR Section 18.1.17 Plant Safety Parameter Display System

2. OP-131(231)-002 Plant Computer Systems

3. EO-000-102 RPV Control

4. EO-000-103 Primary Containment Control

5. NEI 99-01 SU2 Page 154 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 182 of 281 Category: S -System Malfunction Subcategory: 4 - RCS Activity Initiating Condition: Reactor coolant activity greater than Technical Specification allowable limits EAL:

SU4.1 Unusual Event Offgas pretreatment monitor high-high radiation alarm Mode Applicability:

1 - Power Operations, 2 - Startup, 3 - Hot Shutdown Definition(s):

None Basis:

The Offgas Pretreatment RMS monitors radioactivity in the Offgas system downstream of the Motive Steam Jet 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 (ref. 1). Log rad monitors and trip auxiliary units are located on Panel 10604 in the Upper Relay Room. Instrument Channel 'A' is RITS-D12-1K601A and Instrument Channel 'B' is RITS-D12-1 K601 B. Both channels output to Yokagowa Recorder RR-D12-1 R601 on Main Control Room Panel 1C600 (ref. 2, 3).

OFEGAS HI-HI RADIATION (AR-I106-F03) is located on Panel I1C651. The setpoint is variable based on surveillance procedure (ref. 4).

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 R ICs.

Basis Reference(s):

1. ESAR Section 11.5 Process and Effluent Radiological Monitoring and Sampling Systems

2. Technical Specification 3.7.5 Main Condenser Offgas

3. AR-106(206)-001 F03 Offgas Hi Hi Radiation

4. SC-143(243)-101 Unit I (Unit 2) Main Condenser Air Ejector Monthly Noble Gas Activity

5. OP-179(279)-002 Process Radiation Monitoring System

6. NEI 99-01 SU3 Page 155 of 185

Attachment 1 EP-RM-004 Revision IX]

Page 183 of 281 Category: S - System Malfunction Subcategory: 4 - RCS Activity Initiating Condition: Reactor coolant activity greater than Technical Specification allowable limits EAL:

SU4.2 Unusual Event Coolant activity > 0.2 IpCi/gm dose equivalent 1-131 for > 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> OR Coolant activity > 4.0 IpCi/gm dose equivalent 1-131 at any time Mode Applicability:

1 - Power Operations, 2 - Startup, 3 - Hot Shutdown Definition(s):

None Basis:

The specific iodine activity is limited to _<0.2 IpCi/gm dose equivalent 1-131 (Condition A) with a completion time of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. This limit ensures the source term assumed in the safety analysis for the Main Steam Line Break (MSLB) is not exceeded, so any release of radioactivity to the environment during an MSLB is less than a small fraction of the regulatory limits (ref. 1).

The upper limit of 4.0 pCi/gm dose equivalent 1-131 (Condition B) ensures that the TEDE dose from an MSLB will not exceed the dose guidelines of 10 CFR 50.67 or Control Room operator dose limits specified in GDC 19 of 10 CFR 50, Appendix A (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 ICs FA1 or the Recognition Category R ICs.

Basis Reference(s):

1. Technical Specifications section 3.4.7 RCS Specific Activity

2. NEI 99-01 SU3 Page 156 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 184 of 281 Category: S - System Malfunction Subcategory: 5 - RCS Leakage Initiating Condition: RCS leakage for 15 minutes or longer EAL:

8U5.1 Unusual Event RCS unidentified or pressure boundary leakage > 10 gpm for _>15 min.

OR RCS identified leakage > 25 gpm for _Ž15 mai.

OR Leakage from the RCS to a location outside Primary Containment > 25 gpm for _>15 min.

(Note 1)

Note 1: The EDIRM should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability:

1 - Power Operations, 2 - Startup, 3 - Hot Shutdown Definition(s):

None Basis:

Leakage is monitored by utilizing the following techniques (ref. 1):

• Monitoring changes in water level in the drywell floor drain sumps and drywell equipment drain tank

• Sensing excess flow in piping systems

• Sensing pressure and temperature changes in the primary containment

• Monitoring for high flow and temperature through selected drains,

• Sampling airborne particulate and gaseous radioactivity.

Identified leakage is leakage into the drywell, such as that from pump seals or valve packing, that is captured and conducted to the drywell equipment drain* tank; 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 Unidentified leakage is all leakage into the drywell that is not identified leakage (ref. 2).

Pressure boundary leakage is leakage through a nonisolable fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall (ref. 2).

Two drywell floor drain sumps are located in the primary containment for collection of leakage from vent coolers, control rod drive flange leakage, chilled water drains, cooling water drains, Page 157 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 185 of 281 and overflow from the drywell equipment drain tank. The drywell floor drain sumps are located at the drywell diaphragm slab low point. Unidentified leakages will, by gravity, flow down the slab surface into the floor drain sumps. Water flow rate greater than 0.5 gpm can be detected by monitoring changes of level over a time period. The sump depth of 0-5 in. is displayed on a 0-100 percent recorder chart, which relates to the sump nominal capacity of 0-150 gal.

The drywell equipment drain tank collects identified leakage within the primary containment from reactor head seal leak off, bulkhead drain, refueling bellows drain, RPV head vent, recirculation pump seals, reactor recirculation pump cooler drains, and RPV bottom drain (Unit I only). The measured tank depth of 36 in. is displayed on a 0-100 percent recorder chart. This relates directly to the measured tank capacity of 842 gal.

RCS leakage outside of the Primary Containment that is not considered identified or unidentified leakage per Technical Specifications includes leakage via interfacing systems such as RCS to the Reactor Building Closed Cooling Water (RBCCW system), or systems that directly see RCS pressure outside primary containment such as Reactor Water Cleanup (RWCU), reactor water sampling system and Residual Heat Removal (RHR) system (when in the shutdown cooling mode) (ref. 1, 3).

Indicated changes in drywell sump water level are used to calculate unidentified drywell leakage. Indicated changes in drywell equipment drain tank level are used to calculate identified drywell leakage. SO-100-006 and S0-200-006 are the procedures that specify how to complete these calculations.

Drywell leakage calculations in SO-I100(200)-006 take a finite period of time to complete.

Leakage rates cannot be determined quickly by merely observing an indicator. For this reason, the 15 minutes clock starts after it is determined that leakage rates exceed the entry value.

Upon determination that leakage has increased substantially, effort should be made to quantify this leakage in a timely manner.

ON-I1(2)00-005, "Excessive Drywell Leakage Identification", contains methods of quickly estimating drywell leakage. These methods can be used in lieu of completing the calculations contained in SO-i1(2)00-006.

Means to directly quantify RCS leakage outside containment may not be available. For this reason, judgment must be used for assessment of the 25 gpm leak rate criterion. For example, a short steam plume that does not appreciably change room temperature or room radiation levels can be judged to be less than 25 gpm. A leak that causes room temperature to rise rapidly above maximum safe temperatures could be judged to be greater than 25 gpm in the absence of measurable lea k rates, and thus judgment is an acceptable method to evaluate this criterion.

Escalation of this EAL to the Alert level is via Category F, Fission Product Barrier Degradation, EAL FA1 .1. The note has been added to remind the EAL-user to review Table F-i for possible escalation to higher emergency classifications.

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.

Page 158 of 185

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Page 186 of 281 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 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 Primary Containment, or a location outside of Primary 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.

A stuck-open Safety Relief Valve (SRV) or SRV leakage is not considered either identified or unidentified leakage by Technical Specifications and, therefore, is not applicable to this EAL.

The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible.

Escalation of the emergency classification level would be via ICs of Recognition Category R or F.

Basis Reference(s):

1. FSAR Section 5.2.5 Detection of Leakage Through Reactor Coolant Pressure Boundary

2. Technical Specifications Definitions Section 1.1

3. ON-100(200)-005 Excess Drywell Leakage Identification

4. SO-100(200)-006

5. NEI 99-01 SU4 Page 159 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 187 of 281 Category: S - System Malfunction Su bcategory: 2 - RPS Failure Initiating Condition: Inability to shut down the reactor causing a challenge to RPV water level or RCS heat removal EAL:

SS6.1 Site Area Emergency An automatic or manual scram fails to shut down the reactor AND ALL actions to shut down the reactor are not successful as indicated by reactor power _>5%

AND EITHER:

• RPV level CANNOT BE RESTORED AND MAINTAINED > -179 in. or CANNOT be determined OR

• Suppression pool water temperature AND RPV pressure CANNOT BE MAINTAINED below the Heat Capacity Temperature Limit (Figure - HCTL)

Figure - HCTL Heat Capacity Temperature Limit 260 250 240 230 220

• " 210

,,,200

• 170

a. 1260 110 80 70 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34 35 36 37 38 SUPPRESSION POOL LEVEL (F'r)

Page 160 of 185

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Page 188 of 281 Mode Applicability:

1 - Power Operations, 2 - Startup Definition(s):

None Basis:

This EAL addresses the following:

• Any automatic reactor scram signal followed by a manual scram actions that fail 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 SA6.1I), and

• Indications that either core cooling is extremely challenged or heat removal is extremely challenged.

Reactor shutdown achieved by use of EO-OOO-1 13, Control Rod Insertion, is also credited as a successful scram provided reactor power can be reduced below the APRM downscale trip setpoint before indications of an extreme challenge to either core cooling or heat removal exist.

(ref. 1)

The APRM downscale trip setpoint (5%) is a minimum reading on the power range scale that indicates power production. It also approximates the decay heat which the shutdown systems were designed to remove and is indicative of a condition requiring immediate response to prevent subsequent core damage. Below the APRM downscale trip setpoint, plant response will be similar to that observed during a normal shutdown. Nuclear instrumentation (APRM) indications or other reactor parameters (steam flow, RPV pressure, torus temperature trend) can be used to determine if reactor power is greater than 2% power (ref. 2, 3).

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 ROS 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.

3). 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.

When RPV level cannot be determined, EOPs require entry to E=O-OOO-1 14, RPV Flooding. RPV water level indication provides the primary means of knowing if adequate core cooling is being maintained. 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 EO-OOO-1 14 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 (ref.

4).

Page 161 of 185

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Page 189 of 281 The Heat Capacity Temperature Limit (HCTL) is the highest suppression pool water temperature from which Emergency RPV Depressurization will not raise suppression chamber pressure above the 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 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 SP/T in EO-000-1 03, Primary Containment Control, is reached (ref. 5). This condition addresses loss of functions required for hot shutdown with the reactor at pressure and temperature.

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, all subsequent operator actions to manually shutdown the reactor are unsuccessful, and continued power generation is challenging the capability to adequately remove heat from the core and/or the RCS. This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration of a Site Area Emergency.

In some instances, the emergency classification resulting from this IC/EAL may be higher than that resulting from an assessment of the plant 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 IC and EAL ensures the timely declaration of a Site Area Emergency in response to prolonged failure to shut down the reactor.

The adequacy of reactor shutdown (< 5%) is determined in accordance with applicable Emergency Operating Procedure criteria.

Escalation of the emergency classification level would be via IC RG1 or FGI.

Page 162 of 185

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Page 190 of 281 Basis Reference(s):

1. EO-000-1 13 Control Rod Insertion

2. Technical Specifications Table 3.3.1.1-1

3. EO-000-102 RPV Control

4. EO-000-1 14 RPV Flooding

5. EO-000-103 Primary Containment Control

6. NEI 99-01 SS5 Page 163 of 185

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Page 191 of 281 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 AND Manual scram actions taken at the reactor control console (Manual PBs, Mode Switch, 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 Operations, 2 - Startup Definition(s):

None Basis:

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.

For the purposes of emergency classification, successful manual scram actions are those which can be quickly performed from the reactor control console (i.e., manual scram pushbuttons, mode switch, or ARI initiation in accordance with EO-000-1 02 or EO-000-1 13). Reactor shutdown achieved by use of other control rod insertion methods (e.g. individual control rod insertion) directed by EO-000-1 13 does not constitute a successful manual scram (ref. 2, 3).

For the purposes of this EAL, a successful automatic initiation of ARI that reduces reactor power below 5% is not considered a successful automatic scram. Ifautomatic 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.

The APRM downscale trip setpoint (5%) is a minimum reading on the power range scale that indicates power production. It also approximates the decay heat which the shutdown systems were designed to remove and is indicative of a condition requiring imnmediate response to Page 164 of 185

Attachment 1 EP-RM-004 Revision IX]

Page 192 of 281 prevent subsequent core damage. Below the APRM downscale trip setpoint, plant response will be similar to that observed during a normal shutdown. Nuclear instrumentation (APRM) indications or other reactor parameters (steam flow, RPV pressure, torus temperature trend) can be used to determine if reactor power is greater than 5% power (ref. 1, 3).

Escalation of this event to a General Emergency would be under EAL SG2.1 or Emergency Director/Recovery Manager judgment.

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 reactor is subsequently shutdown by an action taken away from the reactor control donsoles since this event entails a significant failure of the RPS.

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 console (e.g., locally opening breakers). Actions taken at backpanels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control console".

Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action.

The plant response to the failure of an automatic or manual reactor scram will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If the failure to shut down 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 FSI. Absent the plant conditions needed to meet either IC SS6 or FS1, an Alert declaration is appropriate for this event.

It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F ICs; however, this IC and EAL are included to ensure a timely emergency declaration.

The adequacy of reactor shutdown (< 5%) is determined in accordance with applicable Emergency Operating Procedure criteria.

Basis Reference(s):

1. Technical Specifications Table 3.3.1.1-1

2. EO-000-1 13 Control Rod Insertion

3. EO-000-102 RPV Control

4. NEI 99-01 SA5 Page 165 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 193 of 281 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 after any RPS setpoint is exceeded AND A subsequent automatic scram or manual scram action taken at the reactor control console (Manual PBs, Mode Switch, 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 Operations, 2 - Startup Definition(s):

None Basis:

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 Reactor Protection System (RPS) when certain continuously monitored parameters exceed predetermined setpoints (ref. 1).

Following a successful reactor scram, rapid insertion of the control rods occurs. Nuclear power promptly drops to a fraction of the original power level and then decays to a level several decades less with a negative period. The reactor power drop continues until reactor power reaches the point at which the influence of source neutrons on reactor power starts to be observable. A predictable post-scram response from an automatic reactor scram signal should therefore consist of a prompt drop in reactor power as sensed by the nuclear instrumentation and a lowering of power into the source range. A successful scram has therefore occurred when there is sufficient rod insertion from the trip of RPS to bring the reactor power 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., manual scram pushbuttons, mode switch, or ARI initiation in accordance with EO-000-1 02 or EO-000-1 13). Reactor shutdown achieved by use of other control rod insertion methods (e.g. individual control rod insertion) directed by EO-000-1 13 does not constitute a successful manual scram (ref. 2, 3).

Following any automatic RPS scram signal, operating procedures (e.g., EO-000-102) 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 Page 166 of 185

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Page 194 of 281 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 below 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 reactor 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 10OCFR 50.72 should be considered for the transient event.

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic scram is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.

Following the failure on an automatic reactor scram, operators will promptly initiate manual actions at the reactor control 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 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.

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Page 195 of 281 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 IC SA6. Depending upon the plant response, escalation is also possible via IC FAI. Absent the plant conditions needed to meet either IC SA6 or FA1, an Unusual Event declaration is appropriate for this event.

The adequacy of reactor shutdown (< 5%) is determined in accordance with applicable Emergency Operating Procedure criteria.

Should a reactor scram signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied.

• If the signal causes a plant transient that should have included an automatic reactor scram and the RPS fails to automatically shutdown the reactor, then this IC and the

.EALs are applicable, and should be evaluated.

• If the signal does not cause a plant transient and the scram failure is determined through other means (e.g., assessment of test results), then this IC and the EALs are not applicable and no classification is warranted.

Basis Reference(s):

1. Technical Specifications Table 3.3.1.1-1

2. EO-000-1 13 Control Rod Insertion

3. EO-000-102 RPV Control

4. NEI 99-01 SU5 Page 168 of 185

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Page 196 of 281 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 after any manual scram action was initiated AND A subsequent automatic scram or manual scram action taken at the reactor control console (Manual PBs, Mode Switch, 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 Operations, 2 - Startup Definition(s):

None Basis:

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 < 5%) (ref. 1).

Following a successful reactor scram, rapid insertion of the control rods occurs. Nuclear power promptly drops to a fraction of the original power level and then decays to a level several decades less with a negative period. The reactor power drop continues until reactor power reaches the point at which the influence of source neutrons on reactor power starts to be observable. A predictable post-scram response from a manual reactor scram signal should therefore consist of a prompt drop in reactor power as sensed by the nuclear instrumentation and a lowering of power into the source range. A successful scram has therefore occurred when there is sufficient rod insertion from the trip of RPS to bring the reactor power 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., manual scram pushbuttons, mode switch, or ARI initiation in accordance with EO-000-1 02 or EO-000-1 13). Reactor shutdown achieved by use of other control rod insertion methods (e.g. individual control rod insertion) directed by EO-000-1 13 does not constitute a successful manual scram (ref. 2, 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.

Page 169 of 185

Attachment 1 EP-RM-004 Revision XIX Page 197 of 281 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.

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

(< 5%) following a failure of an initial manual scram, the event escalates to an Alert under EAL SA6.1 This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic scram is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.

Following the failure on an automatic reactor scram, operators will promptly initiate manual actions at the reactor control 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 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 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 FA1. Absent the plant conditions needed to meet either IC SA6 or FA1, an Unusual Event declaration is appropriate for this event.

The adequacy of reactor shutdown (< 5%) is determined in accordance with applicable Emergency Operating Procedure criteria.

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.

Page 170 of 185

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Page 198 of 281

• If the signal causes a piant transient that should have included an automatic reactor scram and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and should be evaluated.

• If the signal does not cause a plant transient and the scram failure is determined through other means (e.g., assessment of test results), then this IC and the EALs are not applicable and no classification is warranted.

Basis Reference(s):

1. Technical Specifications Table 3.3.1.1-1

2. EO-OOO-1 13 Control Rod Insertion

3. EO-OOO-102 RPV Control

4. NEI 99-01 SU5 Page 171 of 185

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Page 199 of 281 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-3 onsite communication methods OR Loss of ALL Table S-30ORO communication methods OR Loss of ALL Table S-3 NRC communication methods Table S-3 Communication Methods System Onsite ORO NRC UHF Radio X Plant PA System X Dedicated Conference Lines X Commercial Telephone Systems X X X Cellular Telephone x X FTS-2001 (ENS) X X Mode Applicability:

1 - Power Operations, 2 - Startup, 3 - Hot Shutdown Definition(s):

None Basis:

Onsite/offsite communications include one or more of the systems listed in Table S-3 (ref. 1, 2, 3).

UHF Radio Onsite portable radio communication systems are described in the Susquehanna SES Physical Security Plan and in the Susquehanna SES Emergency Plan. Four UHF channels, each Page 172 of 185

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Page 200 of 281 consisting of two frequencies for duplex operation through one of five in-plant repeaters, provide onsite portable radio communications. Operations is assigned two channels; one channel is assigned to Unit 1 and one to Unit 2. Operators in the plant on rounds and on specific assignments are equipped with handheld two-way radios.

Plant PA System The plant PA system is an intra-plant public address providing the following functions:

• A 5-channel page-talk handset intercom system for on-site communications between plant locations.

• Broadcast accountability and fire alarms designed to warn personnel of emergency conditions.

The system consists of telephone handsets, amplifiers and loudspeakers located at various selected areas throughout the plant.

Dedicated Conference Lines (Centrex Three (3) dicqit dialin~q)

The Dedicated Conference Lines are those normally used to communicate with several offsite agencies at one time (e.g., 191 conference line).

Commercial Telephone Systems Two independent telecommunications networks exist to provide primary and backup telephone communications between ERFs and offsite agencies.

Cellular Telephone Cell phones can be utilized to perform both ORO and NRC communications.

ETS 2001 (ENS)

This system is for NRC offsite communications but may also be used to perform ORO notifications.

This EAL is the hot condition equivalent of the cold condition EAL CU5.1.

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 OROs and the NRC.

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

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 OROs of an emergency declaration. The OROs referred to here are the Commonwealth of Pennsylvania, Luzerne and Columbia County EOCs The third EAL addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.

Page 173 of 185

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Page 201 of 281 Basis Reference(s):

1. EP-RM-007 Emergency Teiephone Instructions and Directory

2. SSES Emergency Plan Section 8

3. ESAR Section 9.5.2

4. NE! 99-01 SU6 Page 174 of 185

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Page 202 of 281 Category: S - System Malfunction Subcategory: 8 - Hazardous Event Affecting Safety Systems Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode EAL:

8A8.1 Alert The occurrence of any Table S-4 hazardous event AND EITHER:

• Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode

• The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode Table S-4 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 Mode Applicability:

1 - Power Operations, 2 - Startup, 3 - Hot Shutdown 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 post-event inspection to determine if the attributes of an explosion are present.

Page 175 of 185

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Page 203 of 281 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 component or structure that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure.

Basis:

• The significance of a seismic event is discussed under EAL HU2.1 (ref. 1, 2).

• Internal FLOODING may be caused by events such as component failures, equipment misalignment, or outage activity mishaps (ref. 3, 4, 5).

• Seismic Category I structures are analyzed to withstand a sustained, design wind velocity of at least 80 mph. (ref. 6, 7).

• Areas containing functions and systems required for safe shutdown of the plant are identified by Fire Zone in the fire response procedure (ref. 8, 9).

• An EXPLOSION that degrades the performance of a SAFETY SYSTEM train or visibly damages a SAFETY SYSTEM component or structure would be classified under this EAL.

This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure containing SAFETY SYSTEM components, needed for the current operating mode.

This condition significantly reduces the margin to a loss or potential loss of a fission product barrier, and therefore represents an actual or potential substantial degradation of the level of safety of the plant.

The first condition 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.

The second condition addresses damage to a SAFETY SYSTEM component that is not in service/operation or readily apparent through indications alone, or to a structure containing SAFETY SYSTEM components. Operators will make this determination based on the totality of Page 176 of 185

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Page 204 of 281 available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage.

Escalation of the emergency classification level would be via IC ES1 or RS1.

Basis Reference(s):

1. ON-000-002 Severe Weather / Natural Phenomena

2. ESAR Section 3.7 Seismic Design

3. ON-169(269)-001 Flooding in Turbine Building

4. ON-169(269)-002 Flooding in Reactor Building

5. ESAR Section 3.4 Water Level (Flood) Design

6. FSAR Section 3.3 Wind and Tornado Loadings

7. FSAR Section 3.5 Missile Projection

8. SSES-FPRR Section 6.2 Fire Area Description

9. ON-013-001 Response to Fire

10. NEI 99-01 SA9 Page 177 of 185

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Page 205 of 281 Category E - Independent Spent Fuel Storagqe Installation (ISFS1)

EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold)

An independent spent fuel storage installation (ISESI) 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 cask/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.

A security event involving the ISESI would be classified in accordance with Category HI security based EALs.

Minor surface damage that does not affect storage cask/canister boundary is excluded from the scope of these EALs.

Page 178 of 185

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Page 206 of 281 Category: E - ISFSI Sub-category: None Initiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARY EAL:

EUI.1 Notification of Unusual Event Damage to a loaded canister CONFINEMENT BOUNDARY as indicated by a radiation reading on a loaded spent fuel cask > any of the following:

• 800 mrem/hr at 3 ft from the HSM surface

• 200 mrem/hr on contact on the outside of the HSM door centerline

• 40 mrem/hr on contact on the end shield wall exterior 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 Susquehanna ISESI, Confinement Boundary is defined as the Dry Shielded Canister (DSC).

Basis:

The SSES Independent Spent Fuel Storage Installation utilizes the standardized NUHMOS Horizontal Modular System. The standardized NUHMOS system is a horizontal canister system composed of a steel dry shielded canister (DSC) and a reinforced concrete horizontal storage module (HSM). The DSC provides confinement and criticality control for the storage and transfer of irradiated fuel. The HSM houses the DSC and provides for heat removal. An HSM is considered loaded when it houses a DSC containing spent fuel. (ref. 1, 2)

The values shown represent 2 times the limits specified in the ISFSI Certificate of Compliance Technical Specification 1.2.7, HSM Dose Rates (ref. 1).

This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of a storage cask containing spent fuel. It applies to irradiated fuel that is licensed for dry storage beginning at the point that the loaded storage cask is sealed. The issues of concern are the creation of a potential or actual release path to the environment, degradation of one or more fuel assemblies due to environmental factors, and configuration changes which could cause challenges in removing the cask or fuel from storage.

The existence of "damage" is determined by radiological survey. The technical specification multiple of "2 times" is used here to distinguish between non-emergency and emergency conditions. The emphasis for this classification is the degradation in the level of safety of the spent fuel cask and not the magnitude of the associated dose or dose rate. It is recognized that in the case of extreme damage to a loaded cask, the fact that the "on-contact" dose rate limit is exceeded may be determined based on measurement of a dose rate at some distance from the cask.

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Page 207 of 281 Security-related events for ISFSIs are covered under ICs HU1 and HAl.

Basis Reference(s):

1. Certification of Compliance No. 1004 for the NUHOMS Storage System

2. TRM B3.10.3 Independent Spent Fuel Storage Installation (ISESI)

3. NEI 99-01 E-HU1 Page 180 of 185

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Page 208 of 281 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 (FC): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets.

B. Reactor Coolant System (RCS): The RCS Barrier is the reactor coolant system pressure boundary and includes the RPV and all reactor coolant system piping up to and including the isolation valves.

C. Containment (PC): The drywell, the wetwell, their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves comprise the PC barrier. Primary Containment Barrier thresholds are used as criteria for escalation of the ECL from Alert to either 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-i (Attachment 2). "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 Site Area Emergency:

Loss or potential loss of any two barriers General Emergency:

Loss of any two barriersand 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 Primary 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 Page 181 of 185

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Page 209 of 281 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 RG1 has been exceeded.

• The fission product barrier thresholds specified within a scheme reflect SSES 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 interfacing system, or outside of the primary containment. The release of liquid or steam mass from the RCS due to the as-designed/expected operation of a relief valve is not considered 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 ROS fission product barriers were both lost, 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/Recovery Manager would have more assurance that there was no immediate need to escalate to a General Emergency.

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Page 210 of 281 Category: Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss of any two barriers and loss or potential loss of the third barrier EAL:

FGI.1 General Emergency Loss of any two barriers AND Loss or potential loss of the third barrier (Table F-I)

Mode Applicability:

1 - Power Operations, 2 - Startup, 3 - Hot Shutdown Definition(s):

None Basis:

Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-I (Attachment

2) 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, ROS 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 Basis Reference(s):

1. NEI 99-01 FG1 Page 183 of 185

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Page 211 of 281 Category: Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss or potential loss of any two barriers EAL:

FSI.1 Site Area Emergency Loss or potential loss of any two barriers (Table F-I)

Mode Applicability:

I - Power Operations, 2 - Startup, 3 - Hot Shutdown Definition(s):

None Basis:

Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-i (Attachment

2) 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/Recovery Manager would have greater assurance that escalation to a General Emergency is less imminent.

Basis Reference(s):

1. NEI 99-01 FS1 Page 184 of 185

Attachment 1 EP-RM-004 Revision [X]

Page 212 of 281 Category: Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Any loss or any potential loss of either Fuel Ciad or RCS EAL:

FAI.1 Alert Any loss or any potential loss of EITHER Fuel Clad or RCS (Table F-I )

Mode Applicability:

I - Power Operations, 2 - Startup, 3 - Hot Shutdown Definition(s):

None Basis:

Fuel Clad, RCS and Primary Containment comprise the fission product barriers. Table F-I (Attachment 2) 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 Basis Reference(s):

I. NEI 99-01 FA1 Page 185 of 185

Attachment 2 EP-RM-004 Revision [X]

Page 213 of 281 ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Introduction Table F-i 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 Loss column) lists the categories (types) of fission product barrier thresholds. The fission product barrier categories are:

A. RPV Level B. RCS Leak Rate C. Primary Containment Conditions

0. Primary Containment Radiation I RCS Activity E. Primary Containment Integrity or Bypass F. ED/RM Jugement Each category occupies a row in Table F-i 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 Loss and Potential Loss column beginning with number one. In this manner, a threshold can be identified by its category title and number. For example, the first Fuel Clad barrier Loss in Category A would be assigned "FC Loss A.I,"~the third Containment barrier Potential Loss would be assigned "PC P-Loss C.3," etc.

If a cell in Table F-I 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-I 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-I, 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 potentially lost. For example, if containment radiation is sufficiently high, a Loss of the Fuel Clad Page 1 of 60

Attachment 2 EP-RM-004 Revision [X]

Page 214 of 281 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 FGI1.1, ESI1.1, 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 category Loss followed by category Potential Loss beginning with Category A, then B,...,. F.

Table F-2 provides a human factors enhancement mechanism to track the threshold conditions that define the Loss and Potential Loss of the three fission product barriers (Fuel Clad, Reactor Coolant System, and Containment) to assist with quickly determining which initiating condition for EALs FGI.1, FS1.1, or FAI.1 is met.

Page 2 of 60

Attachment 2 EP-RM-004 Revision [X]

Page 215 of 281 Table F-I Fission Product Barrier Threshold Matrix Fuel Clad Barrier Reactor Coolant System Barrier Primary Containment Barrier Category Loss Potential Loss Loss Potential Loss Loss Potential Loss

1. RPV level CANNOT BE 1. RPV level CANNOT BE A 1. SAG entry required RESTORED AND MAINTAINED > RESTORED AND MAINTAINED> NoeNn .SGetyrqie

-161 in. -161 in.NoeNn .SGntyrqid RPV Level or CANNOT be determined or CANNOT be determined

1. UNISOLABLE break in any of the 1. UNISOLABLE primary syntem 1. UNISOLABLE primary syntem leakage that reaulte in exceeding leakage that resulte in exceeding following: EITHER of the following: EITHER of the following:

• Main Steam Line

• One or more Max Normal o One or more Max Safe

• HPCI Steam Line Reactor Building Radiation Reactor Building Radiation

• RCIC Steam Line Limita (EO-000-104 TableS9) Limita (EO-000-104 TableS9)

BNone None *

• RWCU Feedwater that can be read in the Control Room (Table F-3) that can be read in the Control Room (Table F-h) None RCS Leak Rate OR OR OR

• One or more Max Normal

• One or more Max Safe

2. Emergency RPV Depresaurization Reactor Building area Reactor Building area is required temperature Limita temperature Limita (EO-000-104 Table 8) that can (EO-000-1 04 Table 8) that be read in the Control Room can be read in the Control (Table F-4) Room (Table F-6)

1. UNPLANNED rapid drop in 1. Primary Containment presaure Primary Containment presaure *53pg Cfollowing Primary Containment OR CoeNoe1. Primary Containment presaure Noepresaure rice 2. Defiagration concentrationa exiat PC NnNoe> 1.72 paig due to RCS leakage NoeOR inaide PC (tH2 L:=6%AND 02 >5%)

Conditions OR

2. Primary Containment presaure reaponse not conalatent with 3. Neat Capacity Temperature Limit LOCA conditiona (HCTL) exceeded D OCRRMrdain .63Rh 1. CIHRRM radiation

• 7.0E+0 R/hr PC Rad / ORNone with indications of RCS leakage None None 1. Cl-RRM radiation > 4.OE+4 R/hr RCS 2. Primary coolant activity>* 300 inaide the drywall Activity pCi/gm 1-131 doan equivalent

1. UNISOLABLE direct downatream pathway to the environment exiata after Primary Containment E isolation aignat PCItgiyNone None None None ORNone or Byass .2. Intentional Primary Containment venting per EP-DS-004 RPV and PC Venting F 1. Any condithon in the opinion of 1. Any condition in the opinion of the 1. Any condition in the opinion of the 1. Any condition in the opinion of the 1.Any condition in the opinion of the 1. Any condition in the opinion of the EDRietor/e covergnyMaae ht Emergency Director/Recovery Emergency Director/Recovery Emergenoy Director/Recovery Emergency Director/Recovery Emergency Director/Recovery indicatec loaa of the fuel clad Manager that indicatea potential Manager that indicates baasof the Manager that indicates potential loaa Manager that indicatec 1oss of the Manager that indicatec potential loaa Judgment barrier loaa of the fuel clad barrier RCS barrier of the ROS barrier Primary Containment barrier of the Primary Containment barrier Page 3 of 60

Attachment 2 EP-RM-004 Revision [X]

Page 216 of 281 Table F-2 Fission Product Barrier Status Table Circle the X's in the FG1: General Emergency FS1: Site Area Emergency FAI: Alert table below for all Loss of any two barriers Loss or potential loss of any two barriers Any loss or any potential loss applicable and loss or potential loss of of EITHER Fuel Clad or RCS situations. Declare third barrier. barrier the EAL based upon all circled X's in any column.

Fuel Clad -Loss X X X X X X X X Fuel Clad -Potential X X X X X X Loss RCS -Loss X X X X X X X X RCS -Potential Loss X X X X X X Primary Containment X X X X X X X

-Loss Primary Containment X X X X X

-Potential Loss Page 4 of 60

Attachment 2 EP-RM-004 Revision [X]

Page 217 of 281 Barrier: Fuel Clad Category: A. RPV Level Degradation Threat: Loss Threshold:

1.SGentry required Definition(s):

N/A Basis:

The Loss threshold represents the EOP requirement for SAG entry. This is identified in the BWROG EPGs/SAGs when the phrase, "SAG Entry Is Required," EOPs specify the requirement for entry to the SAGs 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.

SAG entry is required when (ref. 1, 2):

• RPV water level CANNOT BE RESTORED AND MAINTAINED above -179 in.

(MSCRWL)

AND

• RPV water level above -210 in. (Get pump suction) with at least one core spray loop injecting into the RPV at > 6350 gpm OR

• TSC confirmation that core damage is progressing due to inadequate core cooling The above EOP conditions represent a challenge to core cooling and are the minimum values to assure core cooling without further degradation of the clad.

SAGs entry is also a Potential Loss of the Primary Containment barrier (PC P-Loss A.1 ) which constitutes a Site Area Emergency. The threshold for the RCS barrier (RCS Loss A.1) should also be evaluated for escalation to a General Emergency if SAGs entry results in meeting that threshold.

Basis Reference(s):

1. EP-DS-0O1 Containment Combustible Gas Control

2. EP-DS-002 RPV and Primary Containment Flooding SAG-2

3. NEI 99-01 RPV Water Level Fuel Clad Loss 2.A Page 5 of60

Attachment 2 EP-RM-004 Revision [X]

Page 218 of 281 Barrier: Fuel Clad Category: A. RPV Level Degradation Threat: Potential Loss Threshold:

1. RPV level CANNOT BE RESTORED AND MAINTAINED > -161 in. or CANNOT be determined Definition(s):

N/A Basis:

An RPV water level instrument reading of -161 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.

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.

When RPV water level cannot be determined, EOPs require entry to EO-000-1 14, 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 EO-000-1 14 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.

This threshold is intended to be interpreted the same as in EO-000-1 14, that is, a loss of instrumentation is not, by itself, a loss of ability to determine level.

Note that EO-000-1 13, Level/Power Control, may require intentionally lowering RPV water level to -161 in. and control level between the Minimum Steam Cooling RPV Water Level (MSCRWL) and the top of active fuel (ref. 3). 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 6 of 60

Attachment 2 EP-RM-004 Revision [X]

Page 219 of 281 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 A.1. 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.

In high-power ATWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level to the top of active fuel in order to reduce reactor power. RPV water level is then controlled between the top of active fuel and the Minimum Steam Cooling RPV Water Level (MSCRWL). Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority. For such events, ICs SA6 or SS6 will dictate the need for emergency classification.

Since the loss of ability to determine if adequate core cooling is being provided presents a significant challenge to the fuel clad barrier, a potential loss of the fuel clad barrier is specified.

Basis Reference(s):

1. EO-O00-1 02 RPV Control

2. EO-000-1 14 RPV Flooding

3. EO-000-1 13 Level/Power Control

4. NEI 99-01 RPV Water Level Fuel Clad Potential Loss 2.A Page 7 of 60

Attachment 2 EP-RM-004 Revision [X]

Page 220 of 281 Barrier: Fuel Clad Category: B. RCS Leak Rate Degradation Threat: Loss Threshold:

None Page 8 of 60

Attachment 2 EP-RM-004 Revision [X]

Page 221 of 281 Barrier: Fuel Clad Category: B. ROS Leak Rate Degradation Threat: Potential Loss Threshold:

None Page 9 oft6O

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Page 222 of 281 Barrier: Fuel Clad Category: C. PC Conditions Degradation Threat: Loss Threshold:

None Page 10 of 60

Attachment 2 EP-RM-004 Revision XIX Page 223 of 281 Barrier: Fuel Clad Category~ C. PC Conditions Degradation Threat: Potential Loss Threshold:

None Page 11 of 60

Attachment 2 EP-RM-004 Revision [X]

Page 224 of 281 Barrier: Fuel Clad Category: D. PC Radiation / RCS Activity Degradation Threat: Loss Threshold:

1. C RMradiation > 3.0E+3 R/hr Definition(s):

None Basis:

Two gamma radiation levels ion chambers are located inside the drywell to monitor post-accident radiation levels. The detectors are located in the drywell on elevation 719'. They are always in service and read out on the 0601 panel. Range is 100 to 108 R/hr. Environmentally qualified since they are required for use after an accident. A U-234 bug source is installed in the detector to serve as a self-check for instrument operability. The source provides approximately a 1 R/hr dose rate with the reactor shutdown. When the plant is at 100% power, drywell rad indication is normally about 3-4 R/hr. A reading of 3,000 R/hr indicates the release of reactor coolant into the drywell with elevated activity indicative of fuel clad damage (ref. 2).

For a fuel failure event equivalent to approximately 1% of cladding failure and an instantaneous and complete release of reactor coolant to the primary containment, the response of the Containment High Range Radiation Monitors (CHRRM) in the drywell will be approximately 3,450 R/hr immediately after shutdown (rounded to 3,000 R/hr, which approximates the dose rate 10 minutes after shutdown). This assumes that the release has occurred soon after reactor shutdown, and that the fuel cladding failures produce a coolant source term of 300 #tCi/gm of 1-131 dose equivalent just prior to the release into primary containment. (ref. 2)

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 D.1 since it indicates a loss of both the Fuel Clad Barrier and the RCS Barrier. Note that a combination of the two monitor readings appropriately escalates the emergency classification level to a Site Area Emergency.

There is no Fuel Clad Barrier Potential Loss threshold associated with Primary Containment Radiation.

Page 12 of 60

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Page 225 of 281 Basis Reference(s):

1. General Atomic High-Range Gamma Radiation Monitoring System Manual

2. Calculation EC RADN 0525 Rev 2, "Estimation of Containment High Range Radiation Monitor Response to a Loss of Coolant Accident for Emergency Planning Purposes,"

January 8, 2008

3. NEI 99-01 Primary Containment Radiation Fuel Clad Loss 4.A Page 13 of 60

Attachment 2 EP-RM-004 Revision [X]

Page 226 of 281 Barrier: Fuel Clad Category: D. PC Radiation / RCS Activity Degradation Threat: Loss Threshold:

2. Primary coolant activity > 300 IpCi/gm 1-131 dose equivalent Basis:

The Fuel Clad Barrier Shall be declared "lost" if the Primary coolant activity is determined to be

> 300 IpCi/gm 1-131 dose equivalent. Two separate methods can be used make this determination:

• Sample collection and analysis of reactor coolant activity

• Analysis of plant parameters to determine fuel clad damage > 1%

Sample collection and analysis of reactor coolant activity are accomplished in accordance with CH-ON-007.

It is recognized that sample collection and analysis of reactor coolant with highly elevated activity levels could require several hours to complete. The fuel clad damage analysis methodology in ref. 3 provides an alternative method to determine if the 1-131 dose equivalent activity is > 300 tJCi/gm.

Fuels Engineering determines if > 1% fuel clad damage has occurred based on the analysis methodology in ref. 3. Fuel clad damage equal to 1% corresponds to at least 300 micro-Ci/gm 1-131 dose equivalent in the reactor coolant, and drywell radiation values of at least 3000 R/hr during LOCA events (breach inside primary containment) (ref. 1). However, drywell radiation levels can be significantly lower than 3000 R/hr with a similar amount of fuel damage without a LOCA (no breach inside primary containment), since the fission products remain in the reactor vessel/do not escape into the drywell space and the CHRM is shielded from the radiation source (ref. 2 and 3).

Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.

There is no Potential Loss threshold associated with RCS Activity.

Basis Reference(s):

1. Calculation EC RADN 0525 Rev 2, "Estimation of Containment High Range Radiation Monitor Response to a Loss of Coolant Accident for Emergency Planning Purposes,"

January 8, 2008

2. EWR 1642196 - Investigation of EPLAN 1% Clad Damage vs. Barrier Loss

3. EP-PS-324 - Fuels Lead Engineer/Core Thermal Hydraulics Engineer

4. CH-ON-007 - Emergency Sampling Procedures

5. NEI 99-0 1 RCS Activity Fuel Clad Loss 1I.A Page 14 of 60

Attachment 2 EP-RM-004 Revision [XI Page 227 of 281 Barrier: Fuel Clad Category: 0. PC Radiation / RCS Activity Degradation Threat: Potential Loss Threshold:

None Page 15 of 60

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Page 228 of 281 Barrier: Fuel Clad Category: E. PC Integrity or Bypass Degradation Threat: Loss Threshold:

None Page 16 of 60

Attachment 2 EP-RM-004 Revision [X]

Page 229 of 281 Barrier: Fuel Clad Category: E. PC Integrity or Bypass Degradation Threat: Potential Loss Threshold:

None Page 17 of 60

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Page 230 of 281 Barrier: Fuel Clad Category: F. ED/RM Judgment Degradation Threat: Loss Threshold:

1. Any condition in the opinion of the Emergency Director/Recovery Manager that indicates loss of the Fuel Clad barrier Definition(s):

None Basis:

The Emergency Director/Recovery Manager judgment threshold addresses any other factors relevant to determining if the Fuel Clad barrier is lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

• Imminent barrier degqradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance. The term "imminent" refers to recognition of the inability to reach safety acceptance criteria before completion of all checks.

• Barrier monitorinq capability is decreased if there is a loss or lack of reliable indicators.

This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.

• Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Director/Recovery Manager should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.

Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment Fuel Clad Loss 6.A Page 18 of 60

Attachment 2 EP-RM-004 Revision [X]

Page 231 of 281 Barrier: Fuel Clad Category: F. ED/RM Judgment Degradation Threat: Potential Loss Threshold:

1. Any condition in the opinion of the Emergency Director/Recovery Manager that indicates potential loss of the Fuel Clad barrier Basis:

The Emergency Director/Recovery Manager judgment threshold addresses any other factors relevant to determining if the Fuel Clad barrier is potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

• Imminent barrier deqradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance. The term "imminent" refers to recognition of the inability to reach safety acceptance criteria before completion of all checks.

• Barrier monitorinci capability is decreased if there is a loss or lack of reliable indicators.

This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.

• Dominant accident seqiuences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Director/Recovery Manager should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.

The Emergency Director/Recovery Manager should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment Potential Fuel Clad Loss 6.A Page 19 of 60

Attachment 2 EP-RM-004 Revision [X]

Page 232 of 281 Barrier: Reactor Coolant System Category: A. RPV Level Degradation Threat: Loss Threshold:

1. RPV level CANNOT BE RESTORED AND MAINTAINED > -161 in. or CANNOT be determined Definition(s):

None Basis:

An RPV level instrument reading of -161 in. indicates RPV 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 Primary Containment barriers, and initiation of all ECCS. If RPV level cannot be maintained above TAF, ECCS and other sources of RPV injection have been ineffective or incapable of reversing the decreasing 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.

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.

When RPV water level cannot be determined, EOPs require entry to EO-000-1 14, 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 RCS barrier is threatened and reliance on alternate means of assuring adequate core cooling must be attempted. The instructions in EO-000-1 14 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 loss of the RCS barrier exists. This threshold is intended to be interpreted the same as in EO-000-1 14, that is, a loss of instrumentation is not, by itself, a loss of ability to determine level.

The conditions of this threshold are also a Potential Loss of the Fuel Clad barrier (FC P-Loss A.1). A Loss of the RCS barrier and Potential Loss of the Fuel Clad barrier requires a Site Area Emergency classification.

Note that EO-000-1 13, Level/Power Control, may require intentionally lowering RPV water level to -161 in. and control level between the Minimum Steam Cooling RPV Water Level (MSCRWL) and the top of active fuel (ref. 3). Although such action is a challenge to core cooling and the Page 20 of 60

Attachment 2 EP-RM-004 Revision [X]

Page 233 of 281 Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority. For such events, l~s SA6 or SS6 will dictate the need for emergency classification.

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 A.1.

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 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.

Basis Reference(s):

1. EO-000-1 02 RPV Control

2. EO-000-1 14 RPV Flooding

3. EO-000-1 13 Level/Power Control

4. NEI 99-01 RPV Water Level RCS Loss 2.A Page 21 of 60

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Page 234 of 281 Barrier: Reactor Coolant System Category: A. RPV Level Degradation Threat: Potential Loss Threshold:

None Page 22 of 60

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Page 235 of 281 Barrier: Reactor Cooiant System Category: B. RCS Leak Rate Degradation Threat: Loss Threshold:

1. UNISOLABLE break in ANY of the following:

• Main Steam Line

• HPCI Steam Line

• ROIC Steam Line

• RWCU

• Feedwater Definition(s):

UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.

• The term UNISOLABLE also includes any decision by plant staff or procedure direction to not isolate a primary system.

• Normal leakage past a closed isolation valve is not considered UNISOLABLE leakage.

Basis:

As used in this threshold, local isolation actions can only be credited if isolation can be completed promptly (i.e. within 15 min.).

In the case of a failure of both isolation valves to close but in which no downstream flow path exists, emergency declaration under this threshold would not be required. Similarly, if the emergency response requires the normal process flow of a system outside primary 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 Primary Containment (see PC Loss E.1) barriers and justifies declaration of a Site Area Emergency (i.e.,

Loss or Potential Loss of any two barriers).

High steam flow and high steam tunnel temperature annunciators are an indication of a Main Steam Line break. Either parameter will cause an isolation of the MSIVs. Note that the high steam flow alarm may clear if any of the MSIVs close and flow is reduced below the setpoint. If the high steam flow alarm was received (even though it was subsequently cleared) or there is other indication of high flow along with the high temperature alarm, the entry condition for this threshold has been met.

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. Note:

Each of the two feedwater injection lines is isolated from the reactor vessel via a series of swing Page 23 of 60

Attachment 2 EP-RM-004 Revision [X]

Page 236 of 281 check valves. The ability of these check valves to isolate cannot be determined until after feedwater is no longer injecting into the reactor vessel.

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, the RCS barrier Loss threshold is met.

Basis Reference(s):

1. FSAR Section 5.4.5

2. FSAR Section 6.3

3. FSAR Section 5.4.6

4. FSAR Section 10.4.7

5. FSAR Section 5.4.8

6. P&ID M-141 Nuclear Boiler

7. NEI 99-01 RCS Leak Rate RCS Loss 3.A Page 24 of 60

Attachment 2 EP-RM-004 Revision [X]

Page 237 of 281 Barrier: Reactor Coolant System Category: B. RCS Leak Rate Degradation Threat: Loss Threshold:

2. Emergency RVDepressurization isrequired Definition(s):

N/A Basis:

Plant symptoms requiring Emergency RPV Depressurization per the EOPs are indicative of a loss of the RCS barrier. If emergency depressurization is required, 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 exists due to the diminished effectiveness of the RCS pressure barrier to a release of fission products beyond its boundary.

Basis Reference(s):

1. EO-000-1 02 RPV Control

2. EO-000-103 Primary Containment Control

3. EO-000-104 Secondary Containment Control

4. EO-000-105 Radioactivity Release Control

5. EO-000-1 12 Emergency RPV Depressurization

6. EO-000-1 13 Level Power Control

7. EO-000-1 14 RPV Flooding

8. NEI 99-01 RCS Leak Rate RCS Loss 3.B Page 25 of 60

Attachment 2 EP-RM-004 Revision [X]

Page 238 of 281 Barrier: Reactor Coolant System Category: B. RCS Leak Rate Degradation Threat: Potential Loss Threshold:

1. UNISOLABLE primary system leakage that results in exceeding EITHER of the following:

• One or more Max Normal Reactor Building Radiation Limits (EO-000-1 04 Table 9) that can be read in the control room (Table F-3)

OR

• One or more Max Normal Reactor Building Area Temperature Limits (EO-000-1 04 Table 8) that can be read in the control room (Table F-4)

Table F-3 Max Normal Reactor Building Radiation Limits RB Area ARM Channel Max Norm ARM NumberDecito Elevation (ft) DsrpinRad Limit 818 35 Cask Stor Area Hi Alarm 14 Spent Fuel Crit Mon 15 Refuel Floor North (South U2) 42 Refuel Floor West 47 (44 U2) Spent Fuel Crit Mon 749 8 RWCU Recirc PP Access Hi Alarm 10 Fuel Pool PP Area 11 Rx Bldg Sample St 719 5 CRD North Hi Alarm 6 CRD South 670 16 Remote Shutdown Room Hi Alarm 645 3 HPCI PP & Turbine Room Hi Alarm 2 RCIC PP & Turbine Room 25 RHR A CPP Room 1 RHR B DPPRoom 4 .RB/RW Sump Area Page 26 of 60

Attachment 2 EP-RM-004 Revision [X]

Page 239 of 281 Table F-4 Max Normal Reactor Building Temperature Limits RB Area Max Normal Elevation Area Temperature Temp (ft) (°F) 749 RWCU-Pump Room 120 RWCU-Heat Exch Room 120 RWCU-Penetration Room 120 719 Main Steam Line Tunnel 157 683 HPCI Pipe Routing Area 120 RCIC Pipe Routing Area 120 645 HPCI-Equip Area 120 HPCI-Emerg Area Cooler 120 645 RCIC-Emerg Area Cooler 120 RCIC-Equip Area 120 645 RHR Equip Areal1 110 645 RHR Equip Area 2 110 Definition(s):

UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.

• The term UNISOLABLE also includes any decision by plant staff or procedure direction to not isolate a primary system (e.g. EP-O00-104 has direction not to isolate systems if they are needed for EOP actions or damage control)

• Normal leakage past a closed isolation valve is not considered UNISOLABLE leakage.

Basis:

This RCS Potential Loss threshold is limited to primary system leakage that results in exceeding one or more Max Normal Reactor Building Radiation or Temperature Limits that can be remotely determined from within the control room (ref. 1).

NRC regulations require the SSES 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.

Max Normal conditions shall be assumed to be from RCS leakage until proven otherwise.

1. The 15 minute classification requirement (EAL "trigger") for this threshold begins when one or more of the above Max Normal Reactor Building Radiation or Temperature Limits are exceeded.

Page 27 of 60

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Page 240 of 281

• If subsequent actions taken to isolate the leak are successful within the 15 minute classification period, this EAL should not be declared. (Note: EAL SU5.1 should be evaluated).

• If subsequent investigation, within the 15 minute classification period, reveals that the Max Norm conditions are not due to ROS leakage, this EAL should not be declared.

Note that a RCS leak could cause the fire suppression systems to actuate. If this occurs, the

• potential exists that the fire suppression systems could cause the area temperatures to be lower than the values specified in the EO-000-104 even though there is a RCS leak.

• Once it is known that the cause of exceeding Max Norm temperatures is due to a FIRE or loss of ventilation then this threshold is not met. The applicable FIRE EAL should be evaluated.

• If there is certainty that the fire suppression system actuation was caused by a RCS leak and NOT a fire then it is appropriate to judge that the MAX NORM temperature limits have been met even if the actual area temperatures are lower than those listed in the EO-000-1 04.

The presence of elevated general area temperatures or radiation levels in the secondary containment may be indicative of UNISOLABLE primary system leakage outside the primary containment.

The Max Normal Reactor Building Limit values define this RCS threshold because they are the maximum normal operating values and signify the onset of abnormal system operation. When parameters reach this level, equipment failure or misoperation 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 EO-000-1 04, Secondary Containment Control, Table 8 that can be read in the control room (ref. 2).

Cycling of SRVs to reduce primary system overpressure is not considered reactor coolant leakage. Inventory loss events, such as a stuck open SRV, venting and draining the RCS during cold shutdown or refueling, should not be considered when referring to "RCS leakage" because they are not indications of a break which could propagate. For these events entry into this threshold is not warranted however consideration should be given to RCS Loss - RCS Leak Rate threshold B2.

In general, multiple indications should be used to validate that a primary system is discharging outside Primary Containment. For example, a high area radiation condition 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.

Page 28 of 60

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Page 241 of 281 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.

The indicators reaching the threshold barriers and confirmed to be caused by RCS leakage from a primary system warrant an Alert classification. A primary system is defined to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system.

An UNISOLABLE leak as described above escalates to a Site Area Emergency when combined with Containment Barrier Loss threshold B.1 (after a containment isolation) and a General Emergency when the Fuel Clad Barrier criteria is also exceeded.

Basis Reference(s):

1. NCV 05000387; 388/2013005-04, Inadequate Instrumentation to Implement EALs for Fission Product Barrier Degradation

2. EO-000-104 Secondary Containment Control

3. NEI 99-01 RCS Leak Rate RCS Potential Loss 3.A Page 29 of 60

Attachment 2 EP-RM-004 Revision [X]

Page 242 of 281 Barrier: Reactor Coolant System Category: C. PC Conditions Degradation Threat: Loss Threshold:

1. Primary Containment pressure > 1.72 psig due to RCS leakage Definition(s):

None Basis:

The drywell high pressure scram setpoint is an entry condition to EO-000-1 02, RPV Control, and EO-000-1 03, Primary Containment Control (ref. 1, 2). Normal primary containment pressure control functions (e.g., operation of drywell coolers, vent through SGT, etc.) are specified in EO-000-103 in advance of less desirable but more effective functions (e.g., operation of drywell or suppression pool sprays, etc.).

In the design basis, primary 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 increasing 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 primary 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 primary containment pressure. PC pressure greater than 1.72 psig 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 1.72 psig should not be considered an RCS barrier Loss.

1.72 psig is the drywell high pressure setpoint which indicates a LOCA by automatically initiating the ECCS or equivalent makeup system.

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

Basis Reference(s):

1. EO-000-1 02 RPV Control

2. EO-000-103 Primary Containment Control

3. FSAR Section 6.2.1 Primary Containment Functional Design

4. NEI 99-01 Primary Containment Pressure RCS Loss 1I.A Page 30 of 60

Attachment 2 EP-RM-004 Revision IX]

Page 243 of 281 Barrier: Reactor Coolant System Category: C. PC Conditions Degradation Threat: Potential Loss Threshold:

None Page 31 of 60

Attachment 2 EP-RM-004 Revision [X]

Page 244 of 281 Barrier: Reactor Coolant System Category: D. PC Radiation / RCS Activity Degradation Threat: Loss Threshold:

1. CHRRM radiation > 7.0E+0 R/hr with indication of a RCS leak inside the drywell Definition(s):

N/A Basis:

Two gamma radiation levels ion chambers are located inside the drywell to monitor post-accident rad levels. The detectors are located in the drywell on elevation 719'. They are always in service and read out on the C601 panel. Range is 100 to 108 R/hr. Environmentally qualified since they are required for use after an accident. A U-234 bug source is installed in the detector to serve as a self-check for instrument operability. The source provides approximately a 1 R/hr dose rate with the reactor shutdown (ref. 1). When the plant is at 100% power, drywell rad indication is normally about 3-4 R/hr. Containment High Range Radiation Monitor (CHRRM) readings of approximately 3 R/hr indicate an instantaneous release of reactor coolant at normal operating concentrations of 1-131 to the drywell atmosphere. Adding this value to the normal CHRRM background readings of 3-4 R/hr (100% power normal operation) provides the value of 7 R/hr. (ref. 2)

Indication of a RCS leak into the drywall is added to qualify the radiation monitor indication to avoid declaring the loss of the RCS barrier for situations where the radiation increase is not due to a primary system leak. For situations that involve failure of the Fuel Clad Barrier alone, containment Radiation levels would increase to greater than 30 R/hr potentially giving a false indication of a loss of the RCS barrier. Therefore the EAL contains a qualifier to preclude over classification of the event if only the fuel clad barrier has failed. Indication of a leak should be determined by observing other containment indications such as sump level, drywell pressure and ambient temperature.

The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for Fuel Clad Barrier Loss threshold D.1 since it indicates a loss of the RCS Barrier only.

There is no RCS Barrier Potential Loss threshold associated with Primary Containment Radiation.

Basis Reference(s):

1. General Atomic High-Range Gamma Radiation Monitoring System Manual

2. Calculation EC RADN 0525 Rev 2, "Estimation of Containment High Range Radiation Monitor Response to a Loss of Coolant Accident for Emergency Planning Purposes,"

January 8, 2008 Page 32 of 60

Attachment 2 EP-RM-004 Revision [X]

Page 245 of 281

3. NEI 99-01 Primary Containment Radiation RCS Loss 4.A Page 33 of 60

Attachment 2 EP-RM-004 Revision IX]

Page 246 of 281 Barrier: Reactor Coolant System Category: D. PC Radiation / RCS Activity Degradation Threat: Potential Loss Threshold:

None Page 34 of 60

Attachment 2 EP-RM-004 Revision IX]

Page 247 of 281 Barrier: Reactor Coolant System Category: E. PC Integrity or Bypass Degradation Threat: Loss Threshold:

None Page 35 of 60

Attachment 2 EP-RM-004 Revision [X]

Page 248 of 281 Barrier: Reactor Coolant System Category: E. PC integrity or Bypass Degradation Threat: Potential Loss Threshold:

None Page 36 of 60

Attachment 2 EP-RM-004 Revision [X]

Page 249 of 281 Barrier: Reactor Coolant System Category: F. ED/RM Judgment Degradation Threat: Loss Threshold:

1. Any condition in the opinion of the Emergency Director/Recovery Manager that indicates loss of the RCS barrier Definition(s):

None Basis:

The Emergency Director/Recovery Manager judgment threshold addresses any other factors relevant to determining if the RCS barrier is lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

• Imminent barrier deqiradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance. The term "imminent" refers to the recognition of the inability to reach safety acceptance criteria before completion of all checks.

• Barrier monitorinq capability is decreased if there is a loss or lack of reliable indicators.

This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.

• Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Director/Recovery Manager should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.

Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment RCS Loss 6.A Page 37 of 60

Attachment 2 EP-RM-004 Revision [X]

Page 250 of 281 Barrier: Reactor Coolant System Category: F. EDIRM Judgment Degradation Threat: Potential Loss Threshold:

1. Any condition in the opinion of the Emergency Director/Recovery Manager that indicates potential loss of the RCS barrier Definition(s):

None Basis:

The Emergency Director/Recovery Manager judgment threshold addresses any other factors relevant to determining if the RCS barrier is potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

• , Imminent barrier deqradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance. The term "imminent" refers to the inability to reach final safety acceptance criteria before completing all checks.

• Barrier monitoringq capability is decreased if there is a loss or lack of reliable indicators.

This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.

• Dominant accident seguences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Director/Recovery Manager should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.

• The Emergency Director/Recovery Manager should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment RCS Potential Loss 6.A Page 38 of 60

Attachment 2 EP-RM-004 Revision [X]

Page 251 of 281 Barrier: Primary Containment Category: A. RPV Level Degradation Threat: Loss Threshold:

None Page 39 of 60

Attachment 2 EP-RM-004 Revision [X]

Page 252 of 281 Barrier: Primary Containment Category: A. RPV Level Degradation Threat: Potential Loss Threshold:

1. SAGs entry required Definition(s):

None Basis:

The Potential Loss threshold is identical to the Fuel Clad Loss RPV Water Level threshold A.I.

The Potential Loss requirement for SAG entry is required indicates adequate core cooling CANNOT BE RESTORED AND MAINTAINED and that core damage is possible. EOPs specify the requirement for entry to the SAGs 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.

SAG entry is required when (ref. 1, 2):

• RPV water level CANNOT BE RESTORED AND MAINTAINED above -179 in.

(MSCRWL)

AND

• RPV water level CANNOT BE RESTORED AND MAINTAINED above -210 in. (jet pump suction) with at least one core spray loop injecting into the RPV at > 6350 gpm OR

• TSC confirmation that core damage is progressing due to inadequate core cooling The above EOP conditions represent a challenge to core cooling and are the minimum values to assure core cooling without further degradation of the clad.

SAGs entry is also a Potential Loss of the Fuel Clad barrier (FC P-Loss A. 1) which constitutes a Site Area Emergency. The threshold for the RCS barrier (RCS Loss A. 1) should also be evaluated for escalation to a General Emergency if SAGs entry results in meeting that threshold.

PRA studies indicate that the condition of this Potential Loss threshold could be a core melt sequence which, if not corrected, could lead to RPV failure and increased potential for primary containment failure.

Basis Reference(s):

1. EP-DS-001 Containment Combustible Gas Control

2. EP-DS-002 RPV and Primary Containment Flooding SAG-2

3. NEI 99-01 RPV Water Level PC Potential Loss 2.A Page 40 of 60

Attachment 2 EP-RM.-004 Revision [X]

Page 253 of 281 Barrier: Primary Containment Category: B. RCS Leak Rate Degradation Threat: Loss Threshold:

1. UNISOLABLE primary system leakage that results in exceeding EITHER of the following:

• One or more Max Safe Reactor Building Radiation Limits (EO-000-1 04 Table 9) that can be read in the control room (Table F-5)

OR

• One or more Max Safe Reactor Building area temperature Limits (EO-000-104 Table 8) that can be read in the control room (Table F-6)

Table F-5 Max Safe Reactor Building Radiation Limits Max Safe Rad RB Area ARM ChannelLit Elevation (ft) AR ubrDescription (IR 818 49 Refuel Floor Area 10 749 52 RWCU Recirc PP Access 10 54 Fuel Pool PP Area 719 50 CRD North 10 51 CRD South 670 53 Remote Shutdown Room 10 645 48 HPCI PP & Turbine Room 10 57 ROIC PP & Turbine Room 55 RHR A CPP Room 56 RHR B 0 PP Room Page 41 of 60

Attachment 2 EP-RM-004 Revision [X]

Page 254 of 281 Table F-6 Max Safe Reactor Building Temperature Limits RB Area Max Safe Elevation Area Temperature Temp (ft) (°F) 749 RWCU-Pump Room 147 RWCU-Heat Exch Room 147 RWCU-Penetration Room 131 719 Main Steam Line Tunnel 177 683 HPCI Pipe Routing Area 167 RCIC Pipe Routing Area 167 645 HPCI-Equip Area 167 HPCI-Emerg Area Cooler 167 645 RCIC-Emerg Area Cooler 167 RCIC-Equip Area 167 645 RHR Equip Areal1 142 645 RHR Equip Area 2 142 Definition(s):

UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.

• The term UNISOLABLE also includes any decision by plant staff or procedure direction to not isolate a primary system (e.g. EP-O00-104 has direction not to isolate systems if they are needed for EOP actions or damage control)

• Normal leakage past a closed isolation valve is not considered UNISOLABLE leakage.

Basis:

This Primary Containment Loss threshold is limited to primary system leakage that results in exceeding one or more Max Safe Reactor Building Radiation or Temperature Limits that can be remotely determined from within the control room (ref. 1).

NRC regulations require the SSES 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.

The 15 minute classification requirement (EAL "trigger") for this threshold begins when one or more of the above Max Safe Reactor Building Radiation or Temperature Limits are exceeded.

Page 42 of 60

Attachment 2 EP-RM-004 Revision [X]

Page 255 of 281 Max Safe conditions shall be assumed to be from RCS leakage until proven otherwise.

• If subsequent actions taken to isolate the leak are successful, the threshold for this EAL is still met and must be declared. Once leakage is isolated, downgrading the emergency may be appropriate.

• If subsequent investigation, within the 15 minute classification period, reveals that the Max Safe conditions are not due to RCS leakage, this EAL should not be declared.

Note that a RCS leak could cause the fire suppression systems to actuate. If this occurs the potential exists that the fire suppression systems could cause the area temperatures to be lower than the values specified in the EO-000-104 even though there is a RCS.

• Once it is known that the cause of exceeding MAX SAFE temperatures is due to a FIRE or loss of ventilation then this threshold is not met. The applicable FIRE EAL should be evaluated.

• If there is certainty that the fire suppression system actuation was caused by a RCS leak and NOT a fire then it is appropriate to judge that the MAX NORM temperature limits have been met even if the actual area temperatures are lower than those listed in the EO-000-1 04 The presence of elevated general area temperatures or radiation levels in the secondary containment may be indicative of UNISOLABLE primary system leakage outside the primary containment.

• The Maximum Safe Reactor Building Limit values 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 primary 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 EO-000-1 04, Secondary Containment Control, Table 8 that can be read in the control room (ref. 2).

In general, multiple indications should be used to validate that a primary system is discharging outside Primary Containment. For example, a high area radiation condition 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.

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.

The temperatures and radiation levels should be confirmed to be caused by RCS leakage from a primary system. A primary system is defined to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system.

Page 43 of 60

Attachment 2 EP-RM-004 Revision [X]

Page 256 of 281 In combination with RCS Potential Loss B.1 this threshold would result in a Site Area Emergency.

There is no Potential Loss threshold associated with Primary Containment Isolation Failure.

Basis Reference(s):

1. NOV 05000387; 388/2013005-04, Inadequate Instrumentation to Implement EALs for Fission Product Barrier Degradation

2. EO-000-104 Secondary Containment Control

3. NEI 99-01 RCS Leak Rate PC Loss 3.C Page 44 of 60

Attachment 2 EP-RM-004 Revision XIX Page 257 of 281 Barrier: Primary Containment Category: B. RCS Leak Rate Degradation Threat: Potential Loss Threshold:

None Page 45 of 60

Attachment 2 EP-RM-004 Revision [X]

Page 258 of 281 Barrier: Primary Containment Category: C. PC Conditions Degradation Threat: Loss Threshold:

1. UNPLANNED rapid drop in Primary Containment pressure following Primary 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 primary containment pressure (i.e., not attributable to drywell spray or condensation effects) following an initial pressure increase 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.

Basis Reference(s):

1. NEI 99-01 Primary Containment Conditions PC Loss 1I.A Page 46 of 60

Attachment 2 EP-RM-004 Revision IX]

Page 259 of 281 Barrier: Primary Containment Category: C. PC Conditions Degradation Threat: Loss Threshold:

2. Primary Containment pressure response not consistent with LOCA conditions Definition(s):

None Basis:

The calculated pressure response of the containment is shown in Figure 6.2-2 (ref. 1)

(reproduced on next page). The maximum calculated drywell pressure (63.3 psia or 48.6 psig) is well below the design allowable pressure of 53 psig (ref. 2).

Due to conservatisms in LOCA analyses, actual pressure response is expected to be less than the analyzed response. For example, blowdown mass flowrate may be only 60-80% of the analyzed rate.

Primary containment pressure should increase as a result of mass and energy release into the primary containment from a LOCA. Thus, primary containment pressure not increasing under these conditions 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.

Basis Reference(s):

1. FSAR Figure 6.2-2

2. FSAR Section 6.2.1.1.3.1

3. NEI 99-01 Primary Containment Conditions PC Loss I .B Page 47 of 60

Attachment 2 EP-RM-004 Revision [X]

Page 260 of 281 Short-Term RSLB Pressure Response 70 60]

50

• -40 t730 20

- WW Press 10 V

0 5 10 15 20 25 30 Time (se;)

PSAR REV. 65 SUSQUJEHANNA STEAM ELECTRIC STATION UNIT"S 1 AND 2 FINAL SAFETY ANALYSIS REPORT PRESSURE RESPONSE FOR RECIRCULATION UINE BREAK FIGURE 6.2-2, Rev. 56 Auto Cad: Figuma Fsar G_2._2dwg Page 48 of 60

Attachment 2 EP-RM-004 Revision [X]

Page 261 of 281 Barrier: Containment Category: C. PC Conditions Degradation Threat: Potential Loss Threshold:

1. Primary Containment pressure > 53 psig Definition(s):

None Basis:

When the primary containment exceeds the maximum allowable value (53 psig) (ref. 1), primary containment venting may be required even if offsite radioactivity release rate limits will be exceeded (ref. 2). The drywell and suppression chamber maximum allowable value of 53 psig is based on the primary containment design pressure as identified in the SSES 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.

Basis Reference(s):

1. FSAR Section 6.2.1.1.3.1

2. EO-000-103 Primary Containment Control

3. NEI 99-01 Primary Containment Conditions PC Potential Loss 1I.A Page 49 of 60

Attachment 2 EP-RM-004 Revision [X]

Page 262 of 281 Barrier: Containment Category: C. PC Conditions Degradation Threat: Potential Loss Threshold:

2. Deflagration concentrations exist inside PC (H2 >_6% AND 02 >_5%)

Definition(s):

None Basis:

Deflagration (explosive) mixtures in the primary containment are assumed to be elevated concentrations of hydrogen and oxygen. BWR industry evaluation of hydrogen generation for development of EPGs/SAGs indicates that any hydrogen concentration above minimum detectable is not to be expected within the short term. Post-LOCA hydrogen generation primarily caused by radiolysis is a slowly evolving, long-term condition. Hydrogen concentrations that rapidly develop are most likely caused by metal-water reaction. A metal-water reaction is indicative of an accident more severe than accidents considered in the plant design basis and would be indicative, therefore, of a potential threat to primary containment integrity. Hydrogen concentration of approximately 6% is considered the global deflagration concentration limit (ref. 1).

Except for brief periods during plant startup and shutdown, oxygen concentration in the primary containment is maintained at insignificant levels by nitrogen inertion. The specified values for this Potential Loss threshold are the minimum global deflagration concentration limits (6%

hydrogen and 5% oxygen) and readily recognizable because 6% hydrogen is well above the EO-000-1 03, Primary Containment Control, entry condition (ref. 2, 3). Values above the hydrogen/oxygen concentrations (6% and 5%, respectively) require intentional primary containment venting, which is defined to be a Loss of Containment (PC Loss E.2).

Hydrogen and oxygen concentration in the drywell during normal operation. These monitors are isolated by accident isolation signals. However, monitors CAC-AT-4409 and 4410 will be realigned to the primary containment for post-accident monitoring via an operator actuated isolation signal override circuit when directed by the EOPs.

If hydrogen concentration reaches or exceeds the lower flammability limit, as defined in plant EOPs, in an oxygen rich environment, a potentially explosive mixture exists. If the combustible mixture ignites inside the primary containment, loss of the Containment barrier could occur.

Basis Reference(s):

1. BWROG EPG/SAG Revision 3, Subsection PC/G

2. EO-000-103 Primary Containment Control

3. EP-DS-O01 Containment Combustible Gas Control

4. NEI 99-01 Primary Containment Conditions PC Potential Loss I .B Page 50 of 60

Attachment 2 EP-RM-004 Revision [X]

Page 263 of 281 Barrier: Containment Category: C. PC Conditions Degradation Threat: Potential Loss Threshold:

3. Heat Capacity Temperature Limit (Figure - HCTL) exceeded Figure - HCTL Heat Capacity Temperature Limit Definition(s):

None Basis:

This threshold is met when the final step of section SP/T in EO-000-1 03, Primary Containment Control, is reached (ref. 1).

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.

Basis Reference(s):

1. EO-000-1 03 Primary Containment Control

2. NEI 99-01 Primary Containment Conditions PC Potential Loss 1 .C Page 51 of 60

Attachment 2 EP-RM-004 Revision IX]

Page 264 of 281 Barrier: Primary Containment Category: 0. PC Radiation / ROS Activity Degradation Threat: Loss Threshold:

None Page 52 of 60

Attachment 2 EP-RM-004 Revision [X]

Page 265 of 281 Barrier: Primary Containment Category: D. PC Radiation / RCS Activity Degradation Threat: Potential Loss Threshold:

1. CHRRM radiation > 4.0E+4R/hr Definition(s):

None Basis:

Two gamma radiation levels ion chambers are located inside the drywell to monitor post-accident radiation levels. The detectors are located in the drywell on elevation 719'. They are always in service and read out on the 0601 panel. Range is 100 to 108 R/hr. Environmentally qualified since they are required for use after an accident. A U-234 bug source is installed in the detector to serve as a self-check for instrument operability. The source provides approximately a 1 R/hr dose rate with the reactor shutdown (ref. 1). When the plant is at 100% power, Containment High Range Radiation Monitor (CHRRM) indication is normally about 3-4 R/hr. A reading of 40,000 R/hr indicates the release of reactor coolant into the drywell with elevated activity indicative of 20% fuel clad damage (ref. 2).

An analysis of the Containment High Range Radiation Monitor (CHRRM) response to a Loss-Of-Coolant Accident (LOCA) is given in reference 2. Results are summarized in Reference 2, Section 2 for the case of 1% clad damage. For this threshold a reactor shutdown time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is conservatively assumed. Although CHRRM data is always available for emergency planning, a one hour shutdown time is conservative for times less than one hour and is applicable to the timing of the sequence of events for a LOCA that would result in both the release of reactor coolant activity to containment and damage to 20% of the fuel clad. From reference 2, Section 2, the calculated CHRRM dose rate at one hour after shutdown for a complete loss of reactor coolant activity to containment with 1% clad damage is 2160 R/hr. Multiplying this value by a factor of 20 to account for 20% clad damage results in a CHRRM dose rate of 43,200 R/hr. This value is rounded to 40,000 R/hr for human factors considerations.

A Containment High Range Rad Monitor reading > 40,000 R/hr is a value which indicates significant fuel damage well in excess of that required for loss of RCS and Fuel Clad. A major failure of fuel cladding which allows radioactive material to be released from the core into the reactor coolant could result in a major release of radioactivity requiring offsite protective actions.

Regardless of whether containment is challenged, this amount of activity in containment corresponding to a CHHRM reading > 40,000 R/hr, if released, could have such severe consequences that it is prudent to treat this as a potential loss of containment, such that a General Emergency declaration is warranted.

In order to reach this Containment barrier Potential Loss threshold, a loss of the RCS barrier (RCS Loss D.1 ) and a loss of the Fuel Clad barrier (FC Loss D.1) have already occurred. This threshold, therefore, represents at a General Emergency classification.

Page 53 of 60

Attachment 2 EP-RM-004 Revision [X]

Page 266 of 281 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. 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.

Basis Reference(s):

1. General Atomic High-Range Gamma Radiation Monitoring System Manual

2. Calculation EC RADN 0525 Rev 2, "Estimation of Containment High Range Radiation Monitor Response to a Loss of Coolant Accident for Emergency Planning Purposes,"

January 8, 2008

3. NEI 99-01 Primary Containment Radiation Fuel Clad Potential Loss 1.0 Page 54 of 60

Attachment 2 EP-RM-004 Revision [X]

Page 267 of 281 Barrier: Primary Containment Category: E. PC Integrity or Bypass Degradation Threat: Loss Threshold:

Definition(s):

UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.

• The term UNISOLABLE also includes any decision by plant staff or procedure direction to not isolate a primary system (e.g. EP-OO0-104 has direction not to isolate systems if they are needed for EOP actions or damage control)

• Normal leakage past a closed isolation valve is not considered UNISOLABLE leakage.

Basis:

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 with a pathway directly to the environment indicates a breach of primary containment integrity. Examples include:

• This EAL is applicable if plant operators attempt to close isolation valves before any automatic setpoints are reached, and both valves fail to close AND a downstream pathway to the environment exists. If subsequent actions are successful, downgrading the emergency may be appropriate.

• Main Steam Line breaks with two MSIVs in one line failing to close.

• HPCI, RCIC or RWCU steam line breaks with all isolation valves in one line failing to close.

• UNISOLABLE containment atmosphere vent paths.

• Automatic closure of both isolation valves in one line are disabled AND an isolation signal occurs AND a downstream pathway exits.

The adjective "direct" modifies "release pathway" to discriminate against release paths through interfacing liquid systems. The following examples do not meet the EAL threshold:

• If the main condenser is available with an UNISOLABLE main intact steam line, there may be releases through the steam jet air ejectors and gland seal exhausters. These pathways are monitored, however, and do not meet the intent of an UNISOLABLE direct release path to the environment. These minor releases are assessed using the Category R, Abnormal Rad Release/IRad Effluent, EALs.

• Leakage into a closed system (e.g. RHR, Core Spray) is to be considered only if the closed system is breached and thereby creates a significant pathway to the environment.

• Normal leakage past a closed isolation valve is not considered UNISOLABLE leakage.

Page 55 of 60

Attachment 2 EP-RM-004 Revision [X]

Page 268 of 281

• SCRAM Discharge Volume valves fail to close. The SDV drains to the Reactor Building Sump, which creates the pathway to the environment.

• The existence of an in-line charcoal filter (SGTS) does not make a release path indirect since the filter is not effective at removing fission noble gases. Typical filters have an efficiency of 95-99% removal of iodine. Given the magnitude of the core inventory of iodine, significant releases could still occur. In addition, since the fission product release would be driven by boiling in the reactor vessel, the high humidity in the release stream can be expected to render the filters ineffective in a short period.

• EO-000-103, Primary Containment Control, Section PC/P 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.

Declaration of this EAL threshold constitutes a radiological release in progress.

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).

The existence of a filter is not considered in the threshold assessment. Filters do not remove fission product noble gases. In addition, a filter could become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream.

Following the leakage of RCS mass into primary containment and a rise in primary containment pressure, there may be minor radiological releases associated with allowable primary containment leakage through various penetrations or system components. Minor releases may also occur if a primary containment isolation valve(s) fails to close but the primary containment atmosphere escapes to an enclosed system. These releases do not constitute a loss or potential loss of primary containment but should be evaluated using the Recognition Category R ICs.

Basis Reference(s):

1. EO-000-1 03 Primary Containment Control

2. NEI 99-01 Primary Containment Isolation Failure PC Loss 3.A Page 56 of 60

Attachment 2 EP-RM-004 Revision [X]

Page 269 of 281 Barrier: Primary Containment Category: E. PC Integrity or Bypass Degradation Threat: Loss Threshold:

2. Intentional Primary Containment venting per EP-DS-004 RPV and PC Venting Definition(s):

None Basis:

EO-000-1 03, 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). The threshold is met when the operator begins venting the primary containment in accordance with EP-DS-004, RPV and PC Venting, not when actions are taken to bypass interlocks prior to opening the vent valves (ref. 2).

Because the containment vent valves are not qualified for opening/reclosure in a post-accident environment, there is no guarantee that venting, once initiated, can be terminated. Thus it is assumed that once the vents are opened with a source term, they are not re-closed (ref. 3).

Intentional venting of primary containment for primary containment pressure 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.

Basis Reference(s):

1. EO-000-103 Primary Containment Control

2. EP-DS-004 RPV and PC Venting

3. NL-98-036, SSES Safety Evaluation for EP-DS-004, Primary Containment and RPV Venting

4. NEI 99-01 Primary Containment Isolation Failure PC Loss 3.B Page 57 of 60

Attachment 2 EP-RM-004 Revision XIX Page 270 of 281 Barrier: Primary Containment Category: E. PC Integrity or Bypass Degradation Threat: Potential Loss Threshold:

None Page 58 of 60

Attachment 2 EP-RM-004 Revision [X]

Page 271 of 281 Barrier: Primary Containment Category: F. ED/RM Judgment Degradation Threat: Loss Threshold:

1. Any condition in the opinion of the Emergency Director/Recovery Manager that indicates loss of the Primary Containment barrier Definition(s):

None Basis:

The Emergency Director/Recovery Manager judgment threshold addresses any other factors relevant to determining ifthe Primary Containment barrier is lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

• Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance. The term "imminent" refers to recognition of the inability to reach safety acceptance criteria before completion of all checks.

• Barrier monitoringq capability is decreased if there is a loss or lack of reliable indicators.

This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.

= Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Director/Recovery Manager should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.

Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment PC Loss 6.A Page 59 of 60

Attachment 2 EP-RM-004 Revision [X]

Page 272 of 281 Barrier: Primary Containment Category: F. ED/RM Judgment Degradation Threat: Potential Loss Threshold:

1. Any condition in the opinion of the Emergency Director/Recovery Manager that indicates potential loss of the Primary Containment barrier Definition(s):

None Basis:

The Emergency Director/Recovery Manager judgment threshold addresses any other factors relevant to determining if the Primary Containment barrier is potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

• Imminent barrier deqradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance. The term "imminent" refers to recognition of the inability to reach safety acceptance criteria before completion of all checks.

• Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators.

This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.

• Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Director/Recovery Manager should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.

Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment PC Loss 6.A Page 60 of 60

Attachment 3 EP-RM-004 Revision [X]

Page 273 of 281 ATTACHMENT 3 Safe Operation & Shutdown Areas Tables R-2 & H-2 Bases

Background

NEI 99-01 Revision 6 l~s AA3 (SSES RA3.2) and HA5 (SSES HA5.1) 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-relatedmode applicabilityidentified" should specify those rooms or areasthat contain equipment which require a manual/local action as specified in operatingprocedures used for normal plant operation, cooldown and shutdown. Do not include rooms or areasin which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency conditioh*

such as emergency repairs, corrective measures or emergency operations). In addition, the list should specify the plant mode(s) during which entry would be requiredfor each room or area.

The list should not include rooms or areas for which entry is required solely to perform actions of an administrativeor 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, capabilityto draw air from multiple air intakes at different and separate locations, inner and outer atmospheric boundaries, or the capabilityto acquire and maintain positive pressure within the Control Room envelope.

Site-Specific Analysis The Control Room is excluded from Table R-2 since it is included EAL RA3.1 for all modes.

NEI 99-01 Rev. 6 developer notes state that the control room does not need to be included Table R-2 for this reason.

The Control Room is included in Table H-2 for all modes since adequate engineered safety/design features are not in place to preclude a Control Room evacuation due to the release of a hazardous gas at SSES.

The remaining list of mode dependent rooms in Tables R-2 and H-2 is a list of rooms where actions are absolutely required to be performed to move the plant from normal operations through cool down and to achieve and maintain cold shutdown. These areas are not areas requiring entry to solely meet surveillance requirements. This does not include actions called out for in the procedures that are not absolutely needed to move the plant into and maintain cold shutdown (e.g. shutting down turbine lube oils systems, opening of system drains). In order to transition from normal operations (Modes 1 and 2) to hot shutdown (Mode 3) the only location required is the Control Room since the plant is able to be placed into hot shutdown from the control room without the need for any other room entry.

Page 1 of 3

Attachment 3 EP-RM-004 Revision [X]

Page 274 of 281 The remaining rooms/areas shown in in Tables R-2 and H-2 were determined based on the discussion section above and Modes 1 and 2 were excluded. Therefore, the analysis starting point was a review the steps of GO-i100/200-005 "PLANT SHUTDOWN TO HOT/COLD SHUTDOWN" after the Unit is has reached Mode 3. The analysis then branched to other procedure steps described in the GO's if those procedures steps were not excluded from consideration as noted in the discussion section. The analysis concluded that Reactor Building areas and elevations listed in in Tables R-2 and H-2 contain equipment that must be manipulated in Mode 3 to achieve and maintain cold shutdown. Modes 4 and 5 are also included for these areas to account for swapping loops for long term operation of shutdown cooling.

Tables R-2 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 Unit 1 RA3.2 /HA5.1 Elevation Area(s)

Mode(s) 670 27 3,4,5 683 27, 28 & 29 3,4,5 703 28 &29 3,4,5 719 25 & 29 3,4,5 749 25 &29 3,4,5 729

• 12, 21 1,2,3,4,5,D Unit 2 RA3.2 / HA5.1 Mode(s)

Elevation Area(s) 670 32 3,4,5 683 32, 33 & 34 3,4,5 703 33 & 34 3,4,5 719 30 & 34 3,4,5 749 32 & 33 3,4,5 729

• 12, 21 1, 2,3,4,5,D

• Control Room - only applicable to HAS Page 2 of 3

Attachment 3 EP-RM-004 Revision [X]

Page 275 of 281 Table R-2 & H-2 Results Table R-2 & H-2 Safe Operation & Shutdown Areas Elevation Unit 1 Area(s) ** Unit 2 Area(s) ** Mode(s) 670' RB 27 RB 32 3,4,5 683' RB 27, 28, 29 RB 32, 33, 34 3,4,5 703' RB 28, 29 RB 33, 34 3,4,5 719' RB 25, 29 RB 30, 34 3,4,5 749' RB 25, 29 RB 32, 33 3,4,5 729' CS 12, 21" CSI12, 21" 1, 2, 3,4,5, D

• Control Room - only applicable to HA5

• • See Chart 1 for the specific locations of areas listed in Table R-2 and H-2.

Chart 1- Plant Area Key Plan 70 RADWAS V -

LOW LEV*TE II::LWATER TREATMENT

SPRAY POND

[* ACID AN[ VALVE VAULT IICHLORING EIEI PUMPHOUSE L+/-iBUILDIN( ESSW I PUMPHOUSE UNIT #2 IUI TURBINE BLDG.TR BINE BLDG. I F:tADWASTE 16 15 14 13 4 3 8 37 E

20 19 18 17iI 8I165 0 39 24 23 22 I21 12 11 10 9 45 ~7 25 36 35 a. COND &REF 26 STORAG COND. 33 29 28 44 43 *-~DIESEL GENEPRATOR

_RIIINTAKE IVER STRUCTURE E#2 REACTOR E#1 REACTOR r*E DIESEL

.JGENERATOR 4* T Page 3 of 3