ML15191A164

Emergency Action Level Technical Bases (Redline and Strikeout Version)

ML15191A164
Person / Time
Site:	Comanche Peak
Issue date:	06/24/2015
From:	Luminant Generation Co, Luminant Power
To:	Office of Nuclear Reactor Regulation
Shared Package
ML15191A175	List: CP-201500668, License Amendment Request 15-003 for Revision to Units 1 & 2 Emergency Action Levels ML15191A163 ML15191A164
References
CP-201500668, TXX-15101
Download: ML15191A164 (289)

Text

ATTACHMENT 4 TO TXX-15101 EMERGENCY ACTION LEVEL TECHNICAL BASES (REDLINE AND STRIKEOUT VERSION)

(276 PAGES)

Luminant Comanche Peak Nuclear Power Plant EPP-201 Emergency Action Level Technical Bases Document Prepared by:

Print Name Signature Date Technical Reviewer:

Print Name Signature Date Reviewer:

Print Name Signature Date Approval:

Print Name Signature Date Effective Date:

DRAFT D2 6/24/15 I Page 1 of 276 I

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 2.7 Unit Designation .......................................................................................................... 8 3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS .......................................... 9 3.1 General Considerations .................................................................................................. 9 3.2 Classification Methodology ........................................................................................... 10

4.0 REFERENCES

..................................................................................................................... 13 4.1 Developm ental ................................................................................................................... 13 4.2 Im plementing ..................................................................................................................... 13 5.0 DEINITIONS, ACRONYMS & ABBREVIATIONS ............................................................. 14 6.0 CPNPP TO NEI 99-01 Rev. 6 EAL CROSS-REFERENCE .............................................. 20 7.0 ATTACHM ENTS ................................................................................................................... 24 1 Emergency Action Level Technical Bases ........................................................... 25 Cate-gory R Abnormal Rad Release / Rad Effluent .................................... 25 Category E ISFSI ...................................................................................... 65 Cate-gory C Cold Shutdown / Refueling System Malfunction ...................... 68 Category H Hazards ...................................................................................... 112 Cate-gory S System Malfunction ........... ;........................................................ 155 Cate-gory F Fission Product Barrier Degradation .......................................... 200 2 Fission Product Barrier Loss / Potential Loss Matrix and Bases .................................................................................................... 205 3 Safe Operation & Shutdown Areas Tables R-3 & H-2 Bases ................................. 258 Page 2 of 276

1.0 PURPOSE This document provides an explanation and rationale for each Emergency Action Level (EAL) included in the EAL Upgrade Project for Comanche Peak Nuclear Power Plant (CPNPP). It should be used to facilitate review of the CPNPP EALs and provide historical documentation for future reference. Decision-makers responsible for implementation of EPP-201, "Assessment of Emergency Action Levels, Emergency Classification and Plan Activation,"

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 CFR 50.54(q).

2.0 DISCUSSION 2.1 Background EALs are the plant-specific indications, conditions or instrument readings that are utilized to classify emergency conditions defined in the CPNPP Plant Radiological Emergency Response Plan (RERP).

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), CPNPP conducted an EAL implementation upgrade project that produced the EALs discussed herein Page 3 of 276

2.2 Fission Product Barriers Fission product barrier thresholds represent threats to the defense in depth design concept that precludes the release of radioactive fission products to the environment. This concept relies on multiple physical barriers, any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment.

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

The primary fission product barriers are:

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

B. Reactor Coolant System (RCS): The RCS Barrier includes the RCS primary side and its connections up to and including the pressurizer safety and relief valves, and other connections up to and including the primary isolation valves.

C. Containment (CNTMT): 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 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 potentialloss of either Fuel Clad or RCS barrier Site Area Emergency:

Loss or potentialloss of any two barriers General Emergency:

Loss of any two barriersand loss or potential loss of the third barrier Page 4 of 276

2.4 EAL Organization The CPNPP EAL scheme includes the following features:

0 Division of the EAL set into three broad groups:

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

o EALs applicable only under hot operating modes - This group would only be reviewed by the EAL-user when the plant is in Hot Shutdown, 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.

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

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EAL Groups, Categories and Subcategories EAL Group/Category EAL Subcategory Any Operating Mode:

R - Abnormal Rad Levels / Rad Effluent I - Radiological Effluent 2 - Irradiated Fuel Event 3 - Area Radiation Levels H - Hazards and Other Conditions 1- Security Affecting Plant Safety 2- Seismic Event 3- Natural or Technological Hazard 4- Fire 5- Hazardous Gases 6- Control Room Evacuation 7- Emergency Coordinator Judgment E - ISFSI I - Confinement Boundary Hot Conditions:

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

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

I Page 6 of 276

2.5 Technical Bases Information EAL technical bases are provided in Attachment 1 for each EAL according to EAL group (Any, Hot, Cold), EAL category (R, C, H, S, E and F) 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 Subcategiorv Number & Title Initiating Condition (IC)

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

EAL Identifier (enclosed in rectangle)

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

1. First character (letter): Corresponds to the EAL category as described above (R, C, H, S, E or F)

2. Second character (letter): The emergency classification (G, S, A or U)

G = General Emergency S = Site Area Emergency A = Alert U = Unusual Event

3. Third character (number): Subcategory number within the given category.

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

4. Fourth character (number): The numerical sequence of the EAL within the EAL subcategory. If the subcategory has only one EAL, it is given the number one (1).

Classification (enclosed in rectangle):

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

EAL (enclosed in rectangle)

Exact wording of the EAL as it appears in the EAL Classification Matrix Page 7 of 276

Mode Applicability One or more of the following plant operating conditions comprise the mode to which each EAL is applicable: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 -

Cold Shutdown, 6 - Refueling, D - Defueled, or Any. (See Section 2.6 for operating mode definitions)

Definitions:

If the EAL wording contains a defined term, the definition of the term is included in this section. These definitions can also be found in Section 5.1.

Basis:

A Plant-Specific basis section that provides CPNPP-relevant information concerning the EAL. This is followed by a Generic basis section that provides a description of the rationale for the EAL as provided in NEI 99-01 Rev. 6.

CPNPP Basis Reference(s):

Site-specific source documentation from which the EAL is derived.

2.6 Operating Mode Applicability (ref. 4.1.8) 1 Power Operation Keff greater than or equal to 0.99 and reactor thermal power greater than 5%

2 Startup Keff greater than or equal t 0.99 and reactor thermal power-< 5%

3 Hot Standby Keff less than 0.99 and average coolant temperature greater than or equal to 3501F 4 Hot Shutdown Keff less than 0.99 and average coolant temperature 3501F greater than Tavg greater than 200 OF and all reactor vessel head closure bolts fully tensioned 5 Cold Shutdown Keff less than 0.99 and average coolant temperature < 200OF 6 Refuelinq One or more reactor vessel head closure bolts are less than fully tensioned D Defueled All reactor fuel removed from reactor pressure vessel (full core off load during refueling or extended outage).

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.

2.7 Unit Designation The specific unit designator (1 or 2) is represented within these instructions by the symbol "u".

The appropriate unit digit may be substituted for this symbol to obtain the unit specific equipment number (Example u-FK-121 represents 1-FK-121 for Unit 1 and 2-FK-121 for Unit 2). For equipment or components that are common or non unit-specific the "X" designator is used. (Example X-RE-6272 represents a radiation monitor that is common to both units).

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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. 4.1.11).

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).

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3.1.5 Classification Based on Analysis The assessment of some EALs is based on the results of analyses that are necessary to ascertain whether a specific EAL threshold has been exceeded (e.g., dose assessments, chemistry sampling, RCS leak rate calculation, etc.). For these EALs, the EAL wording or the associated basis discussion will identify the necessary analysis. In these cases, the 15-minute declaration period starts with the availability of the analysis results that show the threshold to be exceeded (i.e., this is the time that the EAL information is first available). The NRC expects licensees to establish the capability to initiate and complete EAL-related analyses within a reasonable period of time (e.g., maintain the necessary expertise on-shift).

3.1.6 Emergency 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. 4.1.11).

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, Clarificationof NRC Guidance for Emergency Notifications During Quickly ChangingEvents (ref. 4.1.2).

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3.2.2 Consideration of Mode Changes During Classification The mode in effect at the time that an event or condition occurred, and prior to any plant or operator response, is the mode that determines whether or not an IC is applicable. If an event or condition occurs, and results in a mode change before the emergency is declared, the emergency classification level is still based on the mode that existed at the time that the event or condition was initiated (and not when it was declared). Once a different mode is reached, any new event or condition, not related to the original event or condition, requiring emergency classification should be evaluated against the ICs and EALs applicable to the operating mode at the time of the new event or condition.

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

3.2.3 Classification of Imminent Conditions Although EALs provide specific thresholds, the Emergency 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.

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

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

An ATWS occurs and the high pressure ECCS systems fail to automatically start. RPV level rapidly decreases and the plant enters an inadequate core cooling condition (a potential loss of both the fuel clad and RCS barriers). If an operator manually starts a high pressure ECCS system in accordance with an EOP step and clears the inadequate core cooling condition prior to an emergency declaration, then the classification should be based on the ATWS only.

It is important to stress that the 15-minute emergency classification assessment period (process clock) is not a "grace period" during which a classification may be delayed to allow the performance of a corrective action that would obviate the need to classify the event.

Emergency classification assessments must be deliberate and timely, with no undue delays.

The provision discussed above addresses only those rapidly evolving situations when an operator is able to take a successful corrective action prior to the Emergency Coordinator completing the review and steps necessary to make the emergency declaration. This provision is included to ensure that any public protective actions resulting from the emergency classification are truly warranted by the plant conditions.

3.2.7 After-the-Fact Discovery of an Emergency Event or Condition In some cases, an EAL may be met but the emergency classification was not made at the time of the event or condition. This situation can occur when personnel discover that an event or condition existed which met an EAL, but no emergency was declared, and the event or condition no longer exists at the time of discovery. This may be due to the event or condition not being recognized at the time or an error that was made in the emergency classification process.

In these cases, no emergency declaration is warranted; however, the guidance contained in NUREG-1022 (ref. 4.1.3) is applicable. Specifically, the event should be reported to the NRC in accordance with 10 CFR § 50.72 (ref. 4.1.4) within one hour of the discovery of the undeclared event or condition. The licensee should also notify appropriate State and local agencies in accordance with the agreed upon arrangements.

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

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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-1022 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 § CFR 50.73 License Event Report System 4.1.6 CPNPP Emergency Plan Appendix E, Complex and Owner Controlled Area 4.1.7 CPNPP FSAR Section 2.1.1 Site Location and Description 4.1.8 Technical Specifications Table 1.1-1 Modes 4.1.9 OPT-408A/B Refueling Containment Penetration Verification 4.1.10 ODA-207 Guidelines on the Preparation and Review of Operations Procedures 4.1.11 NSIR/DPR-ISG-01 Interim Staff Guidance, Emergency Planning for Nuclear Power Plants 4.1.12 IPO-01OA/B Reactor Coolant System Reduced Inventory Operations 4.1.13 Technical Specifications 3.9.4 4.1.14 CPNPP Offsite Dose Calculation Manual (ODCM) 4.2 Implementing 4.2.1 EPP-201, Assessment of Emergency Action Levels, Emergency Classification and Plan Activation 4.2.2 NEI 99-01 Rev. 6 to CPNPP EAL Comparison Matrix 4.2.3 CPNPP EAL Matrix Page 13 of 276

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 process, 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.

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. Containment closure means that all potential escape paths are closed or capable of being closed (ref.4.1.13).

A. All penetrations providing direct access from Containment atmosphere to outside atmosphere are closed except:

" Penetrations with automatic valves capable of being closed by an operable CVI

" Penetrations under administrative controls (e.g., Control Room notified and designated person to close if required by fuel handling accident)

B. Equipment hatch is closed and held in place by 4 bolts, or is capable of being closed and held in place by 4 bolts C. One emergency airlock door is closed D. One personnel airlock door is capable of being closed (ref. 4.1.9)

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

Exclusion Area Boundary Exclusion Area Boundary is a synonymous term for Site Boundary. CPNPP FSAR Section 2.1.1.3 and Figure 2.1-2 define the Exclusion Area Boundary. This boundary is used for establishing effluent release limits with respect to the requirements of IOCFR20 (ref. 4.1.7).

See also CPNPP Emergency Plan Appendix E, Complex and Owner Controlled Area (ref.

4.1.6) and CCNPP ODCM Section 5.0 Design Features (ref. 4.1.14).

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 Page 14 of 276

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.

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.

Floodina A condition where water is enteringq 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 process 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 CPNPP 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 CPNPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area).

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

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

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).

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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.

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

Normal Levels Sappliod to radiological lIdEALs, the highest reading in the past t'A'cty-foue or excluding th.e current peak value.

Owner Controlled Area As shown in CPNPP Emergency Plan Appendix E, Complex and Owner Controlled Area.

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 FSAR Figure 1.2-1 Plot Plan (ref. 4.1.7).

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

Reduced Inventory Plant condition when fuel is in the reactor vessel and Reactor Coolant System level is < 80 inches above core plate (829'8") (ref. 4.1.12).

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 IOCFR50.2):

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

Page 16 of 276

(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 process or have occurred which involve actual or likely maior failures of plant functions needed for protection of the public or hostile actions that result 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 Guidelines exposure levels beyond the site boundary.

Site Boundary See EXCLUSION AREA BOUNDARY Unisolable An open or breached system line that cannot be isolated, remotely or locally.

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

Unusual Event Events are in process or have occurred which indicate a potential degradation in the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.

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

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.

Page 17 of 276

5.2 Abbreviations/Acronyms OF ....................................................................................................... Degrees Fahrenheit 0 .......................................................................................................................... De gre es AC ....................................................................................................... Alternating Current APDG .......................................................................... Alternate Power Diesel Generator ATW S ...................................................................... Anticipated Transient W ithout Scram CPNPP .................................................................. Com anche Peak Nuclear Power Plant CDE ....................................................................................... Com m itted Dose Equivalent CFR ..................................................................................... Code of Federal Regulations CNTM T .......................................................................................................... Containm ent CSFST ....................................................................... Critical Safety Function Status Tree DBA ............................................................................................... Design Basis Accident DC ............................................................................................................... Direct Current EAL ............................................................................................. Em ergency Action Level ECCS ............................................................................ Em ergency Core Cooling System ECL ................................................................................. Em ergency Classification Level EO F .................................................................................. Em ergency O perations Facility EO P ............................................................................... Em ergency Operating Procedure EPA .............................................................................. Environm ental Protection Agency ERG ................................................................................ Em ergency Response G uideline EPIP ................................................................ Em ergency Plan Im plem enting Procedure ESF ........................................................................................ Engineered Safety Feature ESW ........................................................................................ Em ergency Service Water FAA ................................................................................. Federal Aviation Adm inistration FBI ................................................................................... Federal Bureau of Investigation FEMA ............................................................... Federal Em ergency Managem ent Agency FSAR .................................................................................... Final Safety Analysis Report G E ..................................................................................................... General Em ergency IC ......................................................................................................... Initiating Condition IPEEE ................. Individual Plant Examination of External Events (Generic Letter 88-20)

Keff ......................................................................... Effective Neutron Multiplication Factor LCO .................................................................................. Lim iting Condition of Operation LER ............................................................................................... Licensee Event Report LOCA ......................................................................................... Loss of Coolant Accident LW R ................................................................................................... Light W ater Reactor MPC ................................... Maxim um Perm issible Concentration/M ulti-Purpose Canister m R, m Rem , m rem , m REM .............................................. m illi-Roentgen Equivalent Man MSL ........................................................................................................ Main Steam Line MW .................................................................................................................... Megawatt NEI ........................................................ ..................................... Nuclear Energy Institute Page 18 of 276

NESP ................................................................... National Environm ental Studies Project NPP .................................................................................................. Nuclear Power Plant NRC ................................................................................ Nuclear Regulatory Com m ission NSSS ................................................................................ Nuclear Steam Supply System NO RAD ................................................... North Am erican Aerospace Defense Com m and (NO )UE ................................................................................ Notification of Unusual Event O BE ...................................................................................... O perating Basis Earthquake O CA ............................................................................................... Owner Controlled Area O DCM ............................................................................ Off-site Dose Calculation Manual O RO ................................................................................. Offsite Response Organization OTO ............................................................................... Off-Norm al O perating Procedure PA .............................................................................................................. Protected Area PAG ........................................................................................ Protective Action Guideline PRA/PSA ..................... Probabilistic Risk Assessment / Probabilistic Safety Assessment PW R ....................................................................................... Pressurized W ater Reactor PSIG ................................................................................ Pounds per Square Inch Gauge R ........................................................................................................................ Roentgen RCC ............................................................................................ Reactor Control Console RCS ............................................................................................ Reactor Coolant System Rem , rem , REM ....................................................................... Roentgen Equivalent Man RETS ......................................................... Radiological Effluent Technical Specifications RPS ....................................................................................... Reactor Protection System R(P)V ....................................................................................... Reactor (Pressure) Vessel RVLIS ................................................................. Reactor Vessel Level Indicating System SAR ............................................................................................... Safety Analysis Report SBO ......................................................................................................... Station Blackout SCBA ....................................................................... Self-Contained Breathing Apparatus SG ......................................................................................................... Steam Generator SI .............................................................................................................. Safety Injection O DCM ............................................................................. Offsite Dose Calculation Manual SPDS ........................................................................... Safety Param eter Display System SRO ............................................................................................ Senior Reactor O perator TEDE ............................................................................... Total Effective Dose Equivalent TOAF .................................................................................................... Top of Active Fuel TSC ... ...................................................................................... Technical Support Center W O G ................................................................................... W estinghouse Ow ners Group I Page 19 of 276

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

CPNPP NEI 99-01 Rev. 6 EAL IC Example EAL RU1.1 AU1 1,2 RU1.2 AU1 3 RU2.1 AU2 I RA1.1 AA1 1 RA1.2 AA1 2 RAU.3 AA1 3 RA1.4 AA1 4 RA2.1 AA2 1 RA2.2 AA2 2 RA2.3 AA2 3 RA3.1 AA3 1 RA3.2 AA3 2 RS1.1 AS1 1 RS1.2 AS1 2 RS1.3 AS1 3 RS2.1 AS2 1 RGI.1 AG1 1 RG1.2 AG1 2 RG1.3 AG1 3 RG2.1 AG2 1 CU1.1 CU1 1 Page 20 of 276

CPNPP NEI 99-01 Rev. 6 EAL IC Example EAL CU1.2 CUI 2 CU2.1 CU2 1 CU3.1 CU3 1 CU3.2 CU3 2 CU4.1 CU4 1 CU5.1 CU5 1,2,3 CAI.1 CA1 1 CAI.2 CA1 2 CA2.1 CA2 1 CA3.1 CA3 1,2 CA6.1 CA6 1 CS1.1 CS1 1 CS1.2 CS1 2 CS1.3 CS1 3 CGI.1 CG1 2 FAI.1 FA1 I FSI.1 FS1 1 FG1.1 FG1 1 HU1.1 HU1 1 HU1.2 HUI 2 HU1.3 HUI 3 HU2.1 HU2 I HU3.1 HU3 1 HU3.2 HU3 2 HU3.3 HU3 3 Page 21 of 276

CPNPP NEI 99-01 Rev. 6 EAL IC Example EAL HU3.4 HU3 4 HU4.1 HU4 1 HU4.2 HU4 2 HU4.3 HU4 3 HU4.4 HU4 4 HU7.1 HU7 1 HA1.1 HAl 1 HA1.2 HAl 2 HA5.1 HA5 I HA6.1 HA6 1 HA7.1 HA7 1 HSl.1 HS1 I HS6.1 HS6 1 HS7.1 HS7 1 HG1.1 HG1 1 HG7.1 HG7 1 su1.1 Sul 1 SU3.1 SU2 1 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 Page 22 of 276

CPNPP NEI 99-01 Rev. 6 EAL IC Example EAL SA1.1 SA1 1 SA3.1 SA2 1 SA6.1 SA5 1 SA9.1 SA9 1 SS1.1 SS1 1 SS2.1 SS8 1 SS6.1 SS5 1 SGl.1 SG1 1 SG1.2 SG8 1 EUI.1 E-HU1 1 Page 23 of 276

7.0 ATTACHMENTS 7.1 Attachment 1, Emergency Action Level Technical Bases 7.2 Attachment 2, Fission Product Barrier Matrix and Basis Page 24 of 276

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. Radiological Effluent Direct indication of effluent radiation monitoring systems provides a rapid assessment mechanism to determine releases in excess of classifiable limits. Projected offsite doses, actual offsite field measurements or measured release rates via sampling indicate doses or dose rates above classifiable limits.

2. Irradiated Fuel Event Conditions indicative of a loss of adequate shielding or damage to irradiated fuel may preclude access to vital plant areas or result in radiological releases that warrant emergency classification.

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

I Page 25 of 276 1

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 ODCM limits for 60 minutes or longer EAL:

RUI.1 Unusual Event Reading on any Table R-1 effluent radiation monitor greater than column "UE" for greater than or equal to 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 Monitor GE SAE Alert UE Plant Vent X-RE-5567 ............ 6.52E-4 1Cic/,ml PVG384 + PVG385 A+ B Plant Vent (WRGM)X-RE-5570 (WRGM) A+ B 4.0E+7 ý.Ci/sec 4.OE+6 p.Ci/sec 4.0E+5 pCi/sec 4.OE+4 iiCi/sec 0PVF684 + PVF685 S Main Steam MSLu78 u-RE-2325 MSLu79 u-RE-2326 90 pici/ml* 9.0 p.Cil/ml* 0.9 t4ci/ml* 2 x high alarm u-RE-2327 setpoint*

MSLu80 MSLu81 u-RE-2328 Liquid Waste X-RE-5253 2 x high alarm LWE-076 setpoint H* Service Water "J SSei65 u-RE-4269 2 x high alarm u-RE-4270 setpoint SSWu66

• with reactor shutdown Mode Applicability:

All I Page 26 of 276

ATTACHMENT 1 EAL Bases Definition(s):

None CPNPP Basis:

The column "UE" gaseous and liquid release values in Table R-1 represent two times the alarm setpoint of the specified monitors. The setpoints are established to ensure the ODCM release limits are not exceeded. (ref. 1)

Plant Vent Monitors sample both plant vent stacks prior to discharge to the environment. They detect normal operational levels of noble gases. The noble gas detectors (X-RE-5567A, B) can be used as backups to the wide range gas monitors (X-RE-5570A, B). These monitors communicate with the RM-23s in the Control Room. Indication and annunciation are provided in the Control Room for alert and high radiation levels and monitor failure. (ref. 2)

The WRGM system is a gaseous effluent monitoring system composed of two identical monitors used for detection of noble gas releases through the two plant vent stacks. Exhaust from the main turbine gland steam condenser exhauster is routed to the vent stacks for monitoring prior to release. Particulate and iodine grab samples may also be obtained from the WRGM. These monitors also initiate the automatic closure of the gas release valve in the waste gas processing system on detection of high radiation. Indication and annunciation are provided in the Control Room for alert and high radiation levels and monitor failure. (ref. 2)

There are four online Main Steam Line Monitors (MSL) for each steam generator. Each one consists of a shielded, Category II seismic detector mounted adjacent to a main steam line, a remote RM-80 microprocessor and a remote customer interface junction box. The RM-80 associated with the MSL monitor communicates with the PC-1I1 CRT console computer. (ref.

2).

Plant Liquid Waste Processing System (LWPS) discharge is continuously monitored by a shielded gamma sensitive (Nal(T1)) scintillation detector. When a LWPS discharge is required, normally locked-closed control valves can be opened directing flow through a path containing a radiation monitor (X-RE-5253) and a control valve which discharges waste to the circulating water discharge tunnel. The control valves are administratively controlled with a key-operated switch selectable to closed, automatic, or "key-held" open modes. In the automatic position, the valve will close on monitor high radiation alarm or monitor failure signals. Indication and annunciation are provided on the Waste Processing System (WPS) control panel for alert or monitor failure alarm and in the Control Room for alert, high, and monitor failure alarms. (ref. 2)

Service Water monitors are provided to monitor the Service Water System for radiation since leakage from radioactive fluid systems could cause potential radioactive leakage to the environment. A shielded gamma sensitive scintillation (Nal(T1)) detector is located in an off-line sample assembly downstream of each component cooling water heat exchanger to monitor service water being discharged. Indication and annunciation are provided at the Control Room RMS console. (ref. 2)

NEI 99-01 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, Page 27 of 276

ATTACHMENT I EAL Bases monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.

Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls.

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

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

Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL.

I EAL--#-I-This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways.

EAL #t2 This EAL addergenscs radioactioity leases that cause c-lucRt radiation monito readings to oxceed 2 times, the limit established by a radioactivity discharge permit. This EAL Will typically be associated with planned batch rcleases fErom non continuous release pathways (e.g., r-adwaste, waste gas).

EAL= 13 This E=AL addresses uncontrollcd gaseous or liquid relcascs that are detected by sample analyses or enviMromental sur~eys, particularly on unmonitored pathways (e.g., spoills of radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.)-.

Escalation of the emergency classification level would be via IC AAI-RPAI.

CPNPP Basis Reference(s):

1. CPNPP ODCM Unit 1 and 2

2. DBD-EE-023 Radiation Monitoring System

3. EAL Section R Revision 6 Table R-1 Effluent Monitor Classification Thresholds Review

4. NEI 99-01 AU1 Page 28 of 276 1

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 ODCM 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 ODCM limits for greater than or equal to 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 CPNPP Basis:

None NEI 99-01 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.

Cla~ssfication basecd on cfflucnt mon~itor readings assumes that a release path to h envionmet is established. if the effluent flo)w past an effluent monitor is known to havo stopped due to actions to isolatc the release path, thcn the effluent moniRtorF reading is no longer valid fra f purposes.

Page 29 of 276

ATTACHMENT 1 EAL Bases Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL.

EAL #1 This E.A.L Addresses

, .eleas' radioactiVity normally occurring continuous . from monioredgaseous or liquid effluent pathways.

EAL-- -2This EAL addresses radioactiVity releases that c-ause -ffluent rditomntr readings to exceed 2 times the li~mit esta-bli-shed by a radioactiv~ity discharge permit. This ,A~L Wil typically be asSocGiate ithpland-bhatch releases from non continuous release pathways (e.g., radwaste, waste gas).

EAL-#3 -- 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 AA4-RA1.

CPNPP Basis Reference(s):

1. CPNPP ODCM Unit 1 and 2

2. NEI 99-01 AU1 Page 30 of 276

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:

RA1.1 Alert Reading on any Table R-1 effluent radiation monitor greater than column "ALERT" for greater than or equal to 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 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 Plant Vent X-RE-5567 ............. 6.52E-4 pCi/ml PVG384 + PVG385 A+ B Plant Vent (WRGM)X-RE-5570 (WRGM) X- A + B7 4.OE+7 ýCi/sec 4.OE+6 *ci/sec 4.OE+5 pci/sec 4.OE+4 pCi/sec 0 PVF684 + PVF685

' Main Steam MSLu78 u-RE-2325 MSLu79 u-RE-2326 90 pci/ml* 9.0 pci/ml* 0.9 j, ci/ml* 2xhigh alarm MSLu8O u-RE-2327 MSLu81 u-RE-2328 Liquid Waste X-RE-5253 2 x high alarm LWE-076 setpoint M Servic Water u-RE-4269 2 x high alarm

"* SSWu65 u-RE-4270 setpoint SSWu66

• with reactor shutdown Mode Applicability:

All I Page 31 of 276

ATTACHMENT 1 EAL Bases Definition(s):

None CPNPP Basis:

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

0 10 mRem TEDE

• 50 mRem CDE 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 CDE Thyroid) (ref. 1).

NEI 99-01 Basis:

This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).

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

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

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

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

CPNPP Basis Reference(s):

1. EAL Section R Revision 6 Table R-1 Effluent Monitor Classification Thresholds Review

2. NEI 99-01 AA1 Page 32 of 276

ATTACHMENT I 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:

RA1.2 Alert Dose assessment using actual meteorology indicates doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the EXCLUSION AREA BOUNDARY (Notes 3, 4)

Note 3: Ifthe effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.

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

EXCLUSION AREA BOUNDARY- Exclusion Area Boundary is a synonymous term for Site Boundary. CPNPP FSAR Section 2.1.1.3 and Figure 2.1-2 define the Exclusion Area Boundary. This boundary is used for establishing effluent release limits with respect to the requirements of 10CFR20. See also CPNPP Emergency Plan Appendix E, Complex and Owner Controlled Area and CCNPP ODCM Section 5.0 Design Features.

CPNPP Basis:

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

NEI 99-01 Basis:

This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).

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

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

Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have Page 33 of 276

ATTACHMENT 1 EAL Bases stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.

Escalation of the emergency classification level would be via IC AS-14RSI.

CPNPP Basis Reference(s):

1. EPP-303 Operation of Computer Based Dose Assessment System

2. NEI 99-01 AA1 Page 34 of 276

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: I - 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:

RA1.3 Alert Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the EXCLUSION AREA 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):

EXCLUSION AREA BOUNDARY - Exclusion Area Boundary is a synonymous term for Site Boundary. CPNPP FSAR Section 2.1.1.3 and Figure 2.1-2 define the Exclusion Area Boundary. This boundary is used for establishing effluent release limits with respect to the requirements of 10CFR20. See also CPNPP Emergency Plan Appendix E, Complex and Owner Controlled Area and CCNPP ODCM Section 5.0 Design Features.

CPNPP Basis:

Dose assessments based on liquid releases are performed per Offsite Dose Calculation Manual (ref. 1).

NEI 99-01 Basis:

This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).

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

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

Page 35 of 276

ATTACHMENT 1 EAL Bases 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 AS4RS1.

CPNPP Basis Reference(s):

1. CPNPP Offsite Dose Calculation Manual

2. NEI 99-01 AA1 Page 36 of 276

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels I 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:

RA1.4 Alert Field survey results indicate EITHER of the following at or beyond the EXCLUSION AREA BOUNDARY:

" Closed window dose rates greater than 10 mR/hr expected to continue for greater than or equal to 60 min.

" Analyses of field survey samples indicate thyroid CDE greater than 50 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):

EXCLUSION AREA BOUNDARY- Exclusion Area Boundary is a synonymous term for Site Boundary. CPNPP FSAR Section 2.1.1.3 and Figure 2.1-2 define the Exclusion Area Boundary. This boundary is used for establishing effluent release limits with respect to the requirements of 10CFR20. See also CPNPP Emergency Plan Appendix E, Complex and Owner Controlled Area and CCNPP ODCM Section 5.0 Design Features.

CPNPP Basis:

EPP-309 Onsite/In-Plant Radiological Surveys and Offsite Radiological Monitoring provides guidance for emergency or post-accident radiological environmental monitoring (ref. 1).

NEI 99-01 Basis:

This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses Page 37 of 276

ATTACHMENT 1 EAL Bases the spectrum of possible accident events and conditions.

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

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

Escalation of the emergency classification level would be via IC AS4RSI.

CPNPP Basis Reference(s):

1. EPP-309 Onsite/In-Plant Radiological Surveys and Offsite Radiological Monitoring

2. NEI 99-01 AA1 Page 38 of 276 1

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:

RSI.1 Site Area Emergency Reading on any Table R-1 effluent radiation monitor greater than column "SAE" for greater than or equal to 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 RAI.1, RSI.1 and RGI.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE Plant Vent X-RE-5567 PVG384 + PVG385 A+B Plant Vent (WRGM)X-RE-5570 (WRGM)

+- A + B7 4.OE+7 [.Ci/sec 4.OE+6 ýtCi/sec 4.0E+5 pCi/sec 4.0E+4 p.Ci/sec 0 PVF684 + PVF685 C Main Steam MSLu78 u-RE-2325 MSLu79 u-RE-2326 90 [Ci/ml* 9.0 [.ci/ml* 0.9 j.ci/ml* 2 x high alarm setpoint*

MSLu80 u-RE-2327 MSLu81 u-RE-2328 Liquid Waste X-RE-5253 ...............- 2 x high alarm LWE-076 setpoint r" Service Water

• Srcu-RE-4269 x high alarm 22x.hgh.lar S SSWu65 ----.--.-----.

SSw65 u-RE-4270 setpoint SSWu66

• with reactor shutdown Mode Applicability:

All Definition(s):

Page 39 of 276

ATTACHMENT 1 EAL Bases None CPNPP 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 mRem TEDE 0 500 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).

NEI 99-01 Basis:

This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.

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

The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

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

Escalation of the emergency classification level would be via IC AG4RGI.

CPNPP Basis Reference(s):

1. EAL Section R Revision 6 Table R-1 Effluent Monitor Classification Thresholds Review

2. NEI 99-01 AS1 I Page 40 of 276

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 100 mrem TEDE or 500 mrem thyroid CDE EAL:

RS1,2 Site Area Emergency Dose assessment using actual meteorology indicates doses greater than 100 mrem TEDE or 500 mrem thyroid CDE at or beyond the EXCLUSION AREA BOUNDARY (Notes 3, 4)

Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.

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

Mode Applicability:

All Definition(s):

EXCLUSION AREA BOUNDARY - Exclusion Area Boundary is a synonymous term for Site Boundary. CPNPP FSAR Section 2.1.1.3 and Figure 2.1-2 define the Exclusion Area Boundary. This boundary is used for establishing effluent release limits with respect to the requirements of 10CFR20. See also CPNPP Emergency Plan Appendix E, Complex and Owner Controlled Area and CCNPP ODCM Section 5.0 Design Features.

CPNPP Basis:

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

NEI 99-01 Basis:

This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.

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

The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no I Page 41 of 276

ATTACHMENT 1 EAL Bases longer valid for classification purposes.

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

CPNPP Basis Reference(s):

1. EPP-303 Operation of Computer Based Dose Assessment System

2. NEI 99-01 AS1 Page 42 of 276

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 EXCLUSION AREA BOUNDARY:

" Closed window dose rates greater than 100 mR/hr expected to continue for greater than or equal to 60 min.

" Analyses of field survey samples indicate thyroid CDE greater than 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):

EXCLUSION AREA BOUNDARY - Exclusion Area Boundary is a synonymous term for Site Boundary. CPNPP FSAR Section 2.1.1.3 and Figure 2.1-2 define the Exclusion Area Boundary. This boundary is used for establishing effluent release limits with respect to the requirements of 10CFR20. See also CPNPP Emergency Plan Appendix E, Complex and Owner Controlled Area and CCNPP ODCM Section 5.0 Design Features.

CPNPP Basis:

EPP-309 Onsite/In-Plant Radiological Surveys and Offsite Radiological Monitoring provides guidance for emergency or post-accident radiological environmental monitoring (ref. 1).

NEI 99-01 Basis:

This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.

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

Page 43 of 276

ATTACHMENT 1 EAL Bases The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

ClassificAtion b-a*,sed, on effluent monitr* readings as.sum.es. that a releaso path to the enViGronment is established. if the cifluent flo)W past an effluent moenitor is kno)Wn to haVo stepped due to- ac-tions to isolate the relcasc path, then the enffluent monitor rea;ding is no longer valid for classification purposes.

Escalation of the emergency classification level would be via IC AG4-RGI.

CPNPP Basis Reference(s):

1. EPP-309 Onsite/In-Plant Radiological Surveys and Offsite Radiological Monitoring

2. NEI 99-01 AS1 Page 44 of 276

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 CDE EAL:

RGI.1 General Emergency Reading on any Table R-1 effluent radiation monitor greater than column "GE" for greater than or equal to 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 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.

Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE Plant Vent X-RE-5567 ............. 6.52E-4 pCi/ml PVG384 + PVG385 A+ B Plant Vent (WRGM)X-RE-5570 S(WRGM) A+ B X-E-7 4.OE+7 luCi/sec 4.OE+6 liCi/sec 4.OE+5 [iCi/sec 4.OE+4 pCi/sec 0 PVF684 + PVF685 Main Steam MSLu78 u-RE-2325 MSLu79 u-RE-2326 90 p.ci/ml* 9.0 ILci/ml* 0.9 pLCi/ml* 2 x high alarm setpoint*

MSLu80 u-RE-2327 MSLu81 u-RE-2328 Liquid Waste X-RE-5253 ...............- 2 x high alarm LWE-076 setpoint "7 Service Water u-RE-4269 ----..-------.. 2 xx.hgh.lar high alarm

"*SSWu65 SSu65 u-RE-4270 setpoint SSWu66

• with reactor shutdown Mode Applicability:

All Page 45 of 276

ATTACHMENT 1 EAL Bases Definition(s):

None CPNPP 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 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).

NEI 99-01 Basis:

This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.

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

The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

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

CPNPP Basis Reference(s):

1. EAL Section R Revision 6 Table R-1 Effluent Monitor Classification Thresholds Review

2. NEI 99-01 AG1 I Page 46 of 276

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 CDE EAL:

RGI.2 General Emergency Dose assessment using actual meteorology indicates doses greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE at or beyond the EXCLUSION AREA BOUNDARY (Notes 3, 4)

Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.

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

EXCLUSION AREA BOUNDARY- Exclusion Area Boundary is a synonymous term for Site Boundary. CPNPP FSAR Section 2.1.1.3 and Figure 2.1-2 define the Exclusion Area Boundary. This boundary is used for establishing effluent release limits with respect to the requirements of 10CFR20. See also CPNPP Emergency Plan Appendix E, Complex and Owner Controlled Area and CCNPP ODCM Section 5.0 Design Features.

CPNPP Basis:

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

NEI 99-01 Basis:

This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.

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

The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have Page 47 of 276

ATTACHMENT 1 EAL Bases stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.

CPNPP Basis Reference(s):

1. EPP-303 Operation of Computer Based Dose Assessment System

2. NEI 99-01 AG1 Page 48 of 276

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 CDE EAL:

RG1.3 General Emergency Field survey results indicate EITHER of the following at or beyond the EXCLUSION AREA BOUNDARY:

" Closed window dose rates greater than 1,000 mR/hr expected to continue for greater than or equal to 60 min.

" Analyses of field survey samples indicate thyroid CDE greater than 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):

EXCLUSION AREA BOUNDARY- Exclusion Area Boundary is a synonymous term for Site Boundary. CPNPP FSAR Section 2.1.1.3 and Figure 2.1-2 define the Exclusion Area Boundary. This boundary is used for establishing effluent release limits with respect to the requirements of 10CFR20. See also CPNPP Emergency Plan Appendix E, Complex and Owner Controlled Area and CCNPP ODCM Section 5.0 Design Features.

CPNPP Basis:

EPP-309 Onsite/In-Plant Radiological Surveys and Offsite Radiological Monitoring provides guidance for emergency or post-accident radiological environmental monitoring (ref. 1).

NEI 99-01 Basis:

This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.

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

Page 49 of 276

ATTACHMENT 1 EAL Bases The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

Classification based on effluent monitor readings assumes that a rolcase path to the envionmet is established. if the effluent flow past an cifluent monitor is known to ae stopped duo to cin to- isolate the release path, then the effluont moenitor reading is no lo~nger valid for clasSiffication purposes.

CPNPP Basis Reference(s):

1. EPP-309 Onsite/In-Plant Radiological Surveys and Offsite Radiological Monitoring

2. NEI 99-01 AG1 Page 50 of 276 1

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 Table R-2 area radiation monitors Table R-2 SFP & Refueling Cavity Area Radiation Monitors SFP:

• SFP001, LRAM SFP 2 E WALL (X-RE-6272)

" SFP002, LRAM SFP 2 N WALL (X-RE-6273)

" SFP003, LRAM SFP 1 E WALL (X-RE-6274)

• SFP004, LRAM SFP 1 S WALL (X-RE-6275)

Refueling Cavity:

• RFCulO, LRAM W REFUEL CAV860 (u-RE-6251)

• RFCu12, LRAM E REFUEL CAV 860 (u-RE-6253)

Mode Applicability:

All Definition(s):

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

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

CPNPP Basis:

Indication of decreasing level includes Spent Fuel Pool Panel (FB 810) (ref. 2):

" RFL CAVITY 1 LEVEL LO (upender area) (SFP-3.3)

• RFL CAVITY 2 LEVEL LO (upender area) (SFP-3.7)

" SFP 1 LEVEL LO (SFP-3.9)

• SFP 2 LEVEL LO (SFP-3.10)

Page 51 of 276

ATTACHMENT 1 EAL Bases

• RFL CAVITY 1 LEVEL LO (vessel area) (SFP-4.3)

• RFL CAVITY 2 LEVEL LO (vessel area) (SFP-4.7)

• SFP 1 TRANSFER CANAL LEVEL LO (SFP-4.9)

• SFP 2 TRANSFER CANAL LEVEL LO (SFP-4.10)

Allowing level to decrease could result in spent fuel being uncovered, reducing spent fuel decay heat removal and creating an extremely hazardous radiation environment. Technical Specification Section 3.7.15 (ref. 4) requires at least 23 ft of water above the Spent Fuel Pool storage racks (857' 31/2") (ref. 2). Technical Specification Section 3.9.7 (ref. 5) requires at least 23 ft of water above the Reactor Vessel flange in the refueling cavity (856' 11" in refueling cavity or 407" above core plate) (ref. 6). During refueling, this maintains sufficient water level in the fuel transfer canal, refueling cavity, and SFP to retain iodine fission product activity in the water in the event of a fuel handling accident. ABN-909, Spent Fuel Pool/Refueling Cavity Malfunctions, provides appropriate guidance to restore and maintain normal water levels in the fuel transfer canal, refueling cavity, and SFP, and to determine if water levels have dropped below the Technical Specification LCOs (ref. 2). The fuel transfer canal is only of concern in assessing this EAL when irradiated fuel transfer is in progress, in which case the spent fuel pool gates are open and connected to the fuel transfer canal.

NEI 99-01 Basis:

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

A water level decrease will be primarily determined by indications from available level instrumentation. Other sources of level indications may include reports from plant personnel (e.g., from a refueling crew) or video camera observations (if available). A significant drop in the water level may also cause an increase in the radiation levels of adjacent areas that can be detected by monitors in 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 AA2RA2.

CPNPP Basis Reference(s):

1. ABN-908 Fuel Handling Accident

2. ABN-909 Spent Fuel Pool/Refueling Cavity Malfunctions

3. 1-ALB-6B SFPCS TRBL

4. Technical Specifications 3.7.15 Fuel Storage Area Water Level

5. Technical Specifications 3.9.7 Refueling Cavity Water Level

6. RFO-102A/B Refueling Operations Page 52 of 276

ATTACHMENT I EAL Bases

7. NEI 99-01 AU2 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 Unusual Event 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.

CPNPP Basis:

None.

NEI 99-01 Basis:

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

This IC applies to irradiated fuel that is ic~ensed for dry s~torage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causiRg loss of the CONFINEMIIENT BOUNDARY is classified .c-, in... ccord r .A.

ith IC E HUl.

Esc-alation of the emergency would be bhasead on either Recognition Categor,' A Or C EAL-#-#

This EAL escalates from AUJ2-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 Page 53 of 276

ATTACHMENT 1 EAL Bases accordance Recognition Category C during the Cold Shutdown and Refueling modes.EAL4#2 ThN" F.AI .*l .. .* rn!*_..-*.* .nfrirnch'.. r*f,-*miftr~rn mui,,r* h,' ..*.nhinirnl dim..i.* tn iin*v *v irraiatd fel, Damaging eventS May include thc dropping, bumping or binding of a assembly, Or dro)pping a heavy load onto an assembly. A risc OR readingG on radiato M9111101 -SHRUIG 08 GGR1S.-P-FP- i0Iii Wyn W tf vdd o'i"1bi-~tW1 EAL.Hut t vn e.#3aue adigaý -tr F= .f1.43 wt~ ~i tth""i --.--. r iti h

. n~ ...--. n n h ~~ in~n

..-.-. n~~t v

prevent significafnt;do;se con Rsequences from direct gamma radiation to personnel performing emeratinn5G .....-in the 'Anlnit'i J- of the

.----. fuel nee[ ThkS refkr~ts a ".innffictnt

• ............... less of-.

speRnt GE)RditiGR ent Tuel pool water I and thlus it is Also 4

a PrecursorF n

I I fI to I

a less T me anniity d

to aacquately cool the irradiated Tuei assemoigis stored in the pool.

Escalation of the emergency classification level would be via ICs AS-1-RS1 9F (see.A,.

.AS2 DevýeFp-ee)

CPNPP Basis Reference(s):

1. ABN-909 Spent Fuel Pool/Refueling Cavity Malfunctions

2. NEI 99-01 AA2 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 High alarm on any of the following:

" Any Table R-2 area radiation monitors

• CAGu97, CNTMT AIR PIG GAS (u-RE-5503)

• CAPu98, CNTMT AIR PIG PART (u-RE-5502)

• CAIu99, CNTMT AIR PIG IODINE (u-RE-5566)

" FBV088, FB VENT EXH (X-RE-5700)

Table R-2 SFP & Refueling Cavity Area Radiation Monitors SFP:

• SFP001, LRAM SFP 2 E WALL (X-RE-6272)

• SFP002, LRAM SFP 2 N WALL (X-RE-6273)

" SFP003, LRAM SFP I E WALL (X-RE-6274)

" SFP004, LRAM SFP 1 S WALL (X-RE-6275)

Page 54 of 276

ATTACHMENT 1 EAL Bases Refueling Cavity:

• RFCulO, LRAM W REFUEL CAV860 (u-RE-6251)

" RFCu12, LRAM E REFUEL CAV 860 (u-RE-6253)

Mode Applicability:

All Definition(s):

None CPNPP Basis:

The specified radiation monitors are those expected to see increase area radiation levels as a result of damage to irradiated fuel (ref. 1, 2).

The bases for the SFP ventilation radiation High alarm and the SFP and containment area radiation readings are a spent fuel handling accident (ref. 1). In the Fuel Handling Building, a fuel assembly could be dropped in the fuel transfer canal or in the SFP. Should a fuel assembly be dropped in the fuel transfer canal or in the SFP and release radioactivity above a prescribed level, the Fuel Handling Building ventilation monitors sound an alarm, alerting personnel to the problem.

NEI 99-01 Basis:

This IC addresses events that have caused imminent or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool(See -Deve..,..

Netes). These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant.

This IC applies to irradiated fuel that is lic-ensed fo-r dry storage up to the point that the loaded-storFage cask is sealed. Once sealed, damage to a loaded cask causing los6 of the CONFINEMI1ENT BOUNDARYis classified in ac.r.d. ,....with ICE HU.

Escalation of the emergency would be based on either Recognition Category A-Ror C ICs.

EAl lt--hs E,ATL ealates from ,*,.2 in that the loss of level,

.. the affected portion ofthe*

inR REFUELING P'HWY is f sufficin mgiuethaeruldinuncoYer' of irdae fuel. IndicatioNs of irradiated fuel uncover' may include direct or indirect visual l bselVati1 (e.g., reports from personn~el or caer imaes), as well as significant changes in water and radiation levels, or other plant parametes.. o.m.putationa aids may also be used (e.g., a boil off cu.e). Classification

. f aR event using this EA.L should be based n the totality of available indications, reports and obsevations.

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 REF UELING PIADTHWAIAV theFeading may not be

-Areliable indication of whetherornot t~he fu-'el is actually uncovered. To the degree possible, readings should bhe cosidered in combination With o-th-ravailable indicationS of invento' los Adrop in water level above irradiated fuel within the reactor vessel ma be classifiedi accordance Recognition Categor,' C duFrig the Co-ld Shutdo'A' and Refueling modes.

Page 55 of 276

ATTACHMENT 1 EAL Bases 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).EAL i,,Spent fuel poel water level at this value is within the lower end of the level range necessar,. to prevent Sig*i**ant do6e conSequences from direct gamma radiation to perSonnel p.. rrming operations in the vicinity of the spent fuel pool. This condition reflects a signific-ant loss of spent fuel pool wte inventor,' and thus it is also a prec-ursor to a loss of the ability to adequately cool the irdae fuel assembles stored in the pool.

Escalation of the emergency classification level would be via ICes ASI--RS1 9FAS2-(seee-AS22 D'evelepeF-Notes).

CPNPP Basis Reference(s):

1. ABN-908 Fuel Handling Accident

2. ABN-909 Spent Fuel Pool/Refueling Cavity Malfunctions

3. NEI 99-01 AA2 I Page 56 of 276

ATTACHMENT I 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 El. 844.3' (Level 2)

Mode Applicability:

All Definition(s):

None CPNPP Basis:

Post-Fukushima order EA-12-051 (ref. 1) required the installation of reliable SFP level indication capable of identifying normal level (Level 1), 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) (ref. 1).

Level 2 is the level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck. It represents the range of water level where any necessary operations in the vicinity of the spent fuel pool can be completed without significant dose consequences from direct gamma radiation from the stored spent fuel.

Comanche Peak designated as Level 2 the water level 10 feet (+/- 1.0 foot) above the top of the fuel racks (El 844' - 2.75" rounded to 844.3' indicated) (ref. 2).

The enhanced SFP level instruments (X-LI-4876, 4878, 4877, 4879) do not have indication available in the control room and must be read remotely outside of the control room.

NEI 99-01 Basis:

This IC addresses events that have caused imminent or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool- (see De...*e Nte4. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant.

This IC applies to irradiated fuel that is licensed for d*, storage up to the point that the loaded storage cask is sealed. Oncc sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUINDARY isclassified on acc*r*dance with IC E HUE

-- Escalation of the emergency would be based on either Recognition Category A-Ror C 1('q~ PA L44T1,kQ P~A L imi==t= m Ul 2 on- th-n+ +j-P Aflq Af I'PUP' on th'-' 'ff4PL~tp -~.nndon nf ih,-n R=FUELIN 1F=1 HAY, is of sufficient magnitude to hav.e resulted in uncoVery of irrFadiated fuel. Indications of iRradiated fuel uncovery may include direct or indirect visual observation (e.g., reports from personnel or caer iags), as well as significant changes in water and r-adiatien levels, or other plant par-amieteirs. Coamputational aids may also be used (e.g., a boil indicationS, repo-ts and ,bseR,'atkiN4.

Page 57 of 276

ATTACHMENT 1 EAL Bases While an area radiatienR Fmn*itGo

• uld detect aninreae a dose rate due to a l-w-oing ef water level in some portion of the REFUELING PATHlw Y, the reading may not be a reliable indicatio enf whether icnt r the fuel is acually urcgvered. To the degree possible, readings should be onsidered inicofbiatioe with other available indieation9resof i lo gniVefto A drfp in water level above inradiated fuel within the reacutor el ay be clasified in accordance Recognition Categor,' C during the Cold Shu~tdoWn and Refueling moedes.

This EAL addrpesrspes a-rpelpease o-f radioactive maeralcased by mnechan~ical damnage to irradiated fuel. Damaging evenss may include the dropping, bLmping or binding o assembly, or dropping a heavy load onRto an assembly. A rise in readings on radiato monitors should be considered in conjunction With in plant reports or o~bservations of a potential fuel damaging event (e.g., a fuel handling accident).

AL#Spent fuel pool water level at this value is within the lower end of the level range necessary to prevent significant dose consequences from direct gamma radiation to personnel performing operations in the vicinity of the spent fuel pool. This condition reflects a significant loss of spent fuel pool water inventory and thus it is also a precursor to a loss of the ability to adequately cool the irradiated fuel assembles stored in the pool.

Escalation of the emergency classification level would be via ICs AS4-RS1-, A$2 -see-*A9

/- p - A,,.,

I -

Page 58 of 276

ATTACHMENT 1 EAL Bases CPNPP Basis Reference(s):

1. NRC EA-12-51 Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation

2. TXX-13103 Overall Integrated Plan in Response to March 12,2012 Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051), Response to Request for Additional Information 3 NEI 99-01 AA2 Page 59 of 276

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 El. 835.3' (Level 3)

Mode Applicability:

All Definition(s):

None CPNPP Basis:

Post-Fukushima order EA-12-051 (ref. 1) required the installation of reliable SFP level indication capable of identifying normal level (Level 1), 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) (ref. 1).

Level 3 is the level where fuel remains covered and actions to implement make-up water addition should no longer be deferred. Level 3 corresponds nominally (i.e., +/- 1 foot) to the highest point of any fuel rack seated in the spent fuel pool. Level 3 is defined in this manner to provide the maximum range of information to operators, decision makers and emergency response personnel. Comanche Peak designated as Level 3 the water level greater than 1 foot above the top of the fuel storage racks plus the accuracy of the SFP level instrument channel (El. 835' - 2.75" rounded to 835.3' indicated). Designation of this level as Level 3 is conservative; its selection assures that the fuel will remain covered, and at that point there would be no functional or operational reason to defer action to implement the addition of make-up water to the pool (ref. 2).

The enhanced SFP level instruments (X-LI-4876, 4878, 4877, 4879) do not have indication available in the control room and must be read remotely outside of the control room.

NEI 99-01 Basis:

This 1C-EAL addresses a significant loss of spent fuel pool inventory control and makeup capability leading to IMMINENT fuel damage. This condition entails major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.

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

Escalation of the emergency classification level would be via IC AG1 or AG2RG2.

I Page 60 of 276 1

ATTACHMENT 1 EAL Bases CPNPP Basis Reference(s):

1. NRC EA-12-51 Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation

2. TXX-13103 Overall Integrated Plan in Response to March 12,2012 Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051), Response to Request for Additional Information

3. NEI 99-01 AS2 I Page 61 of 276

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Spent fuel pool level cannot be restored to at least the top of the fuel racks for 60 minutes or longer EAL:

RG2.1 General Emergency Spent fuel pool level cannot be restored to at least El. 835.3' (Level 3) for greater than or equal to 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 CPNPP Basis:

Post-Fukushima order EA-12-051 (ref. 1) required the installation of reliable SFP level indication capable of identifying normal level (Level 1), 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) (ref. 1).

Level 3 is the level where fuel remains covered and actions to implement make-up water addition should no longer be deferred. Level 3 corresponds nominally (i.e., +/- 1 foot) to the highest point of any fuel rack seated in the spent fuel pool. Level 3 is defined in this manner to provide the maximum range of information to operators, decision makers and emergency response personnel. Comanche Peak designated as Level 3 the water level greater than 1 foot above the top of the fuel storage racks plus the accuracy of the SFP level instrument channel (El. 835' - 2.75" rounded to 835.3' indicated). Designation of this level as Level 3 is conservative; its selection assures that the fuel will remain covered, and at that point there would be no functional or operational reason to defer action to implement the addition of make-up water to the pool (ref. 2).

The enhanced SFP level instruments (X-LI-4876, 4878, 4877, 4879) do not have indication available in the control room and must be read remotely outside of the control room.

NEI 99-01 Basis:

This I,-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.

Page 62 of 276

ATTACHMENT I EAL Bases CPNPP Basis Reference(s):

1. NRC EA-12-51 Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation

2. TXX-13103 Overall Integrated Plan in Response to March 12,2012 Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051), Response to Request for Additional Information

3. NEI 99-01 AG2 Page 63 of 276

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 3 - Area Radiation Levels Initiating Condition: Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown EAL:

RA3.1 Alert Dose rates greater than 15 mR/hr in EITHER of the following areas:

Control Room CRM048 (X-RE-6281) or CRM049 (X-RE-6282)

OR Central Alarm Station (by survey)

Mode Applicability:

All Definition(s):

None CPNPP Basis:

X-RE-6281 and X-RE-6282 are the installed Control Room area radiation monitors and may be used to assess this EAL threshold (range of 1E-4 to 1E+5 mR/hr). However, no permanently installed area radiation monitoring is installed in the CAS and therefore this threshold must be assessed via local radiation survey (ref. 1).

NEI 99-01 Basis:

This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or impede personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The ...e.ge..y-Di*e*te" Emergqency Coordinator should consider the cause of the increased radiation levels and determine if another IC may be applicable.For EAr.#2, an Alert delara-onj, is warranted if operatinG mode in effect at the time of the elevated radiation levels. The emerenc'-

S;~ ~atcnnir'i~ntinn

~hrh~r ~nt" kint ----....- ini' nr'~rv .J- t th~ tmr~nf hr~

• I IV increased radiation levels. Access should be considered as impeded if extraordinary

.-- .a aa ,ara.,aa

- - co -. .,+a fn I+a H;+ . + ~r

-Fr, arrr .nna Ltnia . l,

+k^ a ad raa rnIa a- Ia r*

installing temporary shielding, reqIu"irig Use of non rou1tine protective equipment, requesting an4 extension in dose limits beyond nremal administrative lmt)

An emergency declaration is not wratdif any of the folloWing conditions apply-.

aThe plant is in an operating mode differet than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time ot Page 64 of 276

ATTACHMENT 1 EAL Bases the elevated radiation levels). For example, the plant is in Mode 1 whcn the radiation increase occurFs, and the proGedures used for normnal opcration, cooldoWn and shutdW-n dGont require ent' ; into the affected room until Mode 4.

alThe incfeased radiation levels are a result of a planed acgtivity that incRlu comnpensator,' measures which addrcss the temporar,' inaGeeSSibility of a room or area (e.g., radiography, spcnt filter or re-sin transfer, etc.).

• The action for which room/area entr,'is required is of an administrative or recorFd keeping nature (e.g., nr9mal rounds or routine inspections).-

• The access control measures are of a conservative or precauitionar,' natur~e, and would1 not actually prevent or impede a required action-.

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

CPNPP Basis Reference(s):

1. DBD-EE-023 Radiation Monitoring System

2. NEI 99-01 AA3 Page 65 of 276

ATTACHMENT I EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 3- Area Radiation Levels Initiating Condition: Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown EAL:

RA3.2 Alert An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to any Table R-3 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-3 Safe Operation & Shutdown Rooms/Areas Room/Area Mode Applicability Charging Pump Rooms 1, 2, 3, 4, 5, 6 CVCS Valve Rooms 1, 2, 3, 4, 5, 6 1 E Switchgear Rooms All RHR Pump Rooms 4, 5, 6 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.

CPNPP Basis:

For this EAL, area or room access is considered impeded if radiation levels require locked high radiation controls to to be imposed.

If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event.

The list of plant rooms or areas with entry-related mode applicability identified specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective Page 66 of 276

ATTACHMENT 1 EAL Bases 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).

NEI 99-01 Basis:

This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or impede personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The ergiRy DirecteFEmergency Coordinator should consider the cause of the increased radiation levels and determine if another IC may be applicable.

I For EAL- #2RA3.2, 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 AR, C or F ICs.

CPNPP Basis Reference(s):

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

2. NEI 99-01 AA3 Page 67 of 276

ATTACHMENT I 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 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.

IPage 68 of 276

ATTACHMENT 1 EAL Bases Category: ISFSI Subcategory: Confinement Boundary Initiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARY EAL:

EUl.1 Unusual Event Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading greater than EITHER:

• 60 mrem/hr (-r + rq) on the top of the overpack

• 600 mrem/hr (T"+ rj on the side of the overpack (excluding inlet and outlet ducts)

Mode Applicability:

All Definition(s):

CONFINEMENT BOUNDARY-. The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As applied to the CPNPP ISFSI, the CONFINEMENT BOUNDARY is defined to be the Multi-Purpose Canister (MPC).

CPNPP Basis:

The ISFSI includes the dry-cask storage system, the cask transfer facility, onsite transporter, and the storage pads. The dry-cask storage system is the HI-STORM 100 System. This is a canister-based storage system that stores spent nuclear fuel in a vertical orientation. It consists of three discrete components: the MPC, the HI-TRAC 125 Transfer Cask, and the HI-STORM 100 System Overpack. The MPC provides the confinement boundary for the stored fuel. The HI-TRAC 125 Transfer Cask provides radiation shielding and structural protection of the MPC during transfer operations, while the storage overpack provides radiation shielding and structural protection of the MPC during storage (ref. 1).

The value shown represents 2 times the limits specified in the ISFSI Certificate of Compliance Technical Specification section 5.7.4 for radiation external to a loaded MPC overpack (ref. 1).

NEI 99-01 Basis:

This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of a storage cask containing spent fuel. It applies to irradiated fuel that is licensed for dry storage beginning at the point that the loaded storage cask is sealed. The issues of concern are the creation of a potential or actual release path to the environment, degradation of one or more fuel assemblies due to environmental factors, and configuration changes which could cause challenges in removing the cask or fuel from storage.

The existence of "damage" is determined by radiological survey. The technical specification I multiple of "2 times", which is also used in Recognition Category A-RIC RAU1, is used here to I Page 69 of 276

ATTACHMENT I EAL Bases 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.

CPNPP Basis Reference(s):

1. HI-2104635, HI-STORM Certificate of Compliance Appendix A Technical Specification Section 5.7.4

2. RPI-792 HI-STORM Overpack Surface Dose Rates

3. NEI 99-01 E-HU1 Page 70 of 276

ATTACHMENT 1 EAL Bases Category C - Cold Shutdown / Refueling System Malfunction EAL Group: Cold Conditions (RCS temperature < 2000 F); EALs in this category are applicable only in one or more cold operating modes.

Category C EALs are directly associated with cold shutdown or refueling system safety functions. Given the variability of plant configurations (e.g., systems out-of-service for maintenance, containment open, reduced AC power redundancy, time since shutdown) during these periods, the consequences of any given initiating event can vary greatly. For example, a loss of decay heat removal capability that occurs at the end of an extended outage has less significance than a similar loss occurring during the first week after shutdown. Compounding these events is the likelihood that instrumentation necessary for assessment may also be inoperable. The cold shutdown and refueling system malfunction EALs are based on performance capability to the extent possible with consideration given to RCS integrity, containment closure, and fuel clad integrity for the applicable operating modes (5 - Cold Shutdown, 6 - Refueling, D - Defueled).

The events of this category pertain to the following subcategories:

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

2. Loss of Emerqency AC Power Loss of emergency plant electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of onsite and offsite power sources for 6.9 KV AC emergency buses.

3. RCS 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 125V DC vital buses.

5. Loss of Communications Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification.

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

Page 71 of 276 1

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: UNPLANNED loss of RCS inventory for 15 minutes or longer EAL:

CUI.1 Unusual Event UNPLANNED loss of reactor coolant results in RCS water level less than a required lower limit for greater than or equal to 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.

CPNPP Basis:

With the plant in Cold Shutdown, RCS water level is normally maintained above the pressurizer low level setpoint of 17% (ref. 1). However, if RCS 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 RCS that is the concern.

With the plant in Refueling mode, RCS water level is normally maintained at or above the reactor vessel flange (Technical Specification LCO 3.9.7 requires at least 23 ft. of water above the top of the reactor vessel flange in the refueling cavity during refueling operations) (ref. 2).

The Reactor Vessel flange level is 834' 1/2" elevation or 132.5 in. above the upper core plate (top) (ref. 3).

NEI 99-01 Basis:

This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor .. .....-v-..-,IRCS {RPWR]

,r RPV- r[{BVR)-level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant.

Refueling evolutions that decrease RCS-water inventory are carefully planned and controlled.

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

This EAL-#4 recognizes that the minimum required (.eaGtO'r-*ves/RCS [P.4'.] or RPV [BW,4..)

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,

-Page72 of 276

ATTACHMENT I EAL Bases specified for the current plant conditions, cannot be maintained for 15 minutes or longer. The minimum level is typically specified in the applicable operating procedure but may be specified in another controlling document.

The 15-minute threshold duration allows sufficient time for prompt operator actions to restore and maintain the expected water level. This criterion excludes transient conditions causing a brief lowering of water level.

EAL #2o addresses a condition ,,,here all means to determ.ine (reactoF vesse.R.S

[PI/lRI or RPV [B3VARj) level have been lost. In tthis, cond-iqtion, operators may determinle that an inventory loss isoccurring by ebseiR'ig changes1inSu1mp anmd/ortank levels. Sump and/eF tank level chng Smut bhe evaluated against other potential sources of water flow to ensur~e th e" are ndi r-t:-;Zt ofp l11e.a ka ge fro m the (re acGtor '.GeS P-VRCGS2 [P- tIP] orF RPV [BWR]).

Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CAl or CA3.

CPNPP Basis Reference(s):

1. ALM-0052A/B Alarm Procedure u-ALB-5B3 (513-3.6)

2. Technical Specification Section 3.9.7 Refueling Cavity Water Level

3. IPO-Ol OAIB Reactor Coolant System Reduced Inventory Operations

4. NEI 99-01 CUI I Page 73 of 276

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: UNPLANNED loss of RCS inventory for 15 minutes or longer EAL:

CUI.2 Unusual Event RCS water level cannot be monitored AND EITHER

• UNPLANNED increase in any Table C-1 sump/tank level due to loss of RCS inventory

• Visual observation of UNISOLABLE RCS leakage Table C-I Sumps / Tanks

• Containment Sump 1

• Containment Sump 2

• Reactor Cavity Sump

• CCW Surge Tank A

• CCW Surge Tank B

• PRT

. RCDT 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.

CPNPP Basis:

In Cold Shutdown mode, the RCS will normally be intact and standard RCS level monitoring means are available.

In this EAL, all water level indication is unavailable and the RCS inventory loss must be detected by indirect leakage indications (Table C-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 RCS leakage. If the make-up rate to the RCS unexplainably rises above the pre-established rate, a loss of RCS inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of Page 74 of 276

ATTACHMENT 1 EAL Bases leakage from systems connected to the RCS that cannot be isolated could also be indicative of a loss of RCS inventory (ref. 1, 2, 3, 4).

NEI 99-01 Basis:

This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor (FeaGter-veseRCS [PWR]

or RP, rl3WRYlevel concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant.

Refueling evolutions that decrease RCS water inventory are carefully planned and controlled.

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

E-=ALI #1 recognizes that the mnumrquired (reactor vessel/RCS [PWRq or RPV

[914qR) level can change several times during the course of a refueling outage as difflerent.

plant configurations and system lineups are implemented. This EAL is met if the minimum level, specified forF the- current plant conditions, cannot be maint-ained- for 15 minutes Or longer.-

The minimum lev~,el is typically specified in the apibloperating procedure but ma" be specified in another controlling document.

The 15 minute threshold duration allows sufficient time8 for prompt operator actions to restore and maintain the-expected water level. Thi-s criterion excludes transient conditinsausn a brief lowering of water level.

This EAL-#-2 addresses a condition where all means to determine (reactor vessel!RCS [PWR 9F RPV {BWRI.) level have been lost. In this condition, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels (Table C-1).

Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the (.eat.. .ess RCS [PW*

.* RF R

[BWR]).

Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CAl or CA3.

CPNPP Basis Reference(s):

1. 1PO-01OA/B Reactor Coolant System Reduced Inventory Operations

2. SOP-I1NB/1 Reactor Coolant System

3. ABN-103 Excessive Reactor Coolant Leakage

4. ABN-108 Shutdown Loss of Coolant

5. NEI 99-01 CUI Page 75 of 276

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: Loss of RCS inventory EAL:

CA1.1 Alert Loss of RCS inventory as indicated by RCS level less than 48 in. above upper core plate (top)

Mode Applicability:

5 - Cold Shutdown, 6 - Refueling Definition(s):

None CPNPP Basis:

When Reactor Vessel water level decreases to 48 in. above the upper core plate (top) (EL 827' 0"), RHR pump cavitation may occur. RCS level can be monitored by one or more of the following (ref. 1, 2, 3, 4, 5, 6):

• RCS Level Wide Range u-LI-3615B

" RCS Level Narrow Range u-LI-3615A

• RCS Extended Wide Range u-LI-3615C

• Mansell Level Monitor System u-LT-3619AIB/C-1, -2

" Plant Computer

• RVLIS

• Ultrasonic Level monitoring (optional)

NEI 99-01 Basis:

This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier). This condition represents a potential substantial reduction in the level of plant safety.

For this EAL-#-I-, a lowering of RCS water level below 48 in. above the upper core plate (top)

(site specific level) ft. indicates that operator actions have not been successful in restoring and maintaining RCS (reactor vesse!IRCS [PWrIRI or RPV,4'[*r-, water level. The heat-up rate of the coolant will increase as the available water inventory is reduced. A continuing decrease in water level will lead to core uncovery.

Although related, this EAL-#4 is concerned with the loss of RCS inventory and not the potential concurrent effects on systems needed for decay heat removal (e.g., loss of a Residul- Decay Heat Removal suction point). An increase in RGS-RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3.

Page 76 of 276

ATTACHMENT 1 EAL Bases For EAL #2, the inability to mon.itor (reactor V.SSel.R. S [PV.R] or RPV [B'A'R]) level may be caused by instrumentation and/orF power failures, Or water level dropping below the range of available instFrumentation. If water level c3-annot bhe monitored, operators m~ay determne that an invento' los is occurri. by obsering c.hanges in sump and/or tank levels.

Sump and/or tank level changes mus t be evaluated against other potenti-al sou1-rces of water flow to ensure they are iniaieof leakage from the (reactor vcssel/RCS [PWVRI or RPV The 15 mninute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CSI If RCS the (reactor vcssel/RCS [PWVR] or RPV~ [BVVR]) inventory water level continues to lower, then escalation to Site Area Emergency would be via IC OSi.

CPNPP Basis Reference(s):

1. IPO-Ol OA/B Reactor Coolant System Reduced Inventory Operations

2. INC-6269 Calibration of the Mansell RCS Measurement System

3. SOP-l0lA/B Reactor Coolant System

4. ABN-1 03 Excessive Reactor Coolant Leakage

5. ABN-104 Residual Heat Removal System Malfunction

6. ABN-108 Shutdown Loss of Coolant

7. NEI 99-01 CAl Page 77 of 276

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: Loss of RCS inventory EAL:

CA1.2 Alert RCS water level cannot be monitored for greater than or equal to 15 min. (Note 1)

AND EITHER

" UNPLANNED increase in any Table C-1 sump/tank level due to loss of RCS inventory

" Visual observation of UNISOLABLE RCS 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-1 Sumps / Tanks

• Containment Sump 1

• Containment Sump 2

• Reactor Cavity Sump

• CCW Surge TankA

• CCW Surge Tank B

• PRT

• RCDT 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.

CPNPP Basis:

In Cold Shutdown mode, the RCS will normally be intact and standard RCS level monitoring means are available.

In the Refuel mode, the RCS is not intact and RPV level may be monitored by different means, including the ability to monitor level visually.

In this EAL, all RCS water level indication would be unavailable for greater than 15 minutes, and the RCS inventory loss must be detected by indirect leakage indications (Table C-1).

Page 78 of 276

ATTACHMENT 1 EAL Bases Surveillance procedures provide instructions for calculating primary system leak rate by manual or computer-based water inventory balances. Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage. If the make-up rate to the RCS unexplainably rises above the pre-established rate, a loss of RCS inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS that cannot be isolated could also be indicative of a loss of RCS inventory (ref. 1, 2, 3).

NEI 99-01 Basis:

This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier). This condition represents a potential substantial reduction in the level of plant safety.

PFo EAL #1, a lowering of water level below (site specific level) indicates, that operatorF actions have not been successful in restGorig and maintainfin~g (reactor vcsscl!RCS [PL4'R oe RPV [*14Aql) wate level. The heat up rate Of the coolant will increase as the available w inventory is reduced. A continuing decrease in water leve! will lead to core uncover;.

Alth.ugh related, #1 is concerned with the loss of RCS inventor.' and not the L*A=

potential concurrent effects On systems needed for decay heat removal (e.g., loss of a Residual Heat Removal suction point). An inrGease in ROS temperature caused by a loss of decay heat removal capability is evaluated under IC CA33.

For this EAL-#2-, the inability to monitor RCS (reactor.. essel.RS [PWý eor RPV [BWR]) level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels.

Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the (eaeter-vesSeRCS [PWR] or RPV The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CS1_

If the (reaGtO.ves....RCS [PW.r or RPV [B\R]) inventory level continues to lower, then escalation to Site Area Emergency would be via IC CS1.

CPNPP Basis Reference(s):

1. ABN-1 03 Excessive Reactor Coolant Leakage

2. ABN-108 Shutdown Loss of Coolant

3. FSAR 5.2.5.2

4. NEI 99-01 CA1 Page 79 of 276

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: Loss of RCS inventory affecting core decay heat removal capability EAL:

CS1.1 Site Area Emergency With CONTAINMENT CLOSURE not established, RCS level less than 27.3 in. above upper core plate (top)

Mode Applicability:

5 - Cold Shutdown, 6 - Refueling Definition(s):

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. Containment closure means that all potential escape paths are closed or capable of being closed:

A. All penetrations providing direct access from Containment atmosphere to outside atmosphere are closed except:

" Penetrations with automatic valves capable of being closed by an operable CVI

" Penetrations under administrative controls (e.g., Control Room notified and designated person to close if required by fuel handling accident)

B. Equipment hatch is closed and held in place by 4 bolts, or is capable of being closed and held in place by 4 bolts C. One emergency airlock door is closed D. One personnel airlock door is capable of being closed CPNPP Basis:

When Reactor Vessel water level decreases to 27.25 in. (rounded to 27.3 in. for instrument readability), 825'-3 11/4" elevation (ref. 1), water level is six inches below the elevation of the bottom of the RCS hot leg penetration. When Reactor Vessel water level drops significantly below the elevation of the bottom of the RCS hot leg penetration, all sources of RCS injection have failed or are incapable of making up for the inventory loss. RCS elevations are illustrated in Figure C-3. RCS level can be monitored by one or more of the following (ref. 1, 2, 3):

• RCS Level Wide Range u-LI-3615B

• RCS Level Narrow Range u-LI-3615A

" RCS Extended Wide Range u-LI-3615C

• Mansell Level Monitor System u-LT-3619A/B/C-1, -2

" Plant Computer

" RVLIS

• Ultrasonic Level monitoring (optional)

Page 80 of 276

ATTACHMENT 1 EAL Bases In Refueling mode, Reactor Vessel water level indication from RVLIS is likely unavailable but alternate means of level indication are normally installed (including visual observation) to assure that the ability to monitor water level will not be interrupted.

The status of Containment closure is tracked if plant conditions change that could raise the risk of a fission product release as a result of a loss of decay heat removal (ref. 4, 5).

NEI 99-01 Basis:

This IC addresses a significant and prolonged loss of (reactor vessel/RGS-RCS_[PWRe] r RPV rf^W])-inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.

Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RGS-eaetei vessel RCS level cannot be restored, fuel damage is probable.

Outage/shutdown contingency plans typically provide for re-establishing or verifying CONTAINMENT CLOSURE following a loss of heat removal or RCS inventory control functions. The difference in the specified RCS/reactor vessel levels of EALs 4-1CS1.1 and 2-bCS2.2 reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment.

in EAL 3.a, the 30 minute criterion is tied to a readily reco~gnizable event start time (i.e., the tetal loss of ability to monitor level), and allows sGfficient time to monitoF, a2s6se -andc iorrel reactor and plant co~nditions to determnine if core Uncovery has actually occurred (i.e., to acco)unt for vaiu accident progression and instFrumentation uncertainties). it also allows sufficient time for pcfo~rmancc of actions to terminate leakage, recover inventor; control/makeup equipment and~or restore leVel monitoring-.

The inability to monitor RCS (reactor vesscl/RGS [PWVR] or RPV [BWVR]) level may be caused by instrumentation and/or power failures, or water level dropping below the range of availab~

ins~trumentation. if water level cannot be monitored, operators ma" determine thata inventor; loss is Gccurring by observing chane in sup and/or tank levels. Sump and/oi tank level changes must be evaluated against other potential sources of water flow to ensure they are indicativ~e of leakage from theRGS (reactor vesseliRCS [PVVARI or RP'. [B'NRI).

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

Escalation of the emergency classification level would be via IC CGI or AG4RG1 CPNPP Basis Reference(s):

1. IPO-01 OA/B Reactor Coolant System Reduced Inventory Operations

2. INC-6269 Calibration of the Mansell RCS Measurement System

3. SOP-l0lA/B Reactor Coolant System

4. Technical Specifications 3.9.4 I Page 81 of 276

ATTACHMENT 1 EAL Bases

5. OPT-408A/B Refueling Containment Penetration Verification

6. NEI 99-01 CS1 Page 82 of 276

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: Loss of RCS inventory affecting core decay heat removal capability EAL:

CS1.2 Site Area Emergency With CONTAINMENT CLOSURE established, RCS level less than or equal to 0 in. above upper core plate (top)

Mode Applicability:

5 - Cold Shutdown, 6 - Refueling Definition(s):

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. Containment closure means that all potential escape paths are closed or capable of being closed:

A. All penetrations providing direct access from Containment atmosphere to outside atmosphere are closed except:

• Penetrations with automatic valves capable of being closed by an operable CVI

" Penetrations under administrative controls (e.g., Control Room notified and designated person to close if required by fuel handling accident)

B. Equipment hatch is closed and held in place by 4 bolts, or is capable of being closed and held in place by 4 bolts C. One emergency airlock door is closed D. One personnel airlock door is capable of being closed CPNPP Basis:

When Reactor Vessel water level drops below 0 in. above upper core plate (top) 823'-0" elevation (ref. 1), core uncovery is about to occur. RCS level can be monitored by one or more of the following (ref. 1, 2, 3):

" RCS Level Wide Range u-LI-3615B

• RCS Level Narrow Range uj-LI-3615A

" RCS Extended Wide Range u-LI-3615C

• Mansell Level Monitor System u-LT-3619A/B/C-1, -2

" Plant Computer

• RVLIS

• Ultrasonic Level monitoring (optional)

Under the conditions specified by this EAL, continued lowering of Reactor Vessel water level is indicative of a loss of inventory control. Inventory loss may be due to a vessel breach, RCS pressure boundary leakage or continued boiling in the Reactor Vessel. The magnitude of this Page 83 of 276

ATTACHMENT 1 EAL Bases loss of water indicates that makeup systems have not been effective and may not be capable of preventing further RCS or Reactor Vessel water level drop and potential core uncovery. The inability to restore and maintain level after reaching this setpoint infers a failure of the RCS barrier and Potential Loss of the Fuel Clad barrier.

The status of Containment closure is tracked if plant conditions change that could raise the risk of a fission product release as a result of a loss of decay heat removal (ref. 4, 5).

NEI 99-01 Basis:

This IC addresses a significant and prolonged loss of (reactor vessel/RS-&RCS_[PWR]-r RPV

[-3,WRI}-inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.

Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RGS~eaetor Yesse4RCS level cannot be restored, fuel damage is probable.

Outage/shutdown contingency plans typically provide for re-establishing or verifying CONTAINMENT CLOSURE following a loss of heat removal or RCS inventory control functions. The difference in the specified RCS/reactor vessel levels of EALs bCS1.1 and 2-.bCS1.2 reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment.

1in EAL 3.a, the 30 m~inu-te c-riterion istied to a readily recognizable event start time (i.e., the total less of ability m to onGitor level),.....

and rntime uffici, to monitor, assess and corelate reoac-tor and plant conditions to determine if coro uncover,' has actually occurred (i.e., to suffiient time for perfoermance of ac-tions; to terminate leakage, recover inventor!

control/makeup equipment and/or restore level moenitoring-.

The inability to mon~itor RCS (reactorF vesscl/RGS [PWVR] or RPV~ [BWR]) level may be caused by instrumentation and/or power failures, or water level dropping belo thag Favailable instrum~entation. if water level cannot be monRitored, operatorns may dote rmine that an inventor,' losi ocuring by obser~ing 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 fromn the (reactor vessel/RCS [PVWR] or RPV [BWRJ).

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

Escalation of the emergency classification level would be via IC CG1 or AG--RG1 CPNPP Basis Reference(s):

1. IPO-01OA/B Reactor Coolant System Reduced Inventory Operations

2. INC-6269 Calibration of the Mansell RCS Measurement System

3. SOP-101A/B Reactor Coolant System Page 84 of 276 1

ATTACHMENT 1 EAL Bases

4. Technical Specifications 3.9.4

5. OPT-408A/B Refueling Containment Penetration Verification

6. NEI 99-01 CS1 Page 85 of 276 1

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: Loss of RCS inventory affecting core decay heat removal capability EAL:

CS1.3 Site Area Emergency RCS water level cannot be monitored for greater than or equal to 30 min. (Note 1)

AND Core uncovery is indicated by any of the following:

" UNPLANNED increase in any Table C-1 sump/tank level of sufficient magnitude to indicate core uncovery

• Erratic Source Range Monitor indication

• greater than 20,000 R/hr on any of the following:

- CTEu16, Containment HRRM (u-RE-6290A)

- CTWu17, Containment HRRM (u-RE-6290B)

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-1 Sumps / Tanks

• Containment Sump 1

• Containment Sump 2

• Reactor Cavity Sump

• CCW Surge TankA

• CCW Surge Tank B

• PRT

• RCDT 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.

CPNPP Basis:

In Cold Shutdown mode, the RCS will normally be intact and standard RCS level monitoring means are available.

Page 86 of 276

ATTACHMENT I EAL Bases In the Refueling mode, the RCS is not intact and RCS level may be monitored by different means, including the ability to monitor level visually.

In this EAL, all RCS water level indication would be unavailable for greater than 30 minutes, and the RCS inventory loss must be detected by indirect leakage indications (Table C-1).

Surveillance procedures provide instructions for calculating primary system leak rate by manual or computer-based water inventory balances. Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage. If the make-up rate to the RCS unexplainably rises above the pre-established rate, a loss of RCS inventory may be occurring even if the source of the leakage cannot be immediately identified (ref. 1, 2).

The RCS inventory loss may be detected by the Containment High Range Radiation Monitor (HRRM) or erratic Source Range Monitor indication. As water level in the Reactor Vessel lowers, the dose rate above the core will rise. The dose rate due to this core shine should result in Containment High Range Radiation Monitor indication greater than 20,000 R/hr (ref.

3). The Containment HRRMs have a range of IE-1 to 1E+8 R/hr.

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 (ref. 4, 5).

NEI 99-01 Basis:

This IC addresses a significant and prolonged loss of (reater-ves, . RCSfPWR-,-RPV WR]r*

.)-inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.

Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reacter vessel level cannot be restored, fuel damage is probable.

CONTAINMENT CLOSURE follGoI g a loss of heat removal or RCS invento**, control functionS. The diffeMrene in the specified RGS/reactor vessel levels of EALS 1.band 2.b reflect the fact that with CONTAINIMlENIT CGLOSURE established, ther Is a lower pro.bability of a fission product release to the envieroment.

In EAL 3.aa-The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.

The inability to monitor RCS (reactor vossel/ROS [PWR] or RPV [BDD 'R]) level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. Ifwater 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 Page 87 of 276

ATTACHMENT I EAL Bases tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RCS (rca.te. v...e...lRCS [PWR] or RPV [BWVVR]).

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

Escalation of the emergency classification level would be via IC CG1 or AG-1-RG1 CPNPP Basis Reference(s):

1. ABN-103 Excessive Reactor Coolant Leakage

2. ABN-108 Shutdown Loss of Coolant

3. Engineering Handbook, Guidelines for Events Beyond Design Basis: Spent Fuel Pools, Figure D "Dose Rate at Elevation 860' above Stored Fuel vs. Water Level Depth in SFP"

4. Severe Accident Management Guidance Technical Basis Report, Volume 1: Candidate High-Level Actions and Their Effects, pgs 2-18, 2-19

5. Nuclear Safety Analysis Center (NSAC), 1980, "Analysis of Three Mile Island - Unit 2 Accident," NSAC-1

6. NEI 99-01 CS1 I Page 88 of 276

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: I - RCS Level Initiating Condition: Loss of RCS inventory affecting fuel clad integrity with containment challenged EAL:

CGI.1 General Emergency RCS level less than or equal to 0 in. above upper core plate (top) for _>30 min. (Note 1)

AND Any Containment Challenge indication, Table C-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 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required.

Table C-2 Containment Challenge Indications

• CONTAINMENT CLOSURE not established (Note 6)

• Containment hydrogen concentration greater than 4%

• Unplanned rise greater than 1 psig in Containment pressure Mode Applicability:

5 - Cold Shutdown, 6 - Refueling Definition(s):

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. Containment closure means that all potential escape paths are closed or capable of being closed:

A. All penetrations providing direct access from Containment atmosphere to outside atmosphere are closed except:

" Penetrations with automatic valves capable of being closed by an operable CVI

" Penetrations under administrative controls (e.g., Control Room notified and designated person to close if required by fuel handling accident)

B. Equipment hatch is closed and held in place by 4 bolts, or is capable of being closed and held in place by 4 bolts C. One emergency airlock door is closed D. One personnel airlock door is capable of being closed Page 89 of 276

ATTACHMENT 1 EAL Bases CPNPP Basis:

When Reactor Vessel water level drops below 0 in. above upper core plate (top) 823'-0" elevation (ref. 1), core uncovery is about to occur. RCS level can be monitored by one or more of the following (ref. 1, 2, 3):

" RCS Level Wide Range u-LI-3615B

" RCS Level Narrow Range u-LI-3615A

• RCS Extended Wide Range u-LI-3615C

• Mansell Level Monitor System u-LT-3619A/B/C-1, -2

" Plant Computer

• RVLIS

• Ultrasonic Level monitoring (optional)

Under the conditions specified by this EAL, continued lowering of Reactor Vessel water level is indicative of a loss of inventory control. Inventory loss may be due to a vessel breach, RCS pressure boundary leakage or continued boiling in the Reactor Vessel. The magnitude of this loss of water indicates that makeup systems have not been effective and may not be capable of preventing further RCS or Reactor Vessel water level drop and potential core uncovery. The inability to restore and maintain level after reaching this setpoint infers a failure of the RCS barrier and Potential Loss of the Fuel Clad barrier.

Three conditions are associated with a challenge to Containment integrity:

1. CONTAINMENT COSURE not established - The status of Containment closure is tracked if plant conditions change that could raise the risk of a fission product release as a result of a loss of decay heat removal (ref. 4, 5). If containment closure is re-established prior to exceeding the 30 minute core uncovery time limit then escalation to GE would not occur.

2. Containment hydrogen greater than 4% - The 4% hydrogen concentration threshold is generally considered the lower limit for hydrogen deflagrations. CPNPP is equipped with a Hydrogen Control System (HCS) which serves to limit or reduce combustible gas concentrations in the Containment. The plant has two hydrogen monitoring systems.

Each monitoring system consists of four sensor modules and one microprocessor analyzer. Two sensors from each Containment are coupled to one of the two hydrogen microprocessors located in the Control Room. Thus each microprocessor analyzer is shared by Units 1 and 2. The analyzer system has a range of 0-10% hydrogen by volume. The detector modules are located on the 905', 873', and 860' elevations in Containment. A fourth detector is located on 832' level across from the loop room entrance for loops 1 and 4. Hydrogen concentration is displayed in the Control Room on u-AI-5506A/B and u-AI-5506C/D. Hydrogen concentration can also be displayed on the Plant Computer. Alarms at -3% are provided for high hydrogen concentration, u-ALB-3A, window 3.7. If a hydrogen concentration value can not be obtained from the hydrogen monitoring system, a grab sample from the containment PIG radiation monitor may be used to determine the hydrogen concentration (ref. 6, 7, 8, 9).

3. UNPLANNED rise in Containment pressure - An unplanned pressure rise in containment while in cold Shutdown or Refueling modes can threaten Containment Page 90 of 276

ATTACHMENT 1 EAL Bases Closure capability and thus Containment potentially cannot be relied upon as a barrier to fission product release.

NEI 99-01 Basis:

This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity.

Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.

Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. IfRCS RCS/reactor vessel level cannot be restored, fuel damage is probable.

With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.

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

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

In ,-EAL*.,--tThe 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 ton monitor (reactor vessel*..S [PAqrI or RPV RCSJr'I'RJ) level may be caused by itrmn atinad/or power failures, or water level dropping below the range of availab-le instrum~entation. ifwater level cannot be monflitored, operators ma" determine that an inventor,' loss is occurring by obseR'ing changsinsm and/or tank levels. Sump and/ot tank level changes must be evaluatedagaistothetl I al sources of water floewto ensu*re the" are indicativ~e of leakage from the (reactor ':ossel! RCS [PW4R] or RPV [BIA'R]).

Thisese EALs addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operationat Commercial NuclearPower Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Page 91 of 276

ATTACHMENT I EAL Bases Management.

IPage 92 of 276

ATTACHMENT 1 EAL Bases CPNPP Basis Reference(s):

1. IPO-010A/B Reactor Coolant System Reduced Inventory Operations

2. INC-6269 Calibration of the Mansell RCS Measurement System

3. SOP-101A/B Reactor Coolant System

4. Technical Specifications 3.9.4

5. OPT-408A/B Refueling Containment Penetration Verification

6. FRC-0.1A/B Response to Inadequate Core Cooling, Attachment 5

7. FSAR Section 6.2.5

8. FSAR Table 7.5-7A

9. CHM-1 11 Primary Chemistry Accident Assessment Sampling Program

10. NEI 99-01 CS1 I Page 93 of 276

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: Loss of RCS inventory affecting fuel clad integrity with containment challenged EAL:

CG1.2 General Emergency RCS level cannot be monitored for greater than or equal to 30 min. (Note 1)

AND Core uncovery is indicated by any of the following:

" UNPLANNED increase in any Table C-1 sump/tank level of sufficient magnitude to indicate core uncovery

• Erratic Source Range Monitor indication

" Greater than 20,000 R/hr on any of the following:

- CTEu16, Containment HRRM (u-RE-6290A)

- CTWu17, Containment HRRM (u-RE-6290B)

AND Any Containment Challenge indication, Table C-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 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required.

Table C-1 Sumps / Tanks

• Containment Sump I

• Containment Sump 2

• Reactor Cavity Sump

• CCW Surge TankA

• CCW Surge Tank B

• PRT

• RCDT Page 94 of 276

ATTACHMENT 1 EAL Bases Table C-2 Containment Challenge Indications

• CONTAINMENT CLOSURE not established (Note 6)

" Containment hydrogen concentration greater than 4%

• Unplanned rise greater than 1 psig in Containment pressure Mode Applicability:

5 - Cold Shutdown, 6 - Refueling Definition(s):

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. Containment closure means that all potential escape paths are closed or capable of being closed:

A. All penetrations providing direct access from Containment atmosphere to outside atmosphere are closed except:

" Penetrations with automatic valves capable of being closed by an operable CVI

" Penetrations under administrative controls (e.g., Control Room notified and designated person to close if required by fuel handling accident)

B. Equipment hatch is closed and held in place by 4 bolts, or is capable of being closed and held in place by 4 bolts C. One emergency airlock door is closed D. One personnel airlock door is capable of being closed 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.

CPNPP Basis:

In Cold Shutdown mode, the RCS will normally be intact and standard RCS level monitoring means are available.

In the Refueling mode, the RCS is not intact and RCS level may be monitored by different means, including the ability to monitor level visually.

In this EAL, all RCS water level indication would be unavailable for greater than 30 minutes, and the RCS inventory loss must be detected by indirect leakage indications (Table C-1).

Surveillance procedures provide instructions for calculating primary system leak rate by manual or computer-based water inventory balances. Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage. Ifthe make-up rate to the RCS Page 95 of 276

ATTACHMENT 1 EAL Bases unexplainably rises above the pre-established rate, a loss of RCS inventory may be occurring even if the source of the leakage cannot be immediately identified (ref. 1, 2).

The RCS inventory loss may be detected by the Containment High Range Radiation Monitor (HRRM) or erratic Source Range Monitor indication. As water level in the Reactor Vessel lowers, the dose rate above the core will rise. The dose rate due to this core shine should result in Containment High Range Radiation Monitor indication greater than 20,000 R/hr (ref.

3). The Containment HRRMs have a range of 1 E-1 to 1 E+8 R/hr.

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 (ref. 4, 5).

Three conditions are associated with a challenge to Containment integrity:

1. CONTAINMENT COSURE not established - The status of Containment closure is tracked if plant conditions change that could raise the risk of a fission product release as a result of a loss of decay heat removal (ref. 6, 7). If containment closure is re-established prior to exceeding the 30 minute core uncovery time limit then escalation to GE would not occur.

2. Containment hydrogen greater than 4% - The 4% hydrogen concentration threshold is generally considered the lower limit for hydrogen deflagrations. CPNPP is equipped with a Hydrogen Control System (HCS) which serves to limit or reduce combustible gas concentrations in the Containment. The plant has two hydrogen monitoring systems.

Each monitoring system consists of four sensor modules and one microprocessor analyzer. Two sensors from each Containment are coupled to one of the two hydrogen microprocessors located in the Control Room. Thus each microprocessor analyzer is shared by Units 1 and 2. The analyzer system has a range of 0-10% hydrogen by volume. The detector modules are located on the 905', 873', and 860' elevations in Containment. A fourth detector is located on 832' level across from the loop room entrance for loops 1 and 4. Hydrogen concentration is displayed in the Control Room on u-AI-5506A/B and u-AI-5506C/D. Hydrogen concentration can also be displayed on the Plant Computer. Alarms at -3% are provided for high hydrogen concentration, u-ALB-3A, window 3.7. If a hydrogen concentration value can not be obtained from the hydrogen monitoring system, a grab sample from the containment PIG radiation monitor may be used to determine the hydrogen concentration (ref. 8, 9, 10, 11).

3. UNPLANNED rise in Containment pressure - An unplanned pressure rise in containment while in cold Shutdown or Refueling modes can threaten Containment Closure capability and thus Containment potentially cannot be relied upon as a barrier to fission product release.

NEI 99-01 Basis:

This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity.

Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.

Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. IfRCS Page 96 of 276

ATTACHMENT 1 EAL Bases 9RCS/Fractor Vosel level cannot be restored, fuel damage is probable.

With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.

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

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

in 2.bt-The 30-minute criterion is tied to a readily recognizable event start time (i.e., the

-=n-lA total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.

The inability to monitor (reactor.. esscl,.RGS [Pr4'R] or RPV-RCS.BrK,) 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 (reaGte .vesse./RNCS [P-K/ or RPV [BI\'R,).

Thisese EALs addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1 449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plantsin the United States; and NUMARC 91-06, Guidelinesfor IndustryActions to Assess Shutdown Management.

Page 97 of 276

ATTACHMENT 1 EAL Bases CPNPP Basis Reference(s):

1. ABN-1 03 Excessive Reactor Coolant Leakage

2. ABN-108 Shutdown Loss of Coolant

3. Engineering Handbook, Guidelines for Events Beyond Design Basis: Spent Fuel Pools, Figure D "Dose Rate at Elevation 860' above Stored Fuel vs. Water Level Depth in SFP"

4. Severe Accident Management Guidance Technical Basis Report, Volume 1: Candidate High-Level Actions and Their Effects, pgs 2-18, 2-19

5. Nuclear Safety Analysis Center (NSAC), 1980, "Analysis of Three Mile Island - Unit 2 Accident," NSAC-1

6. Technical Specifications 3.9.4

7. OPT-408A/B Refueling Containment Penetration Verification

8. FRC-O.1A/B Response to Inadequate Core Cooling, Attachment 5

9. FSAR Section 6.2.5

10. FSAR Table 7.5-7A

11. CHM-1 II Primary Chemistry Accident Assessment Sampling Program

12. NEI 99-01 CG1 Page 98 of 276

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 2 - Loss of Emergency AC Power Initiating Condition: Loss of all but one AC power source to safeguard buses for 15 minutes or longer EAL:

CU2.1 Unusual Event AC power capability, Table C-3, to 6.9 KV safeguard buses uEA1 and uEA2 reduced to a single power source for greater than or equal to 15 min. (Note 1)

AND Any additional single Table C-3 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-3 AC Power Sources Offsite:

• 138 KV switchyard circuit

• 345 KV switchyard circuit Onsite:

" uEG1

• uEG2 Mode Applicability:

5 - Cold Shutdown, 6- Refueling, D - Defueled Definition(s):

SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):

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

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

Page 99 of 276

ATTACHMENT 1 EAL Bases CPNPP Basis:

For emergency classification purposes, "capability" means that an offsite AC power source(s) is available to the emergency buses, whether or not the buses are powered from it.

The condition indicated by this EAL is the degradation of the offsite and onsite power sources such that any additional single failure would result in a loss of all AC power to the emergency buses.

The safeguards AC distribution system power sources consist of the preferred and alternate offsite power sources, and the onsite standby emergency diesel generators uEGI and uEG2.

Offsite power is supplied to the plant switchyards from the transmission network by five 345 KV and two 138 KV transmission lines. From the switchyards, two electrically and physically separated circuits provide AC power through step down startup transformers, to the 6.9 kV safeguard buses. The 138 kV switchyard circuit is the preferred source for Unit 2 and alternate source for Unit 1. The 345 KV circuit is the preferred source for Unit 1 and alternate source for Unit 2. The onsite AC distribution system is divided into redundant trains so that the loss of any one load group does not prevent the minimum safety functions from being performed. Each train has connections to two offsite power sources and a dedicated diesel generator. Each offsite circuit can supply the Unit 1 and Unit 2 6.9 KV safeguard buses. (ref. 1, 2, 3, 4).

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

NEI 99-01 Basis:

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

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

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

An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an 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 all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from the unit main generator.

• A loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from an offsite power source.

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

The subsequent loss of the remaining single power source would escalate the event to an Alert in accordance with IC CA2.

Page 100 of 276

ATTACHMENT 1 EAL Bases CPNPP Basis Reference(s):

1. FSAR Figure 8.3-1

2. FSAR Section 8.2

3. FSAR Section 8.3

4. Technical Specifications B3.8.1

5. ABN-601 Response to a 138/345 KV System Malfunction

6. ABN-602 Response to a 6900/480V System Malfunction

7. NEI 99-01 CU2 Page 101 of 276

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 2 - Loss of Emergency AC Power Initiating Condition: Loss of all offsite and all onsite AC power to safeguard buses for 15 minutes or longer EAL:

CA2.1 Alert Loss of all offsite and all onsite AC power capability, Table C-3, to 6.9 KV safeguard buses uEA1 and uEA2 for greater than or equal to 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 C-3 AC Power Sources Offsite:

• 138 KV switchyard circuit

• 345 KV switchyard circuit Onsite:

" uEG1

• uEG2 Mode Applicability:

5 - Cold Shutdown, 6 - Refueling, D - Defueled CPNPP Basis:

For emergency classification purposes, "capability" means that an offsite AC power source(s) is available to the emergency buses, whether or not the buses are powered from it.

The safeguards AC distribution system power sources consist of the preferred and alternate offsite power sources, and the onsite standby emergency diesel generators uEGI and uEG2.

Offsite power is supplied to the plant switchyards from the transmission network by five 345 KV and two 138 KV transmission lines. From the switchyards, two electrically and physically separated circuits provide AC power through step down startup transformers, to the 6.9 kV safeguard buses. The 138 kV switchyard circuit is the preferred source for Unit 2 and alternate source for Unit 1. The 345 KV circuit is the preferred source for Unit I and alternate source for Unit 2. The onsite AC distribution system is divided into redundant trains so that the loss of any one load group does not prevent the minimum safety functions from being performed. Each train has connections to two offsite power sources and a dedicated diesel generator. Each offsite circuit can supply the Unit 1 and Unit 2 6.9 KV safeguard buses. (ref. 1, 2, 3, 4)

This cold condition EAL is equivalent to the hot condition loss of all offsite AC power EAL SS1.1.

I Page 102 of 276 1

ATTACHMENT 1 EAL Bases NEI 99-01 Basis:

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

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

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

Escalation of the emergency classification level would be via IC CS1 or AS4-RS1.

CPNPP Basis Reference(s):

1. FSAR Figure 8.3-1

2. FSAR Section 8.2

3. FSAR Section 8.3

4. Technical Specifications B3.8.1

5. ABN-601 Response to a 138/345 KV System Malfunction

6. ABN-602 Response to a 6900/480V System Malfunction

7. NEI 99-01 CA2 Page 103 of 276 1

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: UNPLANNED increase in RCS temperature EAL:

CU3.1 Unusual Event UNPLANNED increase in RCS temperature to greater than 200OF due to loss of decay heat removal capability (Note 9)

Note 9: Begin monitoring hot condition EALs concurrently.

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.

CPNPP Basis:

Several instruments are capable of providing indication of RCS temperature with respect to the Technical Specification cold shutdown temperature limit (200'F, ref. 1). These include loop Thot (u-TR-413A/23A, u -TR-433A/43A, u -TI-413A, u -TI-423A) and, if no RCPs are operating, the Core Exit Thermocouples (TCs). The most limiting temperature indication should be used. For example, during heatup, the highest reading temperature indication should be used; during cooldown, the lowest (ref. 2, 3, 4, 5).

In the absence of reliable RCS temperature indication caused by the loss of decay heat removal capability, classification should be based on the RCS pressure increase criteria when the RCS is intact in Mode 5 or based on time to boil data when in Mode 6 or the RCS is not intact in Mode 5.

The note is a reminder that any temperature increase above 2001F is an operating mode change from cold to hot conditions. Since each EAL is associated with operating mode applicability, the set of EALs that must be monitored must now include EALs associated with hot condition operating modes.

NEI 99-01 Basis:

This IC addresses an UNPLANNED increase in RCS temperature above the Technical Specification cold shutdown temperature limit.,or the inability to determinc RCS t.mpc.atur.

aadlevel-and represents a potential degradation of the level of safety of the plant. If the RGS RCS-is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Director Emerqency 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.

Page 104 of 276

ATTACHMENT 1 EAL Bases I EAL-#4This EAL involves a loss of decay heat removal capability, or an addition of heat to the RCS in excess of that which can currently be removed, such that reactor coolant temperature cannot be maintained below the cold shutdown temperature limit specified in Technical Specifications. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation.

During an outage, the level in the reactor vessel will normally be maintained at or above the reactor vessel flange. Refueling evolutions that lower water level below the reactor vessel flange are carefully planned and controlled. A loss of forced decay heat removal at reduced inventory may result in a rapid increase in reactor coolant temperature depending on the time after shutdown.

EAL #2 reflects aonldition whore there has been a Significantloss orf instrmentation capability ecessaryont moritor pcoditions RCS and operators would be unable to moaitor key pa1.reters neessarp' to assure core decay heat removal. During this condition, there is no Mimmednate threat fuelof damnage beause the core decay heat load has been reduced ince the cessation of power operatien.

Fifteen mninutes was selected as a threshold to exclude tr-anSient or moementar,' loes ofIRdeation-Escalation to Alert would be via IC CAl based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.

CPNPP Basis Reference(s):

1. Technical Specifications Table 1.1-1

2. IPO-OO5N/B Plant Cooldown From Hot Standby To Cold Shutdown

3. Technical Specifications 3.4.3

4. OPT-407 RCS Pressure and Temperature Verification

5. IPO-01 OA/B Reactor Coolant System Reduced Inventory Condition

6. NEI 99-01 CU3 Page 105 of 276 1

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: UNPLANNED increase in RCS temperature EAL:

CU3.2 Unusual Event Loss of all RCS temperature and RCS level indication for greater than or equal to 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 CPNPP Basis:

RCS level can be monitored by one or more of the following (ref. 2, 3, 4):

" RCS Level Wide Range u-LI-3615B

• RCS Level Narrow Range u-LI-3615A

• RCS Extended Wide Range u-LI-3615C

• Mansell Level Monitor System u-LT-3619A/B/C-1, -2

• Plant Computer

• RVLIS

• Ultrasonic Level monitoring (optional)

Several instruments are capable of providing indication of RCS temperature with respect to the Technical Specification cold shutdown temperature limit (200'F, ref. 1). These include loop Thot (u-TR-413A/23A, u -TR-433A/43A, u -TI-413A, u -TI-423A) and, if no RCPs are operating, the Core Exit Thermocouples (TCs). The most limiting temperature indication should be used. For example, during heatup, the highest reading temperature indication should be used; during cooldown, the lowest (ref. 5, 6, 7, 8).

In the absence of reliable RCS temperature indication caused by the loss of decay heat removal capability, classification should be based on heat up rate data or additionally if in Mode 5 with RCS intact on pressure increase.

NEI 99-01 Basis:

This IG-EAL addresses an UNPLANNED i en RS temperature abo'... the TechniGal Specific3tio c.ld.shutdown temperature limit, or the inability to determine RCS temperature Page 106 of 276

ATTACHMENT 1 EAL Bases and level, and represents a potential degradation of the level of safety of the plant. Ifthe RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency DiroctorEmerqency Coordinator should also refer to IC CA3.

A momn*, tar,' UNPLANNED ,Xcur"sin above the Technir*c Rpecification cold 4

shutd-n,

+ nnr 4..Ir i n-,. *.,kn 4 k kn 4- .- rn ,-,I.v 4n4.a ,-I.k1 H-. AlC +in i M, +;

E=AL #1 involves a loss of decay heat remoeval capability, or an addition of heat to the RCS in exGess ef that Which can currntly b8 rFemroved, such tha-t rPea"t*Fr coolart temperatuFe cannot be mnaintained below the coldG sh-utdown40 temperature limit specified inTechnic~al Specifications. DUFiRg this conditinR, there is no immediate threat of fuel damage because the core decay heat lead has been reduced s the cessatio of power opeFatinr.

Duarig an outage, the levol in the reactor ve-s-sel will nrmally be Maintained above the reactor vessel flange. Refueling evoluitions that lower water level below the reactor vessel flange are carefully planned and controlled. A loss-o-f forc-ed decay heat Frnemval at reduced inventei' may result in a rapid increase ireactor coolant temperature depending en the time after shutdown.

EAl= #2This EAL reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor RCS conditions and operators would be unable to monitor key parameters necessary to assure core decay heat removal. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation.

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

Escalation to Alert would be via IC CA1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.

CPNPP. Basis Reference(s):

1. Technical Specifications Table 1.1-1

2. IPO-01OA/B Reactor Coolant System Reduced Inventory Operations

3. INC-6269 Calibration of the Mansell RCS Measurement System

4. SOP-101A/B Reactor Coolant System

5. IPO-005A/B Plant Cooldown From Hot Standby To Cold Shutdown

6. Technical Specifications 3.4.3

7. OPT-407 RCS Pressure and Temperature Verification

8. IPO-01OA/B Reactor Coolant System Reduced Inventory Condition

9. NEI 99-01 CU3 Page 107 of 276

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: Inability to maintain plant in cold shutdown EAL:

CA3.1 Alert UNPLANNED increase in RCS temperature to greater than 200OF for greater than Table C-4 duration (Notes 1, 9)

OR UNPLANNED RCS pressure increase greater than 10 psig due to a loss of RCS cooling (This EAL 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: Begin monitoring hot condition EALs concurrently.

Table C-4: RCS Heat-up Duration Thresholds CONTAINMENT RCS Status CLOSURE Status Heat-up Duration Intact (but not REDUCED N/A 60 min.*

INVENTORY)

Not intact Established 20 min.*

OR REDUCED INVENTORY Not established 0 min.

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

Mode Applicability:

5 - Cold Shutdown, 6 - Refueling Definition(s):

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. Containment closure means that all potential escape paths are closed or capable of being closed:

A. All penetrations providing direct access from Containment atmosphere to outside atmosphere are closed except:

" Penetrations with automatic valves capable of being closed by an operable CVI

" Penetrations under administrative controls (e.g., Control Room notified and designated person to close if required by fuel handling accident)

Page 108 of 276

ATTACHMENT 1 EAL Bases B. Equipment hatch is closed and held in place by 4 bolts, or is capable of being closed and held in place by 4 bolts C. One emergency airlock door is closed D. One personnel airlock door is capable of being closed 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.

REDUCED INVENTORY - Plant condition when fuel is in the reactor vessel and Reactor Coolant System level is < 80 inches above core plate (829'8").

CPNPP Basis:

Several instruments are capable of providing indication of RCS temperature with respect to the Technical Specification cold shutdown temperature limit (200'F, ref. 1). These include loop Thot (u-TR-413A/23A, u -TR-433A/43A, u -TI-413A, u -TI-423A) and, if no RCPs are operating, the Core Exit Thermocouples (TCs). The most limiting temperature indication should be used. For example, during heatup, the highest reading temperature indication should be used; during cooldown, the lowest (ref. 2, 3, 4, 5).

In the absence of reliable RCS temperature indication caused by the loss of decay heat removal capability, classification should be based on heat up rate data or additionally if in Mode 5 with RCS intact on pressure increase.

A 10 psig RCS pressure increase can be monitored on u-PI-403A and computer points P6498A and P6499A (ref. 9, 10).

The status of Containment closure is tracked if plant conditions change that could raise the risk of a fission product release as a result of a loss of decay heat removal (ref. 6, 7).

The note is a reminder that any temperature increase above 200OF is an operating mode change from cold to hot conditions. Since each EAL is associated with operating mode applicability, the set of EALs that must be monitored must now include EALs associated with hot condition operating modes.

NEI 99-01 Basis:

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

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

The RGS-RCSHeat-up Duration Thresholds table addresses an increase in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact, or RCS inventory is reduced (e.g., mid-loop operation i-R PWRs). The 20-minute criterion was included to allow time for operator action to address the temperature increase.

The RCS Heat-up Duration Thresholds table also addresses an increase in RCS temperature with the RCS intact. The status of CONTAINMENT CLOSURE is not crucial in this condition since the intact RCS is providing a high pressure barrier to a fission product release. The 60-Page 109 of 276

ATTACHMENT 1 EAL Bases minute time frame should allow sufficient time to address the temperature increase without a substantial degradation in plant safety.

Finally, in the case where there is an increase in RCS temperature, the RCS is not intact or is at reduced inventory-fP-WR4, 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.

EAL4#2The RCS pressure increase threshold provides a pressure-based indication of RCS heat-up in the absence of RCS temperature monitoring capability.

Escalation of the emergency classification level would be via IC CS1 or A-S4RS1.

CPNPP Basis Reference(s):

1. Technical Specifications Table 1.1-1

2. IPO-005A/B Plant Cooldown From Hot Standby To Cold Shutdown

3. Technical Specifications 3.4.3

4. OPT-407 RCS Pressure and Temperature Verification

5. IPO-01OA/B Reactor Coolant System Reduced Inventory Condition

6. Technical Specifications 3.9.4

7. OPT-408A/B Refueling Containment Penetration Verification

8. IPO-010A/B Reactor Coolant System Reduced Inventory Operations

9. IPO-005AIB Plant Cooldown From Hot Standby To Cold Shutdown

10. SOP-101A/B Reactor Coolant System Reduced Inventory

11. NEI 99-01 CA3 I Page 110 of 276

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 Less than 105 VDC bus voltage indications on Technical Specification required 125 VDC buses (uED1, uED2, uED3, uED4) for greater than or equal to 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 CPNPP Basis:

The purpose of this EAL is to recognize a loss of DC power compromising the ability to monitor and control the removal of decay heat during cold shutdown or refueling operations. This EAL is intended to be anticipatory in as much as the operating crew may not have necessary indication and control of equipment needed to respond to the loss. The fifteen minute interval is intended to exclude transient or momentary power losses.

The safeguards 125 VDC buses are the Class 1E buses uED1, uED2, uED3 and uED4 (ref. 1).

The 125 VDC safeguard distribution system is illustrated in Figure C-2 (ref. 2, 3).

Each redundant safeguards 125 VDC system consists of two independent batteries each having one main distribution bus, two static battery chargers (one spare), and local distribution panels. For Unit 1, batteries BT1 ED1 and BT1 ED3 feed all train A load requirements, while batteries BT1 ED2 and BT1 ED4 supply train B load requirements.

For Unit 2, batteries BT2ED1 and BT2ED3 feed all train A load requirements, while batteries BT2ED2 and BT2ED4 supply train B load requirements. There are no bus ties or sharing of power supplies between redundant trains (ref. 1).

Minimum DC bus voltage is 105 VDC (ref. 4). Bus voltage may be monitored from the following indications (ref. 6):

Control Room Panel CP-10 Annunciator u--ALB-10B Plant Computer V-1ED1, 125VDC SWITCH PNL 1ED1 VOLT 1.13 V6501A BATT BT1ED1 VOLT V-1 ED2, 125VDC SWITCH PNL 1ED2 VOLT 2.13 V6502A BATT BT1ED2 VOLT V-1ED3, 125VDC SWITCH PNL 1ED3 VOLT 1.9 V6503A BATT BT1ED3 VOLT V-1 ED4, 125VDC SWITCH PNL 1ED4 VOLT 3.9 V6504A BATT BT1ED4 VOLT This EAL is the cold condition equivalent of the hot condition loss of DC power EAL SS7.1.

Page 111 of 276

ATTACHMENT 1 EAL Bases NEI 99-01 Basis This IC addresses a loss of vital DC power which compromises the ability to monitor and control operable SAFETY SYSTEMS when the plant is in the cold shutdown or refueling mode.

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

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

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

Depending upon the event, escalation of the emergency classification level would be via IC CA1 or CA3, or an IC in Recognition Category AR.

CPNPP Basis Reference(s):

1. FSAR 8.3.2

2. FSAR Figure 8.3-14

3. FSAR Figure 8.3-14A

4. ECA-O.OA/B Loss of All AC Power

5. SOP-605ANB 125 VDC Switchgear and Distribution Systems, Batteries and Battery Chargers

6. ALM-0102A/B Alarm Procedures Manual, u-ALB-1OB, nos. 1.9, 1.13, 2.13, 3.8

7. NEI 99-01 CU4 Page 112 of 276

ATTACHMENT I EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 5 - Loss of Communications Initiating Condition: Loss of all onsite or offsjte communications capabilities EAL:

CU5.1 Unusual Event Loss of all Table C-5 onsite communication methods OR Loss of all Table C-5 offsite communication methods OR Loss of all Table C-5 NRC communication methods Table C-5 Communication Methods System Onsite Offsite NRC Gai-tronics Page/Party (PA) X Plant Radios X PABX X X X Public Telephone X X X Federal Telephone System (FTS) X X Mode Applicability:

5 - Cold Shutdown, 6 - Refueling, D - Defueled Definition(s):

None CPNPP Basis:

Onsite/offsite communications include one or more of the systems listed in Table C-5 (ref. 1, 2).

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

NEI 99-01 Basis:

This IC addresses a significant loss of on-site or offsite communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to OROs and the NRC.

Page 113 of 276

ATTACHMENT 1 EAL Bases 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.).

I EAL-#4The first EAL condition addresses a total loss of the communications methods used in support of routine plant operations.

EAL #2The second EAL condition addresses a total loss of the communications methods used to notify all OROs of an emergency declaration. The offsite (OROs. referred to here are-(see DevelGpe.-N.tee) the State Department of Public Safety, Somervell and Hood County EOCs7 EAL4#3The third EAL addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.

CPNPP Basis Reference(s):

1. FSAR 9.5.2

2. DBD-EE-048 Communication System

3. NEI 99-01 CU5 I Page 114 of 276 1

ATTACHMENT I 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-6 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-6 Hazardous Events

• Seismic event (earthquake)

• Internal or external FLOODING event

• High winds or tornado strike

" FIRE

• EXPLOSION

" Other events with similar hazard characteristics as determined by the Emergency Coordinator Mode Applicability:

5 - 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.

Page 115 of 276

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.

CPNPP 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 (ref. 3, 4).

• Seismic Category I structures are analyzed to withstand a sustained, design wind velocity of at least 80 mph. (ref. 5).

• Areas containing functions and systems required for safe shutdown of the plant are identified by fire area (ref. 6, 7).

• 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.

NEI 99-01 Basis:

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.

EAL 4.b.'The first conditional addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.

Page 116 of 276

ATTACHMENT 1 EAL Bases EAL l1b.2The 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 A-S--RSI.

CPNPP Basis Reference(s):

1. ABN-907 Acts of Nature

2. CPNPP PRA Accident Sequence Analysis "Internal Flooding Sequences"

3. FSAR Section 2.4.3.7 Flood Evaluations for Safe Shutdown Impoundment

4. DBD-CS-071 Maximum Probable Flood

5. FSAR Section 3.3.1.1 Wind Loadings

6. CPNPP Fire Protection Report, Section 5.0 "Fire Safe Shutdown Equipment List"

7. FSAR Section 7.4 Systems Required for Safe Shutdown

8. NEI 99-01 CA6 Page 117 of 276

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 Technolo-gy 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.

Page 118 of 276 1

ATTACHMENT I EAL Bases

7. Emeraencv Coordinator JudQment 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.

Page 119 of 276

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 Supervisor 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 CPNPP 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 CPNPP. 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).

CPNPP Basis:

The security shift supervision is defined as the Security Shift Supervisor.

This EAL is based on the CPNPP Safeguards Contingency Plan (ref. 1).

NEI 99-01 Basis:

This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR § 73.71 or 10 CFR § 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HA1, 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. Classification of these events will initiate appropriate threat-related notifications to plant personnel and Offsite Response Orqanizations.

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Trainingand Qualification Plan, Safeguards Contingency Plan [aP4 ndoondPnt Spent Fuel Storage Installation Se276iy P-rogram.

Page 120 of 276

ATTACHMENT 1 EAL Bases

&AL--#-This EAL references (site speGifiG-the seGU-ity-_shift .s.pewi**R Shift Security Supervisor because these are the individuals trained to confirm that a security event is occurring or has occurred. Training on security event confirmation and classification is controlled due to the nature of Safeguards and 10 CFR § 2.39 information.

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 CPNPP Safeguards Contingency Plan (ref. 1).

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

CPNPP Basis Reference(s):

1. CPNPP Safeguards Contingency Plan (Safeguards)

2. NEI 99-01 HU1 i

Page 121 of 276

ATTACHMENT 1 EAL Bases Category: H - Hazards Subcategory: 1 - Security Initiating Condition: Confirmed SECURITY CONDITION or threat EAL:

HUI.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.

HOSTILE ACTION- An act toward CPNPP 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 CPNPP. 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).

CPNPP Basis:

The security shift supervision is defined as the Security Shift Supervisor.

This EAL is based on the CPNPP Safeguards Contingency Plan (ref. 1).

NEI 99-01 Basis:

This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR § 73.71 or 10 CFR § 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HA1, 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. Classification of these events will initiate appropriate threat-related notifications to plant personnel and Offsite Response Orqanizations.

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and QualificationPlan, Safeguards Contingency Plan [aRd ndpcndent Spent Fuel Storage installation Seur.ty Po.gra.,. EAL #1 references (site security Shift supeiion) because these are the i specific d trained to cnR that security event isocurner has occGurred. Trainin on security event confirmation and Page 122 of 276

ATTACHMENT 1 EAL Bases cla.sifi*ation is cRntrlled due to the nature of Saf.guards and 10 CIFR § 2.39 i*fermation.

EAL #2 This EAL addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with (site-specific procedure).

EA6L#3 addresses the threat from thoe impat* of an aircraft on the plant. The NRI m Ieadguartors Operations Officer (HOO) Will commUnicate to the licensee if the threat innvoVes an aircraft. The status and size of thc plane mnay also be provided by NORA.D through the NRC. Valid-ation of the threat is performed in accordance with (site specific proceduro).

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 IEscalation of the emergency classification level would be via IC HA1.

such as the CPNPP Safequards Contingrency Plan (ref. 1).

CPNPP Basis Reference(s):

1. CPNPP Safeguards Contingency Plan (Safeguards)

2. NEI 99-01 HU1 Page 123 of 276

ATTACHMENT 1 EAL Bases Category: H - Hazards Subcategory: I - Security Initiating Condition: Confirmed SECURITY CONDITION or threat EAL:

HUl.3 Unusual Event A validated notification from the NRC providing information of an aircraft threat Mode Applicability:

All Definition(s):

None CPNPP Basis:

This EAL is based on the CPNPP Safeguards Contingency Plan (ref. 1).

NEI 99-01 Basis:

This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR § 73.71 or 10 CFR § 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HA1, 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. Classification of these events will initiate appropriate threat-related notifications to plant personnel and Offsite Response Orqanizations.

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and QualificationPlan, Safeguards Contingency Plan [a&R EAL-#3This EAL addresses the threat from the impact of an aircraft on the plant. The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may also be provided by NORAD through the NRC. Validation of the threat is performed in accordance with (site-specific procedure).

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 CPNPP Safeguards Continqency Plan (ref. 1).

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

CPNPP Basis Reference(s):

1. CPNPP Safeguards Contingency Plan (Safeguards)

Page 124 of 276

ATTACHMENT 1 EAL Bases

2. NEI 99-01 HU1 Category: H - Hazards Subcategory: 1 - Security Initiating Condition: HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes EAL:

HAI.1 Alert A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the Security Shift Supervisor Mode Applicability:

All Definition(s):

HOSTILE ACTION - An act toward CPNPP 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 CPNPP. 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 - As shown in CPNPP Emergency Plan Appendix E, Complex and Owner Controlled Area.

CPNPP Basis:

The security shift supervision is defined as the Security Shift Supervisor.

NEI 99-01 Basis:

This IC addresses the occurrence of a HOSTILE ACTION within the OWNER CONTROLLED AREA or notification of an aircraft attack thrcFat. 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 uperiAsioR -Supervisor and the Control Room is essential for proper classification of a security-related event.

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan-[aiwd Indopondont Spent Fu~el Storage Installation Se G1rFtY P-ro grFam.

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 I Organizations (OROs), allowing them to be better prepared should it be necessary to consider further actions.

Page 125 of 276

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.

EAL-#!This EAL 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.

EAL #taddresses the threat f9Rom the imnpact of an aircraft on the plant, and the anticipated notific-';ations arc made in a timely mRannr so that plant personnRel and OROsr are in heightened state of readireSS. This EAL is met when the threat related iRnformation has been va-lid,-afted on aG,,,,G l-Jn,'aR w^ith (site sp,,*rfi,, fifa ft-.*Lh* Fr,*redr~ei The NRC Headquarters Operations Officer (HO00) Will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may be provided by NORAID thr.ugh the NRC.

Inn Sme cases, it ma"no* t be readily apparent if an aircraft impat within the OWNER CONTROLLED AREA was intentional (i.e., a HOSTILE ACTION). it is expected, although not certain, that notification A ' an appropriate Federal agency to the site would .larif,' this point.÷ In by this ease, the appropriate federal agency is intended to be NORAID, FBI, FA. or NRC. The emnergency declar-ation, including onRe -b-asedon other l~siFEALs, should not be unduly delaye while awaiting notification by a Federal agcncy.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 CPNPP Safeguards Contingiency Plan (ref. 1).

CPNPP Basis Reference(s):

1. CPNPP Safeguards Contingency Plan (Safeguards)

2. NEI 99-01 HA1 IPage 126 of 276

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

HA1.2 Alert A validated notification from NRC of an aircraft attack threat within 30 min. of the site Mode Applicability:

All Definition(s):

None CPNPP Basis:

None NEI 99-01 Basis:

This IC addresses the occurrence of a HOSTILE ACTION withiR 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.

I Timely and accurate communications between the Security Shift Su pe iSieR-Supervisor and the Control Room is essential for proper classification of a security-related event.

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and QualificationPlan, Safeguards Contingency Plan-[and Independent Spent Fue! Storage Installation SocuritY P-rogram]ý1.

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.

Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72.EAL #1 is applicable for any H"OSTILn E ACTPI"ION oc-curring, or that has

-occured, in the OWNEA R CON*

NTROLLE*D*l E D AR EA This inql-cludep an',' action di*e*ted against an ISSI-that is loc-Gated outside the plant [AREA PROTECTED EAL-#2This EAL addresses the threat from the impact of an aircraft on the plant, and the anticipated arrival time is within 30 minutes. The intent of this EAL is to ensure that threat-related notifications are made in a timely manner so that plant personnel and OROs are in a Page 127 of 276

ATTACHMENT 1 EAL Bases 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 CPNPP Safeguards Contingency Plan (ref. 1).

CPNPP Basis Reference(s):

1. CPNPP Safeguards Contingency Plan (Safeguards)

2. NEI 99-01 HAI Page 128 of 276

ATTACHMENT 1 EAL Bases Category: H - Hazards Subcategory: 1 - Security Initiating Condition: HOSTILE ACTION within the PROTECTED AREA EAL:

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

All Definition(s):

HOSTILE ACTION -,An act toward CPNPP 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 CPNPP. 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 FSAR Figure 1.2-1 Plot Plan.

CPNPP Basis:

The security shift supervision is defined as the Security Shift Supervisor.

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 CPNPP Safeguards Contingency Plan (Safeguards) information. (ref. 1)

NEI 99-01 Basis:

This IC addresses the occurrence of a HOSTILE ACTION within the PROTECTED AREA.

This event will require rapid response and assistance due to the possibility for damage to plant equipment.

Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event.

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 [a&R Independent Spent Fue! Storage. n'stallatioAn Security Pregram].

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 Page 129 of 276

ATTACHMENT 1 EAL Bases sheltering). The Site Area Emergency declaration will mobilize Offsite Response Organization (ORO) resources and have them available to develop and implement public protective actions in the unlikely event that the attack is successful in impairing multiple safety functions.

This 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 HAl. It also does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72.

Emergency plans and implementing procedures are public documents; therefore, EALs 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 CPNPP Safeguards Contingency Plan (ref. 1).

Escalation of the emergency classification level would be via IC HG1.

CPNPP Basis Reference(s):

1. CPNPP Safeguards Contingency Plan (Safeguards)

2. NEI 99-01 HS1 Page 130 of 276

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:

" One or more of the following safety functions cannot be controlled or maintained

- Reactivity control

- Core cooling

- RCS heat removal OR

" Damage to spent fuel has occurred or is IMMINENT Mode Applicability:

All Definition(s):

HOSTILE ACTION - An act toward CPNPP 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 CPNPP. 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 FSAR Figure 1.2-1 Plot Plan.

CPNPP Basis:

The security shift supervision is defined as the Security Shift Supervisor.

NEI 99-01 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 Page 131 of 276 I

ATTACHMENT 1 EAL Bases fuel pool cooling system (e.g., pumps, heat exchangers, controls, etc.) or, 2) loss of spent fuel pool integrity such that sufficient water level cannot be maintained.

Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event.

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fu4el Storage Installation Socbrity P-rogram].

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 CPNPP Safeguards Contingency Plan (ref.1).

CPNPP Basis Reference(s):

1. CPNPP Safeguards Contingency Plan (Safeguards)

2. NEI 99-01 HG1 Page 132 of 276

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 2 - Seismic Event Initiating Condition: Seismic event greater than OBE level EAL:

HU2.1 Unusual Event Seismic event greater than OBE as indicated by annunciator 2A-3.1, OBE EXCEEDED, or yellow OBE light on Seismic Monitoring system panel Mode Applicability:

All Definition(s):

None CPNPP Basis:

Seismic events of this magnitude can result in a areas needed for safe shutdown being subjected to forces beyond design limits, and thus damage may be assumed to have occurred to plant safety systems.

A conservative Safe Shutdown Earthquake (SSE) having a peak horizontal ground acceleration at the top of bedrock of 0.12 g has been selected for design (FSAR Section 2.5.2.6). The Operating Basis Earthquake (OBE) is equal to 1/2 the SSE (ref. 1).

When the seismic recorder indicates that the OBE has been exceeded, System Engineering must evaluate and determine whether the reactor must be shut down and remain shutdown until inspection of the facility shows that no damage has been incurred which would jeopardize safe operation of the facility or until such damage is repaired. CPNPP was designed such that, for ground motion less than the OBE, the features of the plant necessary for continued operation without undue risk to the health and safety of the public will remain functional. Any ground motion in excess of this results in an uncertainty as to the extent of the damage which must be resolved before continued operation can be considered safe (ref. 2).

The seismic trigger, CP1-SIATAS-03, is set to activate the strong motion recording system at an acceleration level slightly above normal ambient vibrations (0.01g) and well below the postulated OBE "free field" ground acceleration (0.06g horizontal). This causes an alarm in the control room to alert the operator. (ref. 2, 3) The seismic recorders (strong motion accelerators) monitor earth vibration and, when triggered, store data in the recorder. Triaxial SMAs are installed at appropriate locations to provide data on the frequency, amplitude, and phase relationship of the seismic response of the containment structure and the seismic input to other seismic Category I structures, systems, and components. The Seismic Instrumentation consists of strong motion accelerograph (triaxial time history accelerograph system), triaxial peak accelerograph recorders, passive response spectrum recorders, a response spectrum switch, and a seismic switch. Except for sensors for the active instrumentation, all electronics for processing and storage of the seismic data are located in the seismic instrumentation panel CPX-ECPRCV-1 1 in the control room. There is no additional seismic instrumentation required for Unit 2. However, alarms from seismic instrumentation in Unit 1 are duplicated in Unit 2. The Page 133 of 276

ATTACHMENT 1 EAL Bases time history accelerograph system is fully operational within 0.1 second after the seismic trigger is actuated.

It will operate continuously during the period in which the earthquake exceeds the seismic trigger threshold (0.01g) plus 5 seconds (minimum) beyond the last seismic trigger signal.

ABN-907 Acts of Nature provides the guidance for determining if the OBE earthquake threshold is exceeded and any required response actions. (ref. 2)

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 CPNPP. Alternatively, near real-time seismic activity can be accessed via the NEIC website:

http://earthquake.usgs.gov/earthquakes/dyfi/archives.php NEI 99-01 Basis:

This IC addresses a seismic event that results in accelerations at the plant site greater than those specified for an Operating Basis Earthquake (OBE). An earthquake greater than an OBE but less than a Safe Shutdown Earthquake (SSE) should have no significant impact on safety-related systems, structures and components; however, some time may be required for the plant staff to ascertain the actual post-event condition of the plant (e.g., performs walk-downs and post-event inspections). Given the time necessary to perform walk-downs and inspections, and fully understand any impacts, this event represents a potential degradation of the level of safety of the plant.

Event verification with external sources should not be necessary during or following an OBE.

Earthquakes of this magnitude should be readily felt by on-site personnel and recognized as a seismic event (e.g., lateral accelerations in excess of 0.08g). The Shift Manager or Emergency IDEeeter-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.

CPNPP Basis Reference(s):

1. FSAR Section 2.5.4.9 Earthquake Design Basis

2. ABN-907 Acts of Nature

3. DBD-EE-077 Seismic Instrumentation

4. 1, 2-ALB-2A-3.1 OBE EXCEEDED

5. DBD-ME-028 Classification of Structures, Systems and Components

6. NEI 99-01 HU2 Page 134 of 276

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technology Hazard Initiating Condition: Hazardous event EAL:

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

All Definition(s):

PROTECTED AREA - 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 FSAR Figure 1.2-1 Plot Plan.

CPNPP Basis:

Response actions associated with a tornado onsite is provided in ABN-907 Acts of Nature (ref.

1).

Ifdamage 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.

NEI 99-01 Basis:

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

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

EAL #2 addresses flooding of a building roomn or area that results in operators isolating powcr to a SAFETY SYSTEM compnenRGt du-e to wgater level oer o-ther wetting cARonerns. Cla~ssificloatio~n.

is also rcguircd if the water level or rclated Wetting causcs an automatic isolation of a SAFEP SYSTEM comnponent from its power source (e.g., a breaker or relay trip). To9 warrn classification, operability of the affected component must be reguircd by Tcehnical Specifications for the current operating moede.-

EAL #3 addresses a hadosmtrlsevent originating at an eff-sitc loc-ation and of sufiietmagnitude to impede the moevement Of peF8rRonnl Within t-hc PROTEC=GTEFD AREA.

EAL #4 addrcsscs a hazardous event that c~auses an on site impedimcnet to Vehicle moVment and Significant enough topo ibthe plant staff from accesSing the site usngprsoal vehcis.Exape ofsuch. an. event include site flooding caused by -ahuMrricamne, hcaVY rains, Page 135 of 276

ATTACHMENT 1 EAL Bases rier wter releases, damn failure, etc., Or an on site train derailment blocking the access up This EAL is R*t inteRded apply to routine impediments such as fog, snow, i.e, or Vehi-le breakdowns or accidents, but rather to mrnee significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the flooding around the Cooper Station during the Midw~est floods of 1993, or the flooding aroun~d Ft. Calhoun Station in 2011.

EAL #5 addresses (site specific description).

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

CPNPP Basis Reference(s):

1. ABN-907 Acts of Nature

2. NEI 99-01 HU3 I Page 136 of 276

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technology Hazard Initiating Condition: Hazardous event EAL:

HU3.2 Unusual Event Internal room or area FLOODING of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component needed for the current operating mode Mode Applicability:

All Definition(s):

FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.

SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):

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

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

CPNPP Basis:

The internal flooding areas of concern are the Safeguards Building and Turbine Building (ref.1). Refer to EAL CA6.1 for internal flooding affecting one or more SAFETY SYSTEM trains.

NEI 99-01 Basis:

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

EAL #1 addresses a tornado striking (touching doWn) Within tho_ P2ROTECQTED ARE=A.

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.

Page 137 of 276

ATTACHMENT 1 EAL Bases EAL #3 address-sc-es aahaza*d*ous materials evPent rigiRnating at an f-fsite lcatio;n and of SUfficieRt magnitude to impede the movement of pcrSOnncl Within thc PROTECTED rAREA.A E:AI #1 addresses a h.zArdou e.....vent that c.au.ses an onsite impediment to vehicle movement and significant cnough to prohibit the plant staff fro~m acc-pessing the site using personRal vehicles. Examples oif such an event include site flooding caused by a hurc*Iane, heavy rains&,I rier wter releases, dam failure, etc., or an on site train derailment blocking the access up This EAL is*not intended apply to routine .uchimpediments as fog, snow, iGe, or vehicle bhre-akdoAwns or accidents, but rather to imopree significant conditions such As6 the- Hurricane Andrew strike on Turkey Point in 1992, the floodin;g around the Cooper Station during the Midwest flo-dis o-f 1Q993, or the flooding around Ft. C-alhoun61 Station in 20-11.

EAL #5 addresses (site specific description).

Escalation of the emergency classification level would be based on ICs in Recognition Categories AR, F, SorC.

CPNPP Basis Reference(s):

1. CPNPP PRA Accident Sequence Analysis "Internal Flooding Sequences"

2. NEI 99-01 HU3 Page 138 of 276

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technology Hazard Initiating Condition: Hazardous event EAL:

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

Mode Applicability:

All Definition(s):

IMPEDE(D) - Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective 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 FSAR Figure 1.2-1 Plot Plan.

CPNPP Basis:

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

NEI 99-01 Basis:

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

EAL 11 addresses a terade stFriking (touching dwnr) within the PROTECTED AREA.

This EA'l addresses flooding of a buildi'g room or area that

.esults in operators isolatirg power to a SAFETY SYSTEM co4r-mponont due to water level or other wetting conceFrn.

Classificatioen is also required if the water level or related wetting causes an automatic isolatio of a SAFETY SYSTEM compoenet froM its power sourcGe (o.g., a breaker or relay trip). ToG warrant classification, operability of the affected component must be required by Technical pecifiations for the Gcurent operating mode.

EAL-#3This 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.

EAL #4 addresses a hazardous event that c~auses an on site impediment to vehicle movemen and significant enough to prohibit the plant staff from accessing the Site using personal vehicloes. Examples Of such an event include site flooding caused by a hurricane, heav' rains&,

up rior wter releases, dam failure, etc., or an on site train derailment blocking the access Page 139 of 276

ATTACHMENT 1 EAL Bases This EAL is notntended apply to routineim;pediments 6uch as fog, snow, ice, or vehicle bre~akdoGwns or accidcnts, but rather to) mor)e Significant conditionRs such as the Hurcn Andrew strike.on Turkey Point in 1992, the flooding around the CooE*pe . r. Qtati-"n during the Midwest floods of 1993, Or the flooding aroun~d Ft. Calhoun Station in 2011.

E=AL #f5 addresses (site specific description).

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

CPNPP Basis Reference(s):

1. NEI 99-01 HU3 Page 140 of 276

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technology Hazard Initiating Condition: Hazardous event EAL:

HU3.4 Unusual Event A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles (Note 7)

Note 7: This EAL does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents.

Mode Applicability:

All Definition(s):

None CPNPP Basis:

None NEI 99-01 Basis:

This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.EAL #1 -,,e,4 a t .... ,,euh. @down) ,i;hi14 the PROTECTED AREA.

This EAL addresses flooding of a building reem o area that results in operatorsi i poveeto a SAFETY SYSTEM comproent due to water level orother wetting concerns.

lassificatine is also required if the water level or related wetting causes an automatic isolation ef a SAFETY SYSTEM comrponent fom its powfr seouce (e.g., a breaker or relay trip). To warrant classification, operability of the affected compoenet must be required by Technical Specfcations for the current operating mode.

EAL #3 addresses a hazardous materials event originating at an offsite location and of sufficient magnitude to impede the mo'erncent of personnel w~ithin the PROTECTED ,AREA.

ELAL44This 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 .EAL #5 addresses (site specific description)

Page 141 of 276

ATTACHMENT 1 EAL Bases Escalation of the emergency classification level would be based on ICs in Recognition I Categories AR, F, S or C.

CPNPP Basis Reference(s):

1. NEI 99-01 HU3 Page 142 of 276

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:

HU4.1 Unusual Event A FIRE is not extinguished within 15 min. of any of the following FIRE detection indications (Note 1):

" Report from the field (i.e., visual observation)

• Receipt of multiple (more than 1) fire alarms or indications

" Field verification of a single fire alarm AND The FIRE is located within any Table H-1 area Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Table H-1 Fire Areas

• u-Containment

• u-Safeguards Building

• X-Auxiliary Building

• X-Electrical & Control Building

• X-Fuel Building

. X-Service Water Intake Structure

• u-Diesel Generator Building

• u-Normal Switchgear Rooms

• u-CST

• u-RWST 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.

CPNPP 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. Actual field reports must be made within the 15 minute Page 143 of 276

ATTACHMENT 1 EAL Bases time limit or a classification must be made. If a fire is verified to be occurring by field report, the 15 minute time limit is from the original receipt of the fire detection alarm.

Table H-1 applies to buildings and areas housing equipment needed for safe shutdown (SAFETY SYSTEMS) (ref. 1, 2).

NEI 99-01 Basis:

This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.

EAL-#4-The-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.EAL #2 This EAL addresses rece.pt of a single fire alarm, and thc existence of a FPRE is not verified (i.e., proyed or disproved) within 30 minutes of the alarm. UJpon receipt, operators will take prompt actions to confirm the validity of a single fire alarm. For E=AL assessment purposes, th~e 30 Minute clock starts at the time that the initial -alar~mw~as received, and not the time that a subsequent verification action was performed.

A single fire alarm, absent other iniainsef a FIRE, may be indicative of equipment failur or a.puiusI . activation, and not an actual FIRE. Fer this reaseo, additional time is allowed to verif;, the validity of the alarm. The 30 minute period is a reasonable amount of time_ to determiRe 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 a.tual FIRE is verified by a report from the field, then EAL #11 is immediately applicable1 and the emergency must be declared if the FIRE is not extinguished within 15-minutes of th reporFt. if the alarm is verified to be due to an equipmnent failure or a Gpurious activation, and this Ve.ificato*" .occur within 30 minutes of the receipt of the alarm, then this EAL is net appicable aRd no emeFrgency declaratin is warranted.

in addition to a FIRE addressed by, EAL #i or EAL #t2, a FIRE within the plant PROTECTED AREA not extinguished within 60 minutes may also potentially degrade the level of plant safety. This basis extends to a FIRE occurring within the PROTECTED AREA of an ISFSi located outside the plant PROTEC-TED ,AREA. [Sentence for plants with an ISFSI outside th~e plant Protec-ted Area]

P::AI if a-FIRE= within the plan*t o IOF Ifor,, plants with an ISF=I outside the plant Protected Area]

PROTECiGTED AREA is of sufficient size to requie a re se by an offeite firefighting agency Page 144 of 276 [

ATTACHMENT 1 EAL Bases (e.g., a local tovn Fire Department.), then the lev~,el of plant safety is potentially degraded. The is needed to actively SUPPort firefighting efforts because the fire i beyond the capability of tho Fire BrFigade to extinguiSh. Declaration is not neesar 1if--- th gnc resources arc placed 9R stand by, or supporting post extinguishment reor;r investigation actiolns.

Basis DRel+nA 0 ... ..r.m.. f**m A .n...ixv 0 Appendix R to 10 CFR 50, states in part:

Criterion 3 of Appendix A to this part specifies that "Structures, systems, and comFponents important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires adepoin.

WheR considering the effects of fire, those systems associated with achieving aRd maintaining safe shutdown conditions assume major importance to safety because damage to themB can lead to core damage resulting from loss of coolant through boil off.

Because fire man" affect safe shutdown systems aRd because the loss of functifn Of systems used to mitigate the consequences of design basis accidents under postfr co~nditions does not per se impact public safety, the need to limit fire damage to systems required to -achieveand mnaintain safe shutdown conditions is greater than the need to-limit fire damage to those systems required to mitigate the consequences Of desig.n basis accidents.

In addition, Appendix

.FR R to 10 50, requires, among other considerations, the use of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> fire barriers for the enclosure of cable and equipment and associated non safety cirut of one-redundant train (G.2.G). As used in EAL #2, the 30 minutes to verify a single alarm iswell within this wor-st ease 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 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.

CPNPP Basis Reference(s):

1. CPNPP Fire Protection Report, Section 5.0 "Fire Safe Shutdown Equipment List"

2. FSAR Section 7.4 Systems Required for Safe Shutdown

3. NEI 99-01 HU4 Page 145 of 276

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-1 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-1 Fire Areas

• u-Containment

• u-Safeguards Building

• X-Auxiliary Building

• X-Electrical & Control Building

" X-Fuel Building

• X-Service Water Intake Structure

" u-Diesel Generator Building

• u-Normal Switchgear Rooms

" u-CST

• u-RWST 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.

CPNPP 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 HU4.1.

Page 146 of 276

ATTACHMENT 1 EAL Bases Table H-1 applies to buildings and areas housing equipment needed for safe shutdown (SAFETY SYSTEMS) (ref. 1, 2).

NEI 99-01 Basis:

This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.

E-AL-#-

The intent of the 15 minute duration is to size the FIRE and to discrimninatc against small FIRES that are readily extinguished (c.g., smoldering waste papcr basket). In addition to ala~rms, other indle~ations of a FIRE could be a drop infire mnain pressure, automatic activatio of a suppression system, etc.

Upon receipt, operators will take prompt actionS to cofirm 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 tha subsequent verification action was performed. Similarly, the fire duration clock alsoe starts at.

the time of receipt of the initial alarm, indication or report.

EAL-#2-This EAL addresses receipt of a single fire alarm, and the existence of a FIRE is not verified (i.e., proved or disproved) within 30-minutes of the alarm. Upon receipt, operators will take prompt actions to confirm the validity of a single fire alarm. For EAL assessment purposes, the 30-minute clock starts at the time that the initial alarm was received, and not the time that a subsequent verification action was performed.

A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failure or a spurious activation, and not an actual FIRE. For this reason, additional time is allowed to verify the validity of the alarm. The 30-minute period is a reasonable amount of time to determine if an actual FIRE exists; however, after that time, and absent information to the contrary, it is assumed that an actual FIRE is in progress.

IIf an actual FIRE is verified by a report from the field, then EAL #1HU4.1 is immediately applicable, and the emergency must be declared if the FIRE is not extinguished within 15-minutes of the report. Ifthe 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.

EALI-#-3 In addition to a FIRE addressed by EAL #1 or EAL #2, a FIRE within the plant PROTECTED AREA not extinguished within 60- minutes may also potentially degrade the level of plant safety. Thi's basis extends to a FIRE= oGccurrg within the PROTECTEDý AREA of an !S1 located outside the plant P-ROTECTED AREA4. [Senten-Ge for plants with an ISFSI outside the plant Pr-etected Area]

EAI-#4 Basis-Related Requirements from Appendix R Appendix R to 10 CFR 50, states in part:

Page 147 of 276

ATTACHMENT 1 EAL Bases 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 EAL-#2HU4.2, the 30-minutes to verify a single alarm is well within this worst-case 1-hour time period.

Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9.

CPNPP Basis Reference(s):

1. CPNPP Fire Protection Report, Section 5.0 "Fire Safe Shutdown Equipment List"

2. FSAR Section 7.4 Systems Required for Safe Shutdown

3. NEI 99-01 HU4 I Page 148 of 276

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 ISFSI or plant PROTECTED AREA not extinguished within 60 min. 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 FSAR Figure 1.2-1 Plot Plan.

CPNPP Basis:

None NEI 99-01 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.

The intent of the 15 minute duration is to size the FIRE and to discrhiminate against small FIRES that are readily exti*gUished (e.g., smoldering waste papcr basket). in addition to alarms, othc indications

. of a FIRE could be a drop infir. main pre..ure, automatic activation of a suppression system+, et+.

UJpon receipt, oper-ators will take prompt actions to co4nfr the validity of an initial fire alarm, indicatiEn, or repeot. For EAL assessment purpoSes, the em.ergency declaration clock starts at the time that the initial alarm, indicatin, Or report was rece.ived, and not the time that a subsequent verification action was perormed. Similarly, the fire duration clock also starts at the time of receipt of the initial alarm, indication or report.

This EA- addresses receipt of a single fire alarm, and the existenRce of FIRE is net Verified (i.e., proved Or disproved) within 30 m*iutes of the alarm. Upon receipt, opeFators will take p.ropt actios* to confirm the validity of a single fire alarm. For EAL assessment purposes,h Page 149 of 276

ATTACHMENT 1 EAL Bases 30 minute clock starts at the time that the initial! alarm was received, and not the time that a subsequent verfification action was performed.

A single fire alarmn, absen-t o-t-her indication(s) of a F=IRE, may be indicative of equipment failr or a spurio~us activation, and not an actuial FIRE. ForF this reason, additional time is allowed to verif' the validity of the alarm. The 30 minute period is a reasonable amount Of tim~e to determnine if an actual F=IRE= exists; however, after that time, and absent informnation to the contrar',, it is assumed that an a-tual IRE is in progress.

Ifan actual FIRE is verified by a repod from the field, then #1

.ALis immediately applicable, and the emergency must be declared if the FIRE is not extinguished within 15 m-iutesef the repodt. if the alarmn is verified to be due to an equipment failure or a spurious activation, and-this verification occurs Within 30 mFinutes of the receipt of the alarm, then this EAL= is not applicable and no) emergency declaration is warranted.

EAL-#3 In addition to a FIRE addressed by EAL #4-HU4.1 or EAL #2HU4.2, a FIRE within the plant PROTECTED AREA not extinguished within 60-minutes may also potentially degrade the level of plant safety. This basis extends to a FIRE occurring within the PROTECTED AREA of an ISFSI located outside the plant PROTECTED AREA. [SentenGe foAplants with an !SFSI out-side the plaRt Pr-tected Ara]eAI"/A if a FIRE within the pl*a*1 *+ [for plants with an ISFS outsido the pl*at Prote.etd Are.a]

PROTECTED AREA is of sufficient size to require a nse by an offsite firefighting agency (e.g., a localI twn1 Fire Depadtment), then the level of plant safet is potentially degraded. The is6 needed to actively suppedt firefighting effod-s because the fire is beyond the capability of the Fire Brigade to extinguish. Declaration is not necessar,' if the agency resources are placed o stand by, or suppoding post- extinguishment recve ;orivestigation actions.

Basis Related 0Re.. {ir .. ... + ,.,m A nn.R. ;,d.

Appendix R to 10 CF=R 50, states in pad:-

Criterion 3 of Appendix Ato this padt specifies that "Structures, systems, and components impodtant to safet shall be designed and located to minimize cnIstent4 with other safety requirements, the probability and effect of fires and explio,-nos."

When considering the effects of fire, those systemns associated with achieving and maintaining safe shutdown conditionRs assumne mjo impedtance to safet becausee damage to themR can lead to core damage resulting fromn loss of coolant through boil off.

Beause fire may affect safe shutdown systems and because the loSs Of funcItion of systemns used to mitigate the consequences of design basis accidents under postfr conditions does not per se impact public safety, the need to limit fire damage to systemns 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 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> fire barriers for the enclosure of cablean-d e-quipment and associated non safety iruits of. one Page 150 of 276

ATTACHMENT I EAL Bases red-undiant train (G.2.c). As used in EAL #2, the 30 minutes to verif' a single alarm is well within this wor.st case 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 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.

CPNPP Basis Reference(s):

1. NEI 99-01 HU4 Page 151 of 276

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 ISFSI or 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 FSAR Figure 1.2-1 Plot Plan.

CPNPP Basis:

None NEI 99-01 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.

EIL-#!

The intent of the 15 minute duration is to size the FIRE and to diso*cRiate against small FIRES that are readily extinguished (e.g., smoldering waste paper basket). In addition to alarms, other indc;ations of a FIRE could be a drop in fire main pressure, automatic activatio of a 6uppression system, etc.

Upon receipt, operateos Will take prompt actoRns to- co'Rfirm the validity of an initial fire alarm,'

indication, or report. For EAL assessment purposes, the emergency declarationcloc*k starts at.

the time that the initial alarmn, indication, or report was rec~eived, and not the time that a kn n,~, + a ; fnln +; Vtttl'J* +; VtaI IfArrr ~*I Q; 1-InI1 +1,. F; ArCl, ,rn

. A ),lc drr !n in -,c 4.+ 4- -4I the time of receipt of the initial alarm, indication Or reporFt.

EIAL-4t2 This EAL addresses receipt of a single fire alarmF, and the existence of a FIRE is no~t Verified 4

rrnnA rj r rlcý rrnxfpA\ iAcn~n2 ri c of +kc',nm r oilhjtpC 1 nc. ranr +nf pnrp+r..rq %.At;! +nLp.

promnpt actions to confirmn the validity of a single fire alarm. ForF E=AL assessmnent purposes,th 30_minuteý clock starts at the time that the initial alarm was received, and not the time that a subsequent verific-ation action Was pe~ore~fld.

I Page 152 of 276 1

ATTACHMENT 1 EAL Bases A single fire alarm, absent other indication(s) of a FIRE, mnay be indicativ~e of equipment failur orsprius activation, and not an actuai PRE~. tr-e 1IS reasGn, auuRIuuai umne is aluoweu to erif, the ,l;4idity of the alaFrm. The 30 minute peFred is a reasonable amou.nt of time to determie if .an atual FIRE exists; however, after that time, and absent i*nforation to the Gontrar,', it is -assu~medthat an actual FIRE is i rges If an actual FIRE iserified by a repoFt from the field, then EAL imm ediatel applicable, and- the emergency must be declredAp-- ifthe- FIRE isnot extingui-shed- 1Aithin 15 m-inutes o~f the report. if the alarm is verified to be due to an equipment failure or a spurious activation, and-this verification occurs-Within 0miue of the receipt of the alaFrm, then this EAL is not applicable and no emergency declaration is warranted.

In addition to a FIRE addressed by EAL #1 or EAL #2, a FIRE within the plant PROTECTED AREA not extinguished xvithin 60 minutes may also potentially degrade the love! of plant safety. This basis extends to a FIRE occUrring within the P-ROTECTEiD AREA of aý IS/S located outside the plant P-ROTECTED9 AREA. [Sentence for plants w.'th anQ 1SFSI obutsd the plant P-rotected Area]

E-AL-#4 If a FIRE within the plant or ISFSI [forplants with an I-SFS! outSide the plant PDotected, Ar..a]

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.

Basi Related~~nr Reau'rcrnts afro(mn AooeRnrJdi Appendix R to 10 CFR 50, states n part--.

Criterion 3 of Appendix A to this part specifies that "Structures, systems, and iriir uui ^npnr m[ruel ni II Iiin:i

- - - J Q:1e'

-..- 1:1

ý-; [ .- me nu-

. ýu - InnL iii - nnn- 11111

-Iinri:i~ - n r-Ti-rt-rI with other safety requirements, the probability ard effect of fires nep-losi*ons" When m,... ,inrin.. onnsideiRng f the effects cndtn5

.----..-.---.--. of fire, those ~ systems m... m. iri ass"oiated

... nr hnet with achieving

'ftx c.....wn ~~~5e.--.-.

a .

damage to them. can lead to core damage resulting fro. lss of colant through boil o.ff.

Be*ause fire may affect safe shutdcOWR systems and because the lo9s Of fLuncrtion Of systems used to mitigate the consequences of design basis accidents under post fr conditions does not per se impact public safety, the need to limnit fire damnage to systems required to achieve and mnaintain safe Shutdown conditions is greater than the need to li.it fire damage to those systems required to .itigate the .cnsequen~esof design_

basis accidents.

In addition, Appendix R to 10 CPR 50, req*iFes, aFmGog other cnsiderations, the use of 1 haou fire barriers for the encloGsure Of cable and equipment and associated non safetywcircuits; of one Page 153 of 276

ATTACHMENT 1 EAL Bases, redundant train (G.2.c;). As used in EAL #2, the 30 min*tes to ve,,;' a single alarm i well within this WorSt caSe 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 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.

CPNPP Basis Reference(s):

1. NEI 99-01 HU4 Page 154 of 276

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)

Note 5: Ifthe equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted.

Table H-2 Safe Operation & Shutdown Rooms/Areas Room/Area Mode Applicability Charging Pump Rooms 1, 2, 3, 4, 5, 6 CVCS Valve Rooms 1, 2, 3, 4, 5, 6 1E Switchgear Rooms All RHR Pump Rooms 4, 5, 6 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).

CPNPP 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 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).

I Page 155 of 276

ATTACHMENT 1 EAL Bases NEI 99-01 Basis:

This IC addresses an event involving a release of a hazardous gas that precludes or impedes access to equipment necessary to maintain normal plant operation, or required for a normal plant cooldown and shutdown. This condition represents an actual or potential substantial degradation of the level of safety of the plant.

An Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect at the time of the gaseous release. The 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

,,nEergencY Directr* 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 1 when the gaseous release occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4.

" The gas release is a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., fire suppression system testing).

" 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.

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, or to intentional in"rting of etainment.. (BVWR -ony).

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

CPNPP Basis Reference(s):

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

2. NEI 99-01 HA5 Page 156 of 276

ATTACHMENT 1 EAL Bases I Page 157 of 276

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 Remote Shutdown Panel (RSP)

Mode Applicability:

All Definition(s):

None CPNPP Basis:

Upon evacuation of the Control Room plant control is established at the Remote Shutdown Panel (RSP). ABN-905A/B "Loss of Control Room Habitability" and ABN-803A/B "Response to a Fire in the Control Room or Cable Spreading Room" provide the instructions for tripping the unit, and maintaining RCS inventory and Hot Shutdown conditions from outside the Control Room. 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, 3, 4, 5).

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

NEI 99-01 Basis:

This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations outside the Control Room. The loss of the ability to control the plant from the Control Room is considered to be a potential substantial degradation in the level of plant safety.

Following a Control Room evacuation, control of the plant will be transferred to alternate shutdown locations. The necessity to control a plant shutdown from outside the Control Room, in addition to responding to the event that required the evacuation of the Control Room, will present challenges to plant operators and other on-shift personnel. Activation of the ERO and emergency response facilities will assist in respondingto these challenges.

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

Page 158 of 276

ATTACHMENT 1 EAL Bases CPNPP Basis Reference(s):

1. DBD-ME-003 Control Room Habitability

2. ABN-905A Loss of Control Room Habitability

3. ABN-905B Loss of Control Room Habitability

4. ABN-803A Response to a Fire in the Control Room or Cable Spreading Room

5. ABN-803B Response to a Fire in the Control Room or Cable Spreading Room

6. NEI 99-01 HA6 Page 159 of 276

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 Remote Shutdown Panel (RSP)

AND Control of any of the following key safety functions is not re-established within 15 min.

(Note 1):

" Reactivity

" Core Cooling

" RCS heat removal 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 CPNPP Basis:

Upon evacuation of the Control Room plant control is established at the Remote Shutdown Panel (RSP). ABN-905A/B "Loss of Control Room Habitability" and ABN-803A/B "Response to a Fire in the Control Room or Cable Spreading Room" provide the instructions for tripping the unit, and maintaining RCS inventory and Hot Shutdown conditions from outside the Control Room. 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, 3, 4, 5).

The intent of this EAL is to capture events in which control of the plant cannot be reestablished in a timely manner. The fifteen minute time for transfer starts when the Control Room begins to be evacuated (not when the ABN is entered). The time interval is based on how quickly control must be reestablished without core uncovery and/or core damage. The determination of whether or not control is established from outside the Control Room is based on Emergency Coordinator judgment. The Emergency Coordinator is expected to make a reasonable, informed judgment that control of the plant from outside the Control Room cannot be established within the fifteen minute interval.

Once the Control Room is evacuated, the objective is to establish control of important plant equipment and maintain knowledge of important plant parameters in a timely manner. Primary Page 160 of 276

ATTACHMENT 1 EAL Bases emphasis should be placed on components and instruments that supply protection for and information about safety functions. Typically, these safety functions are reactivity control (ability to shutdown the reactor and maintain it shutdown), RCS inventory (ability to cool the core), and secondary heat removal (ability to maintain a heat sink).

NEI 99-01 Basis:

This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations, and the control of a key safety function cannot be reestablished in a timely manner. The failure to gain control of a key safety function following a transfer of plant control to alternate locations is a precursor to a challenge to one or more fission product barriers within a relatively short period of time.

The determination of whether or not "control" is established at the remote safe shutdown location(s) is based on Emergency Directer-Coordinator judgment. The Emergency DiieetOr Coordinator is expected to make a reasonable, informed judgment within (the site ,pecifictime feF1t4ansfeF)15 minutes whether or not the operating staff has control of key safety functions from the remote safe shutdown location(s).

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

1. DBD-ME-003 Control Room Habitability

2. ABN-905A Loss of Control Room Habitability

3. ABN-905B Loss of Control Room Habitability

4. ABN-803A Response to a Fire in the Control Room or Cable Spreading Room

5. ABN-803B Response to a Fire in the Control Room or Cable Spreading Room

6. NEI 99-01 HS6 I Page 161 of 276

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 UE EAL:

HU7.1 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 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.

CPNPP Basis:

The Emergency Coordinator is the designated onsite individual having the responsibility and authority for implementing the CPNPP Radiological Emergency Response Plan. The Shift Manager (SM) initially acts in the capacity of the Emergency Coordinator and takes actions as outlined in the Emergency Plan implementing procedures. Ifrequired 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).

NEI 99-01 Basis:

This IC addresses unanticipated conditions-not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Page 162 of 276

ATTACHMENT 1 EAL Bases DireeteF-Coordinator to fall under the emergency classification level description for an NQUEUnusual Event.

Page 163 of 276

ATTACHMENT 1 EAL Bases CPNPP Basis Reference(s):

1. CPNPP Radiological Emergency Response Plan section 1.1.2 Response

2. NEI 99-01 HU7 Page 164 of 276

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 CPNPP 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 CPNPP. 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).

CPNPP Basis:

The Emergency Coordinator is the designated onsite individual having the responsibility and authority for implementing the CPNPP Radiological Emergency Response Plan. The 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).

NEI 99-01 Basis:

This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency I Direotor-Coordinator to fall under the emergency classification level description for an Alert.

I Page 165 of 276

ATTACHMENT 1 EAL Bases CPNPP Basis Reference(s):

1. CPNPP Radiological Emergency Response Plan section 1.1.2 Response

2. NEI 99-01 HA7 Page 166 of 276

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 EXCLUSION AREA BOUNDARY Mode Applicability:

All Definition(s):

EXCLUSION AREA BOUNDARY - Exclusion Area Boundary is a synonymous term for Site Boundary. CPNPP FSAR Section 2.1.1.3 and Figure 2.1-2 define the Exclusion Area Boundary. This boundary is used for establishing effluent release limits with respect to the requirements of 10CFR20. See also CPNPP Emergency Plan Appendix E, Complex and Owner Controlled Area and CCNPP ODCM Section 5.0 Design Features.

HOSTILE ACTION - An act toward CPNPP 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 CPNPP. 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)

CPNPP Basis:

The Emergency Coordinator is the designated onsite individual having the responsibility and authority for implementing the CPNPP Radiological Emergency Response Plan. The 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).

Page 167 of 276

ATTACHMENT 1 EAL Bases NEI 99-01 Basis:

This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Dir-eeter-Coordinator to fall under the emergency classification level description for a Site Area Emergency.

CPNPP Basis Reference(s):

1. CPNPP Radiological Emergency Response Plan section 1.1.2 Response I 2. NEI 99-01 HS7 Page 168 of 276

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 CPNPP 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 CPNPP. 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.

CPNPP Basis:

The Emergency Coordinator is the designated onsite individual having the responsibility and authority for implementing the CPNPP Radiological Emergency Response Plan. The 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.

Page 169 of 276

ATTACHMENT 1 EAL Bases NEI 99-01 Basis:

This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Di-eeter-Coordinator to fall under the emergency classification level description for a General Emergency.

CPNPP Basis Reference(s):

1. CPNPP Radiological Emergency Response Plan section 1.1.2 Response

2. NEI 99-01 HG7 L Page 170 of 276

ATTACHMENT 1 EAL Bases Category S - System Malfunction EAL Group: Hot Conditions (RCS temperature greater than 2001F);

EALs in this category are applicable only in one or more hot operating modes.

Numerous system-related equipment failure events that warrant emergency classification have been identified in this category. They may pose actual or potential threats to plant safety.

The events of this category pertain to the following subcategories:

1. Loss of Emergency AC Power Loss of emergency 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 6.9KV AC safeguard 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. RCS Activity During normal operation, reactor coolant fission product activity is very low. Small concentrations of fission products in the coolant are primarily from the fission of tramp uranium in the fuel clad or minor perforations in the clad itself. Any significant increase from these base-line levels (2% - 5% clad failures) is indicative of fuel failures and is covered under the Fission Product Barrier Degradation category. However, lesser amounts of clad damage may result in coolant activity exceeding Technical Specification limits. These fission products will be circulated with the reactor coolant and can be detected by coolant sampling.

5. RCS Leakage 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 RCS leakage greater than Technical Specification limits indicates potential pipe cracks that may propagate to an extent threatening fuel clad, RCS 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 Page 171 of 276

ATTACHMENT I 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, RCS and containment integrity.

7. Loss of Communications Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification.

8. Containment Failure Failure of containment isolation capability (under conditions in which the containment is not currently challenged) warrants emergency classification. Failure of containment pressure control capability also warrants emergency classification.

9. Hazardous Event Affectinq Safety Systems Various natural and technological events that result in degraded plant safety system performance or significant visible damage warrant emergency classification under this subcategory.

Page 172 of 276 1

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 1 - Loss of Emergency AC Power Initiating Condition: Loss of all offsite AC power capability to safeguard buses for 15 minutes or longer EAL:

SUI.1 Unusual Event Loss of all offsite AC power capability, Table S-1, to 6.9 KV safeguard buses uEA1 and uEA2 for greater than or equal to 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-1 AC Power Sources Offsite:

• 138 KV switchyard circuit

• 345 KV switchyard circuit Onsite:

• uEG1

" uEG2 Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None Basis:

CPNPP Basis:

For emergency classification purposes, "capability" means that an offsite AC power source(s) is available to the safeguard buses, whether or not the buses are powered from it.

The safeguards AC distribution system power sources consist of the preferred and alternate offsite power sources, and the onsite standby emergency diesel generators uEG1 and uEG2.

Offsite power is supplied to the plant switchyards from the transmission network by five 345 KV and two 138 KV transmission lines. From the switchyards, two electrically and physically separated circuits provide AC power through step down startup transformers, to the 6.9 kV safeguard buses. The 138 kV switchyard circuit is the preferred source for Unit 2 and alternate source for-Unit 1. The 345 KV circuit is the preferred source for Unit 1 and alternate source for Unit 2. The onsite AC distribution system is divided into redundant trains so that the loss of any one load group does not prevent the minimum safety functions from being performed. Each train has connections to two offsite power sources and a dedicated diesel generator. Each offsite circuit can supply the Unit 1 and Unit 2 6.9 KV safeguard buses. (ref. 1, 2, 3, 4)

Page 173 of 276

ATTACHMENT 1 EAL Bases The 15-minute interval was selected as a threshold to exclude transient or momentary power losses.

NEI 99-01 Basis:

This IC addresses a prolonged loss of offsite power. The loss of offsite power sources renders the plant more vulnerable to a complete loss of power to AC emergency buses. This condition represents a potential reduction in the level of safety of the plant.

For emergency classification purposes, "capability" means that an offsite AC power source(s) is available to the emergency buses, whether or not the buses are powered from it.

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

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

CPNPP Basis Reference(s):

1. FSAR Figure 8.3-1

2. FSAR Section 8.2

3. FSAR Section 8.3

4. Technical Specifications B3.8.1

5. ABN-601 Response to a 138/345 KV System Malfunction

6. ABN-602 Response to a 6900/480V System Malfunction

7. NEI 99-01 SUW Page 174 of 276 1

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 safeguard buses for 15 minutes or longer EAL:

SAI.1 Alert AC power capability, Table S-1, to 6.9 KV safeguard buses uEA1 and uEA2 reduced to a single power source for greater than or equal to 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-1 AC Power Sources Offsite:

• 138 KV switchyard circuit

• 345 KV switchyard circuit Onsite:

• uEG1

" uEG2 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 ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):

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

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

Basis:

CPNPP Basis:

Page 175 of 276

ATTACHMENT 1 EAL Bases For emergency classification purposes, "capability" means that an offsite AC power source(s) is available to the emergency buses, whether or not the buses are powered from it.

The condition indicated by this EAL is the degradation of the offsite and onsite power sources such that any additional single failure would result in a loss of all AC power to the safeguard buses.

The safeguards AC distribution system power sources consist of the preferred and alternate offsite power sources, and the onsite standby emergency diesel generators uEG1 and uEG2.

Offsite power is supplied to the plant switchyards from the transmission network by five 345 KV and two 138 KV transmission lines. From the switchyards, two electrically and physically separated circuits provide AC power through step down startup transformers, to the 6.9 kV safeguard buses. The 138 kV switchyard circuit is the preferred source for Unit 2 and alternate source for Unit 1. The 345 KV circuit is the preferred source for Unit 1 and alternate source for Unit 2. The onsite AC distribution system is divided into redundant trains so that the loss of any one load group does not prevent the minimum safety functions from being performed. Each train has connections to two offsite power sources and a dedicated diesel generator. Each offsite circuit can supply the Unit 1 and Unit 2 6.9 KV safeguard buses. (ref. 1, 2, 3, 4).

The 15-minute interval was selected as a threshold to exclude transient or momentary power losses. Ifthe capability of a second source of emergency bus power is not restored within 15 minutes, an Alert is declared under this EAL.

This hot condition EAL is equivalent to the cold condition EAL CU2. 1.

NEI 99-01 Basis:

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

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

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

" A loss of 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 ba-k-fed from an offsite power source.

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

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

I Page 176 of 276

ATTACHMENT 1 EAL Bases CPNPP Basis Reference(s):

1. FSAR Figure 8.3-1

2. FSAR Section 8.2

3. FSAR Section 8.3

4. Technical Specifications B3.8.1

5. ABN-601 Response to a 138/345 KV System Malfunction

6. ABN-602 Response to a 6900/480V System Malfunction

7. NEI 99-01 SAl I Page 177 of 276

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: I - Loss of Emergency AC Power Initiating Condition: Loss of all offsite power and all onsite AC power to safeguard buses for 15 minutes or longer EAL:

EAL:

SS1.1 Site Area Emergency Loss of all offsite and all onsite AC power capability, Table S-1, to 6.9 KV safeguard buses uEA1 and uEA2 for greater than or equal to 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-1 AC Power Sources Offsite:

• 138 KV switchyard circuit

• 345 KV switchyard circuit Onsite:

" uEGI

• uEG2 Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None CPNPP Basis:

For emergency classification purposes, "capability" means that an AC power source is available to the safeguard buses, whether or not the buses are powered from it.

The safeguards AC distribution system power sources consist of the preferred and alternate offsite power sources, and the onsite standby emergency diesel generators uEG1 and uEG2.

Offsite power is supplied to the plant switchyards from the transmission network by five 345 KV and two 138 KV transmission lines. From the switchyards, two electrically and physically separated circuits provide AC power through step down startup transformers, to the 6.9 kV safeguard buses. The 138 kV switchyard circuit is the preferred source for Unit 2 and alternate source for Unit 1. The 345 KV circuit is the preferred source for Unit 1 and alternate source for Unit 2. The onsite AC distribution system is divided into redundant trains so that the loss of any one load group does not prevent the minimum safety functions from being performed. Each train has connections to two offsite power sources and a dedicated diesel generator. Each offsite circuit can supply the Unit I and Unit 2 6.9 KV safeguard buses. (ref. 1, 2, 3, 4).

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.

Page 178 of 276

ATTACHMENT 1 EAL Bases NEI 99-01 Basis:

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

In addition, fission product barrier monitoring capabilities may be degraded under these conditions. This IC represents a condition that involves actual or likely major failures of plant functions needed for the protection of the public.

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

Escalation of the emergency classification level would be via ICs AG-1-RG1, FG1 or SGI.

CPNPP Basis Reference(s):

1. FSAR Figure 8.3-1

2. FSAR Section 8.2

3. FSAR Section 8.3

4. Technical Specifications B3.8.1

5. ABN-601 Response to a 138/345 KV System Malfunction

6. ABN-602 Response to a 6900/480V System Malfunction

7. NEI 99-01 SS1 Page 179 of 276

ATTACHMENT 1 EAL Bases Category: S -System Malfunction Subcategory: 1 - Loss of Emergency AC Power Initiating Condition: Prolonged loss of all offsite and all onsite AC power to safeguard buses EAL:

SGI.1 General Emergency Loss of all offsite and all onsite AC power capability, Table S-I,to 6.9 KV safeguard buses uEA1 and uEA2 AND EITHER:

• Restoration of at least one safeguard bus from a Table S-1 source or APDG in less than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not likely (Note 1)

• CSFST 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.

Table S-1 AC Power Sources Offsite:

• 138 KV switchyard circuit

• 345 KV switchyard circuit Onsite:

• uEG1

• uEG2 Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None CPNPP Basis:

This EAL is indicated by the extended loss of all offsite and onsite AC power capability to 6.9 KV safeguard buses uEA1 and uEA2 either for greater then the CPNPP Station Blackout (SBO) coping analysis time (4 hrs.) (ref. 7) or that has resulted in indications of an actual loss of adequate core cooling.

Indication of continuing core cooling degradation is manifested by CSFST Core Cooling RED Path conditions being met. (ref. 8).

For emergency classification purposes, "capability" means that an AC power source is available to the emergency buses, whether or not the buses are powered from it.

The safeguards AC distribution system power sources consist of the preferred and alternate offsite power sources, and the onsite standby emergency diesel generators uEG1 and uEG2.

Page 180 of 276

ATTACHMENT 1 EAL Bases Offsite power is supplied to the plant switchyards from the transmission network by five 345 KV and two 138 KV transmission lines. From the switchyards, two electrically and physically separated circuits provide AC power through step down startup transformers, to the 6.9 kV safeguard buses. The 138 kV switchyard circuit is the preferred source for Unit 2 and alternate source for Unit 1. The 345 KV circuit is the preferred source for Unit 1 and alternate source for Unit 2. The onsite AC distribution system is divided into redundant trains so that the loss of any one load group does not prevent the minimum safety functions from being performed. Each train has connections to two offsite power sources and a dedicated diesel generator. Each offsite circuit can supply the Unit 1 and Unit 2 6.9 KV safeguard buses. (ref. 1, 2, 3, 4).

CPNPP has also provided a set of non-safety related Alternate Power Diesel Generators (APDGs) for each unit with the capability to connect to a safeguards bus one at a time to provide defense-in-depth for safe shutdown of a unit during outages or during extended duration of an inoperable offsite circuit on occurrence of concurrent loss of offsite power and failure of EDGs. The APDGs can provide 3450 kVA to provide long term cooling of each unit (ref. 3).

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

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 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. 8). Critical Safety Function Status Tree (CSFST)

Core Cooling-RED path indicates significant core exit superheating and core uncovery. (ref. 3).

NEI-9901 Basis:

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

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

Escalation of the emergency classification from Site Area Emergency will occur if it is projected that power cannot be restored to at least one AC emergency bus by the end of the analyzed station blackout coping period. Beyond this time, plant responses and event trajectory are subject to greater uncertainty, and there is an increased likelihood of challenges to multiple fission product barriers.

The estimate for restoring at least one emergency bus should be based on a realistic appraisal of the situation. Mitigation actions with a low probability of success should not be used as a basis for delaying a classification upgrade. The goal is to maximize the time available to prepare for, and implement, protective actions for the public.

The EAL will also require a General Emergency declaration if the loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core.

Page 181 of 276

ATTACHMENT 1 EAL Bases CPNPP Basis Reference(s):

1. FSAR Figure 8.3-1

2. FSAR Section 8.2

3. FSAR Section 8.3

4. Technical Specifications B3.8.1

5. ABN-601 Response to a 138/345 KV System Malfunction

6. ABN-602 Response to a 6900/480V System Malfunction

7. FSAR Section 8B Station Blackout

8. FRC-O.1A/B Response to Inadequate Core Cooling

9. NEI 99-01 SG1 Page 182 of 276 1

ATTACHMENT 1 EAL Bases Category: S -System Malfunction Subcategory: 1 - Loss of Emergency 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, Table S-1, to 6.9 KV safeguard buses uEA1 and uEA2 for greater than or equal to 15 min.

AND Less than 105 VDC on all 125 VDC safeguard buses uED1, uED2, uED3 and uED4 for greater than or equal to 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-1 AC Power Sources Offsite:

0 138 KV switchyard circuit

• 345 KV switchyard circuit Onsite:

" uEG1

• uEG2 Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None CPNPP Basis:

This EAL is indicated by the loss of all offsite and onsite emergency AC power capability to 6.9 KV safeguard buses uEA1 and uEA2 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.

For emergency classification purposes, "capability" means that an AC power source is available to the emergency buses, whether or not the buses are powered from it.

The safeguards AC distribution system power sources consist of the preferred and alternate offsite power sources, and the onsite standby emergency diesel generators uEG1 and uEG2.

Offsite power is supplied to the plant switchyards from the transmission network by five 345 KV and two 138 KV transmission lines. From the switchyards, two electrically and physically Page 183 of 276

ATTACHMENT 1 EAL Bases separated circuits provide AC power through step down startup transformers, to the 6.9 kV safeguard buses. The 138 kV switchyard circuit is the preferred source for Unit 2 and alternate source for Unit 1. The 345 KV circuit is the preferred source for Unit 1 and alternate source for Unit 2. The onsite AC distribution system is divided into redundant trains so that the loss of any one load group does not prevent the minimum safety functions from being performed. Each train has connections to two offsite power sources and a dedicated diesel generator. Each offsite circuit can supply the Unit 1 and Unit 2 6.9 KV safeguard buses. (ref. 1, 2, 3, 4).

The safeguards 125 VDC buses are the Class 1E buses uED1, uED2, uED3 and uED4 (ref. 7, 8,9).

Each redundant safeguards 125 VDC system consists of two independent batteries each having one main distribution bus, two static battery chargers (one spare), and local distribution panels. For Unit 1, batteries BT1 ED1 and BT1 ED3 feed all train A load requirements, while batteries BT1 ED2 and BT1 ED4 supply train B load requirements.

For Unit 2, batteries BT2ED1 and BT2ED3 feed all train A load requirements, while batteries BT2ED2 and BT2ED4 supply train B load requirements. There are no bus ties or sharing of power supplies between redundant trains (ref. 7).

Minimum DC bus voltage is 105 VDC (ref. 10). Bus voltage may be monitored from the following indications (ref. 12):

Control Room Panel CP-10 Annunciator u--ALB-10B Plant Computer V-IED1, 125VDC SWITCH PNL lED1 VOLT 1.13 V6501A BATT BT1ED1 VOLT V-1 ED2, 125VDC SWITCH PNL 1ED2 VOLT 2.13 V6502A BATT BT1ED2 VOLT V-I ED3, 125VDC SWITCH PNL IED3 VOLT 1.9 none V-i ED4, 125VDC SWITCH PNL IED4 VOLT 3.9 V6504A BATT BT1ED4 VOLT NEI-9901 Basis:

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

A loss of vital DC power compromises the ability to monitor and control SAFETY SYSTEMS.

A sustained loss of both emergency AC and vital DC power will lead to multiple challenges to fission product barriers.

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

The 15-minute emergency declaration clock begins at the point when both EAL thresholds are met.

Page 184 of 276

ATTACHMENT 1 EAL Bases CPNPP Basis Reference(s):

1. FSAR Figure 8.3-1

2. FSAR Section 8.2

3. PSAR Section 8.3

4. Technical Specifications B3.8.1

5. ABN-601 Response to a 138/345 KV System Malfunction

6. ABN-602 Response to a 6900/480V System Malfunction

7. FSAR 8.3.2

8. FSAR Figure 8.3-14

9. PSAR Figure 8.3-14A

10. ECA-O.OA/B Loss of All AC Power

11. SOP-605A/B 125 VDC Switchgear and Distribution Systems, Batteries and Battery Chargers 12.ALM-0102A/B Alarm Procedures Manual, u-ALB-1OB, nos. 1.9, 1.13, 2.13, 3.8 13.NEI 99-01 SG8 Page 185 of 276

ATTACHMENT 1 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 Less than 105 VDC on all 125 VDC safeguard buses uED1, uED2, uED3 and uED4 for greater than or equal to 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 CPNPP Basis:

The safeguards 125 VDC buses are the Class 1E buses uED1, uED2, uED3 and uED4 (ref. 1, 2,3).

Each redundant safeguards 125 VDC system consists of two independent batteries each having one main distribution bus, two static battery chargers (one spare), and local distribution panels. For Unit 1, batteries BT1 ED1 and BT1 ED3 feed all train A load requirements, while batteries BT1 ED2 and BT1 ED4 supply train B load requirements.

For Unit 2, batteries BT2ED1 and BT2ED3 feed all train A load requirements, while batteries BT2ED2 and BT2ED4 supply train B load requirements. There are no bus ties or sharing of power supplies between redundant trains (ref. 1).

Minimum DC bus voltage is 105 VDC (ref. 4). Bus voltage may be monitored from the following indications (ref. 6):

Control Room Panel CP-10 Annunciator u--ALB-1OB Plant Computer V-IED1, 125VDC SWITCH PNL IED1 VOLT 1.13 V6501A BATT BT1ED1 VOLT V-1ED2, 125VDC SWITCH PNL 1ED2 VOLT 2.13 V6502A BATT BT1ED2 VOLT V-1 ED3, 125VDC SWITCH PNL IED3 VOLT 1.9 V6503A BATT BT1ED3 VOLT V-I ED4, 125VDC SWITCH PNL 1ED4 VOLT 3.9 V6504A BATT BT1ED4 VOLT Page 186 of 276

ATTACHMENT 1 EAL Bases NEI 99-01 Basis:

This IC addresses a loss of vital DC power which compromises the ability to monitor and control SAFETY SYSTEMS. In modes above Cold Shutdown, this condition involves a major failure of plant functions needed for the protection of the public.

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

Escalation of the emergency classification level would be via ICs AG-1RGI, FG1 or SG8SGI.

CPNPP Basis Reference(s):

1. FSAR 8.3.2

2. FSAR Figure 8.3-14

3. FSAR Figure 8.3-14A

4. ECA-O.OA/B Loss of All AC Power

5. SOP-605A/B 125 VDC Switchgear and Distribution Systems, Batteries and Battery Chargers

6. ALM-0102A/B Alarm Procedures Manual, u-ALB-1OB, nos. 1.9, 1.13, 2.13, 3.8

7. NEI 99-01 SS8 I Page 187 of 276 1

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 greater than or equal to 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

" RCS level

• RCS pressure

" Core Exit T/C temperature

" Level in at least one SG

" Auxiliary or emergency feed flow in at least one SG 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.

CPNPP Basis:

SAFETY SYSTEM parameters listed in Table S-2 are monitored in the Control Room through a combination of hard control panel indicators as well as computer based information systems.

The Plant Process Computer, 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, 3, 4).

Page 188 of 276

ATTACHMENT 1 EAL Bases NEI 99-01 Basis:

This IC addresses the difficulty associated with monitoring normal plant conditions without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition is a precursor to a more significant event and represents a potential degradation in the level of safety of the plant.

As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room.

An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.

This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling [PWR] /i RPV level [BW*-. and RCS heat removal.

The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for reactor vessel level [PWRI / RPV water level [.WR] 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 SA2SA3.

CPNPP Basis Reference(s):

1. FSAR Section 7.5

2. DBD-EE-033 Detailed Control Room Design, 5.1.2, Figure 1

3. SOP 906 Plant Process Computer System Guidelines

4. ABN 906 Plant Process Computer System Malfunction

5. NEI 99-01 SU2 Page 189 of 276

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 greater than or equal to 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

• RCS level

• RCS pressure

" Core Exit T/C temperature

• Level in at least one SG

• Auxiliary or emergency feed flow in at least one SG Table S-3 Significant Transients

" Reactor trip

• Runback greater than or equal to 25% thermal power

• Electrical load rejection greater than 25% electrical load

• ECCS actuation Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

1 JNPLANNED - A parameter change or an event that is not 1) the result of an intended tolution or 2) an expected plant response to a transient. The cause of the parameter change 93vent may be known or unknown.

Page 190 of 276 1

ATTACHMENT 1 EAL Bases CPNPP Basis:

SAFETY SYSTEM parameters listed in Table S-1 are monitored in the Control Room through a combination of hard control panel indicators as well as computer based information systems.

The Plant Computer, which displays SPDS required information, serves as a redundant compensatory indicator which may be utilized in lieu of normal Control Room indicators (ref. 1, 2,3,4).

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 or equal to 25%

thermal power change, electrical load rejections of greater than 25% full electrical load or ECCS (SI) injection actuations.

NEI 99-01 Basis:

This IC addresses the difficulty associated with monitoring rapidly changing plant conditions during a transient without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. During this condition, the margin to a potential fission product barrier challenge is reduced. It thus represents a potential substantial degradation in the level of safety of the plant.

As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room.

An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.

This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling [PuD4R] / RPV lev.,el [BWR] and RCS heat removal.

The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters are lost, 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 [PWR] RPV i water level [.1r314RI cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well.

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

Escalation of the emergency classification level would be via ICs FS1 or IC AS-I-RS1 Page 191 of 276

ATTACHMENT 1 EAL Bases CPNPP Basis Reference(s):

1. FSAR Section 7.5

2. DBD-EE-033 Detailed Control Room Design, 5.1.2, Figure 1

3. SOP 906 Plant Process Computer System Guidelines

4. ABN 906 Plant Process Computer System Malfunction

5. NEI 99-01 SA2 Page 192 of 276

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 4 - RCS Activity Initiating Condition: Reactor coolant activity greater than Technical Specification allowable limits EAL:

SU4.1 Unusual Event Reactor coolant Dose Equivalent 1-131 specific activity greater than 60 pCi/gm OR Reactor coolant Dose Equivalent XE-133 specific activity greater than 500 pCi/gm Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None CPNPP Basis:

This EAL addresses reactor coolant samples exceeding Technical Specification LCOs 3.4.16.A and 3.4.16.B which are applicable in Modes 1, 2, and 3 and 4 (ref. 1). The Technical Specification limits accommodate an iodine spike phenomenon that may occur following changes in thermal power. The Technical Specification LCO limits are established to minimize the offsite radioactivity dose consequences in the event of a steam generator tube rupture (SGTR) accident (ref. 2).

NEI 99-01 Basis:

This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant.

Escalation of the emergency classification level would be via ICs FA1 or the Recognition Category A-R ICs.

CPNPP Basis Reference(s):

1. Technical Specifications Section 3.4.16

2. Technical Specifications Section B3.4.16

3. NEI 99-01 SU3 Page 193 of 276

ATTACHMENT I EAL Bases Category: S - System Malfunction Subcategory: 4 - RCS Activity Initiating Condition: Reactor coolant activity greater than Technical Specification allowable limits EAL:

SU4.2 Unusual Event Gross Failed Fuel Monitor, FFLu60 (u-RE-0406), High Alarm (RED)

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None CPNPP Basis:

This EAL addresses reactor coolant letdown line radiation levels sensed by FFLu60 (u-RE-0406) in excess of Technical Specification allowable limits. The High Alarm (RED) setpoint is based on the Technical Specifications maximum allowable concentration of radioactivity in the reactor coolant (ref. 1, 2, 3). A Geiger-Mueller tube is mounted on the reactor coolant letdown line after the letdown heat exchanger to monitor fission-product activity. Detection of increased system activity may be indicative of failed fuel. The monitor initiates Alert and High alarms in the Control Room (PC-1 1 and Plant Computer) (ref. 3, 4, 5, 6, 7, 8).

FFLu60 (u-RE-0406) has a range of 1E 1 E+7 pCi/ml.

NEI 99-01 Basis:

This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant.

Escalation of the emergency classification level would be via ICs FAI or the Recognition Category A-R ICs.

CPNPP Basis Reference(s):

1. Technical Specifications Section 3.4.16

2. ALM-3200 Alarm Procedure DRMS, Channel in High Alarm (RED), pg 54

3. DBD-EE-023 Radiation Monitoring System

4. SWEC-NU(S)-174 Radiation Monitor Alarm Concentrations for Failed Fuel Monitors 1-RE-406 & 2-RE-406

5. ABN-102 High Coolant Activity

6. FSAR Section 11.5.2.7.11

7. FSAR Table 11.5-1

8. CHM-1 11 Primary Chemistry Accident Assessment Sampling Program Page 194 of 276

ATTACHMENT 1 EAL Bases

9. DBD-EE-023 Radiation Monitoring System 10.NEI 99-01 SU3 I Page 195 of 276

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 5 - RCS Leakage Initiating Condition: RCS leakage for 15 minutes or longer EAL:

SU5.1 Unusual Event RCS unidentified or pressure boundary leakage greater than 10 gpm for greater than or equal to 15 min.

OR RCS identified leakage greater than 25 gpm for greater than or equal to 15 min.

OR UNISOLABLE leakage from the RCS to a location outside containment greater than 25 gpm for greater than or equal to 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 CPNPP Basis:

RCS leakage outside of the containment that is not considered identified or unidentified leakage per Technical Specifications includes leakage via interfacing systems such as RCS to the Component Cooling Water, or systems that directly see RCS pressure outside containment such as Chemical & Volume Control System, Nuclear Sampling system and Residual Heat Removal system (when in the shutdown cooling mode) (ref. 3, 6, 8)

Isolating letdown is a standard abnormal operating procedure action and may prevent unnecessary classification when a non-RCS leakage path, such as a CVCS leak, exists.

Unidentified leakage and identified leakage are determined by performance of the RCS water inventory balance. Pressure boundary leakage would first appear as unidentified leakage and can only be positively identified by inspection (ref. 1). OPT-303 (ref. 1) is used to ensure RCS leakage is within Technical Specification limits (ref. 2). ABN-103 Attachments 1 and 3 (ref. 3) are used for excessive RCS leakage.

Technical Specifications (ref. 4) defines RCS leakage as follows:

Identified Leakage:

o 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 Page 196 of 276

ATTACHMENT 1 EAL Bases o 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.

o Reactor Coolant System leakage through a steam generator to the Secondary System (primary to secondary leakage);

" Unidentified Leakage: All leakage (except RCP seal water injection or leakoff) that is not identified leakage.

" Pressure Boundary Leakage: Leakage (except primary to secondary leakage) through a non-isolable fault in an RCS component body, pipe wall, or vessel wall.

Escalation of this EAL to the Alert level is via Category F, Fission Product Barrier Degradation, EAL FA1.1.

NEI 99-01 Basis:

This IC addresses RCS leakage which may be a precursor to a more significant event. In this case, RCS leakage has been detected and operators, following applicable procedures, have been unable to promptly isolate the leak. This condition is considered to be a potential degradation of the level of safety of the plant.

.EAL

. andE.AL.2The

. first and second EAL conditions are focused on a loss of mass from the RCS due to "unidentified leakage", "pressure boundary leakage" or "identified leakage" (as these leakage types are defined in the plant Technical Specifications). EAL-#3The third condition addresses an RCS mass loss caused by an UNISOLABLE leak through an interfacing system. These EAL=s-conditions thus apply to leakage into the containment, a secondary-side system (e.g., steam generator tube leakage iOa PWR) or a location outside of containment.

The leak rate values for each E-AL-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). EAL4#!-The first condition uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage.

The release of mass from the RCS due to the as-designed/expected operation of a relief valve does not warrant an emergency classification. FGF PWRS-,aAn 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). FGF BWRs,-a- stuck open Safety Relief Valve (SRV) oe SRV leakage is not on'sidered either identified or unidentified leakage by Technical Specifications and, therefore, is not applicable te th~sFA--A The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible.

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

Page 197 of 276

ATTACHMENT 1 EAL Bases CPNPP Basis Reference(s):

I. OPT-303 Reactor Coolant System Water Inventory

2. Technical Specifications 3.4.13

3. ABN-103 Excessive Reactor Coolant Leakage

4. Technical Specifications 1.1

5. ABN-108 Shutdown Loss of Coolant

6. FSAR 5.2.5.2

7. FSAR 5.2.5.8

8. ECA-1.2 LOCA Outside Containment

9. NEI 99-01 SU4 Page 198 of 276

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 greater than 5% after any RPS setpoint is exceeded AND A subsequent automatic trip or manual trip action taken at the reactor control consoles (MCB reactor trip switches or deenergizing uB3 and uB4) is successful in shutting down the reactor as indicated by reactor power less than or equal to 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 CPNPP 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. 1, 2).

For the purposes of emergency classification, successful manual trip actions are those which can be quickly performed from the reactor control console; MCB reactor trip switches or deenergizing uB3 and uB4. Reactor shutdown achieved by use of other trip actions specified in FR-S.1 Response to Nuclear Power Generation/ATWS (such as manually insert control rods, opening the reactor trip and bypass breakers in the reactor switchgear, tripping the Rod Drive MG sets in the normal switchgear or emergency boration) do not constitute a successful manual trip (ref. 2).

Page 199 of 276

ATTACHMENT 1 EAL Bases Following any automatic RPS trip signal, E-0.0 (ref. 1) and /FR-S.1 (ref. 2) 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. 2).

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 to or 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.

NEI 99-01 Basis:

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor (trip [PrI, / Scram [81/14R]) that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic (trip fLrWR]--.

scramr[{4R4) 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.

I Following the failure on an automatic reactor (trip [PWR]I scram [BI',1.), operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g.,

I initiate a manual reactor (trip [P,44R] / sc,,ram [,WR.)). 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 [P-KR], / Gscam [,WRI) 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 [PrW1RI / Scram [B"l'R])) using a different switch).

Depending upon several factors, the initial or subsequent effort to manually (trip r**-* / sGrim

[BW-R}) the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor (trip [PWR] / scram [*,qR) signal. If a subsequent manual or automatic (trip .PW.. ,

sra B-"4r*]*) 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[PWR] / scram r],,. This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations Page 200 of 276

ATTACHMENT 1 EAL Bases within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles".

Taking the RoactoFr.Mode SHT-DOWN is considered Sw~itcrh to61 to be a mianual scram tion-.

[BWR]

The plant response to the failure of an automatic or manual reactor (trip [P,.R] / cram. [BVrR])

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 SA5SA6. Depending upon the plant response, escalation is also possible via IC FA1. Absent the plant conditions needed to meet either IC SA5-SA6 or FA1, 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 [PV'RJ / *,*,.am [.. "R])signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied.

• Ifthe signal causes a plant transient that should have included an automatic reactor (trip

[PWVR] / scram [BWR]) and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and should be evaluated.

• Ifthe signal does not cause a plant transient and the (trip [PWRI / scrGam [B,-R]) 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.

CPNPP Basis Reference(s):

1. EOP-O.OA/B Reactor Trip or Safety Injection

2. FR-S.1 Response to Nuclear Power Generation/ATWS 3 NEI 99-01 SU5 Page 201 of 276

ATTACHMENT I 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 greater than 5%

after any manual trip action was initiated AND A subsequent automatic trip or manual trip action taken at the reactor control console (MCB reactor trip switches or deenergizing uB3 and uB4) is successful in shutting down the reactor as indicated by reactor power less than or equal to 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 CPNPP Basis:

This EAL addresses a failure of a manually initiated trip in the absence of having exceeded an automatic RTS trip setpoint and a subsequent automatic or manual trip is successful in shutting down the reactor (reactor power less than or equal to 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 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. 1, 2).

For the purposes of emergency classification, successful manual trip actions are those which can be quickly performed from the reactor control console; MCB reactor trip switches or deenergizing uB3 and uB4. Reactor shutdown achieved by use of other trip actions specified in FR-S.1 Response to Nuclear Power Generation/ATWS (such as manually insert control rods, opening the reactor trip and bypass breakers in the reactor switchgear, tripping the Rod Drive MG sets in the normal switchgear or emergency boration) do not constitute a successful manual trip (ref. 2).

Page 202 of 276

ATTACHMENT 1 EAL Bases Following the failure of any manual trip signal, E-0.0 (ref. 1) and FR-S.1 (ref. 2) prescribe insertion of redundant manual trip signals to back up the RPS trip function and ensure reactor shutdown is achieved. Even if a subsequent automatic trip signal or the first subsequent manual trip signal inserts all control rods to the full-in position immediately after the initial failure of the manual trip, the lowest level of classification that must be declared is an Unusual Event (ref. 2).

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 (less than or equal to 5%) following a failure of an initial manual trip, the event escalates to an Alert under EAL SA6.1.

NEI 99-01 Basis:

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor (trip [PWR] / scram [BAR]) that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic (trip [PWk]-

... am- *,9P) 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.

I Following the failure on an automatic reactor (trip [P,,rIq i scram [9**q), operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g.,

I initiate a manual reactor (trip [PR] i scram [,WR])). 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 [PrK, R, scram [B",,q,) 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 [Pr],/ scram [Bl-r)) using a different switch).

Depending upon several factors, the initial or subsequent effort to manually (t*,p [P,/] / scram

[BW-R) the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor (trip [PKR] i scram- [BIA'J) signal. If a subsequent manual or automatic (trip fP...

scam rBWR4* 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

1. trip[PW'R]K ScGram [BKr* ). This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles".

Taking the Roactor Mode Switch to SHUTDOWN is considered to be a manual scram action-.

fB9WRI The plant response to the failure of an automatic or manual reactor (trip [PWR' / scram [BWR])

will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If 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 SA5SA6. Depending upon the plant Page 203 of 276 1

ATTACHMENT 1 EAL Bases response, escalation is also possible via [C FAI. Absent the plant conditions needed to meet either IC SA5-SA6 or FA1, an Unusual Event declaration is appropriate for this event.

A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.

I Should a reactor (trip [P"WR] ! S...m [B.. R]) 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

[PWR] /*cFrm, [B,,-R]) and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and should be evaluated.

e If the signal does not cause a plant transient and the (trip [PWR] / Scr*am [BWR]) 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.

CPNPP Basis Reference(s):

1. EOP-O.OA/B Reactor Trip or Safety Injection

2. FR-S.1 Response to Nuclear Power Generation/ATWS

3. NEI 99-01 SU5 Page 204 of 276

ATTACHMENT I 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 greater than 5%

AND Manual trip actions taken at the reactor control console (MCB reactor trip switches or deenergizing uB3 and uB4) are not successful in shutting down the reactor as indicated by reactor power greater than 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 CPNPP Basis:

This EAL addresses any automatic or manual reactor trip signal that fails to shut down the reactor (reactor power less than or equal to 5%) 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 (ref. 1).

For the purposes of emergency classification, successful manual trip actions are those which can be quickly performed from the reactor control console; MCB reactor trip switches or deenergizing uB3 and uB4. Reactor shutdown achieved by use of other trip actions specified in FR-S.1 Response to Nuclear Power Generation/ATWS (such as manually insert control rods, opening the reactor trip and bypass breakers in the reactor switchgear, tripping the Rod Drive MG sets in the normal switchgear or emergency boration) do not constitute a successful manual trip (ref. 2).

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, 2).

Page 205 of 276

ATTACHMENT I EAL Bases Escalation of this event to a Site Area Emergency would be under EAL SS6.1 or Emergency Coordinator judgment.

NEI 99-01 Basis:

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor (trip [PWR] / Scram. [,LR]) that results in a reactor shutdown, and subsequent operator manual actions taken at the reactor control consoles to shutdown the reactor are also unsuccessful. This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if the reactor is subsequently shutdown by an action taken away from the reactor control consoles since this event entails a significant failure of the RPS.

A manual action at the reactor control 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[PWR] / scram [9B4WR)). This action does not include manually driving in control rods or implementation of boron injection strategies. If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control consoles (e.g., locally opening breakers). Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles".Taking the RcaD*,-o Mode SWitch to SHUTDOWN i, considered to be a manual scram action. [BWR]

The plant response to the failure of an automatic or manual reactor (trip [PVVRD / scram [BrI/)

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. Ifthe failure to shut-down the reactor is prolonged enough to cause a challenge to the core cooling [PWR] / PV .. ate. level [BWR] or RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC SS65. Depending upon plant responses and symptoms, escalation is also possible via IC FSI. Absent the plant conditions needed to meet either IC SS65 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.

CPNPP Basis Reference(s):

1. EOP-O.OA/B Reactor Trip or Safety Injection

2. FR-S.1 Response to Nuclear Power Generation/ATWS

3. NEI 99-01 SA5 Page 206 of 276

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 RCS heat removal EAL:

SS6.1 Site Area Emergency An automatic or manual trip fails to shut down the reactor as indicated by reactor power greater than 5%

AND All actions to shut down the reactor are not successful as indicated by reactor power greater than 5%

AND EITHER:

• CSFST Core Cooling RED Path conditions met

" CSFST Heat Sink RED Path conditions met Mode Applicability:

1 - Power Operation Definition(s):

None CPNPP 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 RCS barriers.

Reactor shutdown achieved by use of FR-S.1 Response to Nuclear Power Generation/ATWS (such as manually insert control rods, opening the reactor trip and bypass breakers in the reactor switchgear, tripping the Rod Drive MG sets in the normal switchgear or emergency boration) 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, 2).

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 Page 207 of 276

ATTACHMENT 1 EAL Bases 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, 2).

Indication of continuing core cooling degradation is manifested by CSFST Core Cooling RED Path conditions being met. Specifically, Core Cooling RED Path conditions exist if either core exit T/Cs are reading greater than or equal to 12001F (ref. 3).

Indication of inability to adequately remove heat from the RCS is manifested by CSFST Heat Sink RED Path conditions being met. Specifically, Heat Sink RED Path conditions exist if narrow range level in at least one steam generator is not greater than or equal to (43[50]%

ACC) on Unit 1 or (10 [18]% ACC) on Unit 2 and total feedwater flow to the steam generators is less than or equal to 460 gpm (ref. 4).

NEI 99-01 Basis:

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor (trip [P] ,'/ .cram [B,,R) that results in a reactor shutdown, all subsequent operator actions to manually shutdown the reactor are unsuccessful, and continued power generation is challenging the capability to adequately remove heat from the core and/or the RCS. This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration of a Site Area Emergency.

In some instances, the emergency classification resulting from this IC/EAL may be higher than that resulting from an assessment of the plant responses and symptoms against the Recognition 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 AG--RG1 or FG1.

CPNPP Basis Reference(s):

1. EOP-0.OA/B Reactor Trip or Safety Injection

2. FR-S.1 Response to Nuclear Power Generation/ATWS

3. FR-C.1 Response to Inadequate Core Cooling

4. FR-H.1 Response to Loss of Heat Sink

5. NEI 99-01 SS5.

I Page 208 of 276

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 7 - Loss of Communications Initiating Condition: Loss of all onsite or offsite communications capabilities EAL:

SU7.1 Unusual Event Loss of all Table S-4 onsite communication methods OR Loss of all Table S-4 offsite communication methods OR Loss of all Table S-4 NRC communication methods Table S-4 Communication Methods System Onsite Offsite NRC Gai-tronics Page/Party (PA) X Plant Radios X PABX X X X Public Telephone X X X Federal Telephone System (FTS) X X Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None CPNPP Basis:

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

This EAL is the hot condition equivalent of the cold condition EAL CU5.1.

NEI 99-01 Basis:

This IC addresses a significant loss of on-site or offsite communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to OROs and the NRC.

Page 209 of 276

ATTACHMENT 1 EAL Bases This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.).

EAL#4-The first EAL condition addresses a total loss of the communications methods used in support of routine plant operations.

EAL-#2The second EAL condition addresses a total loss of the communications methods used to notify all OROs of an emergency declaration. The offsite (OROs. referred to here are-(see Developer Notes) the State Department of Public Safety, Somervell and Hood County EOCs-EAL4#3The third EAL addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.

CPNPP Basis Reference(s):

1. FSAR 9.5.2

2. DBD-EE-048 Communication System

3. NEI 99-01 SU6 I Page 210 of 276

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 Any penetration is not isolated within 15 min. of a VALID containment isolation signal OR Containment pressure greater than 18 psig with neither Containment Spray system operating per design for greater than or equal to 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:

I - 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.

CPNPP Basis:

The Containment Spray System (CSS) is designed to remove heat from the Containment environment following a LOCA, a main steam line break accident, or a feedwater line break accident. Each unit of the CPNPP is equipped with two redundant Containment spray trains, each designed to provide emergency Containment heat removal in the event of a LOCA. This system, in conjunction with the ECCS, removes post-accident thermal energy from the Containment environment, thereby reducing the Containment pressure and temperature. Each train includes two containment spray pumps, spray headers, nozzles, valves, and piping. Each train is powered from a separate safeguard bus. (ref. 1)

The Containment pressure setpoint (18 psig, ref. 2) is the pressure at which the Containment Spray System should actuate and begin performing its function. The design basis accident

.analyses and evaluations assume the loss of one Containment Spray System train (ref. 1).

NEI 99-01 Basis:

This IC,-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.

Page 211 of 276

ATTACHMENT 1 EAL Bases I For EAL--#-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.

EAL #2The 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 RCS fission product barriers.

CPNPP Basis Reference(s):

1. FSAR Section 6.2.2

2. FRC-Z.1A/B Response to High Containment Pressure

3. NEI 99-01 SU7 IPage 212 of 276

ATTACHMENT I 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 Emergency Coordinator Mode Applicability:

1 - 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 Page 213 of 276

ATTACHMENT 1 EAL Bases 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.

CPNPP 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 (ref. 3, 4).

• Seismic Category I structures are analyzed to withstand a sustained, design wind velocity of at least 80 mph. (ref. 5).

" Areas containing functions and systems required for safe shutdown of the plant are identified by fire area (ref. 6, 7).

" 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.

NEI 99-01 Basis:

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.

EAL 1.b.!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.

EAL .b.2The 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 Page 214 of 276

ATTACHMENT 1 EAL Bases 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 A&4-RS1.

CPNPP Basis Reference(s):

1. ABN-907 Acts of Nature

2. CPNPP PRA Accident Sequence Analysis "Internal Flooding Sequences"

3. FSAR Section 2.4.3.7 Flood Evaluations for Safe Shutdown Impoundment

4. DBD-CS-071 Maximum Probable Flood

5. FSAR Section 3.3.1.1 Wind Loadings

6. CPNPP Fire Protection Report, Section 5.0 "Fire Safe Shutdown Equipment List"

7. FSAR Section 7.4 Systems Required for Safe Shutdown

8. NEI 99-01 SA9 I Page 215 of 276

ATTACHMENT 1 EAL Bases Category F - Fission Product Barrier Degradation EAL Group: Hot Conditions (RCS temperature greater than 2000 F); EALs in this category are applicable only in one or more hot operating modes.

EALs in this category represent threats to the defense in depth design concept that precludes the release of highly radioactive fission products to the environment. This concept relies on multiple physical barriers any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment. The primary fission product barriers are:

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

B. Reactor Coolant System (RCS): The RCS Barrier includes the RCS primary side and its connections up to and including the pressurizer safety and relief valves, and other connections up to and including the primary isolation valves.

C. Containment (CNTMT): 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-1 (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 potentialloss of either Fuel Clad or RCS Site Area Emergency:

Loss or potential loss of any two barriers General Emergency:

Loss of any two barriersand loss orpotential loss of third barrier The logic used for emergency classification based on fission product barrier monitoring should reflect the following considerations:

• The Fuel Clad Barrier and the RCS Barrier are weighted more heavily than the Containment Barrier.

• Unusual Event ICs associated with RCS and Fuel Clad Barriers are addressed under System Malfunction ICs.

Page 216 of 276

ATTACHMENT I EAL Bases

• For accident conditions involving a radiological release, evaluation of the fission product barrier thresholds will need to be performed in conjunction with dose assessments to ensure correct and timely escalation of the emergency classification. For example, an evaluation of the fission product barrier thresholds may result in a Site Area Emergency classification while a dose assessment may indicate that an EAL for General Emergency IC RG1 has been exceeded.

• The fission product barrier thresholds specified within a scheme reflect plant-specific CPNPP design and operating characteristics.

" As used in this category, the term RCS leakage encompasses not just those types defined in Technical Specifications but also includes the loss of RCS mass to any location- inside the primary containment, an interfacing system, or outside of the primary containment. The release of liquid or steam mass from the RCS due to the as-designed/expected operation of a relief valve is not considered to be RCS leakage.

" At the Site Area Emergency level, EAL users should maintain cognizance of how far present conditions are from meeting a threshold that would require a General Emergency declaration. For example, if the Fuel Clad and RCS fission product barriers were both lost, then there should be frequent assessments of containment radioactive inventory and integrity. Alternatively, if both the Fuel Clad and RCS fission product barriers were potentially lost, the Emergency Coordinator would have more assurance that there was no immediate need to escalate to a General Emergency.

I Page 217 of 276

ATTACHMENT 1 EAL Bases Category: Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Any loss or any potential loss of either Fuel Clad or RCS EAL:

FAI.1 Alert Any loss or any potential loss of either Fuel Clad or RCS (Table F-I)

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None CPNPP Basis:

Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 (Attachment

2) lists the fission product barrier thresholds, bases and references.

At the Alert classification level, Fuel Clad and RCS barriers are weighted more heavily than the Containment barrier. Unlike the Containment barrier, loss or potential loss of either the Fuel Clad or RCS barrier may result in the relocation of radioactive materials or degradation of core cooling capability. Note that the loss or potential loss of Containment barrier in combination with loss or potential loss of either Fuel Clad or RCS barrier results in declaration of a Site Area Emergency under EAL FSI.1 NEI 99-01 Basis:

None CPNPP Basis Reference(s):

1. NEI 99-01 FA1 I Page 218 of 276

ATTACHMENT 1 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:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None CPNPP Basis:

Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 (Attachment

2) lists the fission product barrier thresholds, bases and references.

At the Site Area Emergency classification level, each barrier is weighted equally. A Site Area Emergency is therefore appropriate for any combination of the following conditions:

• One barrier loss and a second barrier loss (i.e., loss - loss)

• One barrier loss and a second barrier potential loss (i.e., loss - potential loss)

• One barrier potential loss and a second barrier potential loss (i.e., potential loss -

potential loss)

At the Site Area Emergency classification level, the ability to dynamically assess the proximity of present conditions with respect to the threshold for a General Emergency is important. For example, the existence of Fuel Clad and RCS Barrier loss thresholds in addition to offsite dose assessments would require continual assessments of radioactive inventory and Containment integrity in anticipation of reaching a General Emergency classification. Alternatively, if both Fuel Clad and RCS potential loss thresholds existed, the Emergency Coordinator would have greater assurance that escalation to a General Emergency is less imminent.

NEI 99-01 Basis:

None CPNPP Basis Reference(s):

1. NEI 99-01 FS1 Page 219 of 276

ATTACHMENT 1 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:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None CPNPP Basis:

Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 (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, RCS and Containment barriers

" Loss of Fuel Clad and RCS barriers with potential loss of Containment barrier

" Loss of RCS and Containment barriers with potential loss of Fuel Clad barrier

" Loss of Fuel Clad and Containment barriers with potential loss of RCS barrier NEI 99-01 Basis:

None CPNPP Basis Reference(s):

1. NEI 99-01 FG1 Page 220 of 276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Introduction Table F-1 lists the threshold conditions that define the Loss and Potential Loss of the three fission product barriers (Fuel Clad, Reactor Coolant System, and Containment). The table is structured so that each of the three barriers occupies adjacent columns. Each fission product barrier column is further divided into two columns; one for Loss thresholds and one for Potential Loss thresholds.

The first column of the table (to the left of the Fuel Clad Loss column) lists the categories (types) of fission product barrier thresholds. The fission product barrier categories are:

A. RCS or SG Tube Leakage B. Inadequate Heat removal C. CNTMT Radiation / RCS Activity D. CNTMT Integrity or Bypass E. Emergency Coordinator Judgment Each category occupies a row in Table F-1 thus forming a matrix defined by the categories.

The intersection of each row with each Loss/Potential Loss column forms a cell in which one or more fission product barrier thresholds appear. If NEI 99-01 does not define a threshold for a barrier Loss/Potential Loss, the word "None" is entered in the cell.

Thresholds are assigned sequential numbers within each Loss and Potential Loss column beginning with number one. In this manner, a threshold can be identified by its category title and number. For example, the first Fuel Clad barrier Loss in Category A would be assigned "FC Loss A.I," the third Containment barrier Potential Loss in Category C would be assigned "CNTMT P-Loss C.3," etc.

If a cell in Table F-1 contains more than one numbered threshold, each of the numbered thresholds, if exceeded, signifies a Loss or Potential Loss of the barrier. It is not necessary to exceed all of the thresholds in a category before declaring a barrier Loss/Potential Loss.

Subdivision of Table F-1 by category facilitates association of plant conditions to the applicable fission product barrier Loss and Potential Loss thresholds. This structure promotes a systematic approach to assessing the classification status of the fission product barriers.

When equipped with knowledge of plant conditions related to the fission product barriers, the EAL-user first scans down the category column of Table F-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. Ifa 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 RCS barriers and a Potential Loss of the Containment barrier can occur. Barrier Page 221 of 276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Losses and Potential Losses are then applied to the algorithms given in EALs FG1.1, FSI.1, and FA1.1 to determine the appropriate emergency classification.

In the remainder of this Attachment, the Fuel Clad barrier threshold bases appear first, followed by the RCS barrier and finally the Containment barrier threshold bases. In each barrier, the bases are given according category Loss followed by category Potential Loss beginning with Category A, then B,..., E.

Page 222 of 276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Table F-1 Fission Product Barrier Threshold Matrix Fuel Clad (FC) Barrier Reactor Coolant System (RCS) Barrier Containment (CNTMT) Barrier Category Loss Potential Loss Loss Potential Loss Loss Potential Loss

1. An automatic or manual ECCS 1. Operation of a standby charging A (SI) actuation required by EITHER:

pump is required by EITHER:

• UNISOLABLE RCS leakage 1. A leaking or RUPTURED SG is None RCS or None None UNISOLABLE RCS

• SG tube leakage FAULTED outside of containment SG Tube leakage 2. CSFST Integrity-RED Path Leakage SG tube RUPTURE conditions met

1. CSFST Core Cooling-ORANGE B Path conditions met 1. CSFST Heat Sink-RED Path 1. CSFST Core Cooling-RED Path conditions met Inadequate 1. CSFST Core Cooling-RED 2. CSFST Heat Sink-RED Path conditions met Path conditions met conditions met None AND None AND Heat AND Heat sink is required Restoration procedures not Removal RemovalHeat sink is required Heat sink is required effective within 15 min. (Note 1)

I. Containment radiation greater than 85 RJhr CTEuo16 Containment HRRM (L-RE-6290A), or 1. Containment radiation greater than 1. Containment radiation greater than C CTWu_17 Containment HRRM 5 R/hr 2,110 R/hr CNTMT W-RE-6290B) CTEu16 Containment HRRM CTEu16 Containment HRRM R NT 2. Dose equivalent 1-131coolant None (L-RE-6290A), or None None (L-RE-6290A), or Radiation activity greater than 300 CTWu17 Containment HRRM CTWu17 Containment HRRM I RCS jCilco (L-RE-6290B) Lu-RE-6290B)

Activity 3. Gross Failed Fuel Monitor, FFLu60 (Q-RE-0406),

radiation greater than 1.0E04 pCi/cc

1. Containment isolation is required AND EITHER: 1. CSFST Containment-RED Path has conditions met Containment integrity D been lost based on Emergency Coordinator

2. Containment hydrogen concentratior greater than 4%

CNTMT None None None None judgment 3. Containment pressure greater than Integrity UNISOLABLE pathwiay from 18 psig Withneither Containment or Bypass Containment to the environment Spray system train operating exists greater than or equal to 15 rain.

2. Indications of RCS leakage (Note 1) outside of Containment E 1. Any condition in the opinion of 1. Any condition in the opinion of 1. Any condition in the opinion of 1. Any condition in the opinion of the 1. Any condition in the opinion of 1. Any condition in the opinion of the the Emergency Coordinator that the Emergency Coordinator that the Emergency Coordinator that Emergency Coordinator that the Emergency Coordinator that Emergency Coordinator that EC indicates loss of the fuel clad indicates potential loss of the fuel indicates loss of the RCS barrier indicates potential loss of the RCS indicates loss of the Containment indicates potential loss of the Judgment barrier clad barrier barrier barrier Containment barrier

[Document No.] Rev. 6 Page 223 of 276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: A. RCS or SG Tube Leakage Degradation Threat: Loss Threshold:

None

[Document No.] Rev. 6 Page 224 of 276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: A. RCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold:

LNone

((Document No.] Rev. 6 1 Page 225 of 276

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:

Plant-Specific 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 Plant Computer (ref. 1).

Generic This reading indicates temperatures within the core are sufficient to cause significant superheating of reactor coolant.

CPNPP Basis Reference(s):

1. FRC-0.1A/B Response to Inadequate Core Cooling

2. NEI 99-01 Inadequate Heat Removal Fuel Clad Loss 2.A S[Document No.] I Rev. 6 1 Page 226 of 276

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:

Plant-Specific 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 Plant Computer (ref. 1, 2).

Generic This reading indicates a reduction in reactor vessel water level sufficient to allow the onset of heat-induced cladding damage.

CPNPP Basis Reference(s):

1. FRC-0.1A/B Response to Inadequate Core Cooling

2. FRC-0.2A/B Response to Degraded Core Cooling

3. NEI 99-01 Inadequate Heat Removal Fuel Clad Loss 2.A I[Document No.] I Rev. 6 Page 227 of 276 1

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:

Plant-Specific In combination with RCS 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 Plant Computer (ref. 1).

The phrase "and heat sink required" precludes the need for classification for conditions in which RCS pressure is less than SG pressure or Heat Sink-RED path entry was created through operator action directed by an ERG. For example, FRH-0.1 is entered from CSFST Heat Sink-Red. Step 1 tells the operator to determine if heat sink is required by checking that RCS pressure is greater than any non-faulted SG pressure and RCS temperature 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 and place RHR 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).

Generic This condition indicates an extreme challenge to the ability to remove RCS-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.

I[Document No.] I Rev. 6 1 Page 228 of276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases CPNPP Basis Reference(s):

1. FRH-O.1A/B Response to Loss of Secondary Heat Sink

2. NEI 99-01 Inadequate Heat Removal Fuel Clad Loss 2.B

[Document No.] Rev. 6 Page 229 of 276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: C. CNTMT Radiation / RCS Activity Degradation Threat: Loss Threshold:

1. Containment radiation greater than 85 R/hr CTEu16 Containment HRRM (u-RE-6290A), or CTWu17 Containment HRRM (u-RE-6290B)

Definition(s):

None Basis:

Plant-Specific Containment radiation monitor readings greater than 85 R/hr indicate the release of reactor coolant, with elevated activity indicative of fuel damage, into the Containment. The reading is derived assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with a 2% clad failures into the Containment atmosphere.

Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations (including iodine spiking) allowed within technical specifications and are therefore indicative of fuel damage. This value is higher than that specified for RCS Loss C.1 (ref. 2, 3).

Per NUS-1 74, the design basis CPNPP RCS specific activity for 1% fuel defects is 340 tICi/gm; therefore, a threshold corresponding to 2% fuel clad damage correlates to a coolant activity of 680 pCi/gm. VL-03-000032 Figure 2A/2B (CRM2) corresponds to approximately 50% clad damage released to the containment atmosphere. Figure 2A/2B provides several potential limits depending on the pressure of the RCS and the presence of containment spray.

The high RCS pressure with containment spray is the most limiting threshold; however, per NEI 99-01, the fuel clad barrier loss threshold should represent a loss of both the fuel clad and RCS barriers. Therefore, the value of curve representing low RCS pressure with spray was used. The change in dose rates based on amount of fuel defects is a linear function; therefore, the threshold at 2% fuel defects is: 2120R/hr *(2% / 50%) = 85 R/hr (ref. 2).

The Containment High Range Radiation Monitors (HRRMs) provide indication of radiation levels in Containment during and after postulated accidents. The monitors are two ion chamber detectors located on the 905' level of Containment approximately 900 apart. The range of each monitor is 1 to 108 R/hr. The output of each detector is fed to an RM-80 located outside Containment. The RM-80 provides monitoring, alarming, and recording functions for the monitor channel. The RM-80 works in conjunction with the PC-1 1, RM-21, and RM-23 assemblies. (ref. 1)

I [Document No.] Rev. 6 Page 230 of 276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Generic The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals 300 pCi/gm dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.

The radiation monitor reading in this threshold is higher than that specified for RCS Barrier Loss threshold ,-.AC.1 since it indicates a loss of both the Fuel Clad Barrier and the RCS Barrier. Note that a combination of the two monitor readings appropriately escalates the mrn..gc..Y classification I"ve'ECL to a Site Area Emergency.

CPNPP Basis Reference(s):

1. DBD-EE-023 Radiation Monitoring System

2. Evaluation performed by Design Engineering & Analysis (Andrea Lemons) (AI-CR-2014-012646-15)

3. EPP-312 Core Damage Assessment

4. NEI 99-01 CNTMT Radiation / RCS Activity Fuel Clad Loss 3.A i[Document No.] I Rev. 6 1 Page 231 of 276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: C. CNTMT Radiation / RCS Activity Degradation Threat: Loss Threshold:

2. Dose equivalent 1-131 coolant activity greater than 300 pCi/cc Definition(s):

None Basis:

Plant-Specific None Generic This threshold indicates that RCS radioactivity concentration is greater than 300 pCi/gm dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.

There is no Potential Loss threshold associated with RCS Activity / Containment Radiation.

CPNPP Basis Reference(s):

1. NEI 99-01 CNTMT Radiation / RCS Activity Fuel Clad Loss 3.B

[Document No.] Rev. 6 Page 232 of276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: C. CNTMT Radiation / RCS Activity Degradation Threat: Loss Threshold:

3. Gross Failed Fuel Monitor, FFLu60 Lu-RE-0406), radiation greater than 1.0E04 pCi/mlI Definition(s):

None Basis:

Plant-Specific The normal Chemical and Volume Control System (CVCS) charging and letdown flow path allows purification of the reactor coolant and control of the RCS volume while maintaining a continuous feed and bleed flow between the RCS and the CVCS. Reactor coolant is first "letdown" from the RCS through a regenerative heat exchanger, which minimizes heat losses from the RCS. Additional cooling takes place in a letdown heat exchanger that acts as the heat sink for the system. Downstream of the letdown heat exchanger pressure control valve and upstream of the mixed bed demineralizers, the letdown stream passes by a Geiger-Mueller radiation detector, FFLu60 (u-RE-0406), mounted on the reactor coolant letdown line to monitor coolant activity and warn of fission products in the letdown coolant if a fuel element failure occurs. Detection of increased coolant activity may be indicative of failed fuel. The monitor initiates Alert and High alarms in the Control Room (PC-1 1 and Plant Computer). (ref.

1).

Core Damage Assessment Guidelines (VL-03-000032) which was incorporated into EPP-312 "Core Damage Assessment" provides the basis for loss of the Fuel Cladding as monitored by the Gross Failed Fuel Monitor. The setpoint recommended by Westinghouse is 1E+04 pCi/ml (ref. 2, 3).

FFLu60 (u-RE-0406) has a range of 1E 1E+7 pCi/ml.

Generic None CPNPP Basis Reference(s):

1. DBD-EE-023 Radiation Monitoring System

2. Evaluation performed by Design Engineering & Analysis (Andrea Lemons) (AI-CR-2014-012646-15)

3. EPP-312 Core Damage Assessment

4. NEI 99-01 Other Indications Fuel Clad Loss 5.A

[Document No.] Rev. 6 Page 233 of 276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: C. CNTMT Radiation / RCS Activity Degradation Threat: Potential Loss Threshold:

None

[Document No.] Rev. 6 Page 234 of 276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: D. CNTMT Integrity or Bypass Degradation Threat: Loss Threshold:

None I [Document No.] Rev. 6 1 Page 235 of 276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: D. CNTMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:

None I [Document No.] Rev. 6 Page 236 of 276

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 Definit ion(s):

None Basis:

Plant-Specific 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 degradation exists if the degradation will likely occur within relatively short period of time 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.

Generic This threshold addresses any other factors that are to be used by the Emergency DireotO-Coordinator in determining whether the Fuel Clad barrier is lost CPNPP Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment Fuel Clad Loss 6.A I [Document No.] Rev. 6 Page 237 of 276

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:

Plant-Specific 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 degradation exists if the degradation will likely occur within relatively short period of time 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.

Generic This threshold addresses any other factors that are to be used by the Emergency CoordinatorDireGtei in determining whether the Fuel Clad barrier is potentially lost. The Emergency D!reetE)FCoordinator should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

CPNPP Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment Potential Fuel Clad Loss 6.A

[Document No.] Rev. 6 Page 238 of 276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: A. RCS or SG Tube Leakage Degradation Threat: Loss Threshold:

1. An automatic or manual ECCS (SI) actuation required by EITHER:

• UNISOLABLE RCS 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:

Plant-Specific ECCS (SI) actuation is caused by (ref. 1):

• Pressurizer low pressure less than 1820 psig

• Steamline low pressure less than 610 psig

• Containment high pressure greater than 3.0 psig Generic This threshold is based on an UNISOLABLE RCS 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 RCSBarrier.

This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacing system. The mass loss may be into any location - inside containment, to the secondary-side (i.e., steam generator tube leakage) or outside of 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 I.A will also be met.

CPNPP Basis Reference(s):

1. EOP-0.0A/B Reactor Trip or Safety Injection

2. EOP-3.OA/B Steam Generator Tube Rupture

3. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Loss I.A

[Document No.] I Rev. 6 Page 239 of 276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: A. RCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold:

1. Operation of a standby charging pump is required by EITHER:

• UNISOLABLE RCS leakage

• SG tube leakage Definition(s):

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

Basis:

Plant-Specific The Chemical and Volume Control System (CVCS) includes three charging pumps (one positive displacement pump and two centrifugal charging pumps) that take suction from the volume control tank and return the cooled, purified reactor coolant to the RCS. The centrifugal charging pumps in the CVCS also serve as the high-head safety injection pumps in the Emergency Core Cooling System. Positive displacement pump capacity is 98 gpm. The capacity of each centrifugal pump is 150 gpm. A second charging pump being required (positive displacement or centrifugal) is indicative of a substantial RCS leak. (ref. 1, 2, 3)

Generic This threshold is based on an UNISOLABLE RCS leak that results in the inability to maintain pressurizer level within specified limits by operation of a normally used charging (makeup) pump, but an ECCS (SI) actuation has not occurred. The threshold is met when an operating procedure, or operating crew supervision, directs that a standby charging (makeup) pump be placed in service to restore and maintain pressurizer level.

This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacing system. The mass loss may be into any location - inside containment, to the secondary-side (i.e., steam generator tube leakage) or outside of containment.

If a leaking steam generator is also FAULTED outside of containment, the declaration escalates to a Site Area Emergency since the Containment Barrier Loss threshold I.A will also be met.

CPNPP Basis Reference(s):

1. FSAR 9.3.4

2. FSAR Table 9.3-7

3. SOP-103A/B Chemical and Volume Control System

4. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Potential Loss I.A

[Document No.] Rev. 6 Page 240 of 276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: A. RCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold:

2. CSFST Integrity-RED Path conditions met Definition(s):

None Basis:

Plant-Specific Critical Safety Function Status Tree (CSFST) RCS Integrity-RED path indicates the RCS barrier is under significant challenge (ref. 1).

Generic This condition indicates an extreme challenge to the integrity of the RCS pressure boundary I due to pressurized thermal shock - a transient that causes rapid RCS cooldown while the RCS is in Mode 3 or higher (i.e., hot and pressurized).

CPNPP Basis Reference(s):

1. FRP-0.1A/B Response to Imminent Pressurized Thermal Shock Condition

2. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Potential Loss I.B I [Document No.] Rev. 6 1 Page 241 of 276

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

[Document No.] Rev. 6 Page 242 of2

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: B. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold:

1. CSFST Heat Sink-RED path conditions met AND Heat sink is required Definition(s):

None Basis:

Plant-Specific 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 Plant Computer (ref. 1).

The phrase "and heat sink required" precludes the need for classification for conditions in which RCS pressure is less than SG pressure or Heat Sink-RED path entry was created through operator action directed by an ERG. For example, FRH-0.1 is entered from CSFST Heat Sink-Red. Step 1 tells the operator to determine if heat sink is required by checking that RCS pressure is greater than any non-faulted SG pressure and RCS temperature is greater than 350 0F. If these conditions exist, Heat Sink is required. Otherwise, the operator is to either return to the procedure and step in effect and place RHR 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).

Generic This condition indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the RCS 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 2B.2; both will be met. This condition warrants a

[Document No.] I Rev. 6 Page 243 of 276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Site Area Emergency declaration because inadequate RCS heat removal may result in fuel heat-up sufficient to damage the cladding and increase RCS pressure to the point where mass will be lost from the system.

CPNPP Basis Reference(s):

1. FRH-O.1A/B Response to Loss of Secondary Heat Sink

2. NEI 99-01 Inadequate Heat Removal RCS Loss 2.B

[Document No.] Rev. 6 Page 244 of 276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: C. CNTMT Radiation/ RCS Activity Degradation Threat: Loss Threshold:

1. Containment radiation greater than 5 R/hr CTEu16 Containment HRRM (u-RE-6290A), or CTWu17 Containment HRRM (u-RE-6290B)

Definition(s):

N/A Basis:

Plant-Specific As part of the elimination of the Post-Accident Sampling system, Westinghouse performed analysis for Comanche Peak on Core Damage Assessment Guidelines (VL-03-000032) which was incorporated into EPP-312 "Core Damage Assessment". For setpoint CRM1, Westinghouse assumptions match the requirements of NEI 99-01 for RCS barrier loss with the exception that the level of radioactivity in the RCS is assumed to be at 10% of Technical Specifications levels rather than 100% as recommended by NEI 99-01. However, this adds conservatism to this threshold. The limiting maximum value found in Figure 1A is 4.75 R/hr.

This value is time dependent and corresponds to an hour after shutdown. This value has been rounded to 5 R/hr for instrument readability (ref. 1, 3, 4, 5).

The Containment High Range Radiation Monitors (HRRMs) provide indication of radiation levels in Containment during and after postulated accidents. The monitors are two ion chamber detectors located on the 905' level of Containment approximately 90' apart. The range of each monitor is 1 to 108 R/hr. The output of each detector is fed to an RM-80 located outside Containment.

The RM-80 provides monitoring, alarming, and recording functions for the monitor channel.

The RM-80 works in conjunction with the PC-1 1, RM-21, and RM-23 assemblies. (ref. 2)

Generic 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 ,AC.1 since it indicates a loss of the RCS Barrier only.

There is no Potential Loss threshold associated with RCS Activity I Containment Radiation.

I [Document No.] Rev. 6 Page 245 of 276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases CPNPP Basis Reference(s):

1. Technical Specifications Table 3.3.3-1

2. DBD-EE-023 Radiation Monitoring System

3. Evaluation performed by Design Engineering & Analysis (Andrea Lemons) (AI-CR-2014-012646-15)

4. Technical Specifications B3.3.3

5. EPP-312 Core Damage Assessment

6. NEI 99-01 CNTMT Radiation / RCS Activity RCS Loss 3.A I [Document No.] Rev. 6 Page 246 of 276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: B. CNTMT Radiation/ RCS Activity Degradation Threat: Potential Loss Threshold:

None

[Document No.] Rev. 6 1 Page 247 of 276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: D. CNTMT Integrity or Bypass Degradation Threat: Loss Threshold:

None I [Document No.] I Rev. 6 Page 248 of 276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: D. CNTMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:

None L[Document No.] I Rev. 6 Page 249 of 276

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:

Plant-Specific The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the RCS barrier is lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

" Imminent barrier degradation exists if the degradation will likely occur within relatively short period of time 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.

Generic This threshold addresses any other factors that may be used by the Emergency QieGte Coordinator in determining whether the RCS Barrier is lost.

CPNPP Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment RCS Loss 6.A

[Document No.] Rev. 6 Page 250 of 276

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 RCS barrier Definition(s):

None Basis:

Plant-Specific The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the RCS barrier is potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

• Imminent barrier degradation exists if the degradation will likely occur within relatively short period of time 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.

Generic This threshold addresses any other factors that may be used by the Emergency Direet4 Coordinator in determining whether the RCS Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

CPNPP Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment RCS Potential Loss 6.A I [Document No.] Rev. 6 1 Page 251 of 276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: A. RCS or SG Tube Leakage Degradation Threat: Loss Threshold:

1. A leaking or RUPTURED SG is FAULTED 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:

Plant-Specific None.

Generic 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 RCS Barrier Potential Loss 4-.A.1 and Loss 4-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

[Document No.] Rev. 6 Page 252 of276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases containment barrier is analogous to the lower bound criteria specified in IC SU4 for the fuel clad barrier (i.e., RCS activity values) and IC SU5 for the RCS barrier (i.e., RCS 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.

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 A-R ICs.

The emergenYcy clasification levelECLs resulting from primary-to-secondary leakage, with or without a steam release from the FAULTED SG, are summarized below.

I [Document No.] I Rev. 6 Page 253 of 276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases 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 Unusual Event per Unusual Event per Greater than 25 gpm SU4SU5.1 SJ4SU5.1 Requires operation of a standby Site Area Emergency per charging (makeup) pump (RCS Alert per FAI.1 FS1.1 BarrierPotentialLoss)

Requires an automatic or manual Site Area Emergency per ECCS (SI) actuation (RCSBarrier Alert per FAI.1 FS1.1 Loss)

There is no Potential Loss threshold associated with RCS or SG Tube Leakage.

CPNPP Basis Reference(s):

1. EOP-3.0 Steam Generator Tube Rupture

2. EOP-2.OA/B Faulted Steam Generator Isolation

3. NEI 99-01 RCS or SG Tube Leakage Containment Loss I.A

[Document No.] Rev. 6 Page 254 of 276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: A. RCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold:

None I [Document No.] Rev. 6 Page 255 of 276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: B. Inadequate heat Removal Degradation Threat: Loss Threshold:

None I [Document No.] Rev. 6 Page 256 of 276

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)

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

Definition(s):

None Basis:

Plant-Specific 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 Plant Computer (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).

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 1,2000 F (ref. 1), Fuel Clad barrier is also lost.

Generic This threshold addresses any other factors that may be used by the Emergency ir-eeGt Coordinator in determining whether the RCS Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

CPNPP Basis Reference(s):

1. FRC-O.1A/B Response to Inadequate Core Cooling

2. NEI 99-01 Inadequate Heat Removal Containment Potential Loss 2.A I[Document No.] I Rev. 6 - Page 257 of 276 ý

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: C. CNTMT Radiation/RCS Activity Degradation Threat: Loss Threshold:

None

[Document No.] Rev. 6 Page 258 of 276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: C. CNTMT Radiation/RCS Activity Degradation Threat: Potential Loss Threshold:

1. Containment radiation greater than 1,110 R/hr CTEul 6 Containment HRRM (u-RE-6290A), or CTWu17 Containment HRRM (u-RE-6290B)

Definition(s):

None Basis:

Plant-Specific Containment radiation monitor readings greater than 1,110 R/hr indicate significant fuel damage well in excess of that required for loss of the RCS barrier and the Fuel Clad barrier.

Regardless of whether the Containment barrier itself is challenged, this amount of activity in containment could have severe consequences if released. It is, therefore, prudent to treat this as a Potential Loss of the Containment barrier. (ref. 2, 3)

The readings are higher than that specified for Fuel Clad Loss C.3 and RCS Loss C.1.

Containment radiation readings at or above the Containment barrier Potential Loss threshold, therefore, signify a loss of two fission product barriers and Potential Loss of a third, indicating the need to upgrade the emergency classification to a General Emergency.

The analysis performed in VL-03-00032 was used to determine the potential containment loss threshold. Since the containment potential loss threshold also assumes a loss of the RCS barrier, the Figure 3A (CRM3) curve which represents the dose response at low RCS pressure with sprays present was used. The setpoint was developed on the assumption of 100% fuel rod rupture with 100% of the noble gas and 50% of the iodine and cesium in the RCS released to containment. With containment spray operating, the containment inventory of all fission products except the noble gases are reduced by a factor of 100. Per Figure 3A, the value one hour after shutdown for 100% rod rupture is 5560 R/hr; therefore, the EAL threshold at 20%

fuel defects is: 5,560 R/hr *20% = 1,112 R/hr (rounded to 1,110 R/hr for instrument readability)

(ref. 2, 3).

The Containment High Range Radiation Monitors (HRRMs) provide indication of radiation levels in Containment during and after postulated accidents. The monitors are two ion chamber detectors located on the 905' level of Containment approximately 90' apart. The range of each monitor is 1 to 108 R/hr. The output of each detector is fed to an RM-80 located outside Containment. The RM-80 provides monitoring, alarming, and recording functions for the monitor channel. The RM-80 works in conjunction with the PC-1 1, RM-21, and RM-23 assemblies. (ref. 1).

I [Document No.] I Rev. 6 Page 259 of 276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Generic 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 RCS Barrier Loss thresholds.

NUREG-1228, Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents, indicates the fuel clad failure must be greater than approximately 20% in order for there to be a major release of radioactivity requiring offsite protective actions. For this condition to exist, there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential loss of containment which would then escalate the emergencY claSificatilo; levelECL to a General Emergency.

CPNPP Basis Reference(s):

1. DBD-EE-023 Radiation Monitoring System

2. Evaluation performed by Design Engineering & Analysis (Andrea Lemons) (AI-CR-2014-012646-15)

3. EPP-312 Core Damage Assessment

4. NEI 99-01 CNTMT Radiation / RCS Activity Containment Potential Loss 3.A I [Document No.] Rev. 6 1 Page 260 of 276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: D. CNTMT Integrity or Bypass Degradation Threat: Loss Threshold:

1. Containment isolation is required AND EITHER:

• Containment integrity has been lost based on Emergency Coordinator 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:

Plant-Specific None Generic 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 4.A.1 and 4.A.2.

4-A41First 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 RCS 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 DirectEr 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 9-F--41. Two simplified examples are provided. One is leakage from a penetration and the other is leakage from an in-service system valve.

I [Document No.] Rev. 6 1 Page 261 of 276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases 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 RCS 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 RCS 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 A--R ICs.

4-A.2Second 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 8-F--41. 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.

I [Document No.] Rev. 6 1 Page 262 of 276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Refer to the bottom piping run of Figure 8-F--4-1. 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. Ifthe pump developed a leak that allowed steam/water to enter the Auxiliary Building, then second threshold-44B 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 4-A-_--to be met as well.

Following the leakage of RCS 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 A-RICs.

The status of the containment barrier during an event involving steam generator tube leakage is assessed using Loss Threshold 4-.A.1.

CPNPP Basis Reference(s):

1. NEI 99-01 CNTMT Integrity or Bypass Containment Loss 4.A

[Document No.] Rev. 6 Page 263 of 276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: D. CNTMT Integrity or Bypass Degradation Threat: Loss Threshold:

2. Indications of RCS leakage outside of containment Definition(s):

None Basis:

Plant-Specific ECA-1.2A/B LOCA Outside Containment (ref. 1) provides instructions to identify and isolate a LOCA outside of the containment. Potential RCS leak pathways outside containment include (ref. 1):

" Residual Heat Removal

• Safety Injection

" Chemical & Volume Control

" RCP seals

" PZR/RCS Loop sample lines Generic 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 RCS 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 RCS 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 RCS mass is being lost outside of the containment.

Refer to the middle piping run of Figure 9-F--41. 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 4AD.1 to be met as well.

To ensure proper escalation of the emergency classification, the RCS leakage outside of containment must be related to the mass loss that is causing the RCS Loss and/or Potential Loss threshold 4-.A.1 to be met.

[Document No.] Rev. 6 Page 264 of 276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases CPNPP Basis Reference(s):

1. ECA-1.2A/B LOCA Outside Containment

2. NEI 99-01 CNTMT Integrity or Bypass Containment Loss

[Document No.] Rev. 6 Page 265 of 276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Figure 1: Containment Integrity or Bypass Examples Effluent o

-ease ing Monitor 9O

wa I [Document No.] Rev. 6 Page 266 of 276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: D. CNTMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:

1. CSFST Containment-RED Path conditions met Definition(s):

None Basis:

Plant-Specific Critical Safety Function Status Tree (CSFST) Containment-RED path is entered if containment pressure is greater than or equal to 50 psig and represents an extreme challenge to safety function. The CSFSTs are normally monitored using the SPDS display on the Plant Computer (ref. 1).

50 psig is the containment design pressure and is the pressure used to define CSFST Containment Red Path conditions (ref. 2).

Generic 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 RCS 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.

CPNPP Basis Reference(s):

1. FRC-Z.1A/B Response to High Containment Pressure

2. FSAR Table 6.2.1-1

3. NEI 99-01 CNTMT Integrity or Bypass Containment Potential Loss 4.A I [Document No.] Rev. 6 1 Page 267 of 276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: D. CNTMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:

2. Containment hydrogen concentration greater than 4%

Definition(s):

None Basis:

Plant-Specific 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 gasses in 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. A combustible mixture can be formed when hydrogen gas concentration in the Containment atmosphere is greater than 4% by volume. All hydrogen measurements are referenced to concentrations in dry air even though the actual Containment environment may contain significant steam concentrations. The plant has two hydrogen monitoring systems. Each monitoring system consists of four sensor modules and one microprocessor analyzer. Two sensors from each Containment are coupled to one of the two hydrogen microprocessors located in the Control Room. Thus each microprocessor analyzer is shared by Units 1 and 2. The analyzer system has a range of 0-10% hydrogen by volume. The detector modules are located on the 905',

873', and 860' elevations in Containment. A fourth detector is located on 832' level across from the loop room entrance for loops 1 and 4. Hydrogen concentration is displayed in the Control Room on u-AI-5506A/B and u-AI-5506C/D.

Hydrogen concentration can also be displayed on the Plant Computer. Alarms at -3% are provided for high hydrogen concentration, u-ALB-3A, window 3.7. If a hydrogen concentration value can not be obtained from the hydrogen monitoring system, a grab sample from the containment PIG radiation monitor may be used to determine the hydrogen concentration (ref.

1,2,3,4).

To generate such levels of combustible gas, loss of the Fuel Clad and RCS 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.

Generic 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 potential loss of the Containment Barrier.

[Document No.] Rev. 6 Page 268 of 276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases CPNPP Basis Reference(s):

1. FRC-O.1A/B Response to Inadequate Core Cooling, Attachment 5

2. FSAR Section 6.2.5

3. FSAR Table 7.5-7A

4. CHM-1 11, Primary Chemistry Accident Assessment Sampling Program

7. NEI 99-01 CNTMT Integrity or Bypass Containment Potential Loss 4.B I[Document No.] Rev. 6 1 Page 269 of 276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: D. CNTMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:

3. Containment pressure greater than 18 psig with neither Containment Spray system train operating per design for greater than or equal to 15 min. (Note 1)

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

Definition(s):

None Basis:

Plant-Specific This threshold represents a Potential Loss of the Containment barrier because the Containment heat removal and depressurization equipment (but not including Containment venting strategies) is either lost or degraded. The Containment Spray System (CSS) is designed to remove heat from the Containment environment following a LOCA, a main steam line break accident, or a feedwater line break accident. Each unit of the CPNPP is equipped with two redundant Containment spray trains, each designed to provide emergency Containment heat removal in the event of a LOCA. This system, in conjunction with the ECCS, removes postaccident thermal energy from the Containment environment, thereby reducing the Containment pressure and temperature. Each train includes two containment spray pumps, spray headers, nozzles, valves, and piping. Each train is powered from a separate safeguard bus. (ref. 1)

The Containment pressure setpoint (18 psig, ref. 2) is the pressure at which the Containment Spray System should actuate and begin performing its function. The design basis accident analyses and evaluations assume the loss of one Containment Spray System train (ref. 1).

Generic 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 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.

I [Document No.] I Rev. 6 Page 270 of 276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases CPNPP Basis Reference(s):

1. FSAR Section 6.2.2

2. FRC-Z.1A/B Response to High Containment Pressure

3. NEI 99-01 CNTMT Integrity or Bypass Containment Potential Loss 4.C

[Document No.] Rev. 6 Page 271 of 276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: E. 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:

Plant-Specific 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 degradation exists if the degradation will likely occur within relatively short period of time 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.

Generic This threshold addresses any other factors that may be used by the Emergency Qikeeter Coordinator in determining whether the Containment Barrier is lost.

CPNPP Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment PC Loss 6.A

[Document No.] Rev. 6 Page 272 of 276

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment 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 Containment barrier Definition(s):

None Basis:

Plant-Specific 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.

" Imminent barrier degradation exists if the degradation will likely occur within relatively short period of time 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.

Generic This threshold addresses any other factors that may be used by the Emergency DiQeeto Coordinator in determining whether the Containment Barrier is lost.

CPNPP Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment PC Potential Loss 6.A I [Document No.] I Rev. 6 1 Page 273 of 276

ATTACHMENT 3 Safe Operation & Shutdown Areas Tables R-3 & H-2 Bases

Background

NEI 99-01 Revision 6 ICs 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-specificlist of plant rooms or areas with entry-relatedmode applicabilityidentified" should specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normalplant operation, cooldown and shutdown. Do not include rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs,corrective measures or emergency operations). In addition, the list should specify the plant mode(s) during which entry would be required for each room or area.

The list should not include rooms or areasfor which entry is requiredsolely to perform actions of an administrative or record keeping nature (e.g., normal rounds or routine inspections).

Further, as specified in IC HA5:

The list need not include the Control Room if adequate engineered safety/design features are in place to preclude a Control Room evacuation due to the release of a hazardous gas.

Such features may include, but are not limited to, capabilityto draw airfrom multiple air intakes at different and separate locations, inner and outer atmospheric boundaries,or the capability to acquire and maintain positive pressure within the Control Room envelope.

I [Document No.] Rev. 6 Page 274 of 276

ATTACHMENT 3 Safe Operation & Shutdown Areas Tables R-3 & H-2 Bases CPNPP Table R-3 and H-2 Bases A review of station operating procedures identified the following mode dependent in-plant actions and associated areas that are required for normal plant operation, cooldown or shutdown:

Location- Modes- Modes-Safe Shutdown Area 1, 2 3, 4, 5, or 6 Charging Pump Rooms SDC Equipment. Shut Down Cooling (SDC)

- No entry required - No entry required Inventory Control Equipment Inventory Control Equipment

-Entry required during pump - Entry required during pump starts and stops starts and stops Reactivity Control.

- Entry required during pump starts and stops

-Containment Spray Pumps A Post-accident Containment Post-accident Containment and B Pressure Control Pressure Control (modes 3 and

- No entry required 4)

- No entry required

-SI Pumps A and B Post-accident ECCS Post-accident ECCS

- No entry required - No entry required

-Residual Heat Removal Pumps Post-accident ECCS Decay Heat Removal (Modes 4, 5, A and B - No entry required and 6

- Entry required for pumps starts and stops

- CVCS Valve Rooms, Auxiliary Inventory Control Equipment Inventory Control Equipment Building 810' and 822' - Entry required during - Entry required during pump pump starts and stops starts and stops Reactivity Control. Reactivity Control.

- Entry required during - Entry required during pump pump starts and stops starts and stops

-Service Water Intake Structure Ultimate Heat Sink Equipment for Ultimate Heat Sink Equipment for Habitability Control, Habitability Control, Containment Containment Temperature, Temperature, and Shutdown and Shutdown Cooling Cooling

- No entry required No entry required 1E Switchgear Rooms 810', Electrical Power. Electrical Power.

832', and 852' - Entry required for manual - Entry required for manual breaker manipulations on breaker manipulations on component operations, component operations reactor startup and shutdown Control Building 807'Cable Electrical Power. Electrical Power.

Spreading Room - No entry required - No entry required Control Building 792' UPS and Electrical Power. Electrical Power.

Battery Rooms - No entry required Emergency Diesel Generators A Electrical Power. Electrical Power.

&B - No entry required - No entry required Emergency Diesel Generators Electrical Power. Electrical Power.

Day Tank Rooms - No entry required - No entry required Control Building 830' Control Continuously occupied, Continuously occupied, capable Room capable of Ventilation Isolation of Ventilation Isolation mode, mode, covered under H-6 covered under H-6 Control Building 840' Technical - No entry required No entry required Support Center

, [Document No.] Rev. 6 Page 275 of 276

ATTACHMENT 3 Safe Operation & Shutdown Areas Tables R-3 & H-2 Bases Location- Modes- Modes-Safe Shutdown Area 1, 2 3, 4, 5, or 6 Control Building 778' Safety - No entry required - No entry required Chiller Rooms Auxiliary Building 790' - No entry required - No entry required Auxiliary Building 810' other - No entry required - No entry required than CVCS Valave Rooms and Charging Pump Rooms Auxiliary Building 830' - No entry required - No entry required Auxiliary Building 852' - No entry required - No entry required Auxiliary Building 873' - No entry required - No entry required Auxiliary Building 886' - No entry required - No entry required Safeguards 790' - No entry required - No entry required Safeguards 810' - No entry required - No entry required Safeguards 831' - No entry required - No entry required Safeguards 852' - No entry required - No entry required Safeguards 873' - No entry required - No entry required Turbine Building Elevations - No entry required - No entry required Aux. Feedwater Pump Rooms Steam Generator Heat Removal Steam Generator Heat Removal A, B, and Turbine Driven - No entry required - No entry required Table R-3 & H-2 Results Table R-3/H-2 Safe Operation & Shutdown Rooms/Areas Room/Area Mode Applicability Charging Pump Rooms 1, 2, 3, 4, 5, 6 CVCS Valve Rooms 1,2, 3, 4, 5, 6 1E Switchgear Rooms All RHR Pump Rooms 4, 5, 6 Plant Operating Procedures Reviewed

1. IPO-003A/B

2. IPO-005A/B

3. IPO-001A/B

4. IPO-002A/B

5. SOP-1 03

6. SOP-104

7. SOP-102 I [Document No.] Rev. 6 Page 276 of 276

ATTACHMENT 5 TO TXX-15101 CPNPP RADIOLOGICAL EFFLUENT EAL VALUES (7 PAGES)

EAL Section R Revision 6 Table R-1 Effluent Monitor Classification Thresholds Review Submitted By:

Signature: /JA Date:

CmrygWiecheringEieerg'ercyb~'nning Review~ed By:

Signature: JA.f4 . Hrf Nt 4, Date: 6 /*/[s Jeffery Hull, Emergency IFlgrhiing Manager Signature: 16y ( " ae r-Ky Fishencord, E-m g-enc-y Pla'n'ning Signaure:cr1 I Date: __2 3 De6 O' onnor, Radiation Protection Manager Signature:

-- -,Ig* ........

Date:

Andre'a Lemons, Senior Engineer I of 7 06/18/2015

EAL Section R Revision 6 Table R-I Effluent Monitor Classification Thresholds Table R- 1 Effluent Monitor Classification Thresholds Type Release Point Monitor GE SAE ALERT UE Gaseous Plant Vent X-RE-5567 A+B --- - - - - - - - - 6.52E-4*

PVG-384 +PVG-385 uCi/cc Plant Vent X-RE-5570 A+B 4.0E07

• 4.0E06

• 4.0E05

• 4.0E04*

(WRGM) uCi/sec uCi/sec uCi/sec uCi/sec PVF-684 + PVF-685 u Main Steam 90 uCi/cc 9 uCi/cc .9 uCi/cc 2 X High MSL-a78 M-RE-2325 Alarm Setpoint MSL-i79 u-RE-2326 MSL-*180 U-RE-2327 MSL-u8l P-RE-2328 Liquid Liquid Waste 2 X High LWE-076 X-RE-__53 Alarm Setpoint Service Water u-RE.4269 - - -- ---- 2 X High SSW..,65 u-RE-4270 Alarm Setpoint SSW_-A66 I I I

• Total of the two monitors equal this value or greater.

Reference Material for Table R-1 Effluent Monitor Classification Thresholds Type Release Point Monitor High Source for Range**

Alarm Alarm Gaseous Plant Vent X-RE-5567 A or B 3.26E-4 CLI-744-3 IE-06 to IE-02 PVG-384 +PVG-385 uCi/cc uCilcc Plant Vent X-RE-5570 A or B 2.0E04 CLI-744-3 IE-06 to IE+05 (WRGM) uCi/sec uCifcc PVF-684 + PVF-685 I E-04 to 1E+12 uCi/Sec***

Main Steam u-RE-2325 2.OE-1 PC-11 IE-01 to IE+03 MSL--78 u-RE-2326 uCi/cc uCi/cc MSL-U79 MSL-8U0 u--RE-2327 MSL-1u81 u-RE-2328 Liquid Liquid Waste X-RE-5253 2.48E-04 CLI-744-1 1E-05 to 5E-02 LWE-076 uCi/sec uCi/cc Service Water u-RE-4269 4.OE-06 PC-I1 IE-05 to 5E-02 SSw%-65 u-RE-4270 uCi/sec uCi/cc SSW-3j66 I I II

• • Range DBD-EE-023 Specified Instrument Range unless noted.

• • • Range is from the PC-11.

2 of 7 06/18/2015

EAL Section R Revision 6 Table R-1 Effluent Monitor Classification Thresholds Dose Projection Assumptions The dose projections used the following:

Only one unit was used to make dose projections.

One mile was used for Exclusion Area Boundary.

Reactor has been Shutdown for one hour.

The event start time is same as Reactor Shutdown.

Accident Type for Vent release was a LOCA.

Filtration: No Stability Class: D Wind Speed: 10 miles per hour Plant Vent UE Assumptions PVG-384 +PVG-385 X-RE-5567 A+B (This monitor is used for an UE only)

High Alarm Setpoint: 3.26E-4 uCi/cc Plant Vent Total: 6.52E-04 uCi/cc Flow Rate: 140,000 scfm Reactor has been Shut Down for one hour.

Projected Dose using CPAMPEDE 8.0 PROJECTED DOSE I Mile 2 Mile 5 Mile 10 Mile TEDE (Rem) 0.002 0.0008 0.0002 0.0001 CDE (Rem) 0.009 0.003 0.0009 0.0003 Plume Shine (R/hr) 0.0002 0.0001 0.0000 0.0000 3 of 7 06/18/2015

EAL Section R Revision 6 Table R-1 Effluent Monitor Classification Thresholds Plant Vent (WRGM) UE Assumptions Plant Vent (WRGM) PVF-684 + PVF-685 X-RE-5570 A+B High Alarm Setpoint: 2.0E04 uCi/sec CPAMPEDE Value U1 Value U2 2E4 uCi/Sec 2E4 uCi/Sec Plant Vent Total: 4E04 uCi/sec Projected Dose using CPAMPEDE 8.0 PROJECTED DOSE I Mile 2 Mile 5 Mile 10 Mile TEDE (Rem) 0.001 0.0005 0.0001 0.0000 CDE (Rem) 0.006 0.002 0.0005 0.0002 Plume Shine (R/hr) 0.00 0.00 0.00 0.00 Plant Vent (WRGM) ALERT Assumptions Plant Vent (WRGM) PVF-684 + PVF-685 X-RE-5570 A+B 10 mRem TEDE or 50 mRem thyroid CDE High Alarm Setpoint: 2.0E04 uCi/sec CPAMPEDE Value UI I Value U2 2E5 uCi/Sec 2E5 uCi/Sec Plant Vent Total: 4.0E05 uCi/sec Projected Dose using CPAMPEDE 8.0 PROJECTED DOSE I Mile 2 Mile 5 Mile 10 Mile TEDE (Rem) 0.013 0.005 0.001 0.0005 CDE (Rem) 0.061 0.022 0.006 0.002 Plume Shine (R/hr) 0.001 0.00 0.00 0.00 4 of 7 06/18/2015

EAL Section R Revision 6 Table R-1 Effluent Monitor Classification Thresholds Plant Vent (WRGM) SAE Assumptions Plant Vent (WRGM) PVF-684 + PVF-685 X-RE-5570 A+B 100 mRem TEDE or 500 mRem thyroid CDE High Alarm Setpoint: 2.0E04 uCi/sec ICPAMPEDE Value U I Value U2 2E6 uCi/Sec 2E6 uCi/Sec Plant Vent Total: 4E06 uCi/sec Projected Dose using CPAMPEDE 8.0 PROJECTED DOSE I Mile 2 Mile 5 Mile 10 Mile TEDE (Rem) 0.132 0.047 0.012 0.005 CDE (Rem) 0.577 0.208 0.054 0.020 Plume Shine (R/hr) 0.010 0.003 0.001 0.00 Plant Vent (WRGM) GE Assumptions Plant Vent (WRGM) PVF-684 + PVF-685 X-RE-5570 A+B 1000 mRem TEDE or 5000 mRem thyroid CDE High Alarm Setpoint: 2.0E04 uCi/sec CPAMPEDE Value U 1 Value U2 2E7 uCi/Sec 2E7 uCi/Sec Plant Vent Value: 4.0E07 uCi/sec Projected Dose using CPAMPEDE 8.0 PROJECTED DOSE 1 Mile 2 Mile 5 Mile 10 Mile TEDE (Rem) 1.320 0.474 0.124 0.045 CDE (Rem) 5.780 2.080 0.542 0.1 99 Plume Shine (R/hr) 0.102 10.037 0.010 0.004 5 of 7 06/18/2015

EAL Section R Revision 6 Table R-1 Effluent Monitor Classification Thresholds Main Steam UE Assumptions The setpoints differ on the main steam lines so an absolute number cannot be derived.

Following are the current setpoints from the PC-1 1:

Main Steam Line Monitor High (uCi/cc) [ Alert(uCi/ec) 2.OE-01 1.OE-01 Dose Projection Assumptions The dose projections used the following:

Only one unit was used to make dose projections.

One mile was used for Exclusion Area Boundary.

Reactor has been Shutdown for one hour.

The event start time is same as Reactor Shutdown.

Accident Type for Main Steam Line monitors release was a SGTR.

Flow Rate 120,000 lb/hr Filtration: No Stability Class: D Wind Speed: 10 miles per hour Main Steam UE Assumptions High Alarm Setpoint: 2.OE-1 uCi/cc MSL Total: 0.4 uCi/cc Projected Dose using CPAMPEDE 8.0 PROJECTED DOSE 1 Mile 2 Mile 5 Mile 10 Mile TEDE (Rem) 0.002 0.0006 0.0002 0.0001 CDE (Rem) 0.020 0.007 0.002 0.0007 Plume Shine (R/hr) 0.00 0.00 0.00 0.00 6 of 7 06/18/2015

EAL Section R Revision 6 Table R-1 Effluent Monitor Classification Thresholds Main Steam ALERT Assumptions 10 mRem TEDE or 50 mRem thyroid CDE Plant Vent Value: 0.9 uCi/cc Projected Dose using CPAMPEDE 8.0 PROJECTED DOSE I Mile 2 Mile 5 Mile 10 Mile TEDE (Rem) 0.004 0.001 0.0004 0.0001 CDE (Rem) 0.046 0.017 0.004 0.002 Plume Shine (R/hr) 0.003 0.001 0.00 0.00 Main Steam SAE Assumptions 100 mRem TEDE or 500 mRem thyroid CDE Plant Vent Value: 9.0 uCi/cc Projected Dose using CPAMPEDE 8.0 PROJECTED DOSE I Mile 2 Mile 5 Mile 10 Mile TEDE (Rem) 0.041 0.015 0.004 0.001 CDE (Rem) 0.461 0.166 0.043 0.016 Plume Shine (R/hr) 0.002 0.001 10.00 0.00 Main Steam GE Assumptions 1000 mRem TEDE or 5000 mRem thyroid CDE Plant Vent Value: 90 uCi/ec Projected Dose using CPAMPEDE 8.0 PROJECTED DOSE I Mile 2 Mile 5 Mile 10 Mile TEDE (Rem) 0.405 0.146 0.038 0.014 CDE (Rem) 4.610 1.660 0.433 0.158 Plume Shine (R/hr) 0.020 0.007 10.002 0.001 7 of 7 06/18/2015

ATTACHMENT 6 TO TXX-15101 EMERGENCY ACTION LEVEL WALLCHARTS FOR CPNPP (3 PAGES)