Text

Response to Request for Additional Information and Emergency Action Level Technical Bases Document, Revision 1, Clean Version. Part 4 of 4
	ML16055A229
Person / Time
Site:	Catawba
Issue date:	02/19/2016
From:	Henderson K; Duke Energy Carolinas
To:	; Office of Nuclear Reactor Regulation
Shared Package
ML16055A223	List: CNS-16-010, Response to Request for Additional Information and Emergency Action Level Technical Bases Document, Revision 1, Clean Version. Part 4 of 4; ML16055A226; ML16055A227; ML16055A228; ML16055A230;
References
	CNS-16-010, TAC MF6166, TAC MF6167
	Download: ML16055A229 (248)
	v • d • e

V Enclosure 3 Catawba Nuclear Station Emergency Action Level Technical Bases Document Revision 1 (Clean Version)

~ENERGY CATAWBA NUCLEAR STATION EMERGENCY ACTION LEVEL TECHNICAL BASES Revision I Clean Version IRP/0/AI5000/001I Rev. I Page 1 of 2471

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 ............................................................................. 10 3.2 Classification Methodology.......................................................................... 10

4.0 REFERENCES

......................................................................................... 14 4.1 Developmental.................................................... ,.................................... 14 4.2 Implementing ......................................................................................... 14 5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS.................................................. 15 6.0 CNS TO NEI 99-01 Rev. 6 EAL CROSS-REFERENCE............................................ 24 7.0 ATTACHMENTS........................................................................................ 28 1 Emergency Action Level Technical Bases .................................................. 29 Categiory R Abnormal Rad Release / Rad Effluent .................................. 29 Catecqorv C Cold Shutdown / Refueling System Malfunction ....................... 71 Catecqory H Hazards .................................................................. 108 Cateqory S System Malfunction ..................................................... 153 Cate~qory E ISFSI ..................................................................... 197 Categiory F Fission Product Barrier Degradation .................................. 200 2 Fission Product Barrier Loss / Potential Loss Matrix and Bases ............................................................................ 205 3 Safe Operation & Shutdown Roomns/Areas Tables R-2 & H-2 Bases ................... 253 SRP/0/A/5000/001 Rev. I Page 2o°247

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

Decision-makers responsible for implementation of RP/0/A/5000/001 Classification of Emergency, may use this document as a technical reference in support of EAL interpretation.

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

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

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

2.0 DISCUSSION 2.1 Background EALs are the plant-specific indications, conditions or instrument readings that are utilized to classify emergency conditions defined in the CNS 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.

4.1.1), CNS conducted an EAL implementation upgrade project that produced the EALs discussed herein.

SRP/0/A/5000/001I Rev. 1 Page 3 of 247I

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 (NCS): The NCS Barrier includes the NCS primary side and its connections up to and including the pressurizer safety and relief valves, and other connections up to and including the primary isolation valves.

C. Containment (CMT): The Containment Barrier includes the containment building and connections up to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the containment building up to and including the outermost secondary side isolation valve.

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 NCS barrier Site Area Emerqency:

Loss or potentialloss of any two barriers General Emnergqency:

Loss of any two barriers and loss or potential loss of the third barrier SRP/0/A/5000/001 Rev. I Page 4of 247

2.4 EAL Organization The CNS 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 hot operating modes - This group would only be reviewed by the EAL-user when the plant is in Hot Shutdown, Hot Standby, 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 CNS EAL categories are aligned to and represent the NEI 99-01"Recognition Categories." Subcategories are used in the CNS scheme as necessary to further divide the EALs of a category into logical sets of possible emergency classification thresholds. The CNS EAL categories and subcategories are listed below.

IRP/0/A/5000/001 Rev. 1 Page 5 of 247

EAL Groups, Categories and Subcategories 3 - Area Radiation Levels H - Hazards and Other Conditions 1 - Security Affecting Plant Safety 2 - Seismic Event 3 - Natural or Technological Hazard 4 -Fire 5 - Hazardous Gas 6 - Control Room Evacuation 7 - Emergency Coordinator Judgment E - Independent Spent Fuel Storage I-CnieetBudr Installation (ISFSI) 1-CnieetBudr Hot Conditions:

S - System Malfunction 1 - Loss of Essential AC Power 2- Loss of Vital DC Power 3- Loss of Control Room Indications 4- NCS Activity 5- NCS Leakage 6 - RPS Failure 7 - Loss of Communications 8 - Containment Failure 9 - Hazardous Event Affecting Safety Systems F - Fission Product Barrier Degradation None Cold Conditions:

C - Cold Shutdown / Refueling System I- NCS Level Malfunction 2- Loss of Essential AC Power 3- NCS Temperature 4- Loss of Vital DC Power 5- Loss of Communications 6- Hazardous Event Affecting Safety Systems The primary tool for determining the emergency classification level is the EAL Classification Matrix. The user of the EAL Classification Matrix may (but is not required to) consult the EAL Technical Bases Document in order to obtain additional information concerning the EALs under classification consideration. The user should consult Section 3.0 and Attachments I & 2 of this document for such information.

2.5 Technical Bases Information IRPIOIA/50001001 Rev. I Page 6 of 247

EAL technical bases are provided in Attachment 1 for each EAL according to EAL group (Any, 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:

Category Letter & Title Subcateqory Number & Title Initiating Condition (IC)

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

EAL Identifier (enclosed in rectanqle)

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 rectangle):

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

EAL (enclosed in rectangle)

Exact wording of the EAL as it appears in the EAL Classification Matrix IRPIO/AI5000I001I Rev. 1I Page 7 of 247

Mode Applicability One or more of the following plant operating conditions comprise the mode to which each EAL is applicable: I - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 -

Cold Shutdown, 6 - 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 aiso be found in Section 5.1.

Basis:

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

CNS Basis Reference(s):

Site-specific source documentation from which the EAL is derived 2.6 Operating Mode Applicability (ref. 4.1.7) 1 Power Operation Keff > 0.99 and reactor thermal power > 5%

2 Startup Keff > 0.99 and reactor thermal power < 5%

3 Hot Standby Keff < 0.99 and average coolant temperature > 350°F 3 Hot Shutdown Keff < 0.99 and average coolant temperature 350°F > Tav > 200 °F 4 Cold Shutdown Keff < 0.99 and average coolant temperature < 200°F 5 Refuelingq One or more reactor vessel head closure bolts are less than fully tensioned D Defueled Reactor vessel contains no irradiated fuel The plant operating mode that exists at the time that the event occurs (prior to any protective system or operator action being initiated in response to the condition) should be compared to the mode applicability of the EALs. If a lower or higher plant operating mode is reached before the emergency classification is made, the declaration shall be based on the mode that existed at the time the event occurred.

SRPIOIAI50001001I Rev. 1 Page 8 of 247I

3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS 3.1 General Considerations When making an emergency classification, the Emergency Coordinator 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-ISG-01, "Interim Staff Guidance, Emergency Planning for Nuclear Power Plants" (ref. 4.1.12).

3.1.2 Valid Indications All emergency classification assessments shall be based upon valid indications, reports or conditions. A valid indication, report, or condition, is one that has been verified through appropriate means such that there is no doubt regarding the indicator's operability, the condition's existence, or the report's accuracy. For example, verification could be accomplished through an instrument channel check, response on related or redundant indicators, or direct observation by plant personnel.

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

3.1.3 Imminent Conditions For ICs and EALs that have a stipulated time duration (e.g., 15 minutes, 30 minutes, etc.), the Emergency Coordinator 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).

IRP/0/A/5000/001 Rev. 1 Page 9 of 247

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, NOS 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 Coordinator 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 Coordinator 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 Coordinator 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. 4.1.14).

3.2.1 Classification of Multiple Events and Conditions When multiple emergency events or conditions are present, the user will identify all met or exceeded EALs. The highest applicable ECL identified during this review is declared. For example:

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

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

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

Related guidance concerning classification of rapidly escalating events or conditions is provided in Regulatory Issue Summary (RIS) 2007-02, Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events (ref. 4.1.2).

3.2.2 Consideration of Mode Changes During Classification SRP/0/A/5000/001 Rev. 1I Page 10 of 247I

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 Coordinator 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 Coordinator, 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. 4.1.2).

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 trip the reactor followed by a successful manual trip.

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.

EAL momentarily met durinqi expected plant response - In instances where an EAL is briefly met during an expected (normal) plant response, an emergency declaration is not warranted IRP/0/A/5000/001 Rev. 1 Page 11 of 247

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 - Ifan 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 EGGS 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 NCS barriers), If an operator manually starts a high pressure EGGS 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 Goordinator 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 Gondition 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.'1.3) is applicable. Specifically, the event should be reported to the NRG in accordance with 10 GFR § 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 NRG is discussed in NUREG-1 022 (ref. 4.1.3).

IRPIOIAI50001001 Rev. 1I Page 12 of 247

4.0 REFERENCES

4.1 Developmental 4.1.1 NEI 99-01 Revision 6, Methodology for the Development of Emergency Action Levels for Non-Passive Reactors, ADAMS Accession Number ML12326A805 4.1.2 RIS 2007-02 Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events, February 2, 2007.

4.1.3 NUREG-1 022 Event Reporting Guidelines: 10CFR50.72 and 50.73 4.1.4 10 § CFR 50.72 Immediate Notification Requirements for Operating Nuclear Power Reactors 4.1.5 10 § CER 50.73 License Event Report System 4.1.6 CNS-SLC-16.11-16 Figure 16.11-16-1 Unrestricted Area and Site Boundary for Radioactive Effluents 4.1.7 CNS UFSAR Figure 1-20 Plot Plan 4.1.8 Technical Specifications Table 1.1-1 Modes 4.1.9 OP/0/A16100/014 Penetration Control for Modes 5, 6 and NO Mode - Enclosure 4.7 Setting, Maintaining and Securing from Containment Penetration Control 4.1.10 PRO-NGGC-0201 NGG Procedure Writers Guide 4.1.11 NSIR/DPR-ISG-01 Interim Staff Guidance, Emergency Planning for Nuclear Power Plants 4.1.12 CNS ISFSI Certificates of Compliance 4.1.13 CNS Emergency Plan 4.2 Implementing 4.2.1 RP/0/A/5000/001I Classification of Emergency 4.2.2 NEI 99-01 Rev. 6 to ONS EAL Comparison Matrix 4.2.3 CNS EAL Matrix IRP/0/A/5000/001 ' Rev. 1 Page 13 of 247I

5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS 5.1 Definitions (ref. 4.1.1 except as noted)

Selected terms used in Initiating Condition and Emergency Action Level statements are set in all capital letters (e.g., ALL CAPS). These words are defined terms that have specific meanings as used in this document. The definitions of these terms are provided below.

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

Confinement Boundary The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the CNS ISFSI, Confinement Boundary is defined as the Transportable Storage Canister (TSC) for both NAC-UMS and MAGNASTAR storage systems.

Containment Closure The procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.

As applied to CNS, Containment Closure is established when the requirements of OP/0/A/6100/014 Penetration Control for Modes 5, 6 and NO Mode - Enclosure 4.7 Setting, Maintaining and Securing from Containment Penetration Control are met (ref. 4.1.9).

Emergency Action Level (EAL)

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

Emergency Classification Level (ECL)

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

Unusual Event (UE)

Alert

Site Area Emergency (SAE)

General Emergency (GE)

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

Explosion IRP/0/A/5000/001 'Rev. 1 Page 14 of 247 j

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.

Faulted The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.

Fission Product Barrier Threshold A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier.

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.

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 actions that result 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.

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 CNS 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 CNS. 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

]RPIOIAI50001001 Rev. I Page 15 of 247I

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

lmpede(d)

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

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.

Independent Spent Fuel Storage Installation (ISFSI)

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

Initiating Condition (IC)

An event or condition that aligns with the definition of one of the four emergency classification levels by virtue of the potential or actual effects or consequences.

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

IRPIOIAI5000/O01 Rev. 1I Page 16 of 247

Normal Levels As applied to radiological IC/EALs, the highest reading in the past twenty-four hours excluding the current peak value.

Owner Controlled Area Area outside the PROTECTED AREA fence that immediately surrounds the plant. Access to this area is generally restricted to those entering on official business. (ref. 4.1.13).

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

Protected Area An area encompassed by physical barriers and to which access is controlled. The Protected Area refers to the designated security area around the process buildings and is depicted in CNS UFSAR Figure 1-20 Plot Plan (ref. 4.1.7).

NCS Intact The NCS should be considered intact when the NCS pressure boundary is in its normal condition for the cold shutdown mode of operation (e.g., no freeze seals or nozzle dams).

Refueling Pathway The reactor refueling cavity, spent fuel pool and fuel transfer canal comprise the refueling pathway.

Ruptured The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.

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 and following design basis events to assure:

(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

Security Condition Any security event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A security condition does not involve a hostile action.

Site Area Emergency Events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; 1) toward site personnel or equipment that could lead to the likely failure of or; 2) that prevent effective access to, equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA PAG exposure levels beyond the SITE BOUNDARY.

Site Boundary Area as depicted in CNS-SLC-16.11-16 Figure 16.11-16-1 Unrestricted Area and Site Boundary for Radioactive Effluents (ref. 4.1.6).

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

Unplanned A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

Unusual Event Events are in progress or have occurred which indicate a potential degradation 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.

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

Visible Damage Damage to a 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.

IRP/0/A/5000/001I Rev. 1I Page 18 of 247

5.2 Abbreviations/Acronyms

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

° ............................................................................. Degrees AC ........................................................................... Alternating Current AP ............................................................. Abnormal Operating Procedure ATWS ................................................... Anticipated Transient Without Scram CA .......................................................................... Auxiliary Feedwater CNS................................................................... Catawba Nuclear Station CDE ............................................................... Committed Dose Equivalent CFR .............................................................. Code of Federal Regulations CSFST.................................................... Critical, Safety Function Status Tree DBA...................................................................... Design Basis Accident DC ................................................................................. Direct Current EAL.................................................................... Emergency Action Level ECCS........................................................ Emergency Core Cooling System EC...................................................................... Emergency Coordinator ECL ........................................................... Emergency Classification Level EOF ............................................................ Emergency Operations Facility EOP.......................................................... Emergency Operating Procedure EPA ......................................................... Environmental Protection Agency ERG .......................................................... Emergency Response Guideline EPIP............................................... Emergency Plan Implementing Procedure ESF ................................................................. Engineered Safety Feature FAA............................................................ Federal Aviation Administration FBI............................................................. Federal Bureau of Investigation FEMA.............................................. Federal Emergency Management Agency FSAR ............................................................. Final Safety Analysis Report GE.......................................................................... General Emergency 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 LER ..................................................................... Licensee Event Report LOCA ................................................................. Loss of Coolant Accident LWR ........................................................................ Light Water Reactor MPC ..................................................................... Multi-Purpose Canister MSIV ............................................................... Main Steam Isolation Valve MSL ............................................................................ Main Steam Line mR, mRem, mrem, mREM .................................. milli-Roentgen Equivalent Man IRP/0/A/5000/001 Rev. 1 Page 19 of 247

MW. Megawatt NCS.................................................................. Reactor Coolant System 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........................................................ Off-site Dose Calculation Manual ORO ........................................................... Offsite Response Organization PA ................................................................................ Protected Area PAG.....................Protective Action Guideline PRA/PSA................ Probabilistic Risk Assessment / Probabilistic Safety Assessment PWR................................................................ Pressurized Water Reactor PSIG .......................................................... Pounds per Square Inch Gauge R ........................................................................................ Roentgen Rem, rem, REM .................................................... Roentgen Equivalent Man RPS................................................................. Reactor Protection System RV................................................................................ Reactor Vessel RVLIS.................Reactor Vessel Level Indicating System SAR ..................................................................... Safety Analysis Report SBGTS...................................................... Stand-By Gas Treatment System SB5 0.........................................................

................... Station Blackout SCBA .................................................... Self-Contained Breathing Apparatus SG............................................................................. Steam Generator SI ................................................................................. Safety Injection SLC........................................................... Selected Licensee Commitment SPDS....................................................... Safety Parameter Display System SRO ................................................................... Senior Reactor Operator SSF...................................................................... Safe Shutdown Facility TEDE ........... .............................................. Total Effective Dose Equivalent TOAF.......................................................................... Top of Active Fuel TSC.................................................................. Technical Support Center WOG............................................................. Westinghouse Owners Group SRPIOIAI5000/O01 Rev. 1 Page 20 of 247

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

CNS NEI 99-01 Rev. 6 EAL IC Example EAL ICEAL RUI.2 AU1 3 RU2.1 AU21 RAI.1 AA1 1 RA1.2 AAI 2 RA1.3 AA1 3 RA1.4 AA1 4 RA2.1 AA2 I RA2.2 AA2 2 RA2.3 AA2 3 RA3.1 AA3 1 RA3.2 AA3 2 RSI.1 ASI 1 RS1.2 AS1 2 RS1.3 AS1 3 RS2.1 AS2 1 RGI.1 AG1 1 RG1.2 AGI 2 RG1.3 AGI 3 RG2.1 AG2 I SRP/0/A/5000/001 Rev. 1 Page 21 of 247I

CNS NEI 99-01 Rev. 6 EAL IC Example EAL CU1.1 GUI 1 CU1.2 GUi 2 CU2.1 CU2 1 CU3.1 CU3 I CU3.2 CU3 2 CU4.1 GU4 I CU5.1 CU5 1, 2, 3 CA1.1 CA1 I CA1.2 CA1 2 CA2.1 CA2 I CA3.1 CA3 1,2 CA6.1 CA6 I CS1.1 CSl 3 CGI.1 CGI 2 FAI.1 FA1 I FSl.1 FS1 1 FG1.1 FG1 I HUI.1 HU1 1,2 3 HU2.1 HU2 1 HU3.1 HU3 I HU3.2 HU3 2 HU3.3 HU3 3 HU3.4 HU3 4 HU4.1 HU4 1 HU4.2 HU4 2 IRPIOIA/5O000'O01 IRev. I Page822of 2471

CNS NEI 99-01 Rev. 6 EAL IC Example EAL ICEAL HU4.3 HU4 3 HU4.4 HU4 4 HU7.1 HU7 I HAI.1 HAl 1, 2 HA5.1 HA5 1 HA6.1 HA6 1 HA7.1 HA7 1 HSI.1 HS1 I HS6.1 HS6 1 HS7.1 HS7 1 HG1.1 HGI 1 HG7.1 HG7 1 SU1.1 SU1 1 SU3.1 SU2 I SU4.1 SU3 1 SU4.2 SU3 2 SU5.1 SU4 1,2, 3 SU6.1 SU5 I SU6.2 SU5 2 SU7.1 SU6 1,2, 3 SU8.1 SU7 1,2 SA1.1 SA1 1 SA3.1 SA2 1 SA6.1 SA5 1 SA9.1 SA9 1 IRP/0/A/5000/001 IRev. I Page 23 of 2471

CNS NEI 99-01 Rev. 6 EAL IC Example EAL ICEAL SS1.1 SS1 I SS2.1 SS8 1 SS6.1 SS5 1 SGI.1 SGI 1 SG1.2 SG8 1 EU1.1 E-HU1 I SRP/O/A/5000/O01 Rev. 1I Page 24 of 2471

7.0 ATTACHMENTS 7.1 Attachment 1, Emergency Action Level Technical Bases 7.2 Attachment 2, Fission Product Barrier Matrix and Basis IRP/0/A/5000/001I Rev. 1 Page 25 of 247I

ATTACHMENT 1 EAL Bases Category R - Abnormal Rad Release / Rad Effluent EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)

Many EALs are based on actual or potential degradation of fission product barriers because of the elevated potential for offsite radioactivity release. Degradation of fission product barriers though is not always apparent via non-radiological symptoms. Therefore, direct indication of elevated radiological effluents or area radiation levels are appropriate symptoms for emergency classification.

At lower levels, abnormal radioactivity releases may be indicative of a failure of containment systems or precursors to more significant releases. At higher release rates, offsite radiological conditions may result which require offsite protective actions. Elevated area radiation levels in plant may also be indicative of the failure of containment systems or preclude access to plant vital equipment necessary to ensure plant safety.

Events of this category pertain to the following subcategories:

1. Radiologqical 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.

}RPI0/A/50001001 IRev. 1 Page 26 of 247

ATTACHMENT 1 EAL Bases Category: R -Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity > 2 times the SLCITS limits for 60 minutes or longer EAL:

RUI.1 Unusual Event Reading on any Table R-1 effluent radiation monitor > column "UE" for ->60 min.

(Notes 1,2, 3)

Note 1: The Emergency Coordinator 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 1Monitor [ GE ] SAE Alert UE ]

-* Unit Vent Noble Gas Low 1/2EMF36L ..... 4.18E+6 cprn 5.75E+3 cpm 0

0) o* Unit Vent Noble Gas High 1/2EMF36H 2.21E+4 cpm 2.22E+3 cpm 2.42E+2 cpm ..

-. Liquid Waste Effluent Line 0EMF49L ....... 4.50E+6 cpm

" Monitor Tank Discharge 0EMF57L ....... 4.97E+5 cpm Mode Applicability:

All Definition(s):

None Basis:

The column "UE" gaseous and liquid release values in Table R-1 represent two times the appropriate SLC and Technical Specification release rate and concentration limits associated with the specified monitors (ref. 2, 3, 4, 7).

SRP/0/A/5000/001 Rev. 1 Page 27 of 247I

ATTACHMENT 1 EAL Bases Gaseous Releases Instrumentation that may be used to assess this EAL is listed below (ref. 1, 5):

Unit Vent Noble Gas Low Range - 1/2EMF36L has a range of 101 - i0 7 cpm Liquid Releases Instrumentation that may be used to assess this EAL is listed below (ref. 1, 6):

Liquid Waste Effluent Line Monitor - 0EMF49L (batch release) has a range of 101 - 107 cpm

Monitor Tank Discharge Monitor - 0EMF57L has a range of 101 - I107 cpm 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 contro~ls.

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.

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

CNS Basis Reference(s):

1. CNS ODOM Section 3.0 Setpoint Calculations

2. CNS-SLC 16.11-1 Liquid Effluents

3. CNS-SLC 16.11-6 Gaseous Effluents

4. EP-EALCALC-CNS-1401 CNS Radiological Effluent EAL Values, Rev. 0

5. UFSAR Table I11-20 Airborne Process Radiation Monitoring Equipment

6. UFSAR Table 11-19 Liquid Process Radiation Monitoring Equipment

7. Technical Specifications Section 5.5.5 Radioactive Effluent Controls Program

ATTACHMENT 1 EAL Bases

8. NEI 99-01 AU1 SRP/0/A/5000/00 1Re.IPg29o24 Rev. 1 Page 29 of 247 I

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the SLC/TC limits for 60 minutes or longer.

EAL:

RUI.2 Unusual Event Sample analysis for a gaseous or liquid release indicates a concentration or release rate

> 2 x SLC/TC limits for ->60 min. (Notes 1, 2)

Note 1: The Emergency Coordinator 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:

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.

IRP/0/A/5000/001 ' Rev. 1 Page 30 of 247I

ATTACHMENT 1 EAL Bases CNS Basis Reference(s):

1. CNS Offsite Dose Calculation Manual

2. CNS-SLC 16.11-1 Liquid Effluents

3. CNS-SLC 16.11-6 Gaseous Effluents

4. Technical Specifications Section 5.5.5 Radioactive Effluent Controls Program

5. AD-RP-ALL-2003 Investigation of Unusual Radiological Occurences

6. NEI 99-01 AUI SRP/0/A/5000/001I Rev. I Page 31 of 247I

ATTACHMENT 1 EAL Bases 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 thyroid CDE EAL:

Note 1: The Emergency Coordinator 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 RA1 .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.

Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE ] SAE ] Alert [ UE

Unit Vent Noble Gas Low 1/2EMF36L ...... 4.18E+6 cpm 5.75E+3 cpm 0

o* Unit Vent Noble Gas High 1/2EMF36H 2.21 E+4 cpm 2.22E+3 cpm 2.42E+2 cpm ..

-o Liquid Waste Effluent Line 0EMF49L ......... 4.50E+6 cprn

.m Monitor Tank Discharge 0EMF57L ........ 4.97E+5 cpm Mode Applicability:

All SRP/O/AI5000/O01 Rev. 1 Page 32 of 247

ATTACHMENT 1 EAL Bases Definition(s):

None Basis:

This EAL address gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to site boundary doses that exceed either:

10 mRemTEDE

50 mRem ODE Thyroid The column "ALERT" gaseous effluent release values 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) (ref. 3, 4).

Instrumentation that may be used to assess this EAL is listed below (ref. 1, 2):

Unit Vent Noble Gas High Range - EMF36H has a range of 101 - 106 cpm 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-mnonitored 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.

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.

IRP/0/A/5000/001 Rev. 1I Page 33 of 247

ATTACHMENT I EAL Bases CNS Basis Reference(s):

1. CNS ODCM Section 3.0 Setpoint Calculations

2. UFSAR Table 11-20 Airborne Process Radiation Monitoring Equipment

3. EP-EALCALC-CNS-1401 CNS Radiological Effluent EAL Values, Rev. 0

4. SDQA-70400-COM Unified RASCAL Interface (URI)

5. NEI 99-01 AA1 SRP/0/AI5000/001 Rev. 1I Page 34 of 247

ATTACHMENT 1 EAL Bases 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 thyroid ODE EAL:

RA1.2 Alert Dose assessment using actual meteorology indicates doses > 10 mrem TEDE or 50 mrem thyroid ODE at or beyond the SITE BOUNDARY (Note 4)

Note 4: The pre-calculated effluent monitor values presented in EALs RA1 .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):

.S/TE BOUNDARY-Area as depicted in CNS-SLC-16.11-16 Figure 16.11-16-1 Unrestricted Area and Site Boundary for Radioactive Effluents.

Basis:

Dose assessments are performed by computer-based methods (ref. 1, 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).

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.

Rev. 1I IRP/0/A/5000/001 I I Page 35 of 247

=

ATTACHMENT 1 EAL Bases 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 RS1, CNS Basis Reference(s):

1. HP/0/B/1 009/02 6 On-Shift Offsite Dose Assessment

2. AD-EP-ALL-0202, Emergency Response Offsite Dose Assessment

3. NEI 99-01 AA1 SRP/0/A/5000/001 Rev.I1 Page 36 of 247 10 mrem TEDE or 50 mrem thyroid ODE at or beyond the SITE BOUNDARY for 60 min. of exposure (Notes 1, 2)

Note 1: The Emergency Coordinator 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):
SITE BOUNDARY - Area as depicted in CNS-SLC-1 6.11-16 Figure 16.11-16-1 Unrestricted Area and Site Boundary for Radioactive Effluents.
Basis:
Dose assessments based on liquid releases are performed per Offsite Dose Calculation Manual (ref. 1).
This EAL addresses a release of 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 I1%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 RSI.
CNS Basis Reference(s):
1. ONS Offsite Dose Calculation Manual
2. NEI 99-01 AA1 IRP/0/A/5000/001 Rev. 1 Page 37 of 247I
ATTACHMENT I EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent 8ubcategory: 1 - Radiological Effluent Initiating Conditi'on: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid ODE EAL:
RA1.4 Alert Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:
Closed window dose rates > 10 mR/hr expected to continue for >- 60 min.
Analyses of field survey samples indicate thyroid ODE > 50 mrem for 60 mmi. of inhalation.
(Notes 1, 2)
Note 1: The Emergency Coordinator 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):
.SITE BOUNDARY- Area as depicted in CNS-SLC-16.11-16 Figure 16.11-16-1 Unrestricted Area and Site Boundary for Radioactive Effluents.
Basis:
HP/O/B/1009/004, Environmental Monitoring for Emergency Conditions Within the Ten Mile Radius of CNS provides guidance for emergency or post-accident radiological environmental monitoring (ref. 1).
This IC addresses a release Of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to I1%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 RSI.
IRP/01A150001001 Rev. I Page 38 of 247
ATTACHMENT 1 EAL Bases CNS Basis Reference(s):
I. H P/0/B/1 009/004 Environmental Monitoring for Emergency Conditions Within the Ten Mile Radius of CNS
2. NEI 99-01 AA1 RP/0/A/5000/001 Rev. I Page 39 of 247I
ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent S ubcategory: 1- Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL:
RSI.1 Site Area Emergency Reading on any Table R-1 effluent radiation monitor > column "SAE" for ->15 min.
(Notes 1,2, 3,4)
Note 1: The Emergency Coordinator 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 RAlI.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.
Table R-1 Effluent Monitor Classification Thresholds______
__- Release Point Monitor [ GE SAE Alert J UE SUnit Vent Noble Gas Low 1/2EMF36L ..... 4.18E+6 cpm 5.75E+3 cpm 0
SUnit vent Noble Gas High 1/2EMF36H 2.21 E+4 cpm 2.22E+3 cprn 2.42E+i2 cpm
.'9 Liquid Waste Effluent Line 0EMF49L --- 4.50E+6 cprn C."
"*Monitor Tank Discharge 0EMF57L ----- -- 4.97E+5 cpm Mode Applicability:
All Definition(s):
None RPI/OA/5000/O01 Rev. 1 Page 40 of 247
ATTACHMENT I EAL Bases Basis:
This EAL address gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to site boundary doses that exceed either:
100 mRemTEDE
500 mRem CDE Thyroid The column "SAE" gaseous effluent release value in Table R-1 corresponds to calculated doses of 10% of the EPA Protective Action Guidelines (TEDE or CDE Thyroid) (ref. 1, 2).
Instrumentation that may be used to assess this EAL is Unit Vent Noble Gas High Range Monitor - EMF36H and has a range of 101 - 106 cpm (ref 3, 4).
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.
CNS Basis Reference(s):
1. EP-EALCALC-CNS-1401 ONS Radiological Effluent EAL Values, Rev. 0
2. SDQA-70400-COM Unified RASCAL Interface (URI)
3. CNS ODCM Section 3.0 Setpoint Calculations
4. UFSAR Table 11-20 Airborne Process Radiation Monitoring Equipment
5. NEI 99-01 ASI IRP/OIAI5000/O01 Rev. 1 Page 41 of 247t
ATTACHMENT 1 EAL Bases 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 thyroid CDE EAL:
RSI1.2 Site Area Emergency Dose assessment using actual meteorology indicates doses > 100 mrem TEDE or 500 mrem thyroid CDE at or beyond the SITE BOUNDARY (Note 4)
Note 4: The pre-calculated effluent monitor values presented in EALs RA1.1, 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):
SITE BOUNDARY- Area as depicted in CNS-SLC-16.1 1-16 Figure 16.1 1-16-1 Unrestricted Area and Site Boundary for Radioactive Effluents.
Basis:
Dose assessments are performed by computer-based methods (ref. 1, 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 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.
CNS Basis Reference(s):
1. SDQA-70400-COM Unified RASCAL Interface (URI)
2. AD-EP-ALL-0202, Emergency Response Offsite Dose Assessment
3. NEI 99-01 ASI IRPIOIAI5000IO01 Rev. 1I- Page 42 of 247I
ATTACHMENT 1 EAL Bases IRPIOIA/5000/OO1 Rev. 1 Page 43 of 247I
ATTACHMENT 1 EAL Bases 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 thyroid CDE EAL:
RS1 .3 Site Area Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:
Closed window dose rates > 100 mR/hr expected to continue for > 60 min.
Analyses of field survey samples indicate thyroid CDE > 500 mrem for 60 min. of inhalation.
(Notes 1,2)
Note 1: The Emergency Coordinator 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):
.SITE BOUNDARY- Area as depicted in CNS-SLC-16.11-16 Figure 16.11-16-1 Unrestricted Area and Site Boundary for Radioactive Effluents.
Basis:
HP/0/B/1009/004, Environmental Monitoring for Emergency Conditions Within the Ten Mile Radius of CNS provides guidance for emergency or post-accident radiological environmental monitoring (ref. 1).
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.
'lRP/0/A/5000/001 Rev. 1 Page 44 of 247
ATTACHMENT 1 EAL Bases 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.
Escalation of the emergency classification level would be via IC RG1.
CNS Basis Reference(s):
1. HP/0/B/1009/004 Environmental Monitoring for Emergency Conditions Within the Ten Mile Radius of CNS.
2. NEI 99-01 ASI RPIOIN150001001 {Rev. I Page 45 of 247I
ATTACHMENT 1 EAL Bases 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 thyroid ODE EAL:
RGI.1 General Emergency Reading on any Table R-1 effluent radiation monitor > column "GE" for ->15 min.
(Notes 1,2, 3,4)
Note 1: The Emergency Coordinator 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 RAl1.1, RS1 .1 and RGI1.1 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 Release Point Monitor ] GE SAE Alert UE SUnit Vent Noble Gas Low 1/2EMF36L ........- 4.18E+6 cpm 5.75E+3 cpm SUnit Vent Noble Gas High 1/2EMF36H 2.21 E+4 cpm 2.22E+3 cpm 2.42E+2 cpm ..
,' Liquid Waste Effluent Line 0EMF49L ........... 4.50E+6 cpm "Monitor Tank Discharge 0EMF57L ......... 4.97E+5 cpm Mode Applicability:
All Definition(s):
None Basis:
This EAL address gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to site boundary doses that exceed either:
1000 mRemTEDE
5000 mRem CDE Thyroid IRP/0/A/5000/001 Rev. 1I Page 46 of 247
ATTACHMENT I EAL Bases 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 Thyroid) (ref. 1, 2).
Instrumentation that may be used to assess this EAL is Unit Vent Noble Gas High Range Monitor - EMF36H and has a range of 101 - 106 cpm (ref 3, 4).
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.
CNS Basis Reference(s):
1. EP-EALCALC-CNS-1401 CNS Radiological Effluent EAL Values, Rev. 0
2. SDQA-70400-COM Unified RASCAL Interface (URI)
3. CNS ODCM Section 3.0 Setpoint Calculations
4. UFSAR Table 1 1-20 Airborne Process Radiation Monitoring Equipment
5. NEI 99-01 AGI IRP/O/AI5000/O01 Rev. I Page 47 of 247I
ATTACHMENT I EAL Bases Category: R - Abnormal Rad Levels I Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE EAL:
RGI.2 General Emergency Dose assessment using actual meteorology indicates doses > 1,000 mrem TEDE or 5,000 mrem thyroid ODE at or beyond the SITE BOUNDARY (Note 4)
Note 4: The pre-calculated effluent monitor values presented in EALs RAI .1, RSI .1 and RG1I.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
Mode Applicability:
All Definition(s):
SITE BOUNDARY- Area as depicted in CNS-SLC-16.11-16 Figure 16.11-16-1 Unrestricted Area and Site Boundary for Radioactive Effluents.
Basis:
Dose assessments are performed by computer-based methods (ref. 1, 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 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..
CNS Basis Reference(s):
1. SDQA-70400-COM Unified RASCAL Interface (URI)
2. AD-EP-ALL-0202, Emergency Response Offsite Dose Assessment
3. NEI 99-01 AGI SRP/01A15000/001 Rev. I Page 48 of 247
ATTACHMENT 1 EAL Bases 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 thyroid, ODE EAL:
RG1.3 General Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:
Closed window dose rates > 1,000 mR/hr expected to continue for ->60 min.
Analyses of field survey samples indicate thyroid CDE > 5,000 mrem for 60 min. of inhalation.
(Notes 1,2)
Note 1: The Emergency Coordinator 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):
SITE BOUNDARY Area and - Area as depicted in CNS-SLC-1 6.11-16 Figure 16.11-16-1 Unrestricted Site Boundary for Radioactive Effluents.
Basis:
HPIOIBI1009I004, Environmental Monitoring for Emergency Conditions Within the Ten Mile Radius of CNS provides guidance for emergency or post-accident radiological environmental monitoring (ref. 1).
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.
CNS Basis Reference(s):
1. HP/0/B/1 009/004 Environmental Monitoring for Emergency Conditions Within the Ten Mile Radius of CNS SRP/0/A/5000/001 Rev. 1I Page 49 of 247
ATTACHMENT 1 EAL Bases
2. NEI 99-01 AGI IRP/0/A/5000/001I Rev. I Page 50 of 247
ATTACHMENT 1 EAL Bases 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 low water level alarm or indication AND UNPLANNED rise in corresponding area radiation levels as indicated by any of the following radiation monitors:
IEMFI5 (2EMF4) Spent Fuel Building Refueling Bridge
1 EMF1 7 (2EMF2) Reactor Building Refueling Bridge Mode Applicability:
All Definition(s):
UNPLANNED-. A parameter change or an event that is not I) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
REFUELING PATHWAY-. The reactor refueling cavity, spent fuel pool and fuel transfer canal comprise the refueling pathway.
Basis:
The spent fuel pool low water level alarm setpoint is actuated by I (2)KFPS51 20 at a setpoint of 39' (ref. 1). Water level restoration instructions are performed in accordance with AOPs (ref.
2, 3).
The specified radiation monitors are those expected to see increase area radiation levels as a result of a loss of REFUELING PATHWAY inventory (ref. 2, 3). Increasing radiation indications on these monitors in the absence of indications of decreasing REFUELING CAVITY level are not classifiable under this EAL.
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 reactor vessel and spent fuel pool.
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.
,IRP/O/A15000/O01" Rev. 1 Page 51 of 247
ATTACHMENT 1 EAL Bases 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.
CNS Basis Reference(s):
1. OP/I1(2)/B/6100/OION E/2 Spent Fuel Pool Level Hi/Lo
2. API/1(2)/A/5500/026 Loss of Refueling Canal Level
3. APII1(2)1A/55001041 Loss of Spent Fuel Cooling or Level
4. NEI 99-01 AU2 SRP/0/A/5000/001 Rev. 1 Page 52 of 247
ATTACHMENT 1 EAL Bases 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 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 thle 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.
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.
CNS Basis Reference(s):
1. AP/1 (2)/A/5500/026 Loss of Refueling Canal Level
2. AP/1(2)/A/5500/041 Loss of Spent Fuel Cooling or Level
3. NEI 99-01 AA2 SRP/0/A/5000/001 Rev. 1 Page 53 of 247I
ATTACHMENT I EAL Bases RPIOIA/5000IOOI Rev. 1 Page 54 of 247I
ATTACHMENT 1 EAL Bases 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 A Trip 2 radiation alarm on any of the following radiation monitor indications:
1EMFI5 (2EMF4) Spent Fuel Building Refueling Bridge
1EMF1 7 (2EMF2) Reactor Building Refueling Bridge
1EMF42 (2EMF42) spent Fuel Pool Ventilation
1EMF39L (2EMF39L) Containment Noble Gas Mode Applicability:
All Definition(s):
None Basis:
The specified radiation monitors are those expected to see increase area radiation levels as a result of damage to irradiated fuel (ref. 1).
The Trip 2 alarm setpoints for the radiation monitors are set to be indicative of significant increases in area and/or airborne radiation (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.
This EAL applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with EAL EUlI.1.
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 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).
ATTACHMENT 1 EAL Bases Escalation of the emergency classification level would be via IC RS1.
CNS Basis Reference(s):
1. APII1(2)/A/55001025 Damaged Spent Fuel
2. HP/0/B/1000/010 Determination of Radiation Monitor Setpoints
3. NEI 99-01 AA2 IRP/0/A/5000/001 Rev. 1I Page 56 of 247
ATTACHMENT 1 EAL Bases 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 24.5 ft. (Level 2) on I (2)KFP5780 or 1(2)NVP8790 Mode Applicability:
All Definition(s):
None Basis:
Post-Fukushima order EA-12-051 (ref.1) required the installation of reliable SFP level indication capable of identifying normal level (Level I), SFP level 10 ft. above the top of the fuel racks (Level 2) and SFP level at the top of the fuel racks (Level 3).
SEP level indicators 1(2)KFP5780 (radar) or 1(2)NVP8790 (pressure) located on the back of 2MC7 provide continuous wide range SEP level indication to the top of the spent fuel racks (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.
Escalation of the emergency would be based on either Recognition Category R or C l~s.
Spent fuel pool water level at this value is within the lower end of the level range necessary to prevent significant dose consequences from direct gamma radiation to personnel performing operations in the vicinity of the spent fuel pool. This condition reflects a significant loss of spent fuel pool water inventory and thus it is also a precursor to a loss of the ability to adequately cool the irradiated fuel assembles stored in the pool.
Escalation of the emergency classification level would be via IC RS1.
SRP/0/A/5000/001 Rev. 1 Page 57 of 247I
ATTACHMENT 1 EAL Bases CNS Basis Reference(s):
1. NRC EA-I12-51 Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation
2. EC109413 3 NEI 99-01 AA2 SRP/O/A/500O/001 Rev. I Page 58 of 247
ATTACHMENT 1 EAL Bases 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 14.5 ft. (Level 3) on 1(2)KFP5780 or 1(2)NVP8790 Mode Applicability:
All Definition(s):
None Basis:
Post-Fukushima order EA-1 2-051 (ref.1) required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SEP level 10 ft. above the top of the fuel racks (Level 2) and SEP level at the top of the fuel racks (Level 3).
SEP level indicators 1(2)KFP5780 (radar) or 1 (2)NVP8790 (pressure) located on the back of 2MC7 provide continuous wide range SEP level indication to the top of the spent fuel racks (ref. 2).
This EAL addresses a significant loss of spent fuel pool inventory control and makeup capability leading to IMMINENT fuel damage. This condition entails major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.
It is recognized that this IC would likely not be met until well after another Site Area Emergency IC was met; however, it is included to provide classification diversity.
Escalation of the emergency classification level would be via IC AGI or RG2.
CNS Basis Reference(s):
1. NRC EA-12-51 Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation
2. EC109413
3. NEI 99-01 AS2 IRP/0/A/5000/001I Rev. 1 Page 59 of 247
ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels I Rad Effluent Subcategory: 2 - Irradiated Fuel Event initiating Condition: Spent fuel pool level cannot be restored to at least the top of the fuel racks for 60 minutes or longer EAL:
RG2.1 General Emergency Spent fuel pool level cannot be restored to at least 14.5 ft. (Level 3) on 1(2)KFP5780 or I (2)NVP8790 for > 60 min. (Note 1)
Note 1: The Emergency Coordinator 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:
Post-Fukushima order EA-12-051 (ref.1) required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SEP level l0 ft, above the top of the fuel racks (Level 2) and SEP level at the top of the fuel racks (Level 3).
SEP level indicators 1(2)KFP5780 (radar) or 1(2)NVP8790 (pressure) located on the back of 2MC7 provide continuous wide range SEP level indication to the top of the spent fuel racks (ref. 2).
This EAL addresses a significant loss of spent fuel pool inventory control and makeup capability leading to a prolonged uncovery of spent fuel. This condition will lead to fuel damage and a radiological release to the environment.
It is recognized that this IC would likely not be met until well after another General Emergency IC was met; however, it is included to provide classification diversity.
CNS Basis Reference(s):
1. NRC EA-12-51 Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation
2. EC109413
3. NEI 99-01 AG2 Category: R - Abnormal Rad Levels I 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:
IRPIOIAI50001001 Rev. I Page 60 of 247
ATTACHMENT 1 EAL Bases RA3.1 Alert Dose rates > 15 mR/hr in EITHER of the following areas:
Control Room (EMF12)
OR Central Alarm Station (by survey)
Mode Applicability:
All Definition(s):
IMPEDE(D) - Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).
Basis:
Areas that meet this threshold include the Control Room and the Central Alarm Station (CAS).
EMF Channel 12 monitors the Control room for area radiation (ref. 1). The GAS is included in this EAL because of its' importance to permitting access to areas required to assure safe plant operations.
There is no permanently installed CAS area radiation monitors that may be used to assess this EAL threshold. Therefore this threshold must be assessed via local radiation survey for the GAS.
This IC addresses elevated radia'tion levels in certain plant rooms/areas sufficient to preclude or impede personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Coordinator 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.
CNS Basis Reference(s):
1. OP/1(2)/B/6100/010Z C/2 Control Room
2. NEI 99-01 AA3 IRP/0/A/5000/001 Rev. 1 Page 61 of 247I
ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels I 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:
RA3.2 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 Rooms/Areas Bldg. Elevation Unit I RoomlArea Unit 2 RoomlArea Mode Rm 478 (1 EMXA) Rm 469 (2EMXA) 4 Rrn 496 (1 ETA) Rm 486 (2ETA) 4 Auiir 7' Rm 496 (1EMXS) Rm 486 (2EMXS) 4 AB-577', J J-57 (1 MXK) AB-577', J J-57 (2MXK) 4 Rm 330 (1EMXJ) Rm 320 (2EMXJ) 4 Auxiliary 560' Rm 372 (1 ETB) Rm 362 (2ETB) 4 Rrm 372 (1 EMXD) Rrn 362 (2EMXD) 4 Mode Applicability:
All Definition(s):
IMPEDE(D) - Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).
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 rooms or areas with entry-related mode applicability identified specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an IRP/O/AI5000/O01 Rev. I Page 62 of 247
ATTACHMENT 1 EAL Bases action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations) are not included. In addition, the list specifies the plant mode(s) during which entry would be required for each room or area (ref. 1).
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 Coordinator 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 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.
Escalation of the emergency classification level would be via Recognition Category R, C or F ICs.
CNS Basis Reference(s):
1. Attachment 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases
2. NEI 99-01 AA3 IRP/O/AI5000/O01 Rev. 1 Page 63 of 247I
ATTACHMENT 1 EAL Bases Category C - Cold Shutdown I Refueling System Malfunction EAL Group: Cold Conditions (NCS 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 NCS integrity, containment closure, and fuel clad integrity for the applicable operating modes (5 - Cold Shutdown, 6 - Refueling, D - Defueled).
The events of this category pertain to the following subcategories:
1. NCS 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 4160 VAC emergency buses.
3. NCS Temperature Uncontrolled or inadvertent temperature or pressure increases are indicative of a potential loss of safety functions.
4. Loss of Vital DC Power Loss of emergency plant electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of power to or degraded voltage on the 125 VDC vital buses.
SRP/01A10001001 Rev. I Page 64 of247
ATTACHMENT 1 EAL Bases
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 Affectina Safety Systems Certain hazardous natural and technological events may result in visible damage to or degraded performance of safety systems warranting classification.
SRP/O/AI5000/O01 Rev. 1 Page 65 of 247I
ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1- NCS Level Initiating Condition: UNPLANNED loss of NCS inventory for 15 minutes or longer EAL:
CUl.1 Unusual Event UNPLANNED loss of NCS inventory results in NCS water level less than a required lower limit for >- 15 min. (Note 1)
Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
5 - Cold Shutdown, 6 - 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:
NCS water level less than a required lower limit is meant to be less than the lower end of the level control band being procedurally maintained for the current condition or evolution.
With the plant in Cold Shutdown, NCS water level is normally maintained above the pressurizer low level setpoint of 17% (ref. 1). However, if NCS level is being controlled below the pressurizer low level 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 NCS that is the concern.
With the plant in Refueling mode, NCS water level is normally maintained at or above the reactor vessel flange (Technical Specification LCO 3.9.6 requires at least 23 ft of water above the top of the reactor vessel flange in the refueling cavity during refueling operations) (ref. 2).
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 NCS 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 NCS 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 recognizes that the minimum required NCS 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, IRP/0/A/5000/001 Rev. 1 Page 66 of 247
ATTACHMENT 1 EAL Bases 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 NCS inventory may result in escalation to the Alert emergency classification level via either IC CA1 or CA3.
CNS Basis Reference(s):
1. EP/I (2)/A/5000/FR-1.2 Response to Low Pressurizer Level
2. CNS Technical Specifications Section 3.9.6 Refueling Cavity Water Level
3. NEI 99-01 CUI SRP/0/A/5000/001Re.1Pg67o24 Rev. 1 Page 67 of 247 I
ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1- NCS Level Initiating Condition: UNPLANNED loss of NCS inventory EAL:
CU1.2 Unusual Event NOS water level cannot be monitored AND EITHER
UNPLANNED increase in any Table C-6 sump or tank level due to a loss of NCS inventory
Visual observation of UNISOLABLE NCS leakage
[ Table C-6 Sumps/Tanks
Containment Floor & Equipment Sump
Incore Sump (alarm)
ND/NS sump
NCDT
PRT Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.
UNPLANNED-. A parameter change or an event that is not I) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
Basis:
In Cold Shutdown mode, the NCS will normally be intact and standard NCS level monitoring means are available. NCS level in the Refueling mode is normally monitored using the sight glass.
In this EAL, all water level indication is unavailable and the NCS inventory loss must be detected by indirect leakage indications (ref. 1). Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of NCS leakage. If the make-up rate to the NCS unexplainably rises above the pre-established rate, a loss of NCS inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the NCS that cannot be isolated could also be indicative of a loss of NCS inventory.
The Incore Sump level cannot be monitored in the CR but alarms on high level.
IRP/0/A/5000/001 Rev. I Page 68 of 247I
ATTACHMENT I EAL Bases 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 NCS 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 NCS.
Continued loss of NCS inventory may result in escalation to the Alert emergency classification level via either IC CA1 or CA3.
CNS Basis Reference(s):
1. APII (2)/A/5500/01 0 Reactor Coolant Leak
2. NEI 99-01 CU1 SRP/0/A/5000t001 Rev. 1I Page 69 of 247I
ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: I - NCS Level Initiating Condition: Loss of NCS inventory EAL:
CA1.1 Alert UNPLANNED loss of NCS inventory as indicated by NCS water level < 6.5% (wide range)
Mode Applicability:
5 - Cold Shutdown, 6 - 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:
6.5% wide range NCS level indication is the lowest level to assure adequate net positive suction head and prevent ND pump cavitation for all flow rates (ref. 1).
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 NCS water level below 6.5% indicates that operator actions have not been successful in restoring and maintaining NCS 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 NCS inventory and not the potential concurrent effects on systems needed for decay heat removal (e.g., loss of a Decay Heat Removal suction point). An increase in NCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3.
If NCS water level continues to lower, then escalation to Site Area Emergency would be via IC CS1.
IRP/O/A/5000/O01 Rev. I Page 70 of 247
ATTACHMENT 1 EAL Bases CNS Basis Reference(s):
1. OP/I1(2)/A/61 50/006 Draining the Reactor Coolant System
2. NEI 99-01 CA1 IRP/0/A/5000/001I Rev. I Page 71 of 247I
ATTACHMENT I EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - NCS Level Initiating Condition: Loss of NCS inventory EAL:
CA1.2 Alert NCS water level cannot be monitored for ->15 min. (Note 1)
AND EITHER
UNPLANNED increase in any Table C-6 sump or tank level due to a loss of NCS inventory
Visual observation of UNISOLABLE NOS leakage Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Table C-6 Sumps/Tanks
Containment Floor & Equipment Sump
lncore Sump (alarm)
ND/NS sump
NCDT
PRT Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.
UNPLANNED-. A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
Basis:
In Cold Shutdown mode, the NCS will normally be intact and standard NCS monitoring means are available. In the Refuel mode, the NCS is not intact and NCS level may be monitored by different means, including the ability to monitor level visually.
In this EAL, all NCS water level indication would be unavailable for greater than 15 minutes, and the NCS inventory loss must be detected by indirect leakage indications (ref. I). Sump level increases must be evaluated against other potential sources of leakage. If the make-up rate to the NCS unexplainably rises above the pre-established rate, a loss of NCS inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the NCS that cannot be isolated could also be indicative of a loss of NCS inventory.
IRP/0/A/5000/001 Rev. 1I Page 72 of 247I
ATTACHMENT 1 EAL Bases The Incore Sump level cannot be monitored in the CR but alarms on high level.
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 NOS 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 NCS.
The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC OS1.
If the NCS inventory level continues to lower, then escalation to Site Area Emergency would be via IC CS1.
CNS Basis Reference(s):
1. AP/1 (2)/A/5500/01 0 Reactor Coolant Leak
2. NEI 99-01 CA1 SRP/0/A/5000/001 Rev: 1 Page 73 of 247
ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown I Refueling System Malfunction Subcategory: 1 - NCS Level Initiating Condition: Loss of NCS inventory affecting core decay heat removal capability EAL:
C81.1 Site Area Emergency NCS water level cannot be monitored for Ž- 30 mmn. (Note 1)
AND Core uncovery is indicated by any of the following:
UNPLANNED increase in any Table 0-6 sump or tank level due to a loss of NCS inventory
.Visual observation of UNISOLABLE NCS leakage
Reactor Building Refueling Bridge Monitor I EMFI17 (2EMF2) reading > 9,000 mR/hr
Erratic Source Range or Gamma Metric Monitor indication Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Table C-6 SumpslTanks
Containment Floor & Equipment Sump
Incore Sump (alarm)
NDINS sump
NCDT
PRT Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.
UNPLANNED-. A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
SRP/OIA/5000/O01 Rev. I Page 74 of 247I
ATTACHMENT 1 EAL Bases Basis:
The lowest measurable NCS level is the elevation of the NCS hot leg mid-loop. Therefore, NCS inventory loss relative to the NCS level elevation corresponding to the top of active fuel must be detected by indirect leakage indications (ref. 1 ). Sump level increases must be evaluated against other potential sources of leakage. If the make-up rate to the NCS unexplainably rises above the pre-established rate, a loss of NCS inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the NCS in areas outside the containment that cannot be isolated could also be indicative of a loss of NCS inventory (ref. 2).
The Incore Sump level cannot be monitored in the CR but alarms on high level.
In the Refueling Mode, as water level in the reactor vessel lowers, the dose rate above the core will increase. The dose rate due to this core shine should result in indications on installed area radiation monitors. 1EMF1 7 (2EMF2), Reactor Building Refueling Bridge Monitor is located in the containment in proximity to the reactor cavity and is designed to provide monitoring of radiation due to a fuel handling event or loss of shielding during refueling operations. If this radiation monitor reaches and exceeds 9,000 mR/hr (90% of instrument scale), a loss of inventory with potential to uncover the core is likely to have occurred.
Post-TMI accident studies indicated that the installed PWR nuclear instrumentation will operate erratically when the core is uncovered and that this should be used as a tool for making such determinations.
This IC addresses a significant and prolonged loss of reactor vessel/NCS inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a NCS 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 NCS 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 NCS 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 NCS.
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-1 449, 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.
RP/OIAI5000/001,IRev. 1 Page 75 of 247
ATTACHMENT 1 EAL Bases Escalation of the emergency classification level would be via IC CG1 or RGI CNS Basis Reference(s):
1. OP/I (2)1A161 50/006 Draining the Reactor Coolant System
2. AP/1(2)/A/5500/010 Reactor Coolant Leak
3. NEI 99-01 CS1 SRP/0/A/5000/001I Rev. 1I Page 76 of 247I
ATTACHMENT I EAL Bases Category: C - Cold Shutdown I Refueling System Malfunction Subcategory: 1 - NCS Level Initiating Condition: Loss of NCS inventory affecting fuel clad integrity with containment challenged EAL:
CGI.1 General Emergency NOS level cannot be monitored for ->30 mai. (Note I)
AND Core uncovery is indicated by any of the following:
UNPLANNED increase in any Table C-6 sump or tank level due to a loss of NCS inventory
Visual observation of UNISOLABLE NOS leakage
Reactor Building Refueling Bridge Monitor 1EMFI 7 (2EMF2) reading > 9,000 mR/hr
Erratic Source Range or Gamma Metric Monitor indication AND Any Containment Challenge indication, Table C-I Note 1: The Emergency Coordinator 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-estabjished prior to exceeding the 30-minute time limit, declaration of a General Emergency is not' required.
Table C-I Containment Challenge Indications
CONTAINMENT CLOSURE not established (Note 6)
Containment hydrogen concentration > 6%
UNPLANNED rise in containment pressure Table C-6 SumpslTanks
Containment Floor & Equipment Sump
Incore Sump (alarm)
ND/NS sump
NCDT
PRT SRP/OIA/50001001 Rev. I Page 77 of 247
ATTACHMENT 1 EAL Bases Mode Applicability:
5 - Cold Shutdown, 6 - 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 CNS, Containment Closure is established when the requirements of OP/0/A/6100/014 Penetration Control for Modes 5, 6 and NO Mode - Enclosure 4.7 Setting, Maintaining and Securing from Containment Penetration Control are met.
UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.
UNPLANNED-. A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
Basis:
The lowest measurable NCS level is the elevation of the NCS hot leg mid-loop. Therefore, NCS inventory loss relative to the NCS level elevation corresponding to the top of active fuel must be detected by indirect leakage indications (ref. 1). Sump level increases must be evaluated against other potential sources of leakage. If the make-up rate to the NCS unexplainably rises above the pre-established rate, a loss of NCS inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the NCS in areas outside the containment that cannot be isolated Could also be indicative of a loss of NCS inventory (ref. 2).
The Incore Sump level cannot be monitored in the CR but alarms on high level.
In the Refueling Mode, as water level in the reactor vessel lowers, the dose rate above the core will increase. The dose rate due to this core shine should result in indications on installed area radiation monitors. 1EMF1 7 (2EMF2), Reactor Building Refueling Bridge Monitor is located in the containment in proximity to the reactor cavity and is designed to provide monitoring of radiation due to a fuel handling event or loss of shielding during refueling operations. If this radiation monitor reaches and exceeds 9,000 mRlhr (90% of instrument scale), a loss of inventory with potential to uncover the core is likely to have occurred.
Post-TMI accident studies indicated that the installed PWR nuclear instrumentation will operate erratically when the core is uncovered and that this should be used as a tool for making such determinations.
Three conditions are associated with a challenge to containment integrity:
CONTAINMENT CLOSURE is not established (Ref. 3).
In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive mixture of dissolved gases in the containment. However, containment monitoring and/or sampling should be performed to verify this assumption and a General Emergency declared if it is determined that an explosive mixture exists. An explosive mixture can be formed when hydrogen gas IRP/0/A/5000/001 Rev. 1 Page 78 of 247I
ATTACHMENT 1 EAL Bases concentration in the containment atmosphere is greater than 6% by volume in the presence of oxygen (>5%).
Any unplanned increase in containment pressure in the Cold Shutdown or Refueling mode indicates a potential loss of containment closure capability. Unplanned containment pressure increases indicates containment closure cannot be assured and the containment cannot be relied upon as a barrier to fission product release.
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 NCS 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 hydrogeh 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 NCS 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 theNCS.
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-1 449, Shutdown SRP/0/A/5000/001 Rev. 1I Page 79 of 247I
ATTACHMENT 1 EAL Bases and Low-Power Operation at CommercialNuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.
CNS Basis Reference(s):
1. OP/1 (2)/A/61 50/006 Draining the Reactor Coolant System
2. AP/1(2)/A15500/010 Reactor Coolant Leak
3. OP/0/A/6100/014 Penetration Control for Modes 5, 6 and NO Mode. Enclosure 4.7 Setting, Maintaining and Securing from Containment Penetration Control
4. NEI 99-01 CG1 SRPIoIAI50Ioo1/o Rev. I Page 80of 2471
ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction S ubcategory: 2 - Loss of Essential AC Power Initiating Condition: Loss of all but one AC power source to essential buses for 15 minutes or longer EAL:
CU2.1 Unusual Event AC power capability, Table C-2, to essential 41 60V buses 1(2)ETA and 1(2)ETB reduced to a single power source for > 15 mmn. (Note 1)
AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Table C-2 AC Power Sources Offsite:
ATC (Train A)
SATA (Train A)
(if already aligned)
ATD (Train B)
SATB (Train B) (if already aligned)
Onsite:
D/G A(TrainA)
D/G B (Train B)
Mode Applicability:
5 - Cold Shutdown, 6 - Refueling, D - Defueled SRP/0/A/5000/001 Rev. I Page 81 of 247
ATTACHMENT 1 EAL Bases Definition(s):
SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the EGGS. These are typically systems classified as safety-related (as defined in I0CFR50.2):
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 4160 VAC System provides the power requirements for operation and safe shutdown of the plant. The essential switchgear are buses ETA (Train A) and ETB (Train B) (ref. 1).
The essential buses are normally powered from the 6.9KV offsite power system through their respective 6.9KV/4160V Normal Auxiliary Transformers (ATC & ATD). Additionally, a standby source of power to each 41 60V essential bus is provided from the 6.9KV offsite power system via two separate and independent 6.9KV/41 60V transformers (SATA & SATB). These transformers are shared between the two units (ref. 1, 2). However, alignment of SATA or SATB to an essential bus takes longer than 15 minutes and therefore should only be credited if already aligned.
Each essential bus has a dedicated diesel generator (DIG A & DIG B) to supply an onsite emergency source of power to safe shutdown loads in the event of a loss of the normal power source or loss of off-site power. The DIGs will automatically start and tie onto the essential buses if the normal power source or off-site power is lost (ref. 1).
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 EOPs, and capable of supplying required power to an essential bus. Some examples of this condition are presented below.
A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).
A loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being fed from an offsite power source.
RP/0/A15000/001 IRev. 1I Page 82 of 247
ATTACHMENT 1 EAL Bases 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.
CNS Basis Reference(s):
1. UFSAR Section 8.0 Electric Power
2. AP/1(2)/A/5500/007 Loss of Normal Power
3. NEI 99-01 CU2 SRP/0/A/5000/001 Rev. I Page 83 of 247I
ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 2 - Loss of Essential AC Power Initiating Condition: Loss of all offsite and all onsite AC power to essential buses for 15 minutes or longer EAL:
0A2.1 Alert Loss of all offsite and all onsite AC power capability to essential 4160V buses 1(2)ETA and 1I(2)ETB for > 15 min. (Note 1)
Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
5 - Cold Shutdown, 6 - Refueling, D - Defueled Basis:
The 4160 VAC System provides the power requirements for operation and safe shutdown of the plant. The essential switchgear are buses ETA (Train A) and ETB (Train B) (ref. 1).
The essential buses are normally powered from the 6.9KV offsite power system through their respective 6.9KV14160V Normal Auxiliary Transformers (ATC & ATD). Additionally, a standby source of power to each 41 60V essential bus is provided from the 6.9KV offsite power system via two separate and independent 6.9KV/41 60V transformers (SATA & SATB). These transformers are shared between the two units (ref. 1, 2). However, alignment of SATA or SATB to an essential bus takes longer than 15 minutes and therefore should only be credited if already aligned.
Each essential bus has a dedicated diesel generator (DIG A & DIG B) to supply an onsite emergency source of power to safe shutdown loads in the event of a loss of the normal power source or loss of off-site power. The D/Gs will automatically start and tie onto the essential buses if the normal power source or off-site power is lost (ref. 1).
This cold condition EAL is equivalent to the hot condition loss of all offsite AC power EAL SSI.1.
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 essential bus to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition represents an actual or potential substantial degradation of the level of safety of the plant.
Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.
Escalation of the emergency classification level would be via IC CS1 or RSI.
IRPIOIA/50001O01 Rev. 1I Page 84 of 247
ATTACHMENT 1 EAL Bases CNS Basis Reference(s):
1. UFSAR Secti on 8.0 Electric Power
2. AP/1(2)/A/5500/007 Loss of Normal Power
3. ECA-0.0 EP/II(2)I5000IECA-0.0 Loss of All AC Power
4. NEI 99-01 CA2 I RPIOIAI50001001 Rev. 1 Page 85 of 247 I RP0A50001 e.I ae85o 4
ATTACHMENT I EAL Bases Category: C - Cold Shutdown I Refueling System Malfunction Subcategory: 3 - NCS Temperature Initiating Condition: UNPLANNED increase in NCS temperature EAL:
CU3.1 Unusual Event UNPLANNED increase in NCS temperature to > 200°F due to loss of decay heat removal capability Mode Applicability:
5 - Cold Shutdown, 6 - 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:
Several instruments are capable of providing indication of NCS temperature with respect to the Technical Specification cold shutdown temperature limit (200°F, ref. 1) including both hot leg and cold leg RTDs and core exit T/Cs (ref. 2, 3).
In the absence of reliable NCS temperature indication caused by a loss of decay heat removal capability, classification should be based on EAL CU3.2 should NCS level indication be subsequently lost.
This IC addresses an UNPLANNED increase in NCS temperature above the Technical Specification cold shutdown temperature limitand represents a potential degradation of the level of safety of the plant. If the NCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Coordinator should also refer to IC CA3.
A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification.
This EAL involves a loss of decay heat removal capability, or an addition of heat to the NCS 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.
Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.
IRP/0/A/5000/001 Rev. I Page 86 of 247I
ATTACHMENT I EAL Bases Escalation to Alert would be via IC CA1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.
CNS Basis Reference(s):
1. CNS Technical Specifications Table 1.1-1
2. CNS UFSAR Section 7.0 Instrumentation and Controls
3. API1(2)/A/5500/019 Loss of Residual Heat Removal System
4. NEI 99-01 CU3 SRP/0/A/5000/001 Rev. I Page 87 of 247
ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - NGS Temperature Initiating Condition: UNPLANNED increase in NCS temperature EAL:
CU3.2 Unusual Event Loss of all NCS temperature and NCS level indication for > 15 min. (Note 1)
Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
5 Cold Shutdown, 6- Refueling Definition(s):
None Basis:
Several instruments are capable of providing indication of NCS temperature with respect to the Technical Specification cold shutdown temperature limit (200°F, ref. 1) including both hot leg and cold leg RTDs and core exit T/Cs (ref. 2, 3).
NCS water level is normally monitored using various instruments including NC System narrow range and wide range monitors, RVLIS, NC System sightglass, tygon tube and Pressurizer level instruments (ref. 4).
This EAL addresses the inability to determine NCS temperature and level, and represents a potential degradation of the level of safety of the plant. If the NCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Coordinator should also refer to IC CA3.
This EAL reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor NCS 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 CA3 based on exceeding plant configuration-specific time criteria.
CNS Basis Reference(s):
1. CNS Technical Specifications Table 1.1-1
2. CNS UFSAR Section 7.0 Instrumentation and Controls
3. AP/1(2)/N/5500/019 Loss of Residual Heat Removal System SRP/0/AI5000/001 Rev. 1 Page 88 of 247
ATTACHMENT 1 EAL Bases
4. OP/I1(2)/A/6150/006 Draining the Reactor Coolant System
5. NEI 99-01 CU3 IRP/O/A/5000/O01 Rev. 1 Page 89 of 247I
ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - NCS Temperature Initiating Condition: Inability to maintain plant in cold shutdown EAL:
CA3.1 Alert UNPLANNED increase in NCS temperature to > 200°F for > Table C-3 duration (Notes 1,9)
OR UNPLANNED NCS pressure increase > 10 psig due to a loss of NCS cooling (this does not apply during water-solid plant conditions)
Note 1: The Emergency Coordinator should declare the event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
Note 9: In the absence of reliable NCS temperature indication caused by the loss of decay heat removal capability, classification should be based on time to boil data when in Mode 5 and 6.
Table C-3: NCS Heat-up Duration Thresholds NCS StatusContainment Closure Ha-pDrto NCS Status ~Status Ha-pDrto Intact (but not reduced NA6 i.
inventory) N/A___60__min.* _
Not intact established 20 min.*
OR At reduced inventory not established 0 min.
If an NCS heat removal system is in operation within this time frame and NCS temperature is being reduced, the EAL is not applicable.
Mode Applicability:
5 - Cold Shutdown, 6 - 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 CNS, Containment Closure is established when the requirements of OP/0/A/6100/014 Penetration Control for Modes 5, 6 and NO Mode - Enclosure 4.7 Setting, Maintaining and Securing from Containment Penetration Control are met.
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:
ATTACHMENT 1 EAL Bases Several instruments are capable of providing indication of NCS temperature with respect to the Technical Specification cold shutdown temperature limit (200°F, ref. 1) including both hot leg and cold leg RTDs and core exit T/Cs (ref. 2, 3).
A 10 psig RPV pressure increase can be read on various instruments such as NCPT5141 C-Loop NIR, 0 - 800 psig and Point #4 on SMCR581 0 (CR chart recorder, 0 - 600 psi). (ref. 4, 5).
In the absence of reliable NCS temperature indication caused by the loss of decay heat removal capability, classification should be based on the NCS pressure increase criteria when in Mode 5 or based on time to boil data when in Mode 6.
RCS reduced inventory condition exists when NCS level is < 16% (ref. 7).
This IC addresses conditions involving a loss of decay heat removal capability or an addition of heat to the NCS 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 NCS Heat-up Duration Thresholds table addresses an increase in NCS temperature when CONTAINMENT CLOSURE is established but the NCS is not intact, or NCS inventory is reduced (e.g., mid-loop operation). The 20-minute criterion was included to allow time for operator action to address the temperature increase.
The NCS Heat-up Duration Thresholds table also addresses an increase in NCS temperature with the NCS intact. The status of CONTAINMENT CLOSURE is not crucial in this condition since the intact NCS 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 NCS temperature, the NCS is not intact or is at reduced inventory, and CONTAINMENT CLOSURE is not established, no heat-up duration is allowed (i.e., 0 minutes). This is because 1) the evaporated reactor coolant may be released directly into the containment atmosphere and subsequently to the environment, and
2) there is reduced reactor coolant inventory above the top of irradiated fuel.
The NCS pressure increase threshold provides a pressure-based indication of NCS heat-up in the absence of NCS temperature monitoring capability.
Escalation of the emergency classification level would be via IC CSI or RSI.
CNS Basis Reference(s):
1. CNS Technical Specifications Table 1.1-1
2. CNS UFSAR Section 7.0 Instrumentation and Controls
3. AP/1(2)/A/5500/019 Loss of Residual Heat Removal System
4. IP/1(2)/B/3121/011IA
5. IP/1 (2)/A/31 22/055A
6. OPIO/AI6100/014 Penetration Control for Modes 5, 6 and NO Mode. Enclosure 4.7 Setting, Maintaining and Securing from Containment Penetration Control IRP/0/A/5000/001 .I'Rev. I Page 91 of 247
ATTACHMENT I EAL Bases
7. OP/I1(2)/A/6 150/006 Draining the Reactor Coolant System
8. NEI 99-01 CA3
}RPIO/AI50001001 Rev. 1I Page 92 of 247I
ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 4 - Loss of Vital DC Power Initiating Condition: Loss of Vital DC power for 15 minutes or longer EAL:
CU4.1 Unusual Event
< 105 VDC bus voltage indications on Technical Specification required 125 VDC buses for > 15 min. (Note 1)
Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
5 - Cold Shutdown, 6 - Refueling Definition(s):
None Basis:
Four 125 VDC distribution centers are provided for the I25VDC Vital Instrumentation and Control Power System. Four distribution centers (EDA, EDC, EDB and EDD), one per load group, supply the four independent channels of vital instrumentation and control, and are each powered directly from an independent 125 volt battery and battery charger. Each of the four distribution centers supplies one DC panel board and one 125VDC-I20VAC static inverter (ref.
1).
The Class 1E DC loads have an operating voltage range of 105 to 135 volts. The minimum battery discharge voltage (requiring opening the degraded battery output breaker) is 105 VDC (ref. 1,2).
This EAL is the cold condition equivalent of the hot condition loss of DC power EAL SS7.1.
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 Train A is out-of-service (inoperable) for scheduled outage maintenance work and Train B is in-service (operable), then a loss of Vital DC power affecting Train B would require the declaration of an Unusual Event. A loss of Vital DC power to Train A would not warrant an emergency classification.
Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.
SRPIO/AI50001001 Rev. 1 Page 93 of 247
ATTACHMENT I EAL Bases Depending upon the event, escalation of the emergency classification level would be via IC CAl or CA3, or an IC in Recognition Category R.
CNS Basis Reference(s):
1. CNS UFSAR Section 8.0 Electrical Power
2. AP/l (2)/A/5500/029 Loss of Vital or Aux Control Power
3. NEI 99-01 CU4 IRPIOIAI50001001I Rev. I Page 94 of 247
ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 5 - Loss of Communications Initiating Condition: Loss of all onsite or offsite communications capabilities EAL:
CU5.1 Unusual Event Loss of all Table C-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 Public Address X Internal Telephones X Onsite Radios X DEMNET X Commercial Telephones X X Satellite Phones X X Cellular Phones X X NRC Emergency Telephone System (ETS) X X Mode Applicability:
5 - Cold Shutdown, 6 - Refueling, 0 - Defueled Definition(s):
None RP/0/A/5000/001 Rev: 1 Page 95 of 247[
ATTACHMENT 1 EAL Bases Basis:
Onsite/offsite communications include one or more of the systems listed in Table 0-4 (ref. 1).
Public Address System The Catawba Plant public address system provides paging and party line communications between stations located throughout the plant. Inside and outside type wall and desk-mounted stations are used to communicate between roaming personnel and fixed work locations. Plant-wide instructions are issued using the paging feature.
Internal Telephone System The Catawba Site PBX telephone system provides communication capability between telephone stations located within the plant by dialing the four-digit telephone station code.
On-site Radio System Radio systems can be used for communication among operators, off-site monitoring teams, the control room, TSC and EQE.
DEMNET DEMNET is the primary means of offsite communication. This circuit allows intercommunication among the EOF, TSC, control room, counties, and states. DEMNET operates as an internet based (VolP) communications system with a satellite back-up. Should the internet transfer rate become slow or unavailable, the DEMNET will automatically transfer to satellite mode.
Commercial Telephones Commercial telephone lines, which supply public telephone communications, are employed by Duke Energy. The local service provider provides primary and secondary power for their lines at the Central Office.
Satellite Phones Portable satellite telephones are available which enable communication when all other phone systems are inoperable, e.g. following a major external event. These portable systems can be powered by internal batteries, external DC sources as well as external AC sources.
Cellular Phones Cellular phones may be used during emergencies if other communications means are not readily available or are inoperable. These phones are not expected to be used in the Control Room or Power Block due to interference with plant equipment and loss of signal to the phone.
SRP/0/A/5000/001 Rev. I Page 96 of 247
ATTACHMENT 1 EAL Bases NRC Emerqency Telephone System The NRC uses a Duke Energy dedicated telephone line which allows direct telephone.
communications from the plant to NRC regional and national offices. The Duke Energy communications line provides a link independent of the local public telephone network.
Telephones connected to this network are located in the Catawba Control Room, Technical Support Center, and Emergency Operations Facility and can be used to establish NRC Emergency Notification System (ENS) and Health Physics Network (HPN) capability.
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 State, York, Gaston and Mecklenburg County EOCs The third EAL addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.
CNS Basis Reference(s):
1. CNS Emergency Plan Section F Emergency Communications
2. NEI 99-01 CU5 SRP/O/AI5000/O01 Rev. 1 Page 97 of 247
ATTACHMENT 1 EAL Bases 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 C-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:
6 - Cold Shutdown, 6 - 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.
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.
SRPIOIAI50001001 Rev. 1 Page 98 of 247
ATTACHMENT 1 EAL Bases 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 seismic events are discussed under EAL HU2.1 (ref. 1).
Internal FLOODING may be caused by events such as component failures, equipment misalignment, or outage activity mishaps (ref. 2).
External flooding may be due to high lake level. CNS plant yard elevation is 593.5 ft MSL.
The minimum external access elevation for the Auxiliary, Turbine and Service Buildings is 594.0 ft MSL (ref. 1, 3).
Seismic Category I structures are analyzed to withstand a sustained, design wind velocity of at least 95 mph. (ref. 4).
Areas containing functions and systems required for safe shutdown of the plant are identified by fire area in the fire response procedure (ref. 5).
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 RP/o/N/5ooo/o01 Rev. 1I Page 99 of 247
ATTACHMENT 1 EAL Bases 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 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.
CNS Basis Reference(s):
I. RPIO/A15000/O07 Natural Disaster and Earthquake
2. AP/0/A/5500/030 Plant Flooding
3. UFSAR Section 3.4 Water Level (Flood) Design
4. Updated FSAR Section 3.3.1 Wind Loadings
5. AP/01A155001045 Plant Fire
6. NEI 99-01 CA6 SRP/0/A/5000/001 Rev. I Page 100 of 247
ATTACHMENT 1 EAL Bases Category H - Hazards and Other Conditions Affecting Plant Safety EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)
Hazards are non-plant, system-related events that can directly or indirectly affect plant operation, reactor plant safety or personnel safety.
1. Security Unauthorized entry attempts into the Protected Area, bomb threats, sabotage attempts, and actual security compromises threatening loss of physical control of the plant.
2. Seismic Event Natural events such as earthquakes have potential to cause plant structure or equipment damage of sufficient magnitude to threaten personnel or plant safety.
3. Natural or Technoloqiy 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 Events that are indicative of loss of Control Room habitability. If the Control Room must be evacuated, additional support for monitoring and controlling plant functions is necessary through the emergency response facilities.
SRP/O/A/5000/O01 Rev. 1I Page 101 of 247
ATTACHMENT 1 EAL Bases
7. Emeraencv Coordinator Judament 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 Coordinator the latitude to classify emergency conditions consistent with the established classification criteria based upon Emergency Coordinator judgment.
SRP/0/A/5000/001 Rev. I Page 102 of 247
ATTACHMENT 1 EAL Bases 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 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 CNS 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 CNS. 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 Duke Energy Physical Security Plan for CNS (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 HAl, HS1 and HGI.
Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event (ref. 2, 3, 4).
Classification of these events will initiate appropriate threat-related notifications to plant personnel and Offsite Response Organizations.
Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan and Independent Spent Fuel Storage Installation Security Program.
This threshold references the Security Shift Supervison because these are the individuals trained to confirm that a security event is occurring or has occurred. Training on security event IRP/0/N5000/001 IRev. I Page 103 of 2471
ATTACHMENT I EAL Bases confirmation and classification is controlled due to the nature of Safeguards and 10 CFR § 2.39 information.
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 Duke Energy Physical Security Plan for CNS (ref. 1).
Escalation of the emergency classification level would be via IC HAl.
CNS Basis Reference(s):
1. Duke Energy Physical Security Plan for CNS
2. AP/0/A/5500/046 Hostile Aircraft Activity
3. RP/O/B/5000/026 Site Response to a Security Threat
4. AP/0/A/5500/048 Extensive Damage Mitigation
5. NEI 99-01 HU1 IRP/0/A/5000/001 Rev. 1 Page 104 of 247
ATTACHMENT 1 EAL Bases Category: H - Hazards Subcategory: 1 - Security Initiating Condition: Confirmed SECURITY CONDITION or threat EAL:"
HUl.2 Unusual Event Notification of a credible security threat directed at the site 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.
Basis:
This EAL is based on the Duke Energy Physical Security Plan for CNS (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 HAl, HS1 and HGI.
Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event (ref. 2, 3, 4).
Classification of these events will initiate appropriate threat-related notifications to plant personnel and Offsite Response Organizations.
Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan and Independent Spent Fuel Storage Installation Security Program.
This threshold addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with the CNS Security Contingency 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 Duke Energy Physical Security Plan for CNS (ref. 1).
Escalation of the emergency classification level would be via IC HAl.
CNS Basis Reference(s):
1. Duke Energy Physical Security Plan for CNS
2. AP/0/A/5500/046 Hostile Aircraft Activity IRP/0/A/5000/001 Rev. I Page 105 of 247I
ATTACHMENT 1 EAL Bases
3. RP/0/B/5000/026 Site Response to a Security Threat
4. API0/A/5500/048 Extensive Damage Mitigation
5. NEI 99-01 HU1 SRP/0/AI5000/001 Rev. 1 Page 106 of 247
ATTACHMENT 1 EAL Bases Category: H - Hazards Subcategory: 1 - Security Initiating Condition: Confirmed SECURITY CONDITION or threat EAL:
HUI.3 Unusual Event 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.
Basis:
This EAL is based on the Duke Energy Physical Security Plan for CNS (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 HAl, HS1 and HGI.
Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event (ref. 2, 3, 4).
Classification of these events will initiate appropriate threat-related notifications to plant personnel and Offsite Response Organizations.
Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan and Independent Spent Fuel Storage Installation Security Program.
This 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 CNS Security Contingency 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 Duke Energy Physical Security Plan for CNS (ref. 1).
Escalation of the emergency classification level would be via IC HAl.
IRP/0/A15000/001 Rev. 1 Page 107 of 247I
ATTACHMENT 1 EAL Bases CNS Basis Reference(s):
1. Duke Energy Physical Security Plan for CNS
2. AP/0/AI5500/046 Hostile Aircraft Activity
3. RP/0/B/5000/026 Site Response to a Security Threat
4. AP/0/A/5500/048 Extensive Damage Mitigation
5. NEI 99-01 HU1 SRP/0/A/5000/001 Rev. 1I Page 108 of 247
ATTACHMENT 1 EAL Bases Category: H - Hazards Subcategory: 1 - Security Initiating Condition: HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat EAL:
HAl.1 Alert A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the Security Shift Supervision Mode Applicability:
All Definition(s):
HOSTILE ACTION - An act toward CNS 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 c*ivil disobedience or felonious acts that are not part of a concerted attack on CNS. 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 -Area outside the PROTECTED AREA fence that immediately surrounds the plant. Access to this area is generally restricted to those entering on official business.
Basis:
This IC addresses the occurrence of a HOSTILE ACTION within the OWNER CONTROLLED AREA or notification of an aircraft 'attack threat. This event will require rapid response and assistance due to the possibility of the attack progressing to the PROTECTED AREA, or the need to prepare the plant and staff for a potential aircraft impact.
Timely and accurate communications between the Security Shift Supervision and the Control Room is essential for proper classification of a security-related event (ref. 2, 3, 4).
Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan and Independent Spent Fuel Storage Installation Security Program.
As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The 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.
SRPIOIAI50001001 Rev. 1 Page 109 of 247I
ATTACHMENT 1 EAL Bases 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.
This 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.
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 Duke Energy Physical Security Plan for CNS (ref. 1).
CNS Basis Reference(s):
1. Duke Energy Physical Security Plan for CNS
2. AP/0/A/5500/046 Hostile Aircraft Activity
3. RP/0/B/5000/026 Site Response to a Security Threat
4. AP/O/A15500/048 Extensive Damage Mitigation
5. NEI 99-01 HA1 IRPIOIAI5000IO01I Rev. 1 Page 110 of 247
ATTACHMENT 1 EAL Bases Category: H - Hazards Subcategory: 1 - Security Initiating Condition: HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat EAL:
HA1.2 Alert 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 CNS 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 CNS. 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 - Area outside the PROTECTED AREA fence that immediately surrounds the plant. Access to this area is generally restricted to those entering on official business.
Basis:
This IC addresses the occurrence of a HOSTILE ACTION within the OWNER CONTROLLED AREA or notification of an aircraft attack threat. This event will require rapid response and assistance due to the possibility of the attack progressing to the PROTECTED AREA, or the need to prepare the plant and staff for a potential aircraft impact.
Timely and accurate communications between the Security Shift Supervision and the Control Room is essential for proper classification of a security-related event (ref. 2, 3,. 4).
Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan and Independent Spent Fuel Storage Installation Security Program.
As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The 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.
SRP/0/A/5000/001 Rev. I Page 111 of 247
ATTACHMENT 1 EAL Bases 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.
This threshold addresses the threat from the impact of an aircraft on the piant, 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 Duke Energy Physical Security Plan for CNS (ref. 1).
CNS Basis Reference(s):
1. Duke Energy Physical Security Plan for CNS
2. AP/0/A/5500/046 Hostile Aircraft Activity
3. RP/0/B/5000/026 Site Response to a Security Threat
4. AP/0/A/5500/048 Extensive Damage Mitigation
5. NEt 99-01 HAl j RP/0/A/5000/001 Rev. 1 I Page 112 of 247I
ATTACHMENT 1 EAL Bases Category: H - Hazards Subcategory: 1 - Security Initiating Condition: Hostile Action within the Protected Area EAL:
HSl.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 CNS 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 CNS. 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 - An area encompassed by physical barriers and to which access is controlled. The Protected Area refers to the designated security area around the process buildings and is depicted in CNS UFSAR Figure 1-20 Plot Plan.
Basis:
These individuals 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 Duke Energy Physical Security Plan for CNS (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 (ref. 2, 3).
Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan and Independent Spent Fuel Storage Installation Security Program.
As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Site Area Emergency declaration will mobilize Offsite Response Organization (ORO) resources and have them available to develop and implement public protective actions in the unlikely event that the attack is successful in impairing multiple safety functions.
SRP/0/A/5000/001 Rev. I Page 113 of 247
ATTACHMENT 1 EAL Bases This IC does not apply to a HOSTILE ACTION directed at an ISFSI PROTECTED AREA located outside the plant PROTECTED AREA; such an attack should be assessed using IC HA1. It also does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72.
Emergency plans and implementing procedures are public documents; therefore, EALs 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 Duke Energy Physical Security Plan for CNS (ref. 1).
Escalation of the emergency classification level would be via IC HG1.
CNS Basis Reference(s):
1. Duke Energy Physical Security Plan for CNS
2. RP/0/B/5000/026 Site Response to a Security Threat
3. AP/0/A/5500/048 Extensive Damage Mitigation
4. NEI 99-01 HSI SRP/0/A/5000/001 Rev. I Page 114 of 247I
ATTACHMENT 1 EAL Bases 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 Supervisor AND EITHER of the following has occurred:
Any of the following safety functions cannot be controlled or maintained
Reactivity
Core cooling
NCS heat removal OR Damage to spent fuel has occurred or is IMMINENT Mode Applicability:
All Definition(s):
HOSTILE ACTION - An act toward CNS 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 CNS. 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 PROTECTED AREA - An area encompassed by physical barriers and to which access is controlled. The Protected Area refers to the designated security area around the process buildings and is depicted in CNS UFSAR Figure 1-20 Plot Plan.
Basis:
This IC addresses an event in which a HOSTILE FORCE has taken physical control of the facility to the extent that the plant staff can no longer operate equipment necessary to maintain key safety functions. It also addresses a HOSTILE ACTION leading to a loss of physical control that results in actual or IMMINENT damage to spent fuel due to 1) damage to a spent fuel pool cooling system (e.g., pumps, heat exchangers, controls, etc.) or, 2) loss of spent fuel pool integrity such that sufficient water level cannot be maintained.
IRP/O/AI5000/O01 Rev. I Page 115 of 247I
ATTACHMENT 1 EAL Bases Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event (ref. 2, 3).
Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan and Independent Spent Fuel Storage Installation Security Program.
Emergency plans and implementing procedures are public documents; therefore, EALs should 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 Duke Energy Physical Security Plan for CNS (ref. 1).
CNS Basis Reference(s):
1. Duke Energy Physical Security Plan for CNS
2. RP101B150001026 Site Response to a Security Threat
3. AP/0/A/5500/048 Extensive Damage Mitigation
4. AP/1(2)/A/5500/017 Loss of Control Room
5. NEI 99-01 HGI SRP/0/A/5000/001 Rev. 1 Page 116 of 247
ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 2 - Seismic Event Initiating Condition: Seismic event greater than OBE levels EAL:
HU2.1 Unusual Event Seismic event > OBE as indicated by OBE EXCEEDED alarm on IAD-4, B/8 Mode Applicability:
All Definition(s):
None Basis:
Ground motion acceleration of 0.08g horizontal or 0.053g vertical is the Operating Basis Earthquake for CNS (ref. 1).
Five strong motion accelerographs are installed within Unit I structures. The seismic instrumentation system also consists of a network control center (NCC), which is used for rapid interrogation of the accelerograph data and for data transfer to a dedicated system computer for subsequent data processing and analysis. The time-history recorded at each accelerograph location can be analyzed to determine its corresponding peak acceleration values and to verify that site Operating Basis Earthquake (OBE) limits have not been exceeded. Immediate control room alarm indication of an earthquake of 0.08 g horizontal or 0.053 g vertical or greater is annunciated through the system's network control center (NCC), following seismic trigger actuation by at least two accelerographs (ref. 2).
RPIOIAI5000/O07 Natural Disaster and Earthquake provides the guidance for determining if the OBE earthquake threshold is exceeded and any required response actions. (ref. 3)
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 based on receipt of the OBE alarm. 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 CNS. Provide the analyst with the following CNS coordinates: 350 03' 04" north latitude, 810 0)4' 10)" west longitude (ref. 4). Alternatively, near real-time seismic activity can be accessed via the NEIC 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-IRP/0/A/5000/001 Rev. 1I Page 117 of 247
ATTACHMENT I EAL Bases downs and post-event inspections). Given the time necessary to perform walk-downs and inspections, and fully understand any impacts, this event represents a potential degradation of the level of safety of the plant.
Event verification with external sources should not be necessary during or following an OBE.
Earthquakes of this magnitude should be readily felt by on-site personnel and recognized as a seismic event (e.g., lateral accelerations in excess of 0.08g). The Shift Manager or Emergency Coordinator 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 SA9.
CNS Basis Reference(s):
1. Updated ESAR Section 3.1 Conformance with General Design Criteria
2. Updated FSAR Section 3.7.4.2 Location and Description of Instrumentation
3. RP/0/A/5000/007 Natural Disaster and Earthquake
4. Updated FSAR section 2.1.1.1 Specification of Location (Unit 1)
5. NEI 99-01 HU2 IRP/O/A/5000/O01 Rev. 1 Page 118 of 247I
ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technological Hazard Initiating Condition: Hazardous event EAL:
HU3.1 Unusual Event A tornado strike within the PROTECTED AREA Mode Applicability:
All Definition(s):
PROTECTED AREA - An area encompassed by physical barriers and to which access is controlled. The Protected Area refers to the designated security area around the process buildings and is depicted in CNS UFSAR Figure 1-20 Plot Plan.
Basis:
Response actions associated with a tornado onsite is provided in RP/0/A/5000/007 Enclosure 4.2 Tornado Warning Issued for York County or Tornado On-Site (ref. 1).
If damage is confirmed visually or by other in-plant indications, the event may be escalated to an Alert under EAL CA6.1 or SA9.1.
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.
CNS Basis Reference(s):
1. RPI0/AI5000I007 Natural Disaster and Earthquake
2. NEl 99-01 HU3 IRP/0/A/5000/001 Rev. 1 IPage119 of 247I
ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technological Hazard Initiating Condition: Hazardous event EAL:
HU3.2 Unusual Event Internal room or area FLOODING of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component 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 I0CFR50.2):
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:
Areas susceptible to internal flooding are Turbine/Service Buildings and Auxiliary/Diesel Buildings from the following systems: Condenser Circulating Water, Fire Protection, Nuclear and Conventional Service Water and Condensate Storage (ref.1). Refer to EAL CA6.1 for internal flooding affecting one or more SAFETY SYSTEM trains.
This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.
This EAL addresses FLOODING of a building room or area that results in operators isolating power to a SAFETY SYSTEM component due to water level or other wetting concerns.
Classification is also required if the water level or related wetting causes an automatic isolation of a SAFETY SYSTEM component from its power source (e.g., a breaker or relay trip). To warrant classification, operability of the affected component must be required by Technical Specifications for the current operating mode.
Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, S or C.
CNS Basis Reference(s):
SRPI0/A/5000/001I Rev. 1I Page 120 of 247
ATTACHMENT 1 EAL Bases I. AP/0/A/5500/030 Plant Flooding
2. NEI 99-01 HU3 RPIO/NJ50001001 Rev. I Page 121 of 247I
ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technological Hazard Initiating Condition: Hazardous event EAL:
HU3.3 Unusual Event Movement of personnel within the PROTECTED AREA is IMPEDED due to an 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 equipment, such as SCBAs, that is not routinely employed).
PROTECTED AREA - An area encompassed by physical barriers and to which access is
controlled. The Protected Area refers to the designated security area around the process buildings and is depicted in ONS UFSAR Figure 1-20 Plot Plan.
Basis:
As used here, the term "offsite" is meant to be areas external to the CNS 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.
CNS Basis Reference(s):
1. NEI 99-01 HU3 SRPIO/AI5000IO01I Rev. 1 Page 122 of 247
ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technological Hazard Initiating Condition: Hazardous event EAL:
HU3.4 Unusual Event A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles (Note 7)
Note 7: This EAL does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents.
Mode Applicability:
All Definition(s):
None Basis:
This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.
This EAL addresses a hazardous event that causes an on-site impediment to vehicle movement and significant enough to prohibit the plant staff from accessing the site using personal vehicles. Examples of such an event include site FLOODING caused by a hurricane, heavy rains, up-river water releases, dam failure, etc., or an on-site train derailment blocking the access road.
This EAL is not intended apply to routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the flooding around the Cooper Station during the Midwest floods of 1993, or the flooding 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.
CNS Basis Reference(s):
1. NEI 99-01 HU3 SRP/0/A/5000/001 Rev. 1 Page 123 of 247I
ATTACHMENT I EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:
HU4.1 Unusual Event A FIRE is not extinguished within 15 min. of any of the following FIRE detection indications (Note I):
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 Emergency Coordinator should declare the even~t promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Table H-I Fire Areas
Reactor Building (Containment)
Auxiliary Building
Diesel Generator Rooms
RN Pump House
Dog Houses
Standby Shutdown Facility (SSF)
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. The alarm is to be validated using available Control Room indications or alarms to prove that it is not spurious, or by reports from the field.
Table H-I Fire Areas are based on CNS-I1465.00-00-0006 Design Basis Specification for the Plant Fire Protection and AP/O/A/5500/045 Plant Fire. Table H-I Fire Areas include those SRP/0/A/5000/001 Rev. I Page 124 of 247I
ATTACHMENT I EAL Bases structures containing functions and systems required for safe shutdown of the plant (SAFETY SYSTEMS) (ref. 1,2).
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 emergency declaration clock starts at the time that the initial alarm, indication, or report was received, and not the time that a subsequent verification action was performed. Similarly, the fire duration clock also starts at the time of receipt of the initial alarm, indication or report.
CNS Basis Reference(s):
1. CNS-1 465.00-00-0006 Design Basis Specification for the Plant Fire Protection
2. AP/0/A/5500/045 Plant Fire
3. NEI 99-01 HU4 SRP/0/A/5000/001 Rev. 1I Page 125 of 247I
ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 -Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:
HU4..2 Unusual Event Receipt of a single fire alarm (i.e., no other indications of a FIRE)
AND The fire alarm is indicating a FIRE within any Table H-I area AND The existence of a FIRE is not verified within 30 min. of alarm receipt (Note 1)
Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Table H-I Fire Areas
Reactor Building (Containment)
Auxiliary Building
Diesel Generator Rooms
RN Pump House
Dog Houses
Standby Shutdown Facility (SSF)
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 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, classification shall be made based on EAL HU4.I.
Table H-I Fire Areas are based on CNS-1 465.00-00-0006 Design Basis Specification for the Plant Fire Protection and APIOIAI55001045 Plant Fire. Table H-I Fire Areas include those structures containing functions and systems required for safe shutdown of the plant (SAFETY SYSTEMS) (ref. 1,2).
SRP/0/A/5000/00I Rev. 1 Page 126 of 247I
ATTACHMENT 1 EAL Bases 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 Rto 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 this EAL, 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 SA9.
CNS Basis Reference(s):
1. CNS-1 465.00-00-0006 Design Basis Specification for the Plant Fire Protection
2. AP/0/A/5500/045 Plant Fire
3. NEI 99-01 HU4 SRPIOIAI50001001 Rev. I1 Page 127 of 247
ATTACHMENT I EAL Bases SRPIOIA/5000IOO01 Rev. 1 Page 128 of 247I
ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 -Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:
HU4.3 Unusual Event A FIRE within the plant PROTECTED AREA not extinguished within 60 mmn. of the initial report, alarm or indication (Note 1)
Note 1: The Emergency Coordinator 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 - An area encompassed by physical barriers and to which access is controlled. The Protected Area refers to the designated security area around the process buildings and is depicted in CNS UFSAR Figure 1-20 Plot Plan.
Basis:
This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.
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 SA9.
CNS Basis Reference(s):
I. NEI 99-01 HU4 SRP/0/A/5000/001 Rev. 1I Page 129 of 247I
ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:
HU4.4 Unusual Event A FIRE within the 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 - An area encompassed by physical barriers and to which access is controlled. The Protected Area refers to the designated security area around the process buildings and is depicted in CNS UFSAR Figure 1-20 Plot Plan.
Basis:
This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.
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 SAg.
CNS Basis Reference(s):
1. NEI 99-01 HU4 SRP/0/A/5000/001 Rev. 1 Page 130 of 247I
ATTACHMENT 1 EAL Bases 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) no emergency classification is warranted.
Table H-2 Safe Operation & Shutdown Rooms/Areas Bldg. Elevation Unit I RoomlArea Unit 2 RoomlArea Mode Rm 478 (1EMXA) Rm 469 (2EMXA) 4 Rm 496 (1 ETA) Rm 486 (2 ETA) 4 Auiir7'Rrn 496 (1 EMXS) Rm 486 (2EMXS) 4 AB-577', JJ-57 (1 MXK) AB-577', JJ-57 (2MXK) 4 Rm 330.(1EMXJ) Rm 320 (2EMXJ) 4 Auxiliary 560' Rm 372 (1 ETB) Rm 362 (2ETB) 4 Rm 372 I(1 EMXD) Rm 362 (2EMXD) 4 Mode Applicability:
4 - 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 equipment, such as SCBAs, that is not routinely employed).
Basis:
Ifthe equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event.
The list of plant rooms or areas with entry-related mode applicability identified specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations) are not included. In addition, the list specifies the plant
ATTACHMENT 1 EAL Bases mode(s) during which entry would be required for each room or area (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 Coordinator'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 I when the gaseous release occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4.
The gas release is a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., fire suppression system testing).
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 normal level of around 19%, 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.
Escalation of the emergency classification level would be via Recognition Category R, C or F ICs.
IRP/0/A/5000/001 Rev. 1I Page 132 of 247I
ATTACHMENT 1 EAL Bases NOTE: IC HA5 mode applicabilityhas been limited to the applicable modes identified in Table H-2 Safe Operation & Shutdown Rooms/Areas. If due to plant operating procedure or plant configuration changes, the applicable plant modes specified in Table H-2 are changed, a correspondingchange to Attachment 3 'Safe Operation & Shutdown Areas Tables R-2 & 1--2 Bases' and to IC HA5 mode applicabilityis required.
CNS Basis Reference(s):
1. Attachment 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases
2. NEI 99-01 HA5 SRPI0/AI5000I001 Rev. 1I Page 133 of 247
ATTACHMENT 1 EAL Bases Category: H - Hazards and Other' Conditions Affecting Plant Safety Subcategory: 6 - Control Room Evacuation Initiating Condition: Control Room evacuation resulting in transfer of plant control to alternate locations EAL:
HA6.1 Alert An event has resulted in plant control being transferred from the Control Room to the Auxiliary Shutdown Panels or Standby Shutdown Facility Mode Applicability:
All Definition(s):
None Basis:
The Shift Manager (SM) determines if the Control Room is inoperable 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 ERG and emergency response facilities will assist in responding to these challenges.
IRPI/0A 50001001 IRev. I Page 134 of 2471
ATTACHMENT 1 EAL Bases CNS Basis Reference(s):
I. AP/1 (2)/A/5500/01 7 Loss of Control Room
2. OP/I1(2)/A/61 00/004 Shutdown Outside the Control Room From Hot Standby to Cold Shutdown.
3. NEI 99-01 HA6 SRP/0/A/5000/001 Rev. 1I Page 135 of 247I
ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 6 - Control Room Evacuation Initiating Condition: Inability to control a key safety function from outside the Control Room EAL:
HS6.1 Site Area Emergency An event has resulted in plant control being transferred from the Control Room to the Auxiliary Shutdown Panels or Standby Shutdown Facility AND Control of any of the following key safety functions is not reestablished within 15 min.
(Note 1):
Reactivity (Modes 1, 2 and 3 only)
Core Cooling
NCS heat removal Note 1: The Emergency Coordinator should declare the event promptly upon determining that time Jimit has been exceeded, or will likely be exceeded.
Mode Applicability:
1 - Power Operations, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 - Cold Shutdown, 6 -
Refueling Definition(s):
None Basis:
The Shift Manager determines if the Control Room is inoperable 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 Coordinator judgment. The Emergency Coordinator 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 EGI or CGI CNS Basis Reference(s):
1. AP/1 (2)1A/5500101 7 Loss of Control Room SRP/0/AI5000/001 Rev. 1 Page 136 of 247I
ATTACHMENT 1 EAL Bases
2. OP/I1(2)/A/6100/004 Shutdown Outside the Control Room From Hot Standby to Cold Shutdown.
3. NEI 99-01 HS6 IRP/0/A/5000/001 Rev. I Page 137 of 247
ATTACHMENT 1 EAL Bases Category: H - Hazar'ds and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Coordinator Judgment Initiating Condition: Other conditions existing that in the judgment of the Emergency Coordinator warrant declaration of a UE EAL:
HU7M Unusual Event Other conditions exist which in the judgment of the Emergency Coordinator 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 IOCFR5O.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 Coordinator is the designated onsite individual having the responsibility and authority for implementing the CNS Emergency Response Plan. The Operations Shift Manager (SM) initially acts in the capacity of the Emergency Coordinator and takes actions as outlined in the Emergency Plan implementing procedures. If required by the emergency classification or if deemed appropriate by the Emergency Coordinator, 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 Coordinator to fall under the emergency classification level description for an Unusual Event.
CNS Basis Reference(s):
SRP/0/A/5000/001 Rev. 1 Page 138 of 247
ATTACHMENT 1 EAL Bases
1. CNS Emergency Plan section 3.0 Site Emergency Organization Section B.2 Emergency Coordinator
2. NEI 99-01 HU7 RP/0/A/5000/00Re.1 Rev. 1 " ae 3 o 4 Page 139 of 247 I
ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Coordinator Judgment Initiating Condition: Other conditions exist that in the judgment of the Emergency Coordinator warrant declaration of an Alert EAL:
HA7.1 Alert Other conditions exist which, in the judgment of the Emergency Coordinator, 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 CNS 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 CNS. 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 Coordinator is the designated onsite individual having the responsibility and authority for implementing the CNS Emergency Response Plan. The Operations Shift Manager (SM.) initially acts in the capacity of the Emergency Coordinator and takes actions as outlined in the Emergency Plan implementing procedures. If required by the emergency classification or if deemed appropriate by the Emergency Coordinator, 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).
IRP/0/A/5000/001 Rev. 1 Page 140 of 247I
ATTACHMENT 1 EAL Bases This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Coordinator to fall under the emergency classification level description for an Alert.
CNS Basis Reference(s):
1. CNS Emergency Plan section 3.0 Site Emergency Organization Section B.2 Emergency Coordinator
2. NEI 99-01 HA7 IRP/0/A/5000/001 Rev. 1 Page 141 of 247I
ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Coordinator Judgment Initiating Condition: Other conditions existing that in the judgment of the Emergency Coordinator warrant declaration of a Site Area Emergency EAL:
HS7.1 Site Area Emergency Other conditions exist which in the judgment of the Emergency Coordinator indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts, (1) toward site personnel or equipment that could lead to the likely failure of or, (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the SITE BOUNDARY.
Mode Applicability:
All Definition(s):
HOSTILE ACTION - An act toward CNS 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 CNS. 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).
SITE BOUNDARY- Area as depicted in CNS-SLC-16.11-16 Figure 16.11-16-1 Unrestricted Area and Site Boundary for Radioactive Effluents.
Basis:
The Emergency Coordinator is the designated onsite individual having the responsibility and authority for implementing the CNS Emergency Response Plan. The Operations Shift Manager (SM) initially acts in the capacity of the Emergency Coordinator and takes actions as outlined in the Emergency Plan implementing procedures. If required by the emergency classification or if deemed appropriate by the Emergency Coordinator, 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 Coordinator to fall under the emergency classification level description for a Site Area Emergency.
SRP/0/A/5000/001 Rev. 1 Page 142 of 247I
ATTACHMENT 1 EAL Bases CNS Basis Reference(s):
I. CNS Emergency Plan section 3.0 Site Emergency Organization Section B.2 Emergency Coordinator
2. NEI 99-01 HS7 IRP/0/A/5000I001I Rev. 1 Page 143 of 247I
ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Coordinator Judgment initiating Condition: Other conditions exist which in the judgment of the Emergency Coordinator warrant declaration of a General Emergency EAL:
HG7,1 General Emergency Other conditions exist which in the judgment of the Emergency Coordinator 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 CNS 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 CNS. 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.
Basis:
The Emergency Coordinator is the designated onsite individual having the responsibility and authority for implementing the CNS Emergency Response Plan. The Operations Shift Manager(SM) initially acts in the capacity of the Emergency Coordinator and takes actions as outlined in the Emergency Plan implementing procedures. If required by the emergency classification or if deemed appropriate by the Emergency Coordinator, 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).
Releases can reasonably be expected to exceed EPA PAG plume exposure levels outside the Site Boundary.
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 SRP/0/A/5000/001 Rev. I Page 144 of 247I
ATTACHMENT 1 EAL Bases Coordinator to fall under the emergency classification level description for a General Emergency.
CNS Basis Reference(s):
1. CNS Emergency Plan section 3.0 Site Emergency Organization Section B.2 Emergency Coordinator
2. NEI 99-01 HG7 SRP/0/A/5000/001 Rev. 1 Page 145 of 247I
ATTACHMENT 1 EAL Bases Category S - System Malfunction EALGroup: Hot Conditions (NCS temperature > 200°F); EALs in this category are applicable only in one or more hot operating modes.
Numerous system-related equipment failure events that warrant emergency classification have been identified in this category. They may pose actual or potential threats to plant safety.
The events of this category pertain to the following subcategories:
1. Loss of Essential 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 4160 V essential 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 vital plant 125 VDC power sources.
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. NCS 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. NCS Leakaqe The reactor 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 NCS leakage greater than Technical Specification limits indicates potential pipe cracks that may propagate to an extent threatening fuel clad, NCS and containment integrity.
6. RPS Failure This subcategory includes events related to failure of the Reactor Protection System (RPS) to initiate and complete reactor trips. In the plant licensing basis, postulated failures of the RPS to complete a reactor trip comprise a specific set of analyzed events referred to as IIRP/OIA/50001001 Rev. 1 Page 146 of 247 I
ATTACHMENT 1 EAL Bases 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, NCS and containment integrity.
7. Loss of Communications Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification.
8. Containment Isolation Failure Failure of containment isolation capability (under conditions in which the containment is not currently challenged) warrants emergency classification.
9. Hazardous Event Affectincq Safety Systems Various natural and technological events that result in degraded plant safety system performance or significant visible damage warrant emergency classification under this subcategory.
IRP/0/A/5000/001 Rev. I Page 147 of 247
ATTACHMENT 1 EAL Bases 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, Table S-1, to essential 41 60V buses 1(2)ETA and I(2)ETB for > 15 mmn. (Note 1)
Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
[Table S-I AC Power SourcesI Offsite:
ATC (Train A)
SATA (Train A) (if already aligned)
ATD (Train B)
SATB (Train B) (if already aligned)
Onsite:
DIG A(Train A)
DIG B (Train B)
Mode Applicability:
I - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
The 4160 VAC System provides the power requirements for operation and safe shutdown of the plant. The essential switchgear are buses ETA (Train A) and ETB (Train B) (ref. 1).
The essential buses are normally powered from the 6.9KV offsite power system through their respective 6.9KV/41 60V Normal Auxiliary Transformers (ATC & ATD). Additionally, a standby source of power to each 41 60V essential bus is provided from the 6.9KV offsite power system via two separate and independent 6.9KV/41 60V transformers (SATA & SATB). These transformers are shared between the two units (ref. 1, 2). However, alignment of SATA or SATB to an essential bus takes longer than 15 minutes and therefore should only be credited if already aligned.
Each essential bus has a dedicated diesel generator (D/G A & D/G B) to supply an onsite emergency source of power to safe shutdown loads in the event of a loss of the normal power SRPIOIAI5000I001 Rev. I Page 148 of 247 1
ATTACHMENT 1 EAL Bases source or loss of off-site power. The DIGs will automatically start and tie onto the essential buses if the normal power source or off-site power is lost (ref. I).
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 essential 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 essential buses, whether or not the buses are powered from it.
Escalation of the emergency classification level would be via IC SAl.
CNS Basis Reference(s):
1. UFSAR Section 8.0 Electric Power
2. AP/1(2)/A/5500/007 Loss of Normal Power
3. NEI 99-01 SUI SRP/0/A/5000/001 Rev. 1I Page 149 of 247
ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 1 - Loss of Emergency AC Power Initiating Condition: Loss of all but one AC power source to essential buses for 15 minutes or longer EAL:
SA1.1 Alert AC power capability, Table S-I, to essential 41 60V buses I (2)ETA and I (2)ETB reduced to a single power source for ->15 min. (Note 1)
AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Table S-I AC Power Sources Offsite:
ATC (Train A)
SATA (Train A) (if already aligned)
ATD (Train B)
SATB (Train B) (if already aligned)
Onsite:
D/G A(Train A)
DIG B (Train B)
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - 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 EGGS. 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:
RPOA100OO e. ag 5 o 4
ATTACHMENT 1 EAL Bases Basis:
For emergency classification purposes, "capability" means that an AC power source is available to the essential buses, whether or not the buses are powered from it.
The 4160 VAC System provides the power requirements for operation and safe shutdown of the plant..The essential switchgear are buses ETA (Train A) and ETB (Train B) (ref. 1).
The essential buses are normally powered from the 6.9KV offsite power system through their respective 6.9KV/41 60V Normal Auxiliary Transformers (ATC & ATD). Additionally, a' standby source of power to each 41 60V essential bus is provided from the 6.9KV offsite power system via two separate and independent 6.9KV/4160V transformers (SATA & SATB). These transformers are shared between the two units (ref. 1, 2). However, alignment of SATA or SATB to an essential bus takes longer than 15 minutes and therefore should only be credited if already aligned.
Each essential bus has a dedicated diesel generator (DIG A & D/G B) to supply an onsite emergency source of power to safe shutdown loads in the event of a loss of the normal power source or loss of off-site power. The D/Gs will automatically start and tie onto the essential buses if the normal power source or off-site power is lost (ref. 1).
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 IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment. This IC provides an escalation path from IC SUI.
An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.
A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).
A loss of 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 fed from an offsite power source.
Escalation of the emergency classification level would be via IC SSI.
CNS Basis Reference(s):
1. UFSAR Section 8.0 Electric Power
2. AP/1 (2)/A/5500/007 Loss of Normal Power
3. NEI 99-01 SA1 IRP/O/AI5000/O01 Rev. 1 Page 151 of 247
ATTACHMENT 1 EAL Bases Category: S - System Malfuniction Subcategory: I - Loss of Emergency 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 capabilityto essential 41 60V buses I1(2)ETA and 1(2)ETB for --15 mai. (Note 1)
Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown D~efinition(s):
None Basis:
This EAL is indicated by the loss of all offsite and onsite AC power capability (Table S-I) to 41 60V essential buses ETA and ETB. The essential switchgear are buses ETA (Train A) and ETB (Train B) (ref. 1). For emergency classification purposes, "capability" means that an AC power source is available to the essential buses, whether or not the buses are powered from it.
The essential buses are normally powered from the 6.9KV offsite power system through their respective 6.9KV/41 60V Normal Auxiliary Transformers (ATC & ATD)). Additionally, a standby source of power to each 41 60V essential bus is provided from the 6.9KV offsite power system via two separate and independent 6.9KV/4160V transformers (SATA & SATB). These transformers are shared between the two units (ref. 1, 2). However, alignment of SATA or SATB to an essential bus takes longer than 15 minutes and therefore should only be credited if already aligned.
Each essential bus has a dedicated diesel generator (DIG A & DIG B) to supply an onsite emergency source of power to safe shutdown loads in the event of a loss of the normal power source or loss of off-site power. The DIGs will automatically start and tie onto the essential buses if the normal power source or off-site power is lost (ref. 1).
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 capability 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.
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.
IRP/O/A/5000/O01 Rev. 1I Page 152 of 247
ATTACHMENT 1 EAL Bases Escalation of the emergency classification level would be via l~s RG1, FG1 or SG1.
CNS Basis Reference(s):
I. UFSAR Section 8.0 Electric Power
2. AP/1(2)/A15500/007 Loss of Normal Power
3. ECA-0.0 EP/II(2)/5000/ECA-0.O Loss of All AC Power
4. NEI 99-01 SS1
[RPIO/AI5000/O01 Rev. 1 Page 153 of 247
ATTACHMENT 1 EAL Bases 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 capabilityto essential 41 60V buses I1(2)ETA and I (2)ETB AND SSF fails to supply NC pump seal injection OR CA supply to SGs AND EITHER:
Restoration of at least one essential bus in < 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is not likely (Note 1)
Core Cooling RED PATH conditions met Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None SRP/O/AI5000/O01 Rev. 1 Page 154 of 247
ATTACHMENT 1 EAL Bases Basis:
This EAL is indicated by the extended loss of all offsite and onsite AC power capability to 41 60V emergency buses ETA and ETB either for greater then the CNS Station Blackout (SBO) coping analysis time (4 hrs.) (ref. 1) or that has resulted in indications of an actual loss of adequate core cooling.
The SSF is capable of providing the necessary functions (reactor coolant pump seal injection and auxiliary feedwater supply to the steam generators) to maintain a hot shutdown condition for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . No fission product barrier degradation would be expected if the SSF is functioning as intended.
Indication of continuing core cooling degradation is manifested by CSFST Core Cooling RED PATH conditions being met. (ref. 2).
The essential buses are normally powered from the 6.9KV offsite power system through their respective 6.9KV/4160V Normal Auxiliary Transformers (ATC & ATD). Additionally, a standby source of power to each 4160V essential bus is provided from the 6.9KV offsite power system via two separate and independent 6.9KV/41 60V transformers (SATA & SATB). These transformers are shared between the two units (ref. 1, 2).
Each essential bus has a dedicated diesel generator (DIG A & D/G B) to supply an onsite emergency source of power to safe shutdown loads in the event of a loss of the normal power source or loss of off-site power. The D/Gs will automatically start and tie onto the essential buses if the normal power source or off-site power is lost (ref. 3).
Four hours is the station blackout coping time (ref 2).
Indication of continuing core cooling degradation must be based on fission product barrier monitoring with particular emphasis on Emergency Coordinator judgment as it relates to imminent Loss or Potential Loss of fission product barriers and degraded ability to monitor fission product barriers. Indication of continuing core cooling degradation is manifested by CSFST Core Cooling RED PATH conditions being met (ref. 2.
This IC addresses a prolonged loss of all power sources to AC essential buses. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A prolonged loss of these buses will lead to a loss of one or more fission product barriers. In addition, fission product barrier monitoring capabilities may be degraded under these conditions.
The EAL should require declaration of a General Emergency prior to meeting the thresholds for IC 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 essential 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 essential bus should be based on a realistic appraisal of the situation. Mitigation actions with a low probability of success should not be used as a basis SRPIOIAI5000/001 Rev. 1 Page 155 of 247
ATTACHMENT 1 EAL Bases 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.
CNS Basis Reference(s):
1. UFSAR Section 8.4.2 Station Blackout Duration
2. EP/1/A/5000/F-O Critical Safety Function Status Tress - Core Cooling
3. UFSAR Section 8.0 Electric Power
4. AP/1 (2)/A/5500/007 Loss of Normal Power
5. ECA-0.0 EP/1I(2)I5000/ECA-0.0 Loss of All AC Power
6. NEI 99-01 SG1 RP/0/A150001001 Rev. I Page 156 of 247
ATTACHMENT I EAL Bases Category: S -System Malfunction Subcategory: 1 - Loss of Essential AC Power Initiating Condition: Loss of all 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 essential 41 60V buses 1(2)ETA and 1(2)ETB for > 15 min.
AND Loss of all 125 VDC power based on battery bus voltage indications < 105 VDC on all vital DC buses EDA, EDD, EDB and EDO forŽ is1 min.
(Note 1)
Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
I - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown SRP/o/0 0o150o10 Rev., 1 Page 157 of 247
ATTACHMENT 1 EAL Bases Definition(s):
None Basis:
This EAL is indicated by the loss of all offsite and onsite emergency AC power capability to 41 60V emergency buses ETA and ETB for greater than 15 minutes in combination with degraded vital DC power voltage. This EAL addresses operating experience from the March 2011 accident at Fukushima Daiichi.
The essential buses are normally powered from the 6.9KV offsite power system through their respective 6.9KV/4160V Station Auxiliary Transformers (IATC & IATD). Additionally, a standby source of power to each 41 60V essential bus is provided from the 6.9KV offsite power system via two Separate and independent 6.9KV/41 60V transformers (SATA & SATB). These transformers are shared between the two units (ref. 1, 2). However, alignment of SATA or SATB to an essential bus takes longer than 15 minutes and therefore should only be credited if already aligned.
Each essential bus has a dedicated diesel generator (D/G A & D/G B) to supply an onsite emergency source of power to safe shutdown loads in the event of a loss of the normal power source or loss of off-site power. The D/Gs will automatically start and tie onto the essential buses if the normal power source or off-site power is lost (ref. 1).
An Alternate AC power source, the Standby Shutdown Diesel Generator, which provides power to the Standby Shutdown System, is located in the Safe Shutdown Facility (SSF). This AC power source must be started locally from the SSF Control Room. The SSF Diesel Generator has sufficient capability to operate equipment necessary to maintain a safe shutdown condition for the 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months SBO event (ref. 1).
Four 125 VDC distribution centers are provided for the 125VDC Vital Instrumentation and Control Power System. Four distribution centers (EDA, EDD, EDB and EDC), one per load group, supply the four independent channels of vital instrumentation and control, and are each powered directly from an independent 125 volt battery and battery charger. Each of the four distribution centers supplies one DC panel board and one 125VDC-120VAC static inverter (ref.
1,3).
The Class IE DC loads have an operating voltage range of 105 to 135 volts. The minimum battery discharge voltage (requiring opening the degraded battery output breaker) is 105 VDC (ref. 1,3).
This IC addresses a concurrent and prolonged loss of both essential AC and Vital DC power.
A loss of all essential 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 essential 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.
IRP/O/A/5000/O01 Rev. 1 Page 158 of 247I
ATTACHMENT I EAL Bases CNS Basis Reference(s):
1. UFSAR Section 8.0 Electric Power
2. API1(2)/A/5500/007 Loss of Normal Power 3 APII1(2)/A/5500/029 Loss of Vital or Aux Control Power
4. ECA-0.0 EP/I (2)/5000IECA-0.0 Loss of All AC Power
5. NEI 99-01 SG8 0
IRP/°/A/5°0/0°I Rev. I Page 159 of 247
ATTACHMENT I EAL Bases Category: S - System Malfunction Subcategory: 2 - Loss of Vital DC Power Initiating Condition: Loss of all vital DC power for 15 minutes or longer EAL:
SS2.1 Site Area Emergency Loss of all 125 VDC power based on battery bus voltage indications < 105 VDC on all vital DC buses EDA, EDC, EDB, EDD and for ->15 mmn. (Note 1)
Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
I - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
Four 125 VDC distribution centers are provided for the 125VDC Vital Instrumentation and Control Power System. Four distribution centers (EDA, EDC, EDB and EDO), one per load group, supply the four independent channels of vital instrumentation and control, and are each powered directly from an independent 125 volt battery and battery charger. Each of the four distribution centers supplies one DC panel board and one 125VDC-120VAC static inverter (ref.
1,2).
The Class 1E DC loads have an operating voltage range of 105 to 135 volts. The minimum battery discharge voltage (requiring opening the degraded battery output breaker) is 105 VDC (ref. 1,2).
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 SG1.
SRP/0/A/5000/001 Rev. 1 Page 160 of 247{
ATTACHMENT 1 EAL Bases CNS Basis Reference(s):
1. UFSAR Section 8.0 Electric Power 2 API/1(2)/A/5500/029 Loss of Vital or Aux Control Power
3. NEI 99-01 SS8 SRP/0/A/5000/001 Rev. 1I Page 161 of 247I
ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 3 - Loss of Control Room Indications Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer EAL:
SU3.1 Unusual Event An UNPLANNED event results in the inability to monitor one or more Table S-2 parameters from within the Control Room for -- 15 min. (Note 1)
Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
(Table S-2 Safety System Parameters
Reactor power
NCS level
NCS pressure
In-core T/C temperature
Level in at least one S/G
Auxiliary or emergency feed flow in at least one SIG Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - 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 Operator Aid Computer (OAC), which displays SPDS required information, serves as a redundant compensatory indicator 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 SRP/0/AI5000/001 Rev. 1I Page 162 of 247
ATTACHMENT I EAL Bases 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, core cooling and NCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for reactor vessel level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well.
Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.
Escalation of the emergency classification level would be via IC SA3.
CNS Basis Reference(s):
1. UFSAR Section 7.5 Safety-Related Display Instrumentation
2. OP/I1(2)/A/6700/003 Operation With the Operator Aid Computer Out of Service
3. NEI 99-01 SU2 IRP/0/A/5000/001 Rev. 1 Page 163 of 247I
ATTACHMENT 1 EAL Bases Category: S -System Malfunction Subcategory: 3 - Loss of Control Room Indications Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress EAL:
SA3.1 Alert An UNPLANNED event results in the inability to monitor one or more Table S-2 parameters from within the Control Room for > 15 min. (Note 1)
AND Any significant transient is in progress, Table S-3 Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Table S-2 Safety System Parameters
Reactor power
NCS level
NCS pressure
In-core T/C temperature
Level in at least one S/G
Auxiliary or emergency feed flow in at least one SIG Table S-3 Significant Transients
Reactor trip
Runback > 25% thermal power
Electrical load rejection > 25%
electrical load
Safety injection actuation Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - 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:
SRPIOIA/50001001 Rev. 1 Page 164 of 247I
ATTACHMENT I EAL Bases 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 Operator Aid Computer (OAC), which displays SPDS required information, serves as a redundant compensatory indicator 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 reactor trips, runbacks involving greater than 25% thermal power change, electrical load rejections of greater than 25% full electrical load or SI injection actuations.
This IC addresses the difficulty associated with monitoring rapidly changing plant conditions during a transient without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. During this condition, the margin to a potential fission product barrier challenge is reduced. It thus represents a potential substantial degradation in the level of safety of the plant.
As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room.
An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022)to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.
This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling and NCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for reactor vessel level 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 ICs FSI or IC RS1.
CNS Basis Reference(s):
I. UFSAR Section 7.5 Safety-Related Display Instrumentation
2. OP/I1(2)/A/6700/003 Operation with the Operator Aid Computer Out of Service
3. NEI 99-01S5A2 SRP/0/A/5000/001 Rev. I Page 165 of 247
ATTACHMENT 1 EAL Bases RP/0/A/5000/001 *Rev. 1 Page 166 of 247 I
[RIOIA500OOIRe.1 ae16t27
ATTACHMENT 1 EAL Bases Category: S - System Malfunction S ubcateg ory: 4 - NCS Activity Initiating Condition: Reactor coolant activity greater than Technical Specification allowable limits EAL:
SU4.1 Unusual Event NCS activity > Technical Specification 3.4.16 limits or Facility Operating License limits (151/159), whichever is more restrictive Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
Technical Specification Section 3.4.16, as modified in the Facility Operating License, limits NC System Dose Equivalent 1-131 to < 0.46 IpCi/gm or -<26 IpCi/gm Dose Equivalent 1-131 instantaneous. Technical Specification Section 3.4.16 also limits NC System Dose Equivalent Xe-I133 to -<280 IJCi/gm. (ref 1, 2).
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 FAI or the Recognition Category R ICs.
CNS Basis Reference(s):
1. CNS Technical Specifications section 3.4.16 RCS Specific Activity
2. Facility Operating License Attachment B
3. NEI 99-01 SU3 SRP/0/A/5000/001 Rev. 1 Page 167 of 247
ATTACHMENT I EAL Bases Category: S - System Malfunction S ubcategory: 5 - NCS Leakage Initiating Condition: NCS leakage for 15 minutes or longer EAL:
SU5.1 Unusual Event NCS unidentified or pressure boundary leakage > 10 gpm for ->15 min.
OR NCS identified leakage > 25 gpm for ->15 min.
OR Leakage from the NCS to a location outside containment > 25 gpm for ->15 min.
(Note 1)
Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
Identified leakage includes leakage such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank, leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary leakage; or NCS leakage through a steam generator to the secondary system (ref. 1).
Unidentified leakage is all leakage (except RCP seal water injection or leakoff) that is not identified leakage (ref. 1).
Pressure Boundary leakage is leakage (except SG leakage) through an unisolable fault in an NCS component body, pipe wall, or vessel wall (ref. 1)
NOS leakage outside of the containment that is not considered identified or unidentified leakage per Technical Specifications includes leakage via interfacing systems such as NCS to the Component Cooling Water (KC), or systems that directly see NCS pressure outside containment such as Chemical & Volume Control System (NV), Nuclear Sampling system (NM) and Residual Heat Removal (ND) system (when in the shutdown cooling mode) (ref. 2)
Escalation of this EAL to the Alert level is via Category F, Fission Product Barrier Degradation, EAL FAI.1.
This IC addresses NCS leakage which may be a precursor to a more significant event. In this case, NCS leakage has been detected and operators, following applicable procedures, have IRP/0/A/5000/001 Rev. 1 Page 168 of 247I
ATTACHMENT 1 EAL Bases been unable to promptly isolate the leak. This condition is considered to be a potential degradation of the level of safety of the plant.
The first and second EAL conditions are focused on a loss of mass from the NCS 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 NCS mass loss caused by an UNISOLABLE leak through an interfacing system. These conditions thus apply to leakage into the containment, a secondary-side system (e.g., steam generator tube leakage) or a location outside of containment.
The leak rate values for each condition were selected because they are usually observable with normal Control Room indications. Lesser values typically require time-consuming calculations to determine (e.g., a mass balance calculation). The first condition uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage.
The release of mass from the NCS due to the as-designed/expected operation of a relief valve does not warrant an emergency classification. An emergency classification would be required if a mass loss is caused by a relief valve that is not functioning as designed/expected (e.g., a relief valve sticks open and the line flow cannot be isolated).
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 l~s of Recognition Category R or F.
CNS Basis Reference(s):
1. CNS Technical Specifications Definitions section 1.1
2. UFSAR Section 5.2.5.2.1 Intersystem Leakage
3. NEI 99-01 SU4 SRP/O/AI5000/O01 Rev. 1 Page 169 of 247
ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 6 - RPS Failure Initiating Condition: Automatic or manual trip fails to shut down the reactor EAL:
SU6.1 Unusual Event An automatic trip did not shut down the reactor as indicated by reactor power > 5% after any RPS setpoint is exceeded AND A subsequent automatic trip or manual trip action taken at the reactor control console (manual reactor trip switches or turbine manual trip) is success in shutting down the reactor as indicated by reactor power < 5% (Note 8)
Note 8: A manual trip action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.
Mode Applicability:
1 - Power Operation 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) trip function. A reactor trip is automatically initiated by the RPS when certain continuously monitored parameters exceed predetermined setpoints (ref. 1).
Following a successful reactor trip, 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 startup rate. 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-trip response from an automatic reactor trip 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 trip has therefore occurred when there is sufficient rod insertion from the trip of RPS to bring the reactor power below the immediate shutdown decay heat level of 5% (ref. 2, 3, 4).
For the purposes of emergency classification, successful manual trip actions are those which can be quickly performed from the reactor control console (i.e., manual trip switches or turbine trip). Reactor shutdown achieved by use of other trip actions specified in EP/II(2)/A/5000/FR-S.1 Response to Nuclear Power Generation/ATWS (such as depressing manual pushbutton on turbine control panel, emergency boration or manually driving control rods) do not constitute a successful manual trip (ref. 4).
SRP/0/A/5000/001 Rev. 1 Page 170 of 247
ATTACHMENT 1 EAL Bases Following any automatic RPS trip signal, EP/I(2)/A/5000/E-0 (ref. 2) and EP/I(2)IAI5000IFR-S.1 (ref. 3) prescribe insertion of redundant manual trip signals to back up the automatic RPS trip function and ensure reactor shutdown is achieved. Even if the first subsequent manual trip signal inserts all control rods to the full-in position immediately after the initial failure of the automatic trip, the lowest level of classification that must be declared is an Unusual Event (ref.
4).
In the event that the operator identifies a reactor trip is imminent and initiates a successful manual reactor trip before the automatic RPS trip setpoint is reached, no declaration is required. The successful manual trip of the reactor before it reaches its automatic trip setpoint or reactor trip signals caused by instrumentation channel failures do not lead to a potential fission product barrier loss. However, if subsequent manual reactor trip 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 trip following receipt of an automatic trip signal and there are no clear indications that the automatic trip failed (such as a time delay following indications that a trip setpoint was exceeded), it may be difficult to determine if the reactor was shut down because of automatic trip or manual actions.
If a subsequent review of the trip actuation indications reveals that the automatic trip did not cause the reactor to be shut down, then consideration should be given to evaluating the fuel for potential damage, and the reporting requirements of 50.72 should be considered for the transient event.
This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic trip is successful in shutting down the reactor.
This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.
Following the failure on an automatic reactor trip , operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip ). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.
If an initial manual reactor trip is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip ) using a different switch). Depending upon several factors, the initial or subsequent effort to manually trip the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor trip signal. If a subsequent manual or automatic trip 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 trip
). 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".
SRP/0/A/5000/001 Rev. 1 Page 171 of 247
ATTACHMENT 1 EAL Bases The plant response to the failure of an automatic or manual reactor trip 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 FAl, an Unusual Event declaration is appropriate for this event.
A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
Should a reactor trip 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 trip and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and should be evaluated.
If the signal does not cause a plant transient and the trip 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.
CNS Basis Reference(s):
1. CNS Technical Specifications section 3.3.1 Reactor Trip System (RTS) Instrumentation
2. EP/I (2)/A/5000/E-0 Reactor Trip or Safety Injection
3. EP/I(2)/A/5000/F-0 Critical Safety Function Status Trees - Subcriticality
4. EP/I (2)/AI5000IFR-S.1 Response to Nuclear Power Generation/ATWS
5. NEI 99-01 SU5 RPIOIAI5000IO01 Rev. I Page 172 of 247
ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 6 - RPS Failure Initiating Condition: Automatic or manual trip fails to shut down the reactor EAL:
SU6.2 Unusual Event A manual trip did not shut down the reactor as indicated by reactor power > 5% after any manual trip action was initiated AND A subsequent automatic trip or manual trip action taken at the reactor control console (manual reactor trip switches or turbine manual trip) is success in shutting down the reactor as indicated by reactor power < 5% (Note 8)
Note 8: A manual trip 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:
I - Power Operation Definition(s):
None Basis:
This EAL addresses a failure of a manually initiated trip in the absence of having exceeded an automatic RPS trip setpoint and a subsequent automatic or manual trip is successful in shutting down the reactor (reactor power < 5%). (ref. 1).
Following a successful reactor trip, 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 startup rate. 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-trip response from a manual reactor trip 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 trip has therefore occurred when there is sufficient rod insertion from the trip of RPS to bring the reactor power below the immediate shutdown decay heat level of 5% (ref. 2, 3 4).
For the purposes of emergency classification, successful manual trip actions are those which can be quickly performed from the reactor control console (i.e., manual trip switches or turbine trip). Reactor shutdown achieved by use of other trip actions specified in EP/II(2)/A/5000/FR-3.1 Response to Nuclear Power Generation/ATWS (such as depressing manual pushbutton on turbine control panel, emergency boration or manually driving control rods) do not constitute a successful manual trip (ref. 4).
IRP/01A150001001 Rev. I Page 173 of247
ATTACHMENT 1 EAL Bases If both subsequent automatic and subsequent manual reactor trip 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 trip, 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 trip that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic trip is successful in shutting down the reactor.
This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.
Following the failure on an automatic reactor trip , operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip ). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.
If an initial manual reactor trip is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip ) using a different switch). Depending upon several factors, the initial or subsequent effort to manually the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor trip signal. If a subsequent manual or automatic trip 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 trip). 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".
The plant response to the failure of an automatic or manual reactor trip 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. Ifsubsequent 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 FAI, an Unusual Event declaration is appropriate for this event.
A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
Should a reactor trip 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 trip and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and should be evaluated.
If the signal does not cause a plant transient and the trip failure is determined through other means (e.g., assessment of test results), then this IC and the EALs are not SRPIOIAI50001001 Rev. 1 Page 174 of 247
ATTACHMENT 1 EAL Bases applicable and no classification is warranted.
CNS Basis Reference(s):
1. CNS Technical Specifications section 3.3.1 Reactor Trip System (RTS) Instrumentation
2. EP/I (2)/A/5000/E-O Reactor Trip or Safety Injection
3. EP/II(2)/A/5000/F-0 Critical Safety Function Status Trees - Subcriticality
4. EP/II(2)/A/5000IFR-S.1 Response to Nuclear Power Generation/ATWS
5. NEl 99-01 SU5 SRP/0/A/5000/001 Rev. I Page 175 of 247
ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 2 -RPS Failure Initiating Condition: Automatic or manual trip 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 trip fails to shut down the reactor as indicated by reactor power
_>5%
AND Manual trip actions taken at the reactor control console (manual reactor trip switches or turbine manual trip) are not successful in shutting down the reactor as indicated by reactor power >_5% (Note 8)
Note 8: A manual trip action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.
Mode Applicability:
1 - Power Operation Definition(s):
None Basis:
This EAL addresses any automatic or manual reactor trip signal that fails to shut down the reactor followed by a subsequent manual trip 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 trip actions are those which can be quickly performed from the reactor control console (i.e., manual trip switches or turbine trip). Reactor shutdown achieved by use of other trip actions specified in EP/II(2)IAI5000IFR-S.1 Response to Nuclear Power Generation/ATWS (such as depressing manual pushbutton on turbine control panel, emergency boration or manually driving control rods) do not constitute a successful manual trip (ref. 4).
5% rated power is a minimum reading on the power range scale that indicates continued 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 5%, plant response will be similar to that observed during a normal shutdown. Nuclear instrumentation can be used to determine if reactor power is greater than 5 % power (ref. 1).
Escalation of this event to a Site Area Emergency would be under EAL SS6.1 or Emergency Coordinator judgment.
}RPI/OA/50001001 Rev. I *Page 176 of 247I
ATTACHMENT 1 EAL Bases This IC addresses a failure of the RTS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and subsequent operator manual actions taken at the reactor control consoles to shutdown the reactor are also unsuccessful. This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if the reactor is subsequently shutdown by an action taken away from the reactor control consoles since this event entails a significant failure of the RTS.
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 trip
). 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".
The plant response to the failure of an automatic or manual reactor trip 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 the core cooling or NCS 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.
A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
CNS Basis Reference(s):
1. CNS Technical Specifications section 3.3.1 Reactor Trip System (RTS) Instrumentation
2. EP/II(2)/A/5000IE-0 Reactor Trip or Safety Injection
3. EP/1(2)/A/5000/F-0 Critical Safety Function Status Trees - Subcriticality
4. EP/II(2)/A/5000IFR-S.1 Response to Nuclear Power Generation/ATWS
5. NEI 99-01 SA5 SRPIO/AI5000/O01 Rev. 1I Page 177 of 247I
ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 2 - RPS Failure Initiating Condition: Inability to shut down the reactor causing a challenge to core cooling or NCS heat removal EAL:
SS6.1 Site Area Emergency An automatic or manual trip fails to shut down the reactor as indicated by reactor power
_>5%
AND All actions to shut down the reactor are not successful as indicated by reactor power
_>5%
AND EITHER:
Core Cooling RED PATH conditions met
Heat Sink RED PATH conditions met Mode Applicability:
1 - Power Operation Definition(s):
None Basis:
This EAL addresses the following:
Any automatic reactor trip signal followed by a manual trip that fails to shut down the reactor to an extent the reactor is producing energy in excess of the heat load for which the safety systems were designed (EAL SA6.1), and
Indications that either core cooling is extremely challenged or heat removal is extremely challenged.
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 NCS barriers.
Reactor shutdown achieved by use of EP/I (2)/A/5000/FR-S.1 Response to Nuclear Power Generation/ATWS (such as depressing manual pushbutton on turbine control panel, emergency boration or manually driving control rods) are also credited as a successful manual trip provided reactor power can be reduced below 5% before indications of an extreme challenge to either core cooling or heat removal exist (ref. 1, 4).
5% rated power is a minimum reading on the power range scale that indicates continued 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 5%, plant response will be similar to that observed during a IIRP/OIAI5000/O01 Rev. 1 Page 178 of 247Jf
ATTACHMENT 1 EAL Bases normal shutdown. Nuclear instrumentation can be used to determine if reactor power is greater than 5 % power (ref. 1, 4).
Indication of continuing core cooling degradation is manifested by CSFST Core Cooling RED PATH conditions being met (ref. 2).
Indication of inability to adequately remove heat from the NCS is manifested by CSFST Heat Sink RED PATH conditions being met (ref. 3).
This IC addresses a failure of the RTS to initiate or complete an automatic or manual reactor trip 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 NCS. 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 shutdown the reactor.
A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
Escalation of the emergency classification level would be via IC RG1 or EGI.
CNS Basis Reference(s):
1. EP/I (2)/A/5000IF-O Critical Safety Function Status Trees - Subcriticality
2. EP/I(2)/A/5000IF-O Critical Safety Function Status Tress - Core Cooling
3. EP/I(2)/A/5000IF-O Critical Safety Function Status Tress - Heat Sink
4. EP/I(2)/A/5000/FR-S.1 Response to Nuclear Power Generation/ATWS
5. NEI 99-01 SS5 Category: S - System Malfunction Subcategory: 7 - Loss of Communications Initiating Condition: Loss of all onsite or offsite communications capabilities EAL:
JRP/O/AI5000I100 Rev. 1I Page 179 of 247
ATTACHMENT I EAL Bases SU7.1 Unusual Event Loss of all Table S-4 onsite communication methods OR Loss of all Table S-40ORO communication methods OR Loss of all Table S-4 NRC communication methods Table S-4 Communication Methods System Onsite ORO NRC Public Address X Internal Telephones X Onsite Radios X DEMNET X Commercial Telephones X X Satellite Phones X X Cellular Phones X X NRC Emergency Telephone System (ETS) X X Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None IRP/O/AI5000/O01 Rev. 1I Page 180 of 247
ATTACHMENT 1 EAL Bases Basis:
Onsite/offsite communications include one or more of the sYstems listed in Table S-4 (ref. 1).
Public Address System The Catawba Plant public address system provides paging and party line communications between stations located throughout the plant. Inside and outside type wall and desk-mounted stations are used to communicate between roaming personnel and fixed work locations. Plant-wide instructions are issued using the paging feature.
Internal Telephone System The Catawba Site PBX telephone system provides communication capability between telephone stations located within the plant by dialing the four-digit telephone station code.
On-site Radio System Radio systems can be used for communication among operators, off-site monitoring teams, the control room, TSC and EOF.
DEMNET DEMNET is the primary means of offsite communication. This circuit allows intercommunication among the EOF, TSC, control room, counties, and states. DEMNET operates as an internet based (VoIP) communications system with a satellite back-up. Should the internet transfer rate become slow or unavailable, the DEMNET will automatically transfer to satellite mode.
Commercial Telephones Commercial telephone lines, which supply public telephone communications, are employed by Duke Energy. The local service provider provides primary and secondary power for their lines at the Central Office.
Satellite Phones A portable satellite telephones are available which enable communication when all other phone systems are inoperable, e.g. following a major external event. These portable systems can be powered by internal batteries, external DC sources as well as external AC sources.
Cellular Phones Cellular phones may be used during emergencies if other communications means are not readily available or are inoperable. These phones are not expected to be used in the Control Room or Power Block due to interference with plant equipment and loss of signal to the phone.
SRP/0/A/5000/001 Rev. I Page 181 of 247
ATTACHMENT 1 EAL Bases NRC Emergency Telephone System The NRC Uses a Duke Energy dedicated telephone line which allows direct telephone communications from the plant to NRC regional and national offices. The Duke Energy communications line provides a link independent of the local public telephone network.
Telephones connected to this network are located in the Catawba Control Room, Technical Support Center, and Emergency Operations Facility and can be used to establish NRC Emergency Notification System (ENS) and Health Physics Network (HPN) capability.
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 State, York, Gaston and Mecklenburg County EOCs The third EAL addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.
CNS Basis Reference(s):
1. CNS Emergency Plan Section F Emergency Communications
2. NEI 99-01 SU6 IRPIOIAI5000I001I Rev. 1 Page 182 of 247
ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 8 - Containment Failure Initiating Condition: Failure to isolate containment or loss of containment pressure control.
EAL:
SU8.1 Unusual Event EITHER:
Any penetration is not isolated within 15 mai. of a VALID containment isolation signal (Note 1)
OR Containment pressure > 3 psig with < one full train of containment cooling operating per design for > i5 min. (Notes 1, 10)
Note 1" The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Note 10: If the loss of containment cooling threshold is exceeded due to loss of both trains of VX-CARF, this EAL only applies if at least one train of VX-CARF is not operating, per design, after the 10 minute actuation delay for greater than or equal to 15 minutes.
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
VALID - An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant~personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.
Basis:
The containment Phase B pressure setpoint (3 psig, ref. 1, 2) is the pressure at which the containment cooling systems should actuate and begin performing their function.
One full train of containment cooling operating per design is considered (ref. 1, 2):
One train of Containment Air Return Fan System (VX-CARF), and
One train of Containment Spray System (NS)
Once the Residual Heat Removal system is taking suction from the containment sump, with containment pressure greater than 3 psig and procedural guidance, one train of containment spray is manually aligned to the containment sump. If unable to place one NS train in service or without an operating train of VX-CARF (the CARE with a 10-minute delay) within 15 minutes this EAL has been exceeded. At this point a significant portion of the ice in the ice condenser SRP/O/AI5000/O01 Rev. 1 Page 183 of 247I
ATTACHMENT 1 EAL Bases would have melted and the NS system would be needed for containment pressure control.
The Unusual Event threshold applies after automatic or manual alignment of the containment spray system has been attempted with containment pressure greater than 3 psig and less than one full train of NS is operating for greater than or equal to 15 minutes.
The Unusual Event threshold also applies if containment pressure is greater than 3 psig and at least one train of VX-CARF is not operating after a 10 minute delay for greater than or equai to 15 minutes. Without a single train of VX-CARF in service following actuation, the Unusual Event should be declared regardless of whether ECCS is in injection or sump recirculation mode after 15 minutes.
This EAL addresses a failure of one or more containment penetrations to automatically isolate (close) when required by an actuation signal. It also addresses an event that results in high containment pressure with a concurrent failure of containment pressure control systems.
Absent challenges to another fission product barrier, either condition represents potential degradation of the level of safety of the plant.
For the first condition, the containment isolation signal must be generated as the result on an off-normal/accident condition (e.g., a safety injection or high containment pressure); a failure resulting from testing or maintenance does not warrant classification. The determination of containment and penetration status - isolated or not isolated - should be made in accordance with the appropriate criteria contained in the plant AOPs and EOPs. The 15-minute criterion is included to allow operators time to manually isolate the required penetrations, if possible.
The second condition addresses a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design. The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible. The inability to start the required equipment indicates that containment heat removal/depressurization systems (e.g., containment sprays or ice condenser fans) are either lost or performing in a degraded manner.
This event would escalate to a Site Area Emergency in accordance with IC FS1 if there were a concurrent loss or potential loss of either the Fuel Clad or NCS fission product barriers.
CNS Basis Reference(s):
1. CNS Technical Specification 3.6.6
2. CNS Technical Specification 3.6.6 Bases
3. CNS Technical Specification 3.3.2
4. UFSAR Section 6.2 Containment Systems
5. NEI 99-01 SU7 IRPIoIAooo0ooOOI Rev. I Page 184 of 247
ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 9 - Hazardous Event Affecting Safety Systems Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode EAL:
SA9.1 Alert The occurrence of any Table S-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 S-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:
I - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - 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.
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 EGOS. These are typically systems SRP/0/A/5000/001I Rev. 1 Page 185 of 247
ATTACHMENT 1 EAL Bases 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 seismic events are discussed under EAL HU2.1 (ref. 1).
Internal FLOODING may be caused by events such as component failures, equipment misalignment, or outage activity mishaps (ref. 2).
External flooding may be due to high lake level. CNS plant yard elevation is 593.5 ft MSL.
The minimum external access elevation for the Auxiliary, Turbine and Service Buildings is 594.0 ft MSL (ref. 1, 3).
Seismic Category I structures are analyzed to withstand a sustained, design wind velocity of at least 95 mph. (ref. 4).
Areas containing functions and systems required for safe shutdown of the plant are identified by fire area in the fire response procedure (ref. 5).
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 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 FS1 or RSI.
CNS Basis Reference(s):
RP0A/0000 ev age 186of 247
ATTACHMENT 1 EAL Bases 1.
RPIO/A/5000I007 Natural Disaster and Earthquake
2. AP/0/A/5500/030 Plant *Flooding
3. UFSAR Section 3.4 Water Level (Flood) Design
4. Updated FSAR Section 3.3.1 Wind Loadings
5. AP/0/A/5500/045 Plant Fire
6. NEI 99-01 SA9
RP/0/A/5000/001 Rev. 1 Page 187 of 247
ATTACHMENT 1 EAL Bases Category E - Independent Spent Fuel Storage Installation (ISFSI)
EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold)
An independent spent fuel storage installation (ISFSI) is a complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. A significant amount of the radioactive material contained within a 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.
The CNS ISFSI is contained wholly within the plant Protected Area. Therefore a security event related to the ISFSI would be applicable to EALs HUI.1, HA1 .land HSI.1 Minor surface damage that does not affect storage cask/canister boundary is excluded from the scope of these EALs.
SRP/0/A/5000/001 Rev. 1 Page 188 of 247
ATTACHMENT 1 EAL Bases 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 an on-contact radiation reading on the surface of a loaded Vertical Storage Cask (VSC) > any of the following:
100 mrem/hr (neutron + gamma) on the side of the VSC
100 mrem/hr (neutron + gamma) on the top of the VSC
200 mrem/hr (neutron + gamma) at the air inlets or outlets of the VSC 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 CNS ISFSI, Confinement Boundary is defined as the Transportable Storage Canister (TSC) for both NAC-UMS and MAGNASTOR storage systems.
Basis:
The CNS ISFSI utilizes two designs for dry spent fuel storage:
The NAC-UMS dry spent fuel storage system
The MAGNASTOR dry spent fuel storage system Both systems consist of a Transportable Storage Canister (TSC) and concrete Vertical Storage Cask (VSC). The TSC is the CONFINEMENT BOUNDARY for both systems. The TSC is welded and designed to provide confinement of all radionuclides under normal, off-normal, and accident conditions (ref. 1, 2).
Confinement boundary is defined as the barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. Therefore, damage to a confinement boundary must be a confirmed physical breach between the spent fuel and the environment for the TSC.
The values shown represent 2 times the limits specified in the ISESI Certificate of Compliance Technical Specification for radiation external to a loaded VSC for a NAC-UMS canister (ref. 1).
The specified ISFSI dose limits are based on surveys taken consistent with the locations specified in the associated Technical Specification (ref. 1, 2).
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 IRP/0/A/5000/001 Rev. I Page 189 of 247I
ATTACHMENT I EAL Bases fuel assemblies due to environmental factors, and configuration changes which could cause challenges in removing the cask or fuel from storage.
The existence of "damage" is determined by radiological survey. The technical specification multiple of "2 times", which is also used in Recognition Category R IC RU1, is used here to distinguish between non-emergency and emergency conditions. The emphasis for this classification is the degradation in the level of safety of the spent fuel cask and not the magnitude of the associated dose or dose rate. It is recognized that in the case of extreme damage to a loaded cask, the fact that the "on-contact" dose rate limit is exceeded may be determined based on measurement of a dose rate at some distance from the cask.
Security-related events for ISFSIs are covered under ICs HU1 and HAl.
CNS Basis Reference(s):
1. NAC-UMS Certificate of Compliance #1015 Technical Specifications
2. MAGNASTOR Certificate of Compliance #1031 Technical Specifications
3. NEI 99-01 E-HU1 SRPIOIAI5000io01Rv Rev. 1 ag 9 f4 Page 190 of 247 I
ATTACHMENT 1 EAL Bases Category F - Fission Product Barrier Degradation EAL Group: Hot Conditions (NCS 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 (NCS): The NCS Barrier includes the NCS primary side and its connections up to and including the pressurizer safety and relief valves, and other connections up to and including the primary isolation valves.
C. Containment (CMT): The Containment Barrier includes the containment building and connections up to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the containment building up to and including the outermost secondary side isolation valve.
Containment Barrier thresholds are used as criteria for escalation of the ECL from Alert to a Site Area Emergency or a General Emergency.
The EALs in this category require evaluation of the loss and potential loss thresholds listed in the fission product barrier matrix of Table F-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 NCS Site Area Emerqency:
Loss or potential loss of any two barriers General Emerqency:
Loss of any two barriersand loss or potentialloss 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 NCS Barrier are weighted more heavily than the Containment Barrier.
Unusual Event l~s associated with NCS and Fuel Clad Barriers are addressed under System Malfunction l~s.
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 SRP/0/A/5000/001 Rev. 1 Page 191 of 247I
ATTACHMENT 1 EAL Bases 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 plant-specific CNS design and operating characteristics.
As used in this category, the term NCS leakage encompasses not just those types defined in Technical Specifications but also includes the loss of NCS 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 NCS due to the as-designedlexpected operation of a relief valve is not considered to be NCS 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 NCS fission product barriers were both lost, then there should be frequent assessments of containment radioactive inventory and integrity. Alternatively, if both the Fuel Clad and NOS fission product barriers were potentially lost, the Emergency Coordinator would have more assurance that there was no immediate need to escalate to a General Emergency.
IRPIOIAI5000IO01I Rev. 1 Page 192 of 247I
ATTACHMENT 1 EAL Bases Category: Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Any loss or any potential loss of either Fuel Ciad or NCS EAL:
FAI.1 Alert Any loss or any potential ioss of either Fuel Clad or NCS (Table F-I)
Mode Applicability:
1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
Fuel Clad, NCS and 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 NCS barriers are weighted more heavily than the Containment barrier. Unlike the Containment barrier, loss or potential loss of either the Fuel Clad or NCS 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 NCS barrier results in declaration of a Site Area Emergency under EAL FSI .1 CNS Basis Reference(s):
1. NE1 99-01 FA1 SRP/0/A/5000/001 Rev. I Page 193 of 247I
ATTACHMENT I EAL Bases 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 Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
Fuel Clad, NCS 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 NCS 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 NCS potential loss thresholds existed, the Emergency Coordinator would have greater assurance that escalation to a General Emergency is less imminent.
CNS Basis Reference(s):
1. NEI 99-01 FS1 SRP/0/A/5000/001 Rev. I Page 194 of 247
ATTACHMENT I EAL Bases Category: Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss of any two barriers and loss or potential loss of third barrier EAL:
FGI.1 General Emergency Loss of any two barriers AND Loss or potential loss of third barrier (Table F-I)
Mode Applicability:
I - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):
None Basis:
Fuel Clad, NCS 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, NCS and Containment barriers
Loss of Fuel Clad and NCS barriers with potential loss of Containment barrier
Loss of NCS and Containment barriers with potential loss of Fuel Clad barrier
Loss of Fuel Clad and Containment barriers with potential loss of NCS barrier CNS Basis Reference(s):
I. NEI 99-01 FGI SRP/0/A/5000/001 Rev. I Page 195 of 247
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. NCS or SG Tube Leakage B. Inadequate Heat removal C. CMT Radiation / NCS Activity D. CMT Integrity or Bypass E. Emergency Coordinator Judgment 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 B would be assigned "FC Loss B.I," the third Containment barrier Potential Loss in Category D would be assigned "CMT P-Loss D.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 and NCS barriers and a Potential Loss of the Containment barrier can occur. Barrier SRP/0/A/5000/001 Rev. I Page 196 of 247
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Losses and Potential Losses are then applied to the algorithms given in EALs EG1 .1, FS1 .1, and FAI .1 to determine the appropriate emergency classification.
In the remainder of this Attachment, the Fuel Clad barrier threshold bases appear first, followed by the NCS 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,..., E.
IRP/°/A15°°°/°°I1 Rev. 1 Page 197 of 247I
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Table F-I Fission Product Barrier Threshold Matrix Fuel Clad (FC) Barrier Reactor Coolant System (NCS) Barrier Containment (CMT) Barrier Category LOSS Potential Loss Loss Potential Loss Loss Potential Loss A 1. An automatic or manual ECCS A (SI) actuation required by EITHER: 1CSFST Integrity-RED Path 1. A leaking or RUPTURED SG is NC rNneNn UNISOLABLE NCS conditions met SG Tube leakage FAULTED outside of containment Nn Leakage
SG tube RUPTURE
1. CSFST Core Cooling-ORANGEE Spath conditions mel i. CSFST Heal Sink-RED Path 1. CSFST Core Cooling-RED Path ndqae 1. CSFST Core Cooling-RED 2. CSFST Heal Sink-RED Path Indqae conditions mel conditions met Path conditions met conditions met None AND None AND Remoal AND Rm alHeal Heat sink is required Restoration procedures not sink is required effective within 15 mai. (Note 1)
C 1. EMF53AIB > Table F-2 column CUT "FC Loss" None 1. EMF53A(B > Table F-2 column Radiation 2. Dose equivalent 1-131 coolant None None 1. CMF53PoBenTable F-L o umn "NCS Loss""MPoetaLss
/ NCS activity>* 300 pCi/gm Activity___________
1. Containment isolation is required AND EITHER: 1. CSFST Containment-RED Path O
Containment integrity has conditions met Dbeen lost based on 2. Containment hydrogen concentratior CMT None None None None Emergency Coordinator 6%
Integrity >
or Bypass judgment 3. Containment pressure > 3 psig
UNISOLABLE pathway from with <one full train of containment Containment to the environment cooling operating per design for exists > 15 min. (Note 1)
2. Indications of NCS leakage outside of containment E 1. Any condition in the opinion of 1.Any condition in the opinion of 1 n odto nteoiino .Any condition in the opinion of the 1.Any condition in the opinion of 1.Any condition in the opinion of the th mrec oriao Emergency htthe Coordinator that the Emergency Coordinator that Emergency Coordinator that the Emergency Coordinator that Emergency Coordinator that EC Indicates loss of the fuel dlad indicates potential loss of the fuel Eme idcthes soften NCoobrdinat rta indicates potential loss of the NCS indicates loss of the containment Judgment barrier clad barrier indicates potential loss of the iniae oso h C anr barrier barrier containment barrier SRP/O/AI5000/O01 Rev. 1 Page 198 of 247I
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: 1. NCS or SG Tube Leakage Degradation Threat: Loss Threshold:
SNone IRPIOIN50001001 Rev. 1 I Page 199 of 247
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: 1. NCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold:
INone IRP/O/AI5000/O01 Rev. 1I Page 200 of 247
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: B. Inadequate Heat Removal Degradation Threat: Loss Threshold:
1. CSFST Core Cooling-RED Path conditions met Definition(s):
None Basis:
Critical Safety Function Status Tree (CSFST) Core Cooling-RED path indicates significant core exit superheating and core uncovery. The CSFSTs are normally monitored using the SPDS display on the Operator Aid Computer (OAC) (ref. 1).
This reading indicates temperatures within the core are sufficient to cause significant superheating of reactor coolant.
CNS Basis Reference(s):
1. EP/II(2)/5000/F-Q Critical Safety Function Status Trees
2. EP/I(2)/A/5000IFR-C,1 Response to Inadequate Core Cooling
3. EP/I(2)/A/5000/FR-C.2 Response to Degraded Core Cooling
4. NEI 99-01 Inadequate Heat Removal Fuel Clad Loss 2.A IRPIOIAI5000io01 Rev. 1 I Page 201 of 247
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: B. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold:
1. CSFST Core Cooling-ORANGE Path conditions met Definition(s):
None Basis:
Critical Safety Function Status Tree (CSFST) Core Cooling-ORANGE path indicates indicates subcooling has been lost and that some fuel clad damage may potentially occur.
The CSFSTs are normally monitored using the SPDS display on the Operator Aid Computer (OAC)
(ref. 1).
This reading indicates a reduction in reactor vessel water level sufficient to allow the onset of heat-induced cladding damage.
CNS Basis Reference(s):
1. EP/1I(2)/5000/F-O Critical Safety Function Status Trees
2. EP/1(2)/A/5000IFR-C.1 Response to Inadequate Core Cooling
3. EP/1(2)/A/5000/FR-C.2 Response to Degraded Core Cooling
4. NEI 99-01 Inadequate Heat Removal Fuel Clad Loss 2.A tRP/o/A1ooo0oo001 Rev. 1 Page 202of 247
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: B. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold:
2. CSFST Heat Sink-RED Path conditions met AND Heat sink is required Definition(s):
None Basis:
In combination with NCS Potential Loss B.1, meeting this threshold results in a Site Area Emergency.
Critical Safety Function Status Tree (CSFST) Heat Sink-RED path indicates the ultimate heat sink function is under extreme challenge and that some fuel clad damage may potentially occur (ref. 1).
The CSFSTs are normally monitored using the SPDS display on the Operator Aid Computer (OAC) (ref. 1).
The phrase "and heat sink required" precludes the need for classification for conditions in which NCS pressure is less than SG pressure or Heat Sink-RED path entry was created through operator action directed by an EOP. For example, FR-H.1 is entered from CSFST Heat Sink-Red. Step 2 tells the operator to determine if heat sink is required by checking that NCS pressure is greater than any non-faulted SG pressure and NCS Thot is greater than 3500 F. If these conditions exist, Heat Sink is required. Otherwise, the operator is to either return to the procedure and step in effect or place ND in service for heat removal. For large LOCA events inside the Containment, the SGs are moot because heat removal through the containment heat removal systems takes place. Therefore, Heat Sink Red should not be required and, should not be assessed for EAL classification because a LOCA event alone should not require higher than an Alert classification. (ref. 2)
This condition indicates an extreme challenge to the ability to remove NCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the Fuel Clad Barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using threshold is not warranted.
CNS Basis Reference(s):
1. EPII(2)/5000/F-O Critical Safety Function Status Trees
2. EP/I(2)/IN5OOO/FR-H.1 Response to Loss of Secondary Heat Sink IRPIO/A15000/O01 Rev. 1I Page 203 of 247
Fission Product BarrierATTACHMENT Loss/Potential 2Loss Matrix and Bases
3. NEI 99-01 Inadequate Heat Removal Fuel Clad Loss 2.B IRP/OIAI5000/O01 Rev. 1I Page 204 of 247
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: C. CMT Radiation / NCS Activity Degradation Threat: Loss Threshold:
1. EMF53A/B > Table F-2 column "FC Loss" Table F-2 Containment Radiation - R/hr (EM F53AJB)
Time After SID CMT Potential (Hs)FC Loss NCS Loss Ls 0-1 550 8.8 5500 1-2 400 8.4 4000 2-8 160 7.0 1600
>8 100 6.2 1000 Definition(s):
None Basis:
The gamma dose rate resulting from a postulated loss of coolant accident (LOCA) is monitored by the containment high range monitors, EMF53A & B. EMF53A & B are located inside containment. The detector range is approximately 1 to 1 E8 R/hr (logarithmic scale). Radiation Monitors EMF53A & B provide a diverse means of measuring the containment for high level gamma radiation. (ref. 1).
The Table F-2 values, column FC Loss represents, based on core damage assessment procedure, the expected containment high range radiation monitor (EMF53A & B) response based on a LOCA, for periods of 1, 2, 8 and 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months after shutdown, no sprays and NCS pressure < 1600 psig with -2% fuel failure (ref. 2).
The value is derived as follows:
RP/0/A/5000/01 5 Figure 3 Containment Radiation Level vs. Time for 100% Clad Damage 1, 2, 8 and 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months after shutdown without spray and NCS pressure < 1600 psig x 0.02 (rounded)
(ref. 2).
The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals 300 lpCi/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since IRP/O/AI5000/O01 Rev. I Page 205 of 247I
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases 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 NCS Barrier Loss threshold C.1 since it indicates a loss of both the Fuel Clad Barrier and the NCS Barrier.
Note that a combination of the two monitor readings appropriately escalates the ECL to a Site Area Emergency.
CNS Basis Reference(s):
1. IP/0/3314/004 Radiation Monitoring System RP-2C High Range Process Channel Calibration
2. RP/O/A/5000/01 5 Core Damage Assessment
3. NEI 99-01 CMT Radiation / ROS Activity Fuel Clad Loss 3.A SRPIOIAI50001001 Rev. I Page 206 of 247
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: C. CMT Radiation / NCS Activity Degradation Threat: Loss Threshold:
2. Dose equivalent 1-131 coolant activity > 300 tpCi/gm Definition(s):
None Basis:
Elevated reactor coolant activity represents a potential degradation in the level of safety of the plant and a potential precursor of more serious problems. The threshold dose equivalent 1-131 concentration is well above that expected for iodine spikes and corresponds to about 2% fuel clad damage. When reactor coolant activity reaches this level the Fuel Clad barrier is considered lost. (ref. 1).
This threshold indicates that NCS radioactivity concentration is greater than 300 pCi/gm dose equivalent I-1 31. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.
There is no Potential Loss threshold associated with NCS Activity / Containment Radiation.
CNS Basis Reference(s):
1. RP/0/A/5000/01 5 Core Damage Assessment
2. NEI 99-01 CMT Radiation / RCS Activity Fuel Clad Loss 3.B SRP/OI/N5000/O01 Rev. 1 Page 207 of 247
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: C. CMT Radiation / NCS Activity Degradation Threat: Potential Loss Threshold:
None SRP/0/A/5000/001 Rev. 1I Page 208 of 247
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Glad Category: D. CMT Integrity or Bypass Degradation Threat: Loss Threshold:
None SRP/0/A/5000/001 Rev. 1I Page 209 of 247
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: D. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:
None Rev. 1 I RP/0/A/5000/001 Re.IPae20 Page 210 off27247 I
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: E. Emergency Coordinator Judgment Degradation Threat: Loss Threshold:
1. Any condition in the opinion of the Emergency Coordinator that indicates loss of the Fuel Clad barrier Definition(s):
None Basis:
The Emergency Coordinator 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 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 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 Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.
This threshold addresses any other factors that are to be used by the Emergency Coordinator in determining whether the Fuel Clad barrier is lost.
CNS Basis Reference(s):
NEI 99-01 Emergency Director Judgment Fuel Clad Loss 6.A SRP/0/A/5000/001I Rev. 1 Page 211 of 247
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: E. Emergency Coordinator Judgment Degradation Threat: Potential Loss Threshold:
1. Any condition in the opinion of the Emergency Coordinator that indicates potential loss of the Fuel Clad barrier Basis:
The Emergency Coordinator 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 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 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 Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.
This threshold addresses any other factors that are to be used by the Emergency Coordinator in determining whether the Fuel Clad barrier is potentially lost. The Emergency Coordinator should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
CNS Basis Reference(s):
1. NEI 99-01 Emergency Director Judgment Potential Fuel Clad Loss 6.A IRP/IOA/5000/100 Rev. 1 Page 212 of 247I
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: A. NCS or SG Tube Leakage Degradation Threat: Loss Threshold:
1. An automatic or manual ECCS (SI) actuation required by EITHER:
UNISOLABLE NCS leakage
SG tube RUPTURE Definition(s):
UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.
RUPTURE - The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.
Basis:
ECCS (SI) actuation is caused by (ref. 1):
Pressurizer pressure < 1845 psig
Containment pressure > 1.2 psig This threshold is based on an UNISOLABLE NOS leak of sufficient size to require an automatic or manual actuation of the Emergency Core Cooling System (ECCS). This condition clearly represents a loss of the NCS Barrier.
This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE NCS leakage through an interfacing system. The mass loss may be into any location - inside containment, to the secondary-side (i.e., steam generator tube leakage) or outside of containment.
A steam generator with primary-to-secondary leakage of sufficient magnitude to require a safety injection is considered to be RUPTURED. If a RUPTURED steam generator is also FAULTED outside of containment, the declaration escalates to a Site Area Emergency since the Containment Barrier Loss threshold 1.A will also be met.
CNS Basis Reference(s):
1. EP/I (2)/N5000/E-0 Reactor Trip or Safety Injection
2. EP/I (2)/A/5000/E-3 Steam Generator Tube Rupture
3. NEI 99-0 1 RCS or SG Tube Leakage Reactor Coolant System Loss 1 .A SRP/OIAI50001001 Rev. 1I Page 213 of 247
2 ATTACHMENT Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: A. NCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold:
1. CSFST Integrity-RED path conditions met Definition(s):
None Basis:
The "Potential Loss" threshold is defined by the CSFST Reactor Coolant Integrity - RED path.
CSFST NCS Integrity - Red Path plant conditions and associated PTS Limit Curve A indicates an extreme challenge to the safety function when plant parameters are to the left of the limit curve following excessive NCS cooldown under pressure (ref. 1, 2).
This condition indicates an extreme challenge to the integrity of the NCS pressure boundary due to pressurized thermal shock - a transient that causes rapid NCS cooldown while the NCS is in Mode 3 or higher (i.e., hot and pressurized).
CNS Basis Reference(s):
1. EP/II(2)/A/5000/F-0 Critical Safety Function Status Trees
2. EP/1(2)/AI5000/FR-P.1 Response to Imminent Pressurized Thermal Shock Condition
3. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Potential Loss 1.B IRPIO/IN5000/O01 Rev. 1 Page 214 of 247I
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: B. Inadequate Heat Removal Degradation Threat: Loss Threshold:
None IRP/O/IN50001001I Rev. 1 Page 215 of 247
ATTACHMENT Fission Product Barrier Loss/Potential 2Loss Matrix and Bases Barrier: Reactor Coolant System Category: B. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold:
Definition(s):
None Basis:
In combination with FC Potential Loss B.2, meeting this threshold results in a Site Area Emergency.
Critical Safety Function Status Tree (CSFST) Heat Sink-RED path indicates the ultimate heat sink function is under extreme challenge and that some fuel clad damage may potentially occur (ref. 1).
The CSFSTs are normally monitored using the SPDS display on the Operator Aid Computer (OAC) (ref. 1).
The phrase "and heat sink required" precludes the need for classification for conditions in which NCS pressure is less than SG pressure or Heat Sink-RED path entry was created through operator action directed by an EOP. For example, FR-H.1 is entered from CSFST Heat Sink-Red. Step 2 tells the operator to determine if heat sink is required by checking that NCS pressure is greater than any non-faulted SG pressure and NCS Thot is greater than 350°F. If these conditions exist, Heat Sink is required. Otherwise, the operator is to either return to the procedure and step in effect or place ND in service for heat removal. For large LOCA events inside the Containment, the SGs are moot because heat removal through the containment heat removal systems takes place. Therefore, Heat Sink Red should not be required and, should not be assessed for EAL classification because a LOCA event alone should not require higher than an Alert classification. (ref. 1, 2).
This condition indicates an extreme challenge to the ability to remove NCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the NCS Barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using threshold is not warranted.
Meeting this threshold results in a Site Area Emergency because this threshold is identical to Fuel Clad Barrier Potential Loss threshold B.2; both will be met. This condition warrants a Site Area Emergency declaration because inadequate NCS heat removal may result in fuel heat-up SRPIOIA/50001001 Rev. 1 Page 216 of 247
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases sufficient to damage the cladding and increase NCS pressure to the point where mass will be lost from the system.
CNS Basis Reference(s):
1. EP/I1(2)/5000/F-O Critical Safety Function Status Trees
2. EP/I(2)/A/5000/FR-H.1 Response to Loss of Secondary Heat Sink
3. NEI 99-01 Inadequate Heat Removal NCS Loss 2.B IRP/0/A/5000/001 Rev. I Page 217 of 247
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category:
C. CMT Radiation!/NCS Activity Degradation Threat: Loss Threshold:
1. EMF53A/B > Table F-2 column "NCS Loss" Table F-2 Containment Radiation - R/hr (EMiF53A/B)
Time After S/D CMT Potential (Hs)FC Loss NCS Loss Ls 0-1 550 8.8 5500 1-2 400 8.4 4000 2-8 160 7.0 1600
>8 100 6.2 1000 Definition(s):
N/A Basis:
The gamma dose rate resulting from a postulated loss of coolant accident (LOCA) is monitored by the containment high range monitors, EMF53A & B. EMF53A & B are located inside containment. The detector range is approximately 1 to I1E8 R/hr (logarithmic scale). Radiation Monitors EMF53A & B provide a diverse means of measuring the containment for high level gamma radiation. (ref. 1).
The value specified represents, based on core damage assessment procedure RP/0/A/5000/01 5 Figure 1, the expected containment high range radiation monitor (EMF53A &
B) response based on a LOCA, for periods of 1, 2, 8 and 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months after shutdown with no fuel failure (ref. 2).
The value is derived as follows:
RP/0/A/5000/015 Figure 1 Containment Radiation Level vs. Time for RCS Release for periods of 1, 2, 8 and 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months after shutdown (rounded) (ref. 2).
The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for Fuel Clad Barrier Loss threshold C.1 since it indicates a loss of the NCS Barrier only.
There is no Potential Loss threshold associated with NOS Activity / Containment Radiation.
IRP/0/A/5000/001 Rev. 1 Page 218 of 247I
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases CNS Basis Reference(s):
1. IP/0/3314/004 Radiation Monitoring System RP-2C High Range Process Channel Cal ibration
2. RP/0/A/5000/015 Core Damage Assessment
3. NEI 99-01 CMT Radiation / RCS Activity NCS Loss 3.A SRP/0/A/5000/001 Rev. 1 Page 219 of 247
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: B. CMT Radiation/ NCS Activity Degradation Threat: Potential Loss Threshold:
None SRP/0/A/5000/001 Rev. 1 Page 220 of 247
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: D. CMT Integrity or Bypass Degradation Threat: Loss Threshold:
None SRP/0/A/5000/001 Rev. 1I Page 221 of 247
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: D. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:
SNone lRP/0/A/5000/001 Rev. 1 Page 222 of 247I
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: E. Emergency Coordinator Judgment Degradation Threat: Loss Threshold:
Definition(s):
None Basis:
The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the NCS barrier is 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 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 seqiuences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.
This threshold addresses any other factors that may be used by the Emergency Coordinator in determining whether the NCS Barrier is lost.
CNS Basis Reference(s):
1. NEI 99-01 Emergency Director Judgment NCS Loss 6.A SRP/0/A/5000/001 Rev. 1 Page 223 of 247
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: E. Emergency Coordinator Judgment Degradation Threat: Potential Loss Threshold:
1. Any condition in the opinion of the Emergency Coordinator that indicates potential loss of the NCS barrier Definition(s):
None Basis:
The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the NCS 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 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 Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.~
This threshold addresses any other factors that may be used by the Emergency Coordinator in determining whether the NCS Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
CNS Basis Reference(s):
1. NEI 99-01 Emergency Director Judgment NOS Potential Loss 6.A SRPI0IAI5000I001I Rev. 1 Page 224 of 247
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: A. NCS or SG Tube Leakage Degradation Threat: Loss Threshold:
I1. Aleaking or RUTRDS sFAULTED outside of containment Definition(s):
FAULTED - The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.
RUPTURED - The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.
Basis:
This threshold addresses a leaking or RUPTURED Steam Generator (SG) that is also FAULTED outside of containment. The condition of the SG, whether leaking or RUPTURED, is determined in accordance with the thresholds for NCS Barrier Potential Loss A.1 and Loss A.1, respectively. This condition represents a bypass of the containment barrier.
FAULTED is a defined term within the NEI 99-01 methodology; this determination is not necessarily dependent upon entry into, or diagnostic steps within, an EOP. For example, if the pressure in a steam generator is decreasing uncontrollably (part of the FAULTED definition) and the FAULTED steam generator isolation procedure is not entered because EOP user rules are dictating implementation of another procedure to address a higher priority condition, the steam generator is still considered FAULTED for emergency classification purposes.
The FAULTED criterion establishes an appropriate lower bound on the size of a steam release that may require an emergency classification. Steam releases of this size are readily observable with normal Control Room indications. The lower bound for this aspect of the containment barrier is analogous to the lower bound criteria specified in IC SU4 for the fuel clad barrier (i.e., NCS activity values) and IC SU5 for the NCS barrier (i.e., NCS leak rate values).
This threshold also applies to prolonged steam releases necessitated by operational considerations such as the forced steaming of a leaking or RUPTURED steam generator directly to atmosphere to cooldown the plant, or to drive an auxiliary (emergency) feed water pump. These types of conditions will result in a significant and sustained release of radioactive steam to the environment (and are thus similar to a FAULTED condition). The inability to isolate the steam flow without an adverse effect on plant cooldown meets the intent of a loss of containment.
IRP/O/AI5000/O01 Rev. 1I Page 225 of 247I
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Steam releases associated with the expected operation of a SG power operated relief valve or safety relief valve do not meet the intent of this threshold. Such releases may occur intermittently for a short period of time following a reactor trip as operators process through emergency operating procedures to bring the plant to a stable condition and prepare to initiate a plant cooldown. Steam releases associated with the unexpected operation of a valve (e.g., a stuck-open safety valve) do meet this threshold.
Following an SG tube leak or rupture, there may be minor radiological releases through a secondary-side System component (e.g., air ejectors, glad seal exhausters, valve packing, etc.). These types of releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R l~s.
The ECLs resulting from primary-to-secondary leakage, with or without a steam release from the FAULTED SG, are summarized below.
Affected SG is FAULTED Outside of Containment?
P-to-S Leak Rate Yes No Less than or equal to 25 gpm No classification No classification Greater than 25 gpm Unusual Event per SU5.1 Unusual Event per SU5.1 Requires operation of a standby charging (makeup) pump (NCS Site Area Emergency per FS1 .1 Alert per FA1 .1 BarrierPotentialLoss)
Requires an automatic or manual Site Area Emergency per ECCS (SI) actuation (NCS Barrier Alert per FA1 .1 FS1 .1 Loss)
There is no Potential Loss threshold associated with NCS or SG Tube Leakage.
CNS Basis Reference(s):
1. EP/II(2)/A/5000/E-0 Reactor Trip or Safety Injection
2. EP/II(2)1A/5000/E-3 Steam Generator Tube Rupture
3. NEI 99-01 RCS or SG Tube Leakage Containment Loss I .A IRP/0/A15000/001 Rev. I Page 226 of 247
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: A. NCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold:
None IRPIO/AI50001001I Rev. 1 Page 227 of 247
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: B. Inadequate heat Removal Degradation Threat: Potential Loss Threshold:
1. CSFST Core Cooling-RED path conditions met AND Restoration procedures not effective within 15 min. (Note 1)
Definition(s):
None Basis:
Critical Safety Function Status Tree (CSFST) Core Cooling-RED path indicates significant core exit superheating and core uncovery. The CSFSTs are normally monitored using the SPDS display on the Operator Aid Computer (OAC) (ref. 1).
The function restoration procedures are those emergency operating procedures that address the recovery of the core cooling critical safety functions. The procedure is considered effective if the temperature is decreasing or if the vessel water level is increasing (ref. 1, 2, 3).
A direct correlation to status trees can be made if the effectiveness of the restoration procedures is also evaluated. If core exit thermocouple (TC) readings are greater than 1,200°F (ref. 1), Fuel Clad barrier is also lost.
This condition represents an IMMINENT core melt sequence which, if not corrected, could lead to vessel failure and an increased potential for containment failure. For this condition to occur, there must already have been a loss of the NCS Barrier and the Fuel Clad Barrier. If implementation of a procedure(s) to restore adequate core cooling is not effective (successful) within 15 minutes, it is assumed that the event trajectory will likely lead to core melting and a subsequent challenge of the Containment Barrier.
The restoration procedure is considered "effective" if core exit thermocouple readings are decreasing and/or if reactor vessel level is increasing. Whether or not the procedure(s) will be effective should be apparent within 15 minutes. The Emergency Coordinator should escalate the emergency classification level as soon as it is determined that the procedure(s) will not be effective.
Severe accident analyses (e.g., NUREG-1 150) have concluded that function restoration procedures can arrest core degradation in a significant fraction of core damage scenarios, and that the likelihood of containment failure is very small in these events. Given this, it is appropriate to provide 15 minutes beyond the required entry point to determine if procedural actions can reverse the core melt sequence.
CNS Basis Reference(s):
1. EP/I (2)/5000/F-0 Critical Safety Function Status Trees IRP/O/A/5000/O01 Rev. 1I Page 228 of 247
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases
2. EP/I(2)/A/5000/FR-C.1 Response to Inadequate Core Cooling
3. EP/I (2)/A/5000/FR-C.2 Response to Degraded Core Cooling
4. NEI 99-01 Inadequate Heat Removal Containment Potential Loss 2.A IRPIOIAN50001001 IRev. 1I Page 229 of 247
ATTACHMENT Fission Product Barrier Loss/Potential 2Loss Matrix and Bases Barrier: Contzainment Category: C. Cl* VIT Radiation/NCS Activity Degradation Threat: Loss Threshold:
SNone SRP/O/AI5000/O01 Rev. 1 Page 230 of 247
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category:
C. CMT Radiation/NCS Activity Degradation Threat: Potential Loss Threshold:
1. EMF53A/B > Table F-2 column "CMT Potential Loss" Table F-2 Containment Radiation - R/hr (EMF53A/B)
Time After SID FCLs C os CMT Potential (Hrs.) Loss 0-1 550 8.8 5500 1-2 400 8.4 4000 2-8 160 7.0 1600
>8 100 6.2 1000 Definition(s):
None Basis:
The gamma dose rate resulting from a postulated loss of coolant accident (LOCA) is monitored by the containment high range monitors, EMF53A & B. EMF53A & B are located inside containment. The detector range is approximately 1 to 1 E8 R/hr (logarithmic scale). Radiation Monitors EMF53A & B provide a diverse means of measuring the containment for high level gamma radiation. (ref. 1).
The Table F-2 values, column CMT Potential Loss represents, based on core damage assessment procedure, the expected containment high range radiation monitor (EMF53A & B) response based on a LOCA, for periods of 1, 2, 8 and 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months after shutdown, no sprays and NCS pressure < 1600 psig with -20% fuel failure (ref. 2).
The value is derived as follows:
RP/0/A/5000/015 Figure 3 Containment Radiation Level vs. Time for 100% Clad Damage 1, 2, 8 and 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months after shutdown with no spray and NCS pressure < 1600 psig x 0.20 (rounded)
(ref. 2).
The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the 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 NCS Barrier Loss thresholds.
IRP/O/AI5000/O01 Rev. I Page 231 of 247
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases NUREG-1 228, 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 NCS 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 ECL to a General Emergency.
CNS Basis Reference(s):
1. IP/0/3314/004 Radiation Monitoring System RP-2C High Range Process Channel Calibration
2. RP/0/A/5000/01 5 Core Damage Assessment
3. NEI 99-01 CMT Radiation / RCS Activity Containment Potential Loss 3.A SRP/O/A/5000/O01 Rev. 1 Page 232 of 247
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: D. CMT Integrity or Bypass Degradation Threat: Loss Threshold:
1. Containment isolation is required AND EITHER:
Containment integrity has been lost based on EC judgment
UNISOLABLE pathway from containment to the environment exists Definition(s):
UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.
Basis:
These thresholds address a situation where containment isolation is required and one of two conditions exists as discussed below. Users are reminded that there may be accident and release conditions that simultaneously meet both bulleted thresholds.
First Threshold - Containment integrity has been lost, i.e., the actual containment atmospheric leak rate likely exceeds that associated with allowable leakage (or sometimes referred to as design leakage). Following the release of NCS mass into containment, containment pressure will fluctuate based on a variety of factors; a loss of containment integrity condition may (or may not) be accompanied by a noticeable drop in containment pressure. Recognizing the inherent difficulties in determining a containment leak rate during accident conditions, it is expected that the Emergency Coordinator will assess this threshold using judgment, and with due consideration given to current plant conditions, and available operational and radiological data (e.g., containment pressure, readings on radiation monitors outside containment, operating status of containment pressure control equipment, etc.).
Refer to the middle piping run of Figure 1. Two simplified examples are provided. One is leakage from a penetration and the other is leakage from an in-service system valve.
Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure.
Another example would be a loss or potential loss of the NCS barrier, and the simultaneous occurrence of two FAULTED locations on a steam generator where one fault is located inside containment (e.g., on a steam or feedwater line) and the other outside of containment. In this case, the associated steam line provides a pathway for the containment atmosphere to escape to an area outside the containment.
Following the leakage of NCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable (design) containment leakage through various penetrations or system components. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R l~s.
IRPI/AIN5000/O01 Rev. I Page 233 of 247
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Second Threshold - Conditions are such that there is an UNISOLABLE pathway for the migration of radioactive material from the containment atmosphere to the environment. As used here, the term "environment" includes the atmosphere of a room or area, outside the containment, that may, in turn, communicate with the outside-the-plant atmosphere (e.g.,
through discharge of a ventilation system or atmospheric leakage). Depending upon a variety of factors, this condition may or may not be accompanied by a noticeable drop in containment pressure.
Refer to the top piping run of Figure 1. In this simplified example, the inboard and outboard isolation valves remained open after a containment isolation was required (i.e., containment isolation was not successful). There is now an UNISOLABLE pathway from the containment to the environment.
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.
Leakage between two interfacing liquid systems, by itself, does not meet this threshold.
Refer to the bottom piping run of Figure 1. In this simplified example, leakage in an RCP seal cooler is allowing radioactive material to enter the Auxiliary Building. The radioactivity would be detected by the Process Monitor. If there is no leakage from the closed water cooling system to the Auxiliary Building, then no threshold has been met. If the pump developed a leak that allowed steam/water to enter the Auxiliary Building, then second threshold would be met. Depending upon radiation monitor locations and sensitivities, this leakage could be detected by any of the four monitors depicted in the figure and cause the first threshold to be met as well.
Following the leakage of NCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable containment leakage through various penetrations or system components. Minor releases may also occur if a containment isolation valve(s) fails to close but the containment atmosphere escapes to an enclosed system. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R ICs.
The status of the containment barrier during an event involving steam generator tube leakage is assessed using Loss Threshold A.1.
CNS Basis Reference(s):
1. NEI 99-01 CMT Integrity or Bypass Containment Loss 4.A Barrier: Containment Category: D. CMT Integrity or Bypass Degradation Threat: Loss Threshold:
2. ndiatonsof CSleakage outside ofcontainment SRPIOIAI50001001 Rev. 1I Page 234 of 247I
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Definition(s):
None Basis:
ECA-1 .2 LOCA Outside Containment (ref. 1) provides instructions to identify and isolate a LOCA outside of the containment. Potential NCS leak pathways outside containment include (ref. 1,2):
Residual Heat Removal (ND)
Safety Injection (NI)
Chemical & Volume Control (NV)
RCP seals (NC)
PZR/NCS Loop sample lines (NM)
Containment sump, temperature, pressure and/or radiation levels will increase if reactor coolant mass is leaking into the containment. If these parameters have not increased, then the reactor coolant mass may be leaking outside of containment (i.e., a containment bypass sequence). Increases in sump, temperature, pressure, flow and/or radiation level readings outside of the containment may indicate that the NCS mass is being lost outside of containment.
Unexpected elevated readings and alarms on radiation monitors with detectors outside containment should be corroborated with other available indications to confirm that the source is a loss of NCS mass outside of containment. If the fuel clad barrier has not been lost, radiation monitor readings outside of containment may not increase significantly; however, other unexpected changes in sump levels, area temperatures or pressures, flow rates, etc.
should be sufficient to determine if NCS mass is being lost outside of the containment.
Refer to the middle piping run of Figure 1. In this simplified example, a leak has occurred at a reducer on a pipe carrying reactor coolant in the Auxiliary Building. Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure and cause threshold D.1 to be met as well.
To ensure proper escalation of the emergency classification, the NCS leakage outside of containment must be related to the mass loss that is causing the NCS Loss and/or Potential Loss threshold A.1 to be met.
CNS Basis Reference(s):
1. EP/I (2)/A/5000/ECA-1 .2 LOCA Outside Containment
2. EP/1(2)/A/5000/E-1 Loss of Reactor or Secondary Coolant
3. NEI 99-01 CMT Integrity or Bypass Containment Loss SRP/O/N5000/001I Rev. I Page 235 of 247
Fission Product BarrierATTACHMENT 2 Loss/Potential Loss Matrix and Bases Figure 1: Containment Integrity or Bypass Examples Cooling RP/O/AI50001001 Rev. 1 Page 236 of 247
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category:
D. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:
I1. CSFST Containment-RED path conditions met Definition(s):
None Basis:
Critical Safety Function Status Tree (CS FST) Containment-RED path is entered if containment pressure is greater than or equal to 15 psig and represents an extreme challenge to safety function. The CSFSTs are normally monitored using the SPDS display on the Operator Aid Computer (OAC) (ref. 1).
15 psig is based on the containment design pressure (ref. 2).
If containment pressure exceeds the design pressure, there exists a potential to lose the Containment Barrier. To reach this level, there must be an inadequate core cooling condition for an extended period of time; therefore, the NCS and Fuel Clad barriers would already be lost. Thus, this threshold is a discriminator between a Site Area Emergency and General Emergency since there is now a potential to lose the third barrier.
CNS Basis Reference(s):
1. EP/I (2)/A/5000IF-0 Critical Safety Function Status Trees
2. UFSAR Section 6.2 Containment Systems
3. NEI 99-01 CMT Integrity or Bypass Containment Potential Loss 4.A SRPI/0N5000/O01 Rev. 1 Page 237 of 247
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: D. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:
2. Containment hyrgnconcentration > 6%
Definition(s):
None Basis:
Following a design basis accident, hydrogen gas may be generated inside the containment by reactions such as zirconium metal with water, corrosion of materials of construction and radiolysis of aqueous solution in the core and sump. (ref. 1).
The Containment Hydrogen Purge and Sample System (VY) is used to monitor the hydrogen concentration inside containment after a severe accident involving core damage. Samples of Containment air are obtained via the containment hydrogen/oxygen sample lines to the Post Accident Containment Sample (PACS) panel located in the auxiliary building. Additionally, the containment hydrogen analyzer system continuously monitors the hydrogen concentration inside containment (ref. 1).
The lower limit of deflagration of hydrogen in air is approximately 6% and is the maximum concentration at which hydrogen igniters can be placed in service (ref. 2).
To generate such levels of combustible gas, loss of the Fuel Clad and NCS barriers must have occurred. With the Potential Loss of the containment barrier, the threshold hydrogen concentration, therefore, will likely warrant declaration of a General Emergency.
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 bum will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a potential loss of the Containment Barrier.
CNS Basis Reference(s):
1. UFSAR Section 6.2 Containment Systems
2. EP/I (2)/A/5000/FR-Z.4 Response to High Containment Hydrogen Concentration
3. NEI 99-01 CMT Integrity or Bypass Containment Potential Loss 4.B RP/O/AI5000/O01 Rev. 1 Page 238 of 247
Fission Product BarrierATTACHMENT Loss/Potential 2Loss Matrix and Bases Barrier: Containment Category: D. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:
3. Containment pressure > 3 psig with < one full train of containment cooling operating per design for > 15 min. (Notes 1, 10)
Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Note 10: If the loss of containment cooling threshold is exceeded due to loss of both trains of VX-CARF, this EAL only applies if at least one train of VX-CARF is not operating, per design, after the 10 minute actuation delay for greater than or equal to 15 minutes.
Definition(s):
None Basis:
The containment Phase B pressure setpoint (3 psig, ref. 1, 2) is the pressure at which the containment cooling systems should actuate and begin performing their function.
One full train of containment cooling Operating per design is considered (ref. 1, 2):
One train of Containment Air Return Fan System (VX-CARF), and
One train of Containment Spray System (NS)
Once the Residual Heat Removal system is taking suction from the containment sump, with containment pressure greater than 3 psig and procedural guidance, one train of containment spray is manually aligned to the containment sump. If unable to place one NS train in service or without an operating train of VX-CARF (the CARE with a 10-minute delay) within 15 minutes a potential loss of containment exists. At this point a significant portion of the ice in the ice condenser would have melted and the NS system would be needed for containment pressure control. The potential loss of containment applies after automatic or manual alignment of the containment spray system has been attempted with containment pressure greater than 3 psig and less than one full train of NS is operating for greater than or equal to 15 minutes.
The potential loss of containment also applies if containment pressure is greater than 3 psig and at least one train of VX-CARF is not operating after a 10 minute delay for greater than or equal to 15 minutes. Without a single train of VX-CARF in service following actuation, the potential loss should be credited regardless of whether ECCS is in injection or sump recirculation mode after 15 minutes.
This threshold describes a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design. The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible. This threshold represents a potential loss of containment in IRP/O/A/5000/O01I Rev. 1 Page 239 of 247I
ATTACHMENT 2 Fission Product Barrier L.oss/Potential Loss Matrix and Bases that containment heat removal/depressurization systems (e.g., containment sprays, ice condenser fans, etc., but not including containment venting strategies) are either lost or performing in a degraded manner.
CNS Basis Reference(s):
1. CNS Technical Specification 3.6.6
2. CNS Technical Specification 3.6.6 Bases
3. CNS Technical Specification 3.3.2
4. UFSAR Section 6.2 Containment Systems
5. NEI 99-01 CMT Integrity or Bypass Containment Potential Loss 4.C IRP/O/A/5000/O01 I Rev. 1 Page 240 of 247 I
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: F. Emergency Coordinator Judgment Degradation Threat: Loss Threshold:
1. Any condition in the opinion of the Emergency Coordinator that indicates loss of the Containment barrier Definition(s):
None Basis:
The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the Primary Containment barrier is 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 Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.
This threshold addresses any other factors that may be used by the Emergency Coordinator in determining whether the Containment Barrier is lost.
CNS Basis Reference(s):
1. NEI 99-01 Emergency Director Judgment PC Loss 6.A RP/0/A/5000/001I Rev. 1 Page 241 of 247I
ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: F. Emergency Coordinator Judgment Degradation Threat: Potential Loss Threshold:
1. Any condition in the opinion of the Emergency Coordinator that indicates potential loss of the Containment barrier Definition(s):
None Basis:
The Emergency Coordinator 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. I
Imminent barrier decqradation 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 monitorincq 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 Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.
This threshold addresses any other factors that may be used by the Emergency Coordinator in determining whether the Containment Barrier is lost.
CNS Basis Reference(s):
1. NEI 99-01 Emergency Director Judgment PC Potential Loss 6.A RP/0/A/5000/001 Rev. 1 Page 242 of 2471
ATTACHMENT 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases
Background
NEI 99-01 Revision 6 l~s AA3 and HA5 prescribe declaration of an Alert based on impeded access to rooms or areas (due to either area radiation levels or hazardous gas concentrations) where equipment necessary for normal plant operations, cooldown or shutdown is located.
These areas are intended to be plant operating mode dependent. Specifically the Developers Notes for AA3 and HA5 states:
The "site-specific list of plant rooms or areas with entry-related mode applicabilityidentified" should specify those rooms or areas that 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 areas in which actions of a con tingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations). In addition, the list should specify the plant mode(s) during which entry would be required for each room or area.
The list should not include rooms or areasfor which entry is required solely to perform actions of an administrative or record keeping nature (e.g., normal rounds or routine inspections).
Further, as specified in IC HA5:
The list need not include the Control Room if adequate engineered safety/design features are in place to preclude a Control Room evacuation due to the release of a hazardousgas.
Such features may include, but are not limited to, capability to draw air from multiple air intakes at different and separatelocations, inner and outer atmosphericboundaries, or the capabilityto acquire and maintain positive pressure within the Contfrol Room envelope.
The review at CNS was completed using the following Controlling Procedures:
OP/1 (2)/A/6100/003 (Controlling Procedure For Unit Operation)
OP/1 (2)/A/6100/002 (Controlling Procedure For Unit Shutdown)
RP/0/A/5000/001I Rev. 1 Page 243 of 247
ATTACHMENT 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases CNS Table 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:
OP/1/A/61 00/003 Coordinate with Chemistry Auxiliary Building (Various Enclosure 4.3 Step and Radwaste while Locations) 3.19: U2 Encl. 4.3 performing NC System Step 3.20 Degas. Radwaste I No continues Degas OPS thru shutdown & cooldown OP/1/AI6100/O03, Align heater vent orifices Turbine Building (Various Enclosure 4.3, Steps per OP/1 (2)1B162501004 Locations) 3.26, 3.27 and 3.28; (Feedwater Heaters, Vents, No U2 Encl. 4.3, Steps Drains and Bleed Systems) 3.28, 3.29 and 3.30 ,Align VI and SP valves associated with CFPTs OP1llA161001003, Align Auxiliary Steam to Turbine Building (Various Enclosure 4.2, Step CFPTs. Locations) INo 3.11: U2 Endl. 4.2 Step 3.10 _________
OP/I1&2/A/6100/003, Align "C" Htr Drain Turbine Building (568')
Enclosure 4.2, Steps Pump per 3.13; U2 Encd. 4.2 OP/i1(2)/B162501004 Step 3.12 (Feedwater Heaters, INo Vents, Drains and Bleed Systems) for removal from service OP/11AI6100/003, Plant activities to ensure Turbine Building (594')
Enclosure 4.2, Step Main Turbine Sealing Steam INo 3.14; U2 Endl. 4.2 system responds as Step 3.13 required.
OPII&2/AI6100/O03, Ensure Moisture Separator Turbine Building (619')
Enclosure 4.2, Step Reheater low load valve 3.18: U2 Encd. 4.2 operation per 1No Step 3.17 OP/i (2)/B/6250/013 (Moisture Separator Reheater Operation)
OP/IllA/6100/003 Secure one Main CFPT per Turbine Building (Mainly Enclosure 4.2 Step OP/i1(2)/A/6250/001 594') 1No 3.19: U2 Endl. 4.2 (Condensate and Step 3.18 Feedwater System)
OP/1/A/6100/003, Secure half Main Outside in Main Enclosure 4.2, Step Transformer Cooling Fans Transformer yard 1No 3.20; U2 Encl. 4.2 and oil pumps Step 3.19 __________
RPIOIAI5000/O01 Rev. 1 Page 244 of 247
ATTACHMENT 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases
.. ..... . ... action.not performe¢ CNS Procedure , .. *"Ste Action=...... Building/Elevation/Room MOde does this prevent and..Step , , .... .. ..... '.... ..... ... ..cooldown/ shutdown?.
OP/i1(2)/A/6100/003 Shutdown the Main Turbine Turbine Building and Enclosure 4.2 Step per OP/I1(2)/B/6300/001 Transformer Yard INo 3.21: (Turbine generator)
OP/i1(2)/A/61 00/003, Bypass 'F" LP heaters Turbine Building (594" LP Enclosure 4.2, Step Htr Panel). INo 3.28 &3.29 OP/i1(2)/A/61 00/003, Transfer of Aux Steam to Service Bldg. (568')
Enclosure 4.2, Step on line Unit per 3.34 OP/0/B/6250/007 A (Auxiliary Steam System 1No Alignment) or place Aux Electric Boiler in service per OP/I1/B/6250/007 B (Auxiliary Electric Boilers)
OP/I1(2)/A/61 00/003, Isolate Unit Related Turbine Building. (594')
Enclosure 4.2, Step Steam supply to Aux 1No 3.37 Steam Header OP/i1(2)/A/61 00/002, Initiate action to reduce Auxiliary Building (577' Enclosure 4.1, Step VCT pressure per Mechanical Pent. Room) 3.4 OP/i1(2)/6200/001 1, 2, 3 No (Chemical and Volume Control System)
OP/1(2)/A/6100/002, Align S/G reverse purge. Both Doghouses IN Enclosure 4.1, Step N 3.9 OP/1(2)/A/6100/002, Align CM system flow for Turbine Building (619') 3N Enclosure 4.1, Step Low Pressure cleanup 3N 3.52 thru Upper Surge Tank.
OP/i1(2)/A/61 00/002 Shutdown Rod Control Auxiliary Building (594' Enclosure 4.2 or 4.7, System per Electrical Pent Room) 3 No Step 3.3 OP/i (2)/61 50/008 (Rod Control)
OP/i1(2)/A/61 00/002 Chemistry obtains Auxiliary Building (543' Enclosure 4.2 or 4.7 samples to ensure boron Sample Lab)
Step 3.10 concentration good to 3 No allow NCS cooldown to begin OP/I1(2)/A/61 00/002 Perform PZR PORVs Auxiliary. Building (577',
Enclosure 4.2 or 4.7, stroke testing per IF performing IWVR Step 3.22 PT/I1(2)/A/4200/023 A Containment 635' as 3 No (NC valve Inservice Test) well)
SRP/0/A/5000/001 JRev. I Page 245 of 247
ATTACHMENT 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases OP/1 (2)/A/61 00/002 Support placing N2 Auxiliary Building Enclosure 4.2 or 4.7 Cover gas on NCDT per (Various Locations on Step 3.23 OP/i (2)/A/6500/01 4 577' & 560')
3 No (Operations Controlled Liquid Waste Systems)
OP/i1(2)/A/61 00/002, Removing CLAs from Auxiliary Building (577' &
Enclosure 4.2 or 4.7 service per 560' Ess. MCC Bkrs) 3N Step 3.31 OP/I (2)/A/6200/009 3N (Cold Leg Accumulator
________________Operation).
OP/i &2/A/6100/002 Remove CAPT and one Auxiliary Building. (577' &
Enclosure 4.2 or 4.7, Motor Driven CA Pump from 560 Ess MCC Bkrs)
Step 3.45.1 service per 4 No OP/i (2)/A/6250/002 (Auxiliary Feedwater System').____________
OP/i1(2)/A/6100/002 Open NCS Loop Suction Auxiliary. Building. (577' Enclosure 4.2 or 4.7, Vlvs for train of ND to be & 560' Ess. MCC Bkrs)
Step 3.46.3 placed in service per 4 Yes OP/i (2)1AI62001004 (Residual Heat removal System).
OP/I1(2)/A/61 00/002 Rack out appropriate NI and Auxiliary Building (577' &
Enclosure 4.2 or 4.7, NV Pump Motor Bkrs per 560 Electrical Pent Step 3.48.2 OP/0/A/6350/0i0 (Operation Rooms) 4 Yes of Station Breakers and Disconnects)
OP/I1(2)/A/61 00/002 Support placing first train of Auxiliary Building (577' or Enclosure 4.2 or 4.7 ND in service per 560' Ess MCC Bkr s)
Step 3.52.2 OP/i (2)/A/6200/004 4 Yes (Residual Heat removal
____ ____ ___ System)__ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _
SRPiOIAI5000/O01 Rev. 1I Page 246 of 247
ATTACHMENT 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases Table R-2 & H-2 Results Table R-2/H-2 Safe Operation & Shutdown Rooms/Areas Bldg. Elevation Unit 1 Room/Area Unit 2 Room/Area Mode Rm 478 (1 EMXA) Rm 469 (2EMXA) 4 Rm 496 (1 ETA) R~m 486 (2 ETA) 4 Auiir 7'Rm 496 (1 EMXS) Rm 486 (2EMXS) 4 AB-577', J J-57 (1 MXK) AB-577', JJ-57 (2MXK) 4 Rm 330 (1 EMXJ) Rm 320 (2EMXJ) 4 Auxiliary 560' Rm 372 (1 ETB) Rm 362 (2ETB) 4 Rm 372 (1 EMXD) Rm 362 (2EMXD) 4 Plant Operating Procedures Reviewed
1. OP/i1(2)/A/6100/003 (Controlling Procedure for Unit Operation)
2. OP/I1(2)/A/6 1001002 (Controlling Procedure for Unit Shutdown)
3. OP/1 (2)1B/6250/004 (Feedwater Heaters, Vents, Drains and Bleed Systems)
4. OP/1 (2)/B16250101 3 (Moisture Separator Reheater Operation)
5. OP/I1(2)/A/6250/001 (Condensate and Feedwater System)
6. OP/I1(2)/B/6300/001 (Turbine generator)
7. OP101B16250/007 A (Auxiliary Steam System Alignment)
8. OP/1 /B/6250/007 B (Auxiliary Electric Boilers)
9. OP/I1(2)/6200/001 (Chemical and Volume Control System)
10. OP/I1(2)/6150/008 (Rod Control)
11. PT/I1(2)/A/4200/023 A (NC valve Inservice Test)
12. OP/I1(2)/A16500/014 (Operations Controlled Liquid Waste Systems)
13. OP/i1(2)/A/6200/009 (Cold Leg Accumulator Operation)
14. OP/I1(2)/A/6250/002 (Auxiliary Feedwater System)
15. OP/I1(2)/A/6200/004 (Residual Heat removal System)
16. OP/OIA16350/010 (Operation of Station Breakers and Disconnects)
SRP/0/A/5000/00I Rev. 1I Page 247 of 247

v t e Letter Sequence Response to RAI
TAC:MF6166, (Open) TAC:MF6167, (Open)
Initiation Acceptance Administration Questions Answers 19 February 2016 19 February 2016 19 February 2016 19 February 2016 19 February 2016 Results Approval... Other:
MONTHYEAR2015-05-04 [Table View]

CNS-16-010, Response to Request for Additional Information and Emergency Action Level Technical Bases Document, Revision 1, Clean Version. Part 4 of 4

Text

4.0 REFERENCES

4.0 REFERENCES

Background

Navigation menu

Search