ML17123A202

Enclosure to WO 17-0042 - WCNOC EAL Technical Basis Document

ML17123A202
Person / Time
Site:	Wolf Creek
Issue date:	04/26/2017
From:	Wolf Creek
To:	Office of Nuclear Reactor Regulation
Shared Package
ML17123A291	List: ML17123A202 WO 17-0042, Operating Corporation Supplement to RAI Response Regarding Adopting New EAL Scheme
References
WO 17-0042
Download: ML17123A202 (229)

Text

ENCLOSURE TO WO 17-0042 WCNOC EAL TECHNICAL BASIS DOCUMENT (228 pages)

APF [XX-XXX-XX]

Revision xxx Follow-up RAI Response 4/10/17 Page 1 of 228 INFORMATION USE

TABLE OF CONTENTS

. SECTION PAGE 1.0 PURPOSE ......... ~ ....................................................................................................................3 2.0 DISCUSS.ION ..........................................................................................................................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 ................................................................................................6 2.6 Operating Mode Applicability ...............................................................................................8 3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS ............................................... 9 3.1 General Considerations .......................................................................................................9 3.2 Classification Methodology ............................................................................................*... 1O 4.0 *REFERENCES ..................................................................................................................... 13 4.1 Developmental ................................................................................................................... 13 4.2 Implementing .......'................................................ ~ ............................................................. 13 5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS ............................................................... 14 6.0 WCGS TO NEI 99-01 Rev. 6 EAL CROSS-REFERENCE .................................................... 21 7.0 ATTACHMENTS ...................................................................................................................25 1 Emergency Action Level Technical Bases ................................................................ 25 Category R Abnormal Rad Release I Rad Effluent.. ........................................ 26 Category C Cold Shutdown I Refueling System Malfunction ... :....................... 61 Category H Hazards ...................................................................................... 102 Category S System Malfunction .................................................................... 136 Category F Fi~sion Product Barrier-Degradation .................................... ~.~ ... 176 2 Fission Product Barrier Loss I Potential Loss Matrix and Bases .................................................................................................... 183 3 Safe Operation & Shutdown Areas Tables R-3 & H-2 Bases ................................. 227 Page2 of 228 INFORMATION USE

1.0 PURPOSE This document provides an explanation and rationale for each Emergency Action Level (EAL) included in the EAL Upgrade Project for Wolf Creek Generating Station (WCGS). Decision-makers responsible for implementation of procedure EPP 06-005 "Emergency Classification" may use this document as a technical reference in support of EAL interpretation. This information may assist the Emergency Manager 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 offsite 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 Manager refers to it during an event), the NRG staff expects that changes to the basis document will be evaluated in accordance with the provisions of 10 CFR 50.54(q). Additionally, changes to plant OFNs and EMGs that may impact EAL bases shall 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 Wolf Creek Station Radiological Emergency Response Plan (RERP) AP 06-002.

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 Revisions 4 and 5 were subsequently issued for industry implementation.

Enhancements over earlier revisions included:

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

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

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

Subsequently, Revision 6 of NEI 99-01 has been issued which incorporates resolutions to numerous implementation issues including the NRC EAL Frequently Asked Questions (FAQs).

Using NEI 99-01 Revision 6, "Methodology for the Development of Emergency Action Levels for Non-Passive Reactors," November 2012 (ADAMS Accession Number ML12326A805) (ref.

4.1.1 ), Wolf Creek Nuclear Operating Company (WCNOC) conducted an EAL implementation upgrade project that produced the EALs discussed herein.

Page 3 of 228 INFORMATION USE ,.

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

Containment Barrier thresholds are used as criteria for escalation of the emergency classification level (EGL) from Alert to a Site Area Emergency or a General Emergency 2.3 Fission Product Barrier Classification Criteria The following criteria are the bases for event classification related to fission product barrier loss or potential loss:

Alert:

Any loss or any potential loss of either Fuel Clad or RCS barrier Site Area Emergency:

Loss or potential loss of any two barriers General Emergency:

Loss of any two barriers and loss or potential loss. of the third barrier Page 4 of 228 INFORMATION USE .

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

• 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.

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

Page 5 of 228 INFORMATION USE

EAL Groups, Categories and Subcategories EAL Group/Category EAL Subcategory I

Any Operating Mode:

R - Abnormal Rad Levels I Rad Effluent 1 - Radiological Effluent 2 - Irradiated Fuel Event 3 - Area Radiation Levels H - Hazards and Other Conditions 1 - Security Affecting Plant Safety 2 - Seismic Event 3 - Natural or Technological Hazard 4- Fire

• 5 - Hazardous Gases 6 - Control Room Evacuation 7 - Emergency Manager Judgment 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- RTS 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 I 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 1 & 2 of this document for such information.

2.5 Technical Basis 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 and F) and EAL subcategory. A summary explanation of Page 6 of 228 INFORMATION USE

each category and subcategory is given at the beginning of the technical bases discussions of the EALs included in the category. For each EAL, the following information is provided:

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

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

EAL Identifier (enclosed in rectangle)

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

1. First character (letter): Corresponds to the EAL category as described above (R, C, H, S, 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 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 basis section that provides WCGS-relevant information concerning the EAL as well as a description of the rationale for the EAL as provided in NEI 99-01 Rev. 6.

I Page 7 of 228 INFORMATION USE

WCGS Basis Reference(s):

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

2 Startup Keff ~ 0.99 and rated thermal power s 5%

3 Hot Standby Keff < 0.99 and average reactor coolant temperature~ 350°F 4 Hot Shutdown Keff < 0.99 and average reactor coolant temperature 350°F > Tavg > 200 °F and all reactor vessel head. closure bolts fully tensioned, except as specified in Technical Specification 5.5.17, "Reactor Vessel i-lead Closure Bolt Integrity."

5 Cold Shutdown Keff < 0.99 and average reactor coolant temperature s 200°F and all reactor vessel head closure bolts fully tensioned, except as specified in Technical Specification 5.5.17, "Reactor Vessel Head Closure Bolt Integrity."

6 Refueling One or more reactor vessel head closure bolts are less than fully tensioned, except as specified in Technical Specification 5.5.17, "Reactor Vessel Head Closure Bolt Integrity."

D Defueled All fuel assemblies have been removed from Containment and placed in the spent fuel pool and the SFP transfer canal gate valve is closed.

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.

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

3.1.1 Classification Timeliness NRC regulations require the licensee to establish and maintain the capability to assess, classify, and declare an emergency condition within 15 minutes after the availability of indications to plant operators that an EAL has been exceeded and ,to promptly declare the emergency condition as soon as possible following identification of the appropriate emergency classification level. The NRG staff has provided guidance on implementing this requirement in NSIR/DPR-ISG-01, "Interim Staff Guidance, Emergerfcy Planning for Nuclear Power Plants" (ref. 4.1.9).

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."

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. The validation of indications should be completed in a manner that supports timely emergency declaration.

An indication, report, or condition is considered to be valid when it is verifie~ 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 Manager should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time. If an ongoing radiological release is detected and the release start time is unknown, it should be assumed that the release duration specified in the IC/EAL has been exceeded, absent data to the contrary.

3.1.4 Planned vs. Unplanned Events A planned work activity that results in an expected event or condition which meets or exceeds

• an EAL does not warrant an emergency declaration provided that: 1) the activity proceeds as planned, and 2) the plant remains within the limits imposed by the operating license. Such activities include planned work to test, manipulate, repair, maintain or modify a system or component. In these cases, the controls associated with the planning, preparation and I Page 9 of 228 INFORMATION USE

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 10CFR 50. 72 (ref.

4.1.4).

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

3.1.6 Emergency Manager Judgment While the EALs have been developed to address a full spectrum of possible events and conditions which may warrant emergency classification, a provision for classification based on operator/management experience and judgment is still necessary. The NEI 99-01 EAL scheme provides the Emergency Manager with the ability to classify events and conditions based upon judgment using EALs that are consistent with the ECL definitions (refer to Category H). The Emergency Manager.will need to determine if the effects or consequences of the event or condition reasonably meet or exceed a particular ECL definition. A similar provision is incorporated in the Fission Product Barrier Tables; judgment may be used to determine the status of a fission product barrier.

3.2 Classification Methodology To make an emergency classification, the user will compare an event or condition (i.e., the relevant plant indications and reports) to an EAL(s) and determine if the EAL has been met or exceeded. The evaluation of an EAL must be consistent with the related Operating Mode Applicability and Notes. If an EAL has been met or exceeded, the associated IC is likewise met, the emergency classification process "clock" starts, and the ECL must be declared in accordance with plant procedures no later than 15 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. 4.1.9).

3.2.1 Classific~tion 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 E.CL identified during this review is declared. For example:

• If an Alert EAL and a* Site Area Emergency EAL are met, a Site Area Emergency should be declared.

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

• If two Aleit EALs are met, an Alert should be declared .

.Page 10 of 228 INFORMATION USE

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

3.2.2 Consideration of Mode Changes During Classification 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 ECL 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 Manager must remain alert to events Or conditions that could lead to meeting or exceeding an EAL within a relatively short period of time (i.e., a change in the ECL is imminent). If, in the judgment of the Emergency Manager, meeting an EAL is imminent, the emergency classification should be made as if the EAL has been met. While applicable to all ECLs, this approach is particularly important at the higher ECLs 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 Page 11 of 228 INFORMATION USE

to a few minutes). The following guidance should be applied to the classification of these conditions.

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

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

An ATWS occurs and the high pressure ECCS 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 EMG step and clears the inadequate core cooling condition prior to an emergency declaration, then tlie 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 Manager completing the review and steps necessary to make the emergency declaration. This provision is included to ensure that any public protective actions resulting from the emergency classification are truly warranted by the plant conditions.

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

condition no longer exists at the time of discovery. This may be due to the event or condition not being recognized at the time or an error that was made in the emergency classification process.

In these cases, no emergency declaration is warranted; however, the guidance contained in NUREG-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 NRG 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 Wolf Creek USAR Figure 2.1-6 Site Features 4.1.6 Wolf Creek USAR Figure 1.2-44 Site Plan 4.1.7 Technical Specifications Table 1.1-1 Modes 4.1.8 GEN 00-008 RCS Level Less Than Reactor Vessel Flange Operations 4.1.9 NSIR/DPR-ISG-01 Interim Staff Guidance, Emergency Planning for Nuclear Power Plants 4.1.1 O AP 06-002 Wolf Creek Radiological Emergency Response Plan (RERP) 4.1.11 OFN EJ-015 Loss of RHR Cooling 4.1.12 STS GP-006 CTMT Closure Verification 4.2 Implementing 4.2.1 EPP 06-005 Emergency Classification 4.2.2 NEI 99-01 Rev. 6 to Wolf Creek EAL Comparison Matrix 4.2.3 WCGS EAL Matrix Page 13 of 228 INFORMATION USE

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

Selected terms used in IC and EAL statements are set in all capital letters (e.g., ALL CAPS).

These words are defined terms that have specific meanings as used in this document. The definitions of these terms are provided below.

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

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

As applied to WCGS, Containment Closure is established when the requirements of STS GP-006 CTMT Closure Verification are met (ref. 4.1.12).

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

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

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.

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 evidence of heat are observed.

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

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

General Emergency Events are in progress or have occurred which involve actual or imminent substantial core degradation or melting with potential for loss of containment integrity or hostile 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 WCGS 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 WCGS. 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.

Page 15 of 228 INFORMATION USE

lmpede(d)

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

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

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

Owner Controlled Area (OCA)

Property contiguous to the reactor site and acquired by fee, title or easement for WCGS for which public access is limited (ref 4.1.10).

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

• Protected Area (PA)

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 USAR Figure 1.2-44 Site Plan (ref. 4.1.6).

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 lower than 3 feet below the Reactor Vessel flange{< 64.1 in.) with fuel in the vessel (ref. 4.1.8).

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.

• Page 16 of 228 INFORMATION USE

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 10 CFR 50.2):

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

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

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

Site Area Emergency Events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or hostile 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 Exclusion Area Boundary is a synonymous term for Site Boundary. The Exclusion Area is defined as the area that encompasses the land surrounding the Plant to a radius of 1,200 meters (3,937 feet) from the midpoint of the Unit 1 Reactor. The exclusion area boundary location coincides with the restricted area boundary. Control of access to this is by virtue of ownership and in accordance with 10 CFR 100. (ref. 4.1.5).

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

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

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

Page 17 of 228 INFORMATION USE

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 components on two or more SAFETY SYSTEM trains, or one or more structures, that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected SAFETY SYSTEM trains in the area. Events that result in visible damage to the components of one SAFETY SYSTEM tr~in, and do not appear to affect the components of other SAFETY SYSTEM trains, do not m~et the intent of this definition as the failure of a component(s) affecting the operability of one SAFETY SYSTEM train, regardless of cause, is well within the operational controls provided by a licensee's Technical Specifications and Operating

• Procedures. However, visible damage to the components of more than one SAFETY SYSTEM train does meet this definition, as well as visible damage to a structure.

Page 18 of 228 INFORMATION USE

5.2 Abbreviations/Acronyms .

°F ....................................................................................................... Degrees Fahrenheit 0

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • ******************************************************* Degrees AC ....................................................................................................... Alternating Current ATWS ................. :.................................................... Anticipated Transient Without Scram COE ....................................................................................... Committed Dose Equivalent CFR ..................................................................................... Code of Federal Regulations GMT ..............................................................................*................................. Containment CSFST ....................................................................... Critical Safety Function Status Tree OBA ............................................................................................... Design Basis Accident DC ............................................................................................................... Direct Current EAL. ............................................................................................ Emergency Action Level ECCS ............................................................................ Emergency Core Cooling System EGL ................................................................................. Emergency Classification Level EMG .............................................................................. Emergency Operating Procedure EOF .................................................................................. Emergency Operations Facility EPA .............................................................................. Environmental Protection Agency ERG ................................................................................ Emergency Response Guideline EPIP ................................................................ Emergency Plan Implementing Procedure ESF ......................................................................................... Engineered Safety Feature ESW ............................................................................................ Essential Service Water FAA .................................................................................. Federal Aviation Administration FBI ................................................................................... Federal Bureau of Investigation FEMA. .............................................................. Federal Emergency Management Agency GE ..................................................................................................... General Emergency IC ......................................................................................................... Initiating Condition IPEEE ................. Individual Plant Examination of External Events (Generic Letter 88-20)

Kett ......................................................................... Effective Neutron Multiplication Factor LCO ................................................................................. Limiting Condition for Operation LER ............................................................................................... Licensee Event Report LOCA ......................................................................................... Loss of Coolant Accident LWR ................................................................................................... Light Water Reactor MPG ................................... Maximum Permissible Concentration/Multi-Purpose Canister mR, mRem, mrem, mREM .............................................. mi Iii-Roentgen Equivalent Man MSL ........................................................................................................ Main Steam Line MW .................................................................................................................... Megawatt NEI .............................................................................................. Nuclear Energy Institute NESP ................................................................... National Environmental Studies Project NPIS ............................................................................. Nuclear Plant Information System NPP .................................................................................................. Nuclear Power Plant Page 19 of 228 INFORMATION USE

NRC ........................................................... ~ .................... Nuclear Regulatory Commission NSSS ................................................................................ Nuclear Steam Supply System NORAD ............................... ~ ................... North American Aerospace Defense Command (NO)UE ................................... :............................................ Notification of Unusual Event QBE ...................................................................................... Operating Basis Earthquake OCA ...............................................................................................Owner Controlled Area ODCM ............................................................................. Offsite Dose Calculation Manual OFN ....................... :....................................................... Off-Normal Operating Procedure PA .............................................................................................................. Protected Area PAG .............................._.......................................................... Protective Action Guideline PRA/PSA ..................... Probabilistic Risk Assessment I Probabilistic Safety Assessment PWR ....................................................................................... Pressurized Water 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 R(P)V ....................................................................................... Reactor (Pressure) Vessel RTS .................................................................................................. Reactor Trip System RVLIS ................................................................. Reactor Vessel Level Indicating System SBO ......................................................................................................... Station Blackout SCBA ................................................... ;................... Self-Contained Breathing Apparatus SG ......................................................................................................... Steam Generator SI .............................................................................................................. Safety Injection SGTR. ............................ :................................................. Steam Generator Tube Rupture SPDS .................................................................. :........ Safety Parameter Display System SRO ............................................................................................ Senior Reactor Operator SSF .................................................................................................Safe Shutdown Facility TEDE ............................................................................... Total Effective Dose Equivalent TOAF .... :... ;........................................................................................... Top of Active Fuel TSC .......................................................................................... Technical Support Center USAR. ............................................................................. Updated Safety Analysis Report WCGS .................... ~ ......................................................... Wolf Creek Generating Station WCNOC ............................................................. Wolf Creek Nuclear Operating Company WOG ............. :............................ ~ ........................................ Westinghouse Owners Group Page 20 of 228 INFORMATION USE

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

WCGS NEI 99-01 Rev. 6 Example EAL IC EAL RU1.1 AU1 1, 2 RU1.2 AU1 3 RU2.1 AU2 1 RA1.1 AA1 1 RA1.2 AA1 2 RA1.3 AA1 3 RA1.4 AA1 4 RA2.1 AA2 1 RA2.2 AA2 2 RA2.3 AA2 3 RA3.1 AA3 1 RA3.2 AA3 2 RS1.1 AS1 1 RS1.2 AS1 2 RS1.3 AS1 3 RS2.1 AS2 1 RG1.1 AG1 1 RG1.2 AG1 2 RG1.3 AG1 3 RG2.1 AG2 1 CU1.1 CU1 1 Page 21 of 228 INFORMATION USE

WCGS NEI 99-01 Rev. 6 Example EAL IC EAL CU1.2 CU1 2 CU2.1 CU2 1 CU3.1 CU3 1 CU3.2 CU3 2 CU4.1 CU4 1 CU5.1 CU5 1, 2, 3 CA1.1 CA1 1 CA1.2 CA1 2 CA2.1 CA2 1 CA3.1 CA3 1, 2 CA6.1 CA6 1 CS1.1 CS1 1 CS1.2 CS1 2 CS1.3 CS1 3 CG1.1 CG1 1 CG1.2 CG1 2 FA1.1 FA1 1 FS1 .1 FS1 1 FG1.1 FG1 1 HU1.1 HU1 1, 2, 3 HU2.1 HU2 1 HU3.1 HU3 1 HU3.2 HU3 2 HU3.3 HU3 3 HU3.4 HU3 4 Page 22 of 228 INFORMATION USE

WCGS NEI 99-01 Rev. 6 Example EAL IC EAL HU4.1 HU4 1 HU4.2 HU4 2 HU4.3 HU4 3 HU4.4 HU4 4 HU7.1 HU? 1 HA1.1 HA1 1, 2 HA5.1 HA5 1 HA6.1 HA6 1 HA?.1 HA? 1

. HS1.1 HS1 1 HS6.1 HS6 1 HS?.1 HS? 1 HG7.1 HG? 1 SU1.1 SU1 1 SU3.1 SU2 1 SU4.1 SU3 2 SU5.1 SU4 1, 2, 3 SU6.1 SU5 1 SU6.2 SU5 2 SU?.1 SU6 1, 2, 3 SU8.1 SU? 1, 2 SA1.1 SA1 1 SA3.1 SA2 1 SA6.1 SA5 1 SA9.1 SA9 1 Page 23 of 228 INFORMATION USE I*

WCGS NEI 99-01 Rev. 6 Example EAL IC EAL SS1.1 SS1 1 SS2.1 SS8 1 SS6.1 SSS 1 SG1.1 SG1 1 SG1.2 SG8 1 Page 24 of 228 INFORMATION USE

7.0 ATTACHMENTS 7 .1 Attachment 1, EAL Bases 7.2 Attachment 2, Fission Product Barrier Loss/Potential Loss Matrix and Basis 7.3 Attachment 3, Safe Operation & Shutdown Areas Tables R-3 & H-2 Bases Page 25 of 228 INFORMATION USE .. I

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

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

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

conditions may result which require offsite protective actions. Elevated area radiation levels in 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.

Page 26 of 228 INFORMATION USE

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

RU1.1 Unusual Event Reading on any Table R-1 effluent radiation monitor> column "UE" for ~ 60 min.

(Notes 1, 2, 3)

Note 1: The Emergency Manager 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 I Release Point I Monitor I GE I SAE I Alert I UE I

Ill Unit Vent (EFF) O-GT-RE-21 B 4.45E+8 µCi/sec 4.45E+ 7 µCi/sec 4.45E+6 µCi/sec 7.85E+5 µCi/sec

I 0

Cll Ill C'CI

(!) Radwaste Vent (EFF) O-GH-RE-1 OB 4.45E+8 µCi/sec 4.45E+ 7 µCi/sec 4.45E+6 µCi/sec 7.85E+5 µCi/sec SG Slowdown Discharge O-BM-RE-52 ---- ---- ---- 6.91 E-3 µCi/ml Turbine Building Drain O-LE-RE-59 ---- ---- ---- 2.00E-5 µCi/ml "C Waste Water Treatment

• 3 System O-HF-RE-95 ---- ---- ---- 2.09E-3 µCi/ml C"

i Liquid Radwaste Discharge O-HB-RE-18 ---- ---- ---- 1.00E-2 µCi/ml Secondary Liquid Waste System O-HF-RE-45 ---- ---- ---- 1.00E-2 µCi/ml Mode Applicability:

All Page 27 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Definition(s):

None Basis:

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

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 radioacti've release to the environment is indicative of degradation in these features and/or controls.

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

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

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

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

• This EAL addresses radioactivity releases that cause effluent radiation monitor readings to exceed 2 times the limit established by a radioactivity discharge permit. This. EAL will typically also be associated with planned batch releases from non-continuous release pathways (e.g.,

radwaste, waste gas).

Escalation of the emergency classification level would be via IC RA 1.

WCGS Basis Reference(s):

1. AP 078-003 Offsite Dose Calculation Manual

2. USAR Section 7.6, All Other instrumentation Systems Required for Safety

3. EP-CALC-WCNOC-1601 Radiological Effluent EAL Values

4. NEI 99-01 AU1 Page 28 of 228 INFORMATION USE

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

RU1.2 Unusual Event Sample analysis for a gaseous or liquid release indicates a concentration or release rate

> 2 x ODCM limits for;::: 60 min. (Notes 1, 2)

Note 1: The Emergency Manager should declare the event promptly upon d~termining that time limit has been exceeded, or will likely be exceeded.

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

Mode Applicability:

All Definition(s):

None Basis:

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

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

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

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

This EAL addresses uncontrolled gaseous or liquid releases that are detected by sample analyses or environmental surveys, particularly on unmonitored pathways (e.g., spills.of radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.).

Escalation of the emergency classification level would be via IC RA 1.

I Page 29 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases WCGS Basis Reference(s):

1. AP 078-003 Offsite Dose Calculation Manual Section

2. NEI 99-01 AU1 Page 30 of 228 INFORMATION USE

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 COE EAL:

RA1.1 Alert Reading on any Table R-1 effluent radiation monitor> column "ALERT" for ~ 15 min.

(Notes 1, 2, 3, 4)

Note 1: The Emergency Manager 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 radiation 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 rn Unit Vent (EFF) O-GT-RE-21 B 4.45E+8 µCi/sec 4.45E+ 7 µCi/sec 4.45E+6 µCi/sec 7.85E+5 µCi/sec

I 0

GI rn C'CI

(!) Radwaste Vent (EFF) O-GH-RE-108 4.45E+8 µCi/sec 4.45E+7 µCi/sec 4.45E+6 µCi/sec 7.85E+5 µCi/sec SG Slowdown Discharge O-BM-RE-52 ---- --- ---- 6.91 E-3 µCi/ml Turbine Building Drain O-LE-RE-59 ---- ---- ---- 2.00E-5 µCi/ml "1J Waste Water Treatment

• s C'" System O-HF-RE-95 ---- --- -- 2.09E-3 µCi/ml

i Liquid Radwaste Discharge O-HB-RE-18 ---- ---- ---- 1.00E-2 µCi/ml Secondary Liquid Waste System O-HF-RE-45 ---- ---- ---- 1.00E-2 µCi/ml Mode Applicability:

All Page 31 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Definition(s):

None Basis:

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

• 10 mRem TEOE

• 50 mRem COE 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 PAGs (TEOE or COE Thyroid) (ref. 1).

This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 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 TEOE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid COE was established in consideration of the 1:5 ratio of the EPA PAG for TEOE and thyroid COE.

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

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

WCGS Basis Reference(s):

1. EP-CALC-WCNOC-1601 Radiological Effluent EAL Values

2. NEI 99-01 AA 1

.Page 32 of 228 INFORMATION USE

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 COE EAL:*

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

Note 4: The pre-calculated effluent monitor values presented in EALs RA1 .1.; RS1 .1 and RG1 .1 should be used for emergency classification assessments until the results from a dose assessment using actual

• meteorology are available.

Mode Applicability:

All Definition(s):

SITE BOUNDARY - Exclusion Area Boundary is a synonymous term for Site Boundary. The Exclusion Area is defined as the area that encompasses the land surrounding the Plant to a radius of 1,200 meters (3,937 feet) from the midpoint of the Reactor. The exclusion area boundary location coincides with the restricted area boundary. Control of access to this is by virtue of ownership and in accordance with 10CFR100.

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 b.asis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

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

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

WCGS Basis* Reference(s):

1. NEI 99-01 AA 1 Category: R- Abnormal Rad Levels I Rad Effluent Subcategory: 1 - Radiological Effluent Page 33 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid COE EAL:

RA1.3 Alert Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses > 10 mrem TEDE or 50 mrem thyroid COE at or beyond the SITE

• BOUNDARY for 60 min. of exposure (Notes 1, 2)

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

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

Mode Applicability:

All Definition(s):

SITE BOUNDARY - - Exclusion Area Boundary is a synonymous term for Site Boundary. The Exclusion Area is defined as the area that encompasses the land surrounding the Plant to a radius of 1,200 meters (3,937 feet) from the midpoint of the Reactor. The exclusion area boundary location coincides with the restricted area boundary. Control of access to this is by virtue of ownership and in accordance with 10CFR100.

Basis:

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

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

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

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

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

WCGS Basis Reference(s):

1. AP 07B-003 Wolf Creek Offsite Dose Calculation Manual Section

2. NEI 99-01 AA 1 Page 34 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Page 35 of 228 INFORMATION USE

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 COE EAL:

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

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

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

(Notes 1, 2)

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

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

Mode Applicability:

All Definition(s):

SITE BOUNDARY - Exclusion Area Boundary is a synonymous term for Site Boundary. The Exclusion Area is defined as the area that encompasses the land surrounding the Plant to a radius of 1,200 meters (3,937 feet) from the midpoint of the Reactor. The exclusion area boundary location coincides with the restricted area boundary. Control of access to this is by virtue of ownership and in accordance with 10CFR100.

Basis:

EPP 06-011, Team Formation provides guidance for emergency or post-accident radiological environmental monitoring (ref. 1).

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

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

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

J Page 36 of 228 . INFORMATION USE

ATTACHMENT 1 EAL Bases Escalation of the emergency classification level would be via IC RS1.

WCGS Basis Reference(s):

1. EPP 06-011, Team Formation

2. NEI 99-01 AA1

. Page 37 of 228 INFORMATION USE

ATTACHMENT 1 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:

RS1.1 Site Area Emergency Reading on any Table R-1 effluent radiation monitor> column "SAE" for ~ 15 min.

(Notes 1, 2, 3, 4)

Note 1: The Emergency Manager 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 radiation 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 RA 1.1, 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 Ill

s Unit Vent (EFF) O-GT-RE-218 4.45E+8 µCi/sec 4.45E+ 7 µCi/sec 4.45E+6 µCi/sec 7.85E+5 µCi/sec 0

Q)

Ill cG C> Radwaste Vent (EFF) O-GH-RE-1 OB 4.45E+8 µCi/sec 4.45E+7 µCi/sec 4.45E+6 µCi/sec 7.85E+5 µCi/sec SG Slowdown Discharge O-BM-RE-52 ---- ---- ---- 6.91 E-3 µCi/ml Turbine Building Drain O-LE-RE-59 ---- ---- ---- 2.00E-5 µCi/ml Waste Water Treatment

• s'CC' System O-HF-RE-95 ---- ---- ---- 2.09E-3 µCi/ml

i Liquid Radwaste Discharge O-HB-RE-18 ---- ---- ---- 1.00E-2 µCi/ml Secondary Liquid Waste System O-HF-RE-45 ---- ---- ---- 1.00E-2 µCi/ml Mode Applicability:

All Definition(s):

None Page 38 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Basis:

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

• 100 mRem TEOE

• 500 mRem COE Thyroid The column "SAE" gaseous effluent release value in Table R-1 corresponds to calculated doses of 10% of the EPA Protective Action Guidelines {TEOE or COE Thyroid) (ref. 1).

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

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

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

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

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

WCGS Basis Reference(s):

1. EP-CALC-WCNOC-1601 Radiological Effluent EAL Values

2. NEI 99-01 AS1 Page 39 of 228 INFORMATION USE

ATTACHMENT 1 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 COE EAL:

RS1.2 Site Area Emergency Dose assessment using actual meteorology indicates doses > 100 mrem TEDE or 500 mrem thyroid COE at or beyond the SITE BOUNDARY (Note 4)

Note 4: The pre-calculated effluent monitor values presented in EALs RA 1.1, RS1 .1 and RG1 .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Mode Applicability:

All Definition(s):

SITE BOUNDARY - Exclusion Area Boundary is a synonymous term for Site Boundary. The Exclusion Area is defined as the area that encompasses the land surrounding the Plant to a radius of 1,200 meters (3,937 feet) from the midpoint of the Reactor. The exclusion area boundary location coincides with the restricted area boundary. Control of access to this is by virtue of ownership and in accordance with 10CFR100.

Basis:

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

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

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

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

WCGS Basis Reference(s):

1. NEI 99-01 AS1 Page 40 of 228 INFORMATION USE

ATTACHMENT 1 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 COE EAL:

RS1.3 Site Area Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:

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

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

(Notes 1, 2)

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

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

Mode Applicability:

All Definition(s):

SITE BOUNDARY - Exclusion Area Boundary is a synonymous term for Site Boundary. The Exclusion Area is defined as the area that encompasses the land surrounding the Plant to a radius of 1 ,200 meters (3,937 feet) from the midpoint of the Reactor. The exclusion area boundary location coincides with the restricted area boundary. Control of access to this is by virtue of ownership and in accordance with 10CFR100.

Basis:

EPP 06-011, Team Formation provides guidance for emergency or post-accident radiological environmental monitoring (ref. 1).

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

• Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions; The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid COE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid COE.

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

Page 41 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases WCGS Basis Reference(s):

1. EPP 06-011, Team Formation

2. NEI 99-01 AS1 Page 42 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: R -Abnormal Rad Levels I Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid COE EAL:

RG1.1 General Emergency Reading on any Table R-1 effluent radiation monitor> column "GE" for ;::; 15 min.

(Notes 1, 2, 3, 4)

Note 1: The Emergency Manager 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 radiation 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 RA 1.1, 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 I I I I I I I UI

l Unit Vent (EFF) O-GT-RE-21 B 4.45E+8 µCi/sec 4.45E+ 7 µCi/sec 4.45E+6 µCi/sec 7.85E+5 µCi/sec 0

Cl)

UI

(!) Radwaste Vent (EFF) O-GH-RE-1 OB 4.45E+8 µCi/sec 4.45E+7 µCi/sec 4.45E+6 µCi/sec 7.85E+5 µCi/sec SG Slowdown Discharge O-BM-RE-52 ---- ---- ---- 6.91 E-3 µCi/ml Turbine Building Drain O-LE-RE-59 ---- ---- ---- 2.00E-5 µCi/ml "O Waste Water Treatment

• O-HF-RE-95 ---- ---- ---- 2.09E-3 µCi/ml er System

i Liquid Radwaste Discharge O-HB-RE-18 ---- ---- ---- 1.00E-2 µCi/ml Secondary Liquid Waste System O-HF-RE-45 ---- ---- ---- 1.00E-2 µCi/ml Mode Applicability:

All Definition(s):

None Basis:

Page 43 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases This EAL address gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to site boundary doses that exceed either:

• 1000 mRem TEOE

• 5000 mRem COE Thyroid The column "GE" gaseous effluent release values in Table R-1 correspond to calculated doses of 100% of the EPA Protective Action Guidelines (TEOE or COE Thyroid) (ref. 1).

This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Adion 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 TEOE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid COE was established in consideration of the 1:5 ratio of the EPA PAG for TEOE and thyroid COE.

Classification based on effluent monitor reac;jings 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.

WCGS Basis Reference(s):

1. EP-CALC-WCNOC-1601 Radiological Effluent EAL Values

2. NEI 99-01 AG1 Page 44 of 228 INFORMATION USE I*

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels I Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid COE EAL:

RG1.2 General Emergency Dose assessment using actual meteorology indicates doses > 1,000 mrem TEDE or 5,000 mrem thyroid COE at or beyond the SITE BOUNDARY (Note 4)

Note 4: The pre-calculated effluent monitor values presented in EALs RA1 .1, RS1 .1 and RG1 .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Mode Applicability:

All Definition(s):

SITE BOUNDARY - Exclusion Area Boundary is a synonymous term for Site Boundary. The Exclusion Area is defined as the area that encompasses the land surrounding the Plant to a radius of 1,200 meters (3,937 feet) from the midpoint of the Reactor. The exclusion area boundary location coincides with the restricted area boundary. Control of access to this is by virtue of ownership and in accordance with 10CFR100.

Basis:

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

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

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

WCGS Basis Reference(s):

1. NEI 99-01 AG1 Page 45 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels I Rad Effluent Subcategory: 1 -: Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid COE EAL:

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

• Closed window dose rates > 1,000 mR/hr expected to continue for~ 60 min.

• Analyses of field survey samples indicate thyroid COE > 5,000 mrem for 60 min. of inhalation.

(Notes 1, 2)

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

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

Mode Applicability:

All Definition(s):

SITE BOUNDARY - Exclusion Area Boundary is a synonymous term for Site Boundary. The Exclusion Area is defined as the area that encompasses the land surrounding the Plant to a radius of 1,200 meters (3,937 feet) from the midpoint of the Reactor. The exclusion area boundary location coincides with the restricted area boundary. Control of access to this is by virtue of ownership and in accordance with 10CFR100.

Basis:

EPP 06-011, Team Formation provides guidance for emergency or post-accident radiological environmental monitoring (ref. 1). .

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

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

The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid COE was.

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

Page 46 of 228 .INFORMATION USE

ATTACHMENT 1 EAL Bases WCGS Basis Reference(s):

1. EPP 06-011, Team Formation

2. NEI 99-01 AG1 Page 4 7 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels I Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating ConcJition: 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 (EC Ll-39A, EC Ll-39B, EC LIT-39, local observation of SFP level)

AND UNPLANNED rise in corresponding area radiation levels as indicated by any Table R-2 radiation monitors Table R-2 Fuel Building & Containment Area Radiation Monitors Fuel Building:

• SD RE-34, Cask Handling Area Radiation

• SD RE-35, New Fuel Storage Area Radiation

• SD RE-36, New Fuel Storage Area Radiation

• SD RE-37, Fuel Pool Bridge Crane Radiation

• SD RE-38, Spent Fuel Pool Area Radiation Containment:

• SD RE-40, Personnel Access Hatch Area Radiation

• SD RE-41, Manipulator Bridge Crane Radiation

• SD RE-42, Containment Building Radiation

• GT RE-59 Containment High Area Radiation Monitor

• GT RE-60 Containment High Area Radiation Monitor 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.

Page 48 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Basis:

The low water level alarm in this EAL refers to the Spent Fuel Pool (SFP) low level alarm (Window Number 00-076D, SFP LEV HI LO) (ref. 1). During the fuel transfer phase of refueling operations, the fuel transfer canal is normally in communication with the spent fuel pool and the refueling pool in the Containment is in communication with the fuel transfer canal when the fuel transfer tube is open. A lowering in water level in the SFP, fuel transfer canal or refueling pool is therefore sensed by the SFP low level alarm. Neither the refueling pool nor the fuel transfer canal is equipped with a low level alarm (ref. 1). The SFP level is monitored in the Control Room by level indicator EC Ll-39A. The level switch initiates high and low level annunciators.

Technical Specification Section 3. 7 .15 (ref. 2) requires at least 23 ,ft of water above the Spent Fuel Pool storage racks. Technical Specification Section 3.9.7 (ref. 3) requires at least 23 ft of water above the Reactor Vessel flange in the refueling pool. During refueling, this maintains sufficient water level in the fuel transfer canal, refueling pool, and SFP to retain iodine fission product activity in the water in the event of a fuel handling accident.

The Table R-2 radiation monitors are those expected to see increase area radiation levels as a result of a loss of REFUELING PATHWAY inventory (ref. 1, 4). Increasing radiation indications on these monitors in the absence of indications of decreasing REFUELING CAVITY level are not classifiable under this EAL.

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

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

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

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

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

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

WCGS Basis Reference(s):

1. ALR 00-076D SFP LEV HI LO

2. Technical Specification Section 3.7.15 Fuel Storage Pool Water Level

3. Technical Specification Section 3.9.7 Refueling Pool Water Level I . Page 49 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases

4. OFN KE-018 Fuel Handling Accident

5. NEI 99-01 AU2 Page 50 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels I Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL:

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

All Definition(s):

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

Basis:

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

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

This EAL escalates from RU2.1 in that the loss of level, in the affected portion of the REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation (e.g., reports from personnel or camera images), as well as significant changes in water and radiation levels, or other plant parameters. Computational aids may also be used (e.g., a boil-off curve). Classification of an event using this EAL should be based on the totality of available indications, reports and observations.

While an area radiation monitor could detect an increase in a dose rate due to a lowering of water level in some portion of the REFUELING PATHWAY, the reading may not be a reliable indication of whether or not the fuel is actually uncovered. To the degree possible, readings should be considered in combination with other available indications of inventory loss.

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

WCGS Basis Reference(s):

1. ALR 00-0760 SFP LEV HI LO

2. OFN KE-018 Fuel Handling Accident

3. NEI 99-01 AA2 Page 51 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels I Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL:

RA2.2 Alert Mechanical damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by HI HI alarm on any of the following:

• Fuel Building atmosphere monitors (GG RE-27 or 28)

• Containment purge monitors (GT RE-22 or 33)

• Containment atmosphere monitors (GG RE-31 or 32)

• Manipulator bridge crane radiation monitor (SD RE-41)

• Fuel Pool Bridge Crane OR Spent Fuel Pool Area radiation monitor (SD RE-37 or 38)

Mode Applicability:

All Definition(s):

None Basis:

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).

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 HI HI alarm and the SFP and containment area radiation high alarms are a spent fuel handling accident (ref. 1, 2). 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 (ref. 1, 2).

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

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

Page 52 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases WCGS Basis Reference(s):

1. OFN EC-046 Fuel Pool Cooling and Cleanup Malfunctions

2. OFN KE-018 Fuel Handling Accident

3. NEI 99-01 AA2 Page 53 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels I 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 120 in. on EC-Ll-0059 or 0060 (Level 2)

Mode Applicability:

All Definition(s):

None Basis:

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

For WCGS, SFP Level 2 is a reading of 120 in. (plant elevation 2031 ft. 1.25 in.), as indicated on EC-Ll-0059 or EC-Ll-0060 (ref. 3).

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

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 RS1 or RS2.

WCGS Basis Reference(s):

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

2. Drawing EID-0004 Pool Parameters

3. Drawing J-481A-00071 Full Range Level Measurement

4. NEI 99-01 AA2 Page 54 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels I 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 elevation 15 in. on EC-Ll-0059 or 0060 (Level 3)

Mode Applicability:

All Definition(s ):

None Basis:

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

For WCGS, SFP Level 3'has been set at a reading of 15 in. (elevation 2022 ft.4.25 in.) as indicated on EC-Ll-0059 or EC-Ll-0060 (ref. 2, 3).

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

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

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

WCGS Basis Reference(s):

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

2. Drawing EID-0004 Pool Parameters

3. Drawing J-481A-00071 Full Range Level Measurement

4. NEI 99-01 AS2

. Page 55 of 228 INFORMATION USE

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

RG2.1 General Emergency Spent fuel pool level cannot be restored to at least 15 in. (Level 3) on EC-Ll-0059 or 0060 for~ 60 min. (Note 1)

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

Mode Applicability:

All Definition(s):

None Basis:

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

For WCGS, SFP Level 3 has been set at a reading of 15 in. (elevation 2022 ft.4.25 in.) as indicated on EC-Ll-0059 or EC-Ll-0060 (ref. 2, 3).

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

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

WCGS Basis Reference(s):

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

2. Drawing EID-0004 Pool Parameters

3. Drawing J-481A-00071 Full Range Level Measurement

4. NEI 99-01 AG2 Page 56 of 228 INFORMATION USE

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

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

Control Room (SD-RE-33)

OR Central Alarm Station (by survey)

Mode Applicability:

All Definition(s ):

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

Basis:

Areas that meet this threshold include the Control Room and the Central Alarm Station (CAS).

SD-RE-33 monitors the Control Room for area radiation (ref. 1). The CAS is included in this EAL because of its' importance to permitting access to areas required to assure safe plant operations.

There is no permanently installed CAS area radiation monitors that may be used to assess this EAL threshold. Therefore this threshold must be assessed via local radiation survey for the CAS (ref. 1).

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

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

WCGS Basis Reference(s):

1. USAR Section 12.3 Table 12.3-2 Area Radiation Monitors*

2. NEI 99-01 AA3 Page 57 of 228 INFORMATION USE

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

RA3.2 Alert An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to any Table R-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 North Electrical Pen. Room A 3,4 South Electrical Pen. Room B 3,4 ESF SWGR Room No. 1 (TRN A) 4 ESF SWGR Room No. 2 (TRN B) 4 Auxiliary Building/West Hall Elev 2000 3,4,5 Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

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

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

Basis:

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

An Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally I Page 58 of 228 INFORMATION USE j

ATTACHMENT 1 EAL Bases 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 5 when the radiation increase occurs, and the procedures used for normal operation, cooldown and shutdown only require entry into the affected room in Modes 1-4.

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

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

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

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

NOTE: EAL RA3.2 mode applicability has been limited to the applicable modes identified in Table R-3 Safe Operation & Shutdown Rooms/Areas. If due to plant operating procedure or plant configuration changes, the applicable plant modes specified in Table R-3 are changed, a corresponding change to Attachment 3 'Safe Operation &

Shutdown Areas Tables R-3 & H-2 Bases' and to EAL RA3.2 mode applicability is required.

WCGS Basis Reference(s):

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

2. NEI 99-01 AA3 Page 59 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Category C - Cold Shutdown I Refueling System Malfunction EAL Group: Cold Conditions (RCS temperature s 200°F); EALs in this category are applicable only in one or more cold operating modes.

Category C EALs are directly associated with cold shutdown or refueling system safety functions. Given the variability of plant configurations (e.g., systems out-of-service for maintenance, containment open, reduced AC power redundancy, time since shutdown) during these periods, the consequences of any given initiating event can vary greatly. For example, a loss of decay heat removal capability that occurs at the end of an extended outage has less significance than a similar loss occurring during the first week after shutdown. Compounding these events is the likelihood that instrumentation necessary for assessment may also be inoperable. The cold shutdown and refueling system malfunction EALs are based on performance capability to the extent possible with consideration given to RCS integrity, CONTAINMENT CLOSURE, and fuel clad integrity for the applicable operating modes (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 Emergency AC Power Loss of emergency plant electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of onsite and offsite power sources for 4.16KV 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.

Page 60 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases

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

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

Page 61 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases.

Category: C - Cold Shutdown I Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: UNPLANNED loss of RCS inventory EAL:

CU1.1 Unusual Event UNPLANNED loss of reactor coolant results in RCS water level less than a required lower limit for~ 15 min. (Note 1)

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

Mode Applicability:

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

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

Basis:

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 (100.1 in.) (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).

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 RCS 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 recognizes that the minimum required RCS level can change several times during the course of a refueling outage as different plant configurations and system lineups are implemented. This EAL is met if the minimum level, specified for the current plant conditions, cannot be maintained for 15 minutes or longer. The minimum level is typically specified in the applicable operating procedure but may be specified in another controlling document.

Page 62 of 228 INFORMATION USE.

ATTACHMENT 1 EAL Bases The 15-minute threshold duration allows sufficient time for prompt operator actions to restore and maintain the expected water level. This criterion excludes transient conditions causing a brief lowering of water level.

Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA 1 or CA3.

WCGS Basis Reference(s):

1. FR-12 Response to Low Pressurizer Level

2. Technical Specification Section 3.9.7 Refueling Pool Water Level

3. Gen 00-008 RCS Level Less Than Reactor Vessel Flange Operations

4. NEI 99-01 CU1 ..

Page 63 of 228 INFORMATION USE

_j

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

CU1.2 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-1 Sumps I Tanks

• Containment Normal Sump

• PRT

• RCDT

• Auxiliary Building Sump

• Liquid Waste Holdup Tank

• Recycle Holdup Tank

• CCW Surge Tank Mode Applicability:

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

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

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

Basis:

In Cold Shutdown mode, the 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. 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 Page 64 of 228 INFORMATION USE

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

This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor RCS level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant.

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

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

This EAL addresses a condition where all means to determine RCS level have been lost. In this condition, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RCS.

Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA 1 or CA3.

WCGS Basis Reference(s):

1. *OFN BB-007 RCS Leakage High

2. STS BB-004 RCS Water Inventory Balance

3. NEI 99-01 CU1 Page 65 of 228 INFORMATION USE I*

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown I 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 < 12 in.

Mode Applicability:

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

None Basis:

A RCS level of 12 inches as measured by BB Ll-53(54)A and/or BB Ll-53(54)B is indicative of a loss of level that is well below the desired RCS water level between 20 and 22 inches for RCS fill _and also below the desired level of 15 to 17 inches for RCS vacuum fill (ref. 1).

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

For this EAL, a lowering of water level below 12. in. indicates that operator actions have not been successful in restoring and maintaining RCS water level. The heat-up rate of the coolant will increase as the available water inventory is reduced. A continuing decrease in water level will lead to core uncovery.

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

If the RCS inventory level continues to lower, then escalation to Site Area Emergency would be via IC CS1.

WCGS Basis Reference(s):

1. GEN 00-008 RCS Level Less Than Reactor Vessel Flange Operations Figure 1 RCS Level Versus RHR Flow

2. NEI 99-01 CA 1 Page 66 of 228 INFORMATION USE

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

CA1.2 Alert RCS water level cannot be monitored for~ 15 min. (Note 1)

AND EITHER

• UNPLANNED increase in any Table C-1 Sump I Tank level.

• Visual observation of UNISOLABLE RCS leakage Note 1: The Emergency Manager should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Table C-1 Sumps I Tanks

• Containment Normal Sump

• PRT

• RCDT

• Auxiliary Building Sump

• Liquid Waste Holdup Tank

.* Recycle Holdup Tank

• CCW Surge Tank Mode Applicability:

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

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

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

Basis:

In Cold Shutdown mode, the 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 67 of 228 INFORMATION USE.

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).

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

For this EAL, the inability to monitor RCS level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RCS.

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 RCS) inventory level continues to lower, then escalation to Site Area Emergency would be via IC CS1.

WCGS Basis Reference(s):

1. OFN BB-007 RCS Leakage High

2. STS BB-004 RCS Water Inventory Balance

3. NEI 99-01 CA 1 Page 68 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown I 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, RVLIS natural circulation range< 72%

Mode Applicability:

5 - Cold Shutdown 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.

As applied to WCGS, Containment Closure is established when the requirements of STS GP-006 CTMT Closure Verification are met.

Basis:

When Reactor Vessel water level lowers to 72%, water level is six inches below the elevation of the bottom of the RCS hot leg penetration. Six inches below the elevation of the bottom of the RCS hot leg penetration can be monitored only by RVLIS with RVLIS in service (ref. 1).

This EAL addresses a significant and prolonged loss of RCS inventory control and makeup capability leading to imminent fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant.

These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.

Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel 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 CS1 .1 and CS1 .2 reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment.

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

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

Page 69 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases WCGS Basis Reference(s):

1. WCOP-24 EMG/OFN Setpoints - Setpoint F.36 and F.37

2. NEI 99-01 CS1 Page 70 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown I 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, RVLIS natural circulation range< 66%

Mode Applicability:

5- Cold Shutdown 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.

As applied to WCGS, Containment Closure is established when the requirements of STS GP-006 CTMT Closure Verification are met.

Basis:

When Reactor Vessel water level lowers to 66%, core uncover is about to occur. The top of the reactor fuel can be monitored only by RVLIS with RVLIS in service (ref. 1).

This EAL addresses a significant and prolonged loss of RCS inventory control and makeup capability leading to imminent fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant.

These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.

Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel 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 CS1 .1 and CS1 .2 reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment.

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

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

WCGS Basis Reference(s):

Page 71 of 228 INFORMATION us~

ATTACHMENT 1 EAL Bases

1. WCOP-24 EMG/OFN Setpoints - Setpoint F.37

2. NEI 99-01 CS1 Page 72 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown I 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 ~ 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

• Manipulator bridge crane radiation monitor SD RE-41 Hi-Hi alarm

• Erratic Source Range Monitor indication Note 1: The Emergency Manager should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Table C-1 Sumps I Tanks

• Containment Normal Sump

• PRT

• RCDT

• Auxiliary Building Sump

• Liquid Waste Holdup Tank

• Recycle Holdup Tank

• CCW Surge Tank Mode Applicability:

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

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

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

Basis:

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

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

Page 73 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases 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 Reactor Vessel inventory loss may be detected by the manipulator bridge crane radiation monitor 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 up-scaled (Hi-Hi alarm) manipulator crane radiation monitor (SD-RE-41) indication (ref.3).

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 ).

This EAL addresses a significant and prolonged loss of RCS inventory control and makeup capability leading to imminent fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant.

These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.

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

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

• control/makeup equipment and/or restore level monitoring.

• The inability to monitor RCS level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RCS.

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

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

WCGS Basis Reference(s):

Page 74 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases

1. OFN BB-007 RCS Leakage High

2. STS BB-004 RCS Water Inventory Balance

3. USAR Table 12.3-2 Area Radiation Monitors

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

5. NEI 99-01 CS1 Page 75 of 228 INFORMATION USE

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

CG1 .1 General Emergency RVLIS natural circulation range < 66% for ~ 30 min. (Note 1)

AND Any Containment Challenge indication, Table C-2 Note 1: The Emergency Manager 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 ~ 4%

• UNPLANNED rise in Containment pressure Mode Applicability:

5 - Cold Shutdown Definition(s):

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

As applied to WCGS, Containment Closure is established when the requirements of STS GP-006 CTMT Closure Verification are met.

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

Basis:

When Reactor Vessel water level lowers to 66%, core uncover is about to occur. The top of

- the reactor fuel can be monitored only by RVLIS with RVLIS in service (ref. 1):

This EAL 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 ..

Page 76 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.

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

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.

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

WCGS Basis Reference(s):

1. WCOP-24 EMG/OFN Setpoints - Setpoint F.37

2. FSAR Section 6.2.5 Combustible Gas Control In Containment

3. NEI 99-01 CG1 Page 77 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown I 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 ;:::: 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 *

• Manipulator bridge crane radiation monitor SD RE-41 Hi-Hi alarm

• Erratic Source Range Monitor indication AND Any Containment Challenge indication, Table C-2 Note 1: The Emergency Manager 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 I Tanks

• Containment Normal Sump

• PRT

• RCDT

• Auxiliary Building Sump

• Liquid Waste Holdup Tank

• Recycle Holdup Tank

• CCW Surge Tank Table C-2 Containment Challenge Indications

• CONTAINMENT CLOSURE not established (Note 6)

• Containment hydrogen concentration;:::: 4%

• UNPLANNED rise in Containment pressure Page 78 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Mode Applicability:

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

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

As applied to WCGS, Containment Closure is established when the requirements of STS GP-006 CTMT Closure Verification are met.

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

Basis:

In Cold Shutdown mode, the RCS will normally be intact and standard RCS level monitoring 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 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 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 balance (ref. 1, 2).

The Reactor Vessel inventory loss may be detected by the manipulator bridge crane radiation monitor 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 up-scaled (Hi-Hi alarm) manipulator crane radiation monitor (SD-RE-41) indication (ref.3).

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 ).

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.

2. Containment hydrogen ~ 4% - The 4% hydrogen concentration threshold is generally considered the lower limit for hydrogen deflagrations. WCGS is equipped with a Hydrogen Control System (HCS) which serves to limit or reduce combustible gas concentrations in the Containment. The HCS is an engineered safety feature with redundant hydrogen recombiners, hydrogen mixing system, hydrogen monitoring subsystem, and a backup hydrogen purge subsystem. The HCS is designed to maintain the Containment hydrogen concentration below 4% by volume (ref.5). Two Containment Page 79 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases hydrogen monitors (GS Al-10 and GS Al-19) with a range of 0% to 10% provide indication on Control Room Panel CL020 and NPIS (ref.6).

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.

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

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

Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel level cannotbe 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.

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 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 Page 80 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases in sump and/or tank levels. Sump and/or tank level. changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RCS, These EALs address concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.

WCGS Basis Reference(s): *

1. OFN BB-007 RCS Leakage High

2. STS BB-004 RCS Water Inventory Balance

3. USAR Table 12.3-2 Area Radiation Monitors

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

5. FSAR Section 6.2.5 Combustible Gas Control In Containment

6. FSAR Table 7A-3 (Sheet 6.4)

7. NEI 99-01 CG1 Page 81 of 228 INFORMATION USE

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

CU2.1 Unusual Event AC power capability, Table C-3, to emergency 4.16KV buses NB01 and NB02 reduced to a single power source for~ 15 min. (Note 1)

AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS Note 1: The Emergency Manager 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:

• ESF XFMR XNB01

• ESF XFMR XNB02 Onsite:

• EOG NE01

• EOG NE02

• SBO OGs (if already aligned)

Mode Applicability:

5 - Cold Shutdown, 6 - Refueling, 0 - Oefueled 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 82 of 228 INFORMATION LI.SE

ATTACHMENT 1 EAL Bases Basis:

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.

4.16KV buses NB01 and NB02 are the emergency (essential) buses. NB01 supplies power to Load Group 1 (Red Train) safety related loads and NB02 supplies power to Load Group 2 (Yellow Train) safety related loads. Each bus has two sources of offsite power. One source is from 4.16KV ESF transformer XNB01 and the other source is from the ESF transformer XNB02. Transformer XNB01 is the normal supply to bus NB01 and the alternate supply to bus NB02; XNB02 is the normal supply to bus NB02 and the alternate supply to bus NB01 (ref. 1, 2).

In addition, NB01 and NB02 each have an emergency diesel generator which supply electrical power to the bus automatically in the event that the preferred source becomes unavailable (ref.

2).

An additional source of power are the SBO diesel generators SBO DGs (ref. 3, 4). Credit can be taken for this source only if they are already aligned because they cannot be aligned within 15 minutes.

This cold condition EAL is equivalent to the hot condition EAL SA 1.1.

If the capability of a second source of emergency bus power is not restored to at least one emergency bus within 15 minutes, an Unusual Event is declared under this EAL.

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

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 OFNs and EMGs, 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.

Page 83 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases The subsequent loss of the remaining single power source would escalate the event to an Alert in accordance with IC CA2.

WCGS Basis Reference(s):

1. USAR figure 8.2-5 Electrical one line diagram of Wolf Creek 345 KV switchyard

2. USAR Section 8.3

3. OFN NB-030 Loss of AC Emergency Bus NB01 (NB02)

4. SYS KU-121 (122) Energizing NB01 (NB02) From Station Blackout Diesel Generators

5. NEI 99-01 CU2 Page 84 of 228 INFORMATION USE

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

CA2.1 Alert Loss of all offsite and all onsite AC power to emergency 4.16KV buses NB01 and NB02 for~ 15 min. (Note 1)

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

Mode Applicability:

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

None Basis:

The emergency 4.16KV AC System provides the power requirements for operation and safe shutdown of the plant. The essential switchgear are buses NB01 and NB02 (ref. 1).

4.16KV buses NB01 and NB02 are the emergency (essential) buses. NB01 supplies power to Load Group 1 (Red Train) safety related loads and NB02 supplies power to Load Group 2 (Yellow Train) safety related loads. Each bus has two sources of offsite power. One source is from 4.16KV ESF transformer XNB01 and the other source is from the ESF transformer XNB02. Transformer XNB01 is the normal supply to bus NB01 and the alternate supply to bus NB02; XNB02 is the normal supply to bus NB02 and the alternate supply to bus NB01 (ref. 1, 2).

In addition, NB01 and NB02 each have an emergency diesel generator which supply electrical power to the bus automatically in the event that the preferred source becomes unavailable (ref.

2).

An additional source of power are the SBO diesel generators SBO DGs (ref. 3, 4 ). Credit can be taken for this source only if they are already aligned because they cannot be aligned within 15 minutes.

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

The interval begins when both offsite and onsite AC power capability are lost.

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

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 Page 85 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases 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 RS1.

WCGS Basis Reference(s}:

1. USAR figure 8.2-5 Electrical one line diagram of Wolf Creek 345 KV switchyard

2. USAR Section 8.3

3. OFN NB-030 Loss of AC Emergency Bus NB01 (NB02)

4. SYS KU-121(122) Energizing NB01(NB02) From Station Blackout Diesel Generators

5. NEI 99-01 CA2 Page 86 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown I 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 > 200°F (Note 10)

Note 10: Begin monitoring hot condition EALs concurrently for any new event or condition not related to the loss of decay heat removal.

Mode Applicability:

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

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

Basis:

Several instruments are capable of providing indication of RCS temperature with respect to the Technical Specification cold shutdown temperature limit (200°F, ref. 1). These include core exit thermocouples (T/Cs) and Wide Range hot leg temperature indications. Plant computer screens are available for monitoring heatup and cooldown. 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).

In the absence of reliable RCS temperature indication caused by a loss of decay heat removal capability, classification should be based on EAL CU3.2 should RCS level indication be subsequently lost.

This EAL addresses an unplanned increase in RCS temperature above the Technical Specification cold shutdown temperature limit and represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Manager should also refer to IC CA3.

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

This EAL involves a loss of decay heat removal capability, or an addition* of heat to the 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 above the reactor vessel flange. Refueling evolutions that lower water level below the reactor vessel flange are Page 87 of 228 INFORMATl_ON USE .I*

ATTACHMENT 1 EAL Bases 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.

Escalation to Alert would be via IC CA 1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.

WCGS Basis Reference(s):

1. Wolf Creek Technical Specifications Table 1.1-1

2. GEN 00-002 Cold Shutdown to Hot Standby

3. USAR Section 7.2.2.3.2

4. NEI 99-01 CU3 Page 88 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown I 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;::: 15 min. (Note 1)

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

Mode Applicability:

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

None Basis:

In cold operating modes RCS water level is normally monitored using the following instruments (ref.2):

• Ll-462, Pressurizer Cold Calibrated Level

• RCS Loop level indications):

• Mid-loop level indicators on RL018:

BB Ll-53A, RCS LEVEL LOOP 4 WR MIDLOOP BB Ll-53B, RCS LEVEL LOOP 4 NR BB Ll-54A, RCS LEVEL LOOP 1 WR MIDLOOP BB Ll-54B, RCS LEVEL LOOP 1 NR

• Computer points:

BBL0053A, RCS LEVEL LOOP 4 WR MIDLOOP BBL0053B, RCS LEVEL LOOP 4 NR BBL0054A, RCS LEVEL LOOP 1 WR MIDLOOP BBL0054B, RCS LEVEL LOOP 1 NR.

• Tygon Hose

• Visual observation (if vessel head is removed)

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 core exit thermocouples (T/Cs) and WideRange hot leg*temperature indications. (ref. 3).

This EAL addresses the inability to determine RCS temperature and level, and represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Manager should also refer to IC CA3.

Page 89 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases This EAL reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor RCS conditions and operators would be unable to monitor key parameters necessary to assure core decay heat removal. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation.

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

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

WCGS Basis Reference(s,):

1. Wolf Creek Technical Specifications Table 1.1-1

2. SYS BB-215 RCS Drain Down with Fuel in Reactor

3. FSAR Section 7.2.2.3.2

4. NEI 99-01 CU3 Page 90 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown I 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> 200°F for> Table C-4 duration (Notes 1, 10)

OR UNPLANNED RCS pressure increase > 10 psig (This EAL does riot apply during water-solid plant conditions)

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

Note 10: Begin monitoring hot condition EALs concurrently for any new event or condition not related to the loss of decay heat removal.

Table C-4: RCS Heat-up Duration Thresholds CONTAINMENT RCS Status Heat-up Duration CLOSURE Status 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 conditions or actions taken to secure Primary or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.

As applied to WCGS, Containment Closure is established when the requirements of STS GP-006 CTMT Closure Verification are met.

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

Page 91 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases REDUCED INVENTORY-. Plant condition when fuel is in the reactor vessel and Reactor Coolant System level is lower than 3 feet below the Reactor Vessel flange(< 64.1 in.).

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 core exit thermocouples (T/Cs) and Wide Range hot leg temperature indications. (ref. 2).

RCS pressure instrument BB Pl-403 and BB Pl-405 are capable of measuring pressure to less than 10 psig (ref. 3).

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.

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

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

The RCS Heat-up Duration Thresholds table addresses an increase in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact, or RCS inventory is reduced (e.g., mid-loop operation in 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-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, and CONTAINMENT CLOSURE is not established, no heat-up duration is allowed (i.e., 0 minutes). This is because 1) the evaporated reactor coolant may be released directly into the Containment atmosphere and subsequently to the environment, and

2) there is reduced reactor coolant inventory above the top of irradiated fuel.

The second condition provides a pressure-based indication of RCS heat-up.

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

WCGS Basis Reference(s):

1. Wolf Creek Technical Specifications Table 1.1-1 Page 92 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases

2. FSAR Section 7.2.2.3.2

3. GEN 00-006 Hot Standby to Cold Shutdown

4. NEI 99-01 CA3 Category: C - Cold Shutdown I Refueling System Malfunction Subcategory: 4 - Loss of Vital DC Power Initiating Condition: Loss of Vital DC power for 15 minutes or longer*

EAL:

CU4.1 Unusual Event

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

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

Mode Applicability:

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

None Basis:

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 vital DC buses are the following 125 VDC Class 1E buses (ref. 1):

Divisi~n 1 (Train A): Division 2 (Train B):

• NK01

• NK02

• NK03

• NK04 There are four, 60 cell, lead-calcium storage batteries (NK11, NK12, NK13 and NK14) that supplement the. output of the battery chargers. They supply DC power to the distribution buses when AC power to the chargers is lost or when transient loads exceed the 300 amp capacity of the battery chargers.

Each battery is designed to have sufficient stored energy to supply the required emergency loads for 240 minutes following a loss of AC power (station blackout) (ref. 2, 3).

Minimum DC bus voltage is 105 VDC (ref. 4 ).

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

This IC addresses a loss of vital DC power which compromises the ability to monitor and control operable Safety Systems when the plant is in the cold shutdown or refueling mode. In these modes, the core decay heat load has been significantly reduced, and coolant system Page 93 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases 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 Bis in-service (operable), then a loss of Vital DC power affecting Train B would require the declaration of an Unusual Event. A loss of Vital DC power to Train A would not warrant an emergency classification.

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

Depending upon the event, escalation of the emergency classification level would be via IC CA 1 or CA3, or an IC in Recognition Category R.

WCGS Basis Reference(s):

1. OFN NK-020 Loss of Vital 125 VDC Bus NK01, NK02, NK03, and NK04

2. FSAR Tables 8.3-2, -3

3. FSAR Section 8.3.2 DC Power Systems

4. Calculation NK-E-001 125 VDC Class 1 E Battery System Sizing, Voltage Drop and Short Circuit Studies

5. NEI 99-01 CU4 Page 94 of 228 INFORMATION USE 1*

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown I Refueling System Malfunction Subcategory: 5 - Loss of Communications Initiating Condition: Loss of all onsite or offsite communications capabilities EAL:

CUS.1 Unusual Event Loss of all Table C-5 onsite communication methods OR Loss of all Table C-5 offsite communication methods OR Loss of all Table C-5 NRC communication methods Table C-5 Communication Methods System Onsite Offsite NRC PA system x Plant Radios x x Site Telephone System x x x Local Telephone Company Direct Lines x x x ENS Line x Mode Applicability:

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

None Basis:

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

1. Public Address (PA) system The system provides five separate independent Communication lines and one general page channel. Communication between parties in the plant and the Control Room can be easily and quickly established using the Control Room page .channel (channel 1).

Communication between. parties within the plant can be easily and quickly established by Page 95 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases using the general page channel. The party line channel is normally used after the page call is completed. As many as five party lines may communicate simultaneously.

2. Plant Radios The Plant Radio System consists of two repeater sites: the Turbine Building and the Meteorological Tower. The sites have multiple repeaters to support communications inside the Plant structures as well as in a 5 mile radius of the Plant. The system provides two-way portable and mobile communications for normal and emergency Plant operation. Portable and mobile users have communications capability with fixed operator positions in the Control Room, TSC and EOF, and with security radio consoles, if desired.

3. Site Telephone System The Touchtone Telephone System consists of telephone stations located throughout the Power Block, in the main Control Room, Security Building, Administration Building and various other buildings around the site. The system has diverse routing to provide multiple routes for both inward and outward dialing. The telephone system is powered through a battery backup system, which can provide about 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of service after loss of offsite power.

4. Local Telephone Company Direct Lines In the event of a total system failure there are multiple direct telephone lines from the local telephone company which remain operational for access to the local Burlington exchange.

5. ENS line The NRC Emergency Notification System (ENS) is a Federal Telephone System (FTS) telephone used for official communications with NRC Headquarters. The NRC Headquarters has the capability to patch into the NRC Regional offices. The primary purpose of this phone is to provide a reliable method for the initial notification of the NRC and to maintain continuous communications with the NRC after initial notification. ENS telephones are located in the Control Room, TSC and EOF.

This EAL is the cold condition equivalent of the hot condition EAL SU? .1.

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

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

• The first condition addresses a total loss of the communications methods used in support of routine plant operations.

The second condition addresses a total loss of the communications methods used to notify all offsite organizations of an emergency declaration. The offsite organizations referred to here are the State and Coffey County EOCs. *

• Page 96 of 228 INFORMATION USE . , .

ATTACHMENT 1 EAL Bases The third condition addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.

WCGS Basis Reference(s):

1. Wolf Creek Plant Radiological Emergency Response Plan (RERP), Section 6.16.1

2. USAR Section 9.5.2

3. NEI 99-01 CU5 Page 97 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown I Refueling System Malfunction Subcategory: 6 - Hazardous Event Affecting Safety Systems Initiating Condition: Hazardous event affecting SAFETY SYSTEMS needed for the current operating mode EAL:

CA6.1 Alert The occurrence of any Table C-6 hazardous event AND Event damage has caused indications of degraded performance onone train of a SAFETY SYSTEM needed for the current operating mode AND EITHER:

• Event damage has caused indications of degraded performance to a second train of a SAFETY SYSTEM needed for the current operating mode

• Event damage has resulted in VISIBLE DAMAGE to a second train of a SAFETY SYSTEM needed for the current operating mode (Notes 11, 12)

Note 11: If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then emergency classification is not warranted.

Note 12: If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted.

Table C-6 Hazardous Events

• Seismic event (earthquake)

• Internal or external FLOODING event

• High winds or tornado strike

• FIRE

• EXPLOSION

• Other events with similar hazard characteristics as determined by the Emergency Manager Mode Applicability:

5 - Cold Shutdown, 6 - Refueling Page 98 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Definition(s):

EXPLOSION-A rapid, violent and catastrophic failure of a piece ofequipment 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 evidence of heat are observed.

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

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

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

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

VISIBLE DAMAGE - Damage to a SAFETY SYSTEM train that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected SAFETY SYSTEM train ..

Basis:

This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the current operating mode. In order to provide the appropriate context for consideration of an ALERT classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train, and there must be either indications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance, and the second .

SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE such that the potential exists for performance issues. Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has indications of degraded performance; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events.

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

Page 99 of 228 INFORMATION USE

~-

ATTACHMENT 1 EAL Bases VISIBLE DAMAGE addresses damage to a SAFETY SYSTEM train that is not in service/operation and that potentially could cause performance issues. Operators will make this determination based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. This VISIBLE DAMAGE should be significant enough to cause concern regarding the .

operability or reliability of the SAFETY SYSTEM train.

Escalation of the emergency classification level would be via IC FS1 or RS1.

WCGS Basis Reference(s):

1. NEI 99-01 CA6 Page 100 of 228 INFORMATION USE

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, credible 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 Technology 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

• s. Hazardous Gases 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 101 of 228 INFORMATION USE I*

ATTACHMENT 1 EAL Bases

7. Emergency Manager Judgment The EALs defined in other categories specify the predetermined symptoms or events that are indicative of emergency or potential emergency conditions and thus warrant classification. While these EALs have been developed to address the full spectrum of possible emergency conditions which may warrant classification and subsequent implementation of the Emergency Plan, a provision for classification of emergencies based on operator/management experience and judgment is still necessary. The EALs of this category provide the Emergency Manager the latitude to classify emergency conditions consistent with the established classification criteria based upon Emergency Manager judgment.

• ~1~~~~~~~~-----'-,-~-P_a_ge~10_2_o_f_2_2_8~~~~~~-'-N_FO~R_M_A_T_IO_N~U_S_E__J

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

HU1.1 Unusual Event A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the Security Shift Lieutenant OR Notification of a credible security threat directed at the site OR A validated notification from the NRG providing information of an aircraft threat Mode Applicability:

All Definition(s):

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

HOSTILE ACTION-An act toward WCGS or its personnel that includes the use of violent force to destroy equipment, take hostages, and/or intimidate the licensee to ach_ieve 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 WCGS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area).

Basis:

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

This EAL is based on the Wolf Creek Generating Station Security Plan (ref. 1).

This IC addresses events that pose a threat to plant personnel or Safety System equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR § 73.71 or 10 CFR § 50.72. Security events assessed as Hostile Actions are classifiable under ICs HA1 and HS1. .

Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event (ref. 2, 3). Classification of these events will initiate appropriate threat-related notifications to plant personnel and Offsite Response Organizations.

Page 103 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan.

The first threshold references the Shift Security Lieutenant because these are the individuals trained to confirm that a security event is occurring or has occurred. Training on security event confirmation and classification is controlled due to the nature of Safeguards and 10 CFR § 2.39 information.

The second threshold addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with the Wolf Creek Generating Station Security Plan.

• The third threshold addresses the threat from the impact of an aircraft on the plant. The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may also be provided by NORAD through the NRC. Validation of the threat is performed in accordance with the Wolf Creek Generating Station Security Plan (ref. 1).

Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Wolf Creek Generating Station Security Plan (ref. 1).

Escalation of the emergency classification level would be via IC HA 1.

WCGS Basis Reference(s):

1. Wolf Creek Generating Station Security Plan (Safeguards)

2. OFN SK-039 Security Event

3. OFN 00-036 Bomb Threat, Sabotage, Medical Emergency/Rescue, and Spills

4. NEI 99-01 HU1 Page 104 of 228 INFORMATION USE

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:

HA1.1 Alert A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the Security Shift Lieutenant OR A validated notification from NRC of an aircraft attack threat within 30 min. of the site Mode Applicability:

All Definition(s):

HOSTILE ACTION - An act toward WCGS 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 WCGS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area).

OWNER CONTROLLED AREA - Area outside the PROTECTED AREA fence that immediately surrounds the plant. Access to this area is generally restricted to those entering on official business.

Basis:

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

This IC addresses the occurrence of a Hostile Action within the Owner Controlled Area or notification of an aircraft attack threat. This event will require rapid response and assistance due to the possibility of the attack progressing to the Protected Area, or the need to prepare the- plant and staff for a potential aircraft impact.

Timely and accurate communications between the Shift Security Lieutenant and the Control Room is essential for proper classification of a security-related event (ref. 2, 3).

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

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, allowing them to be better prepared should it be necessary to consider further actions.

• Page 105 of 228 INFORMATION USE

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.

The first threshold is applicable for any Hostile Action occurring, or that has occurred, in the Owner Controlled Area.

The second threshold addresses the threat from the impact of an aircraft on the plant, and the anticipated arrival time is within 30 minutes. The intent of this EAL is to ensure that threat-related notifications are made in a timely manner so that plant personnel and offsite organizations are in a heightened state of readiness. This EAL is met when the threat-related information has been validated in accordance with OFN SK-039 Security Event (ref. 2).

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 Wolf Creek Generating Station Security Plan (ref. 1).

WCGS Basis Reference(s):

1. Wolf Creek Generating Station Security Plan (Safeguards)

2. OFN SK-039 Security Event

3. OFN 00-036 Bomb Threat, Sabotage, Medical ~mergency/Rescue, and Spills

4. NEI 99-01 HA 1 Page 106 of 228 INFORMATION USE

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

HS1 .1 Site Area Emergency A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the Security Shift Lieutenant Mode Applicability:

All Definition(s):

HOSTILE ACTION - An act toward WCGS 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 WCGS. 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 USAR Figure 1.2-44 Site Plan.

Basis:

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

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 Wolf Creek Plant Security Plan (Safeguards) information. (ref. 1)

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

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

Timely and accurate communications between the Security Shift Lieutenant and the Control Room is essential for proper classificat\on of a security-related event (ref. 2, 3).

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

As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Site Area Emergency declaration will mobilize state and county 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.

J Page 107 of 228 INFORMATION USE

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.

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 Wolf Creek Generating Station Security Plan (ref. 1).

WCGS Basis Reference(s):

1. Wolf'Creek Generating Station Security Plan (Safeguards)

2. OFN SK-039 Security Event

3. OFN 00-036 Bomb Threat, Sabotage, Medical Emergency/Rescue, and Spills

4. NEI 99-01 HS1 Page 108 of 228 INFORMATION USE

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 > OBE as indicated by Seismic Activity Annunciator 00-0980 Mode Applicability:

All Definition(s):

None Basis:

Ground motion acceleration of 0.06 g horizontal or .04 g vertical is the Operating Basis Earthquake for WCGS(ref. 1).

Annunciator 00-0980, OBE will illuminate if the seismic instrument detects ground motion in excess of the OBE threshold (ref. 2).

OFN SG-003, Natural Events provides the guidance for determining if the OBE earthquake threshold is exceeded and any required response actions. (ref. 3)

To avoid inappropriate emergency classification resulting from spurious actuation of the seismic instrumentation or felt motion not attributable to seismic activity, an offsite agency (USGS, National Earthquake Information Center) can confirm that an earthquake has occurred .

in the area of the plant. Such confirmation should not, however, preclude a timely emergency declaration based on receipt of the OBE alarm. The NEIC can be contacted by calling (303) 273-8500. Select option #1 and inform the analyst you wish to confirm recent seismic activity in the vicinity of WCGS. Alternatively, near real-time seismic activity can be accessed via the NEIC website:

http://earthquake.usgs.gov/eqcenter/

This IC addresses a seismic event that results in accelerations at the plant site greater than those specified for an Operating Basis Earthquake (OBE)~ An earthquake greater than an OBE but less than a Safe Shutdown Earthquake (SSE) should have no significant impact on safety-related systems, structures and components; however, some time may be required for the plant staff to ascertain the actual post-event condition of the plant (e.g., performs walk-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. 06g). The Shift Manager or Emergency Manager may seek external verification if deemed appropriate (e.g .*. a call to the Page 109 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases 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.

WCGS Basis Reference(s):

1. FSAR Section 2.5.2.7 Operating Basis Earthquake

2. ALR 00-0980 QBE

3. OFN SG-003, Natural Events

4. NEI 99-01 HU2 Page 110 of 228 INFORMATION USE

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 USAR Figure 1.2-44 Site Plan.

Basis:

Response actions associated with a tornado onsite is provided in OFN SG-003 Natural Events (ref. 1).

If damage is confirmed visually or by other in-plant indications, the event may be escalated to an Alert Linder EAL CA6.1 or SA9.1 ..

A tornado striking (touching down) within the Protected Area warrants declaration of an Unusual Event regardless of the measured wind speed at the meteorological tower. A tornado is defined as a violently rotating column of air in contact with the ground and extending from the base of a thunderstorm.

This EAL addresses a tornado striking (touching down) within the Protected Area ..

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

WCGS Basis Reference(s):

1. OFN SG-:003 Natural Events

2. NEI 99-01 HU3 Page 111 of 228 INFORMATION USE

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.

Basis:

Refer to EAL CA6.1 or SA9.1 for internal or external flooding affecting *one or more SAFETY SYSTEM trains. .

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

This EAL addresses flooding of a building room or area that results in operators isolating power to a Safety System component due to water level or other wetting concerns.

Classification is also required if the water level or related wetting causes an automatic isolation of a Safety System component from its power source (e.g., a breaker or relay trip). To warrant classification, operability of the affected component must be required by Technical Specifications for the current operating mode.

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

• Page 112 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases WCGS Basis Reference(s):

1. NEI 99-01 HU3 Page 113 of 228 INFORMATION USE

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):

JMPEDE(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 USAR Figure 1.2-44 Site Plan.

Basis:

As used here, the term "offsite" is meant to be areas external to the WCGS Protected Area.

This EAL addresses a hazardous materials event originating at an offsite location and of sufficient magnitude to impede the movement of personnel within the Protected Area.

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

WCGS Basis Reference(s):

1. NEI 99-01 HU3 Page 114 of 228 INFORMATION USE

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 Basis:

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

This EAL addresses a hazardous event that causes an on-site impediment to vehicle movement and significant enough to prohibit the plant staff from accessing the site using personal vehicles. Examples of such an event include site flooding caused by a hurricane, heavy rains, up-river water releases, dam failure, etc., or an on-site train derailment blocking the access road.

This EAL is not intended apply to routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the flooding around the Cooper Station during the Midwest floods of 1993, or the flooding around Ft. Calhoun Station in 2011.

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

WCGS Basis Reference(s):

1. NEI 99-01 HU3 Page 115 of 228 INFORMATION USE

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 Manager should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Table H-1 Fire Areas

• Auxiliary Building

• Reactor Building

• Control Building

• Fuel Building

• Emergency Diesel Generator Building

• ESW Pump House

• Refueling Water Storage Tank Valve Room 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 evidence of heat are observed.

Basis:

The 15 minute requirement begins with a credible notification that a fire is occurring, or receipt of multiple valid fire detection system alarms or field validation of a single fire alarm. The alarm is to be validated using available Control Room indications or alarms to prove that it is not spurious, or by reports from the field.

Table H-1 Fire Areas are based on AP 10-100 Fire Protection Program Section 4.15 Safety Related Areas. Table H-1 Fire Areas include those structures containing functions and systems required for safe shutdown of the plant (Safety Systems) (ref. 1).

J Page 116 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases This IC addresses the magnitude and extent of fires that may be indicative of a potential degradation of the level of safety of the plant.

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 a

of suppression system, etc.

Upon receipt, operators will take prompt actions to confirm the validity of initial fire alarms, indications, or report. For EAL assessment purposes, the emergency declaration clock starts at the time that the initial alarms, 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 multiple initial alarms, indication or report.

Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9.

WCGS Basis Reference(s):

1. AP 10-100 Fire Protection Program Section 4.15 Safety Related Areas

2. NEI 99-01 HU4 Page 117 of 228 INFORMATION USE

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 Manager should declare the event promptly upon determining that time limit has been exceeded, or will.likely be exceeded.

Table H-1 Fire Areas

• Auxiliary Building

• Reactor Building

• Control Building

• Fuel Building

• Emergency Diesel Generator Building

• ESW Pump House

• Refueling Water Storage Tank Valve Room 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 evidence of heat are observed.

Basis:

The 30 minute requirement begins upon receipt of a single 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.

Table H-1 Fire Areas are based on AP 10-100 Fire Protection Program Section 4.15 Safety Related Areas. Table H-1 Fire Areas include those structures containing functions and systems required for safe shutdown of the plant (Safety Systems) (ref. 1).

Page 118 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases This IC addresses the magnitude and extent of FIRES that may be indicative of a potential*

degradation of the level of safety of the plant.

This EAL addresses receipt of a single fire alarm, and the existence of a FIRE is not verified (i.e., proved or disproved) within 30-minutes of the alarm. Upon receipt, operators will take prompt actions to confirm the validity of a single fire alarm. For EAL assessment purposes, the 30-minute clock starts at the time that the initial alarm was received, and not the time that a subsequent verification action was performed.

A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failure or a spurious activation, and not an actual FIRE. For this reason, additional time is allowed to verify the validity of the alarm. The 30-minute period is a reasonable amount of time to determine if an actual FIRE exists; however, after that time, and absent information to the contrary, it is assumed that an actual FIRE is in progress.

If an actual FIRE is verified by a report from the field, then EAL HU4.1 is immediately applicable, and the emergency must be declared if the FIRE is not extinguished within 15-minutes of the report. If the alarm is verified to be due to an equipment failure or a spurious activation, and this verification occurs within 30-minutes of the receipt of the alarm, then this EAL is not applicable and no emergency declaration is warranted.

Basis.,.Related Requirements from Appendix R Appendix R to 10 CFR 50, states in part:

Criterion 3 of Appendix A to this part specifies that "Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions."

When considering the effects of FIRE, those systems associated with achieving and maintaining safe shutdown conditions assume major importance to safety because damage to them can lead to core damage resulting from loss of coolant through boil-off.

Because FIRE may affect safe shutdown systems and because the loss of function of systems used to mitigate the consequences of design basis accidents under post-fire conditions does not per se impact public safety, the need to limit fire damage to systems required to achieve and maintain safe shutdown conditions is greater than the need to limit fire damage to those systems required to .mitigate the consequences of design basis accidents.

In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of 1-hour fire barriers for the enclosure of cable and equipment and associated non-safety circuits of one redundant train (G.2.c). As used in HU4.2, the 30-minutes to verify a single alarm is well within this worst-case 1-hour time period.

Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9.

WCGS Basis Reference(s):

Page 119 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases

1. AP 10-100 Fire Protection Program Section 4.15 Safety Related Areas

2. NEI 99-01 HU4 Page 120 of 228 INFORMATION USE I*

ATTACHMENT 1 EAL Bases Category: H ~ Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:

HU4.3 Unusual Event A FIRE within the plant PROTECTED AREA not extinguished within 60 min. of the initial report, alarm or indication (Note 1)

Note 1: The Emergency Manager 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 evidence of 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 USAR Figure 1.2-44 Site Plan.

Basis:

This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.

In addition to a fire addressed by HU4.1 or HU4.2, a FIRE within the plant Protected Area not extinguished within 60-minutes may also potentially degrade the level of plant safety.

Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9.

WCGS Basis Reference(s):

1. NEI 99-01 HU4 Page 121 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:

HU4.4 Unusual Event A FIRE within the plant PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish Mode Applicability:

All Definition(s):

FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and evidence of 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 USAR Figure 1.2-44 Site Plan.

Basis:

This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.

Declaration is not necessary if the agency resources are placed on stand-by, or supporting post-extinguishment recovery or investigation actions. If a FIRE within the plant Protected Area is of sufficient size to require a response by an offsite firefighting agency (e.g., a local town Fire Department), then the level of plant safety is potentially degraded. The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the FIRE is beyond the capability of the Fire Brigade to extinguish. Note that the offsite fire agency is always called to respond to an actual fire within the PROTECTED AREA.

Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9.

WCGS Basis Reference(s):

1. NEI 99-01 HU4 Page 122 of 228 INFORMATION USE

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 that prohibits or IMPEDES access to any Table H-2 rooms or areas (Note 5)

Note 5: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted Table H-2 Safe Operation & Shutdown Rooms/Areas Room/Area Mode Applicability North Electrical Pen. Room A 3, 4 South Electrical Pen. Room B 3, 4 ESF SWGR Room No. 1 (TRN A) 4 ESF SWGR Room No. 2 (TRN B) 4 Auxiliary Building/West Hall Elev 2000 3,4,5 Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

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

Basis:

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 nec~ssary at the time of the release.

Evaluation of the IC and EAL do not require atmospheric sampling; it only requires the Emergency Manager's judgment that the gas concentration in the affected room/area is Page* 123 of 228 INFORMATION USE I

---

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

ATTACHMENT 1 EAL Bases 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 elevated radiation levels). For example, the plant is in Mode 5 when the radiation increase occurs, and the procedures used for normal operation, cooldown and shutdown only require entry into the affected room in Modes 1-4.

• 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.

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

NOTE: EAL HA5.1 mode applicability has been limited to the applicable modes identified in Table H-2 Safe Operation & Shutdown Rooms/Areas. If due to plant operating procedure or plant configuration changes, the applicable plant modes specified in Table H-2 are changed, a corresponding change to Attachment 3 'Safe Operation &

Shutdown Areas Tables R-3 & H-2 Bases' and to EAL HA5.1 mode applicability is required.

WCGS Basis Reference(s):

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

2. NEI ~9-01 M3 Page 124 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 6 - Control Room Evacuation Initiating Condition: Control Room evacuation resulting in transfer of plant control to alternate locations EAL:

HA6.1 Alert An event has resulted in plant control being transferred from the Control Room to the Auxiliary Shutdown Panel (ASP)

Mode Applicability:

All Definition(s):

None Basis:

The Shift Manager (SM) determines if the Control Room is inoperable and requires evacuation.

Control Room inhabitability may be caused by FIRE, dense smoke, noxious vapors or radiation/airborne activity in or adjacent to the Control Room, or other life threatening conditions. OFN RP-013 Control Room Not Habitable and/or OFN RP-017 Control Room Evacuation provides the instructions for tripping the unit, and maintaining RCS inventory and Hot Shutdown conditions from outside the Co~trol Room (Ref. 1, 2).

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

This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations outside the Control Room. The loss of the ability to control the plant from the Control Room is considered to be a potential substantial degradation in the level of plant safety.

Following a Control Room evacuation, control of the plant will be transferred to alternate shutdown locations. The necessity to control C1 plant shutdown from outside the Control Room, in addition to responding to the event that required the evacuation of the Control Room, will

• present challenges to plant operators and other on-shift personnel. Activation of the ERO and emergency response facilities will assist in responding to these challenges.

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

WCGS Basis Reference(s):

1. OFN RP-013 Control Room Not Habitable

2. OFN RP-017 Control Room Evacuation

3. NEI 99-01 HA6 Page 125 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 6 - Control Room Evacuation Initiating Condition: Inability to control a key safety function from outside the Control Room EAL:

HS6.1 Site Area Emergency An event has resulted in plant control being transferred from the Control Room to the Auxiliary Shutdown Panel (ASP)

AND Control of any of the following key safety functions is not re-established within 15 min.

(Note 1):

• Reactivity (Modes 1, 2 and 3 only)

• Core Cooling

• RCS heat removal Note 1: The Emergency Manager 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, 5 - Cold Shutdown, 6 - Refueling Definition(s):

None Basis:

For the purpose of this EAL the 15 minute clock starts when the last licensed operator leaves 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 vapors or radiation/airborne activity in or adjacent to the Control Room, or other life threatening conditions. OFN RP-013 Control Room Not Habitable and/or OFN RP-017 Control Room Evacuation provides the instructions for tripping the unit, and maintaining RCS inventory and Hot Shutdown conditions from outside the Control Room (Ref. 1, 2).

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 last licensed operator leaves the Control Room (not when OFN RP-013 or OFN RP-017 is entered). The time interval is based on how quickly control must be reestablished without core uncovery and/or core damage.

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 emphasis should be placed on components and instruments that supply protection for and Page 126 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases 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).

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 Manager judgment. The Emergency Manager is expected to make a reasonable, informed judgment within (the site-specific.time for transfer) 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.

WCGS Basis Reference(s):

1. OFN RP-013 Control Room Not Habitable

2. OFN RP-017 Control Room Evacuation

3. NEI 99-01 HS6 Page 127 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: H - tiazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Manager Judgment

. Initiating Condition: Other conditions existing that in the judgment of the Emergency Manager warrant declaration of a UE EAL:

HU7.1 Unusual Event Other conditions exist which in the judgment of the Emergency Manager indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated.

No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of SAFETY SYSTEMS occurs.

Mode Applicability:

All Definition(s):

None Basis:

The Emergency Manager is the designated onsite individual having the responsibility and authority for implementing the Wolf Creek Radiological Emergency Response Plan (ref. 1).

The Operations Shift Manager (SM) initially acts in the capacity of the Emergency Manager and takes actions as outlined in the Emergency Plan implementing procedures (ref. 2). If required by the emergency classification or if deemed appropriate by the Emergency Manager, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency.

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

WCGS Basis Reference(s):

1. Wolf Creek Radiological Emergency Response Plan section 5.1 Site Emergency Manager

2. Wolf Creek Radiological Emergency Response Plan section 5. 7 Shift. Manager

3. NEI 99-01 HU?

Page* 128 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Manager Judgment Initiating Condition: Other conditions exist that in the judgment of the Emergency Manager warrant declaration of an Alert EAL:

HA7.1 Alert Other conditions exist which, in the judgment of the Emergency Manager, indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.

Mode Applicability:

All Definition(s):

HOSTILE ACTION-An act toward WCGS 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 WCGS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area).

Basis:

The Emergency Manager is the designated onsite individual having the responsibility and authority for implementing the Wolf Creek Radiological Emergency Response Plan (ref. 1).

The Operations Shift Manager (SM) initially acts in the capacity of the Emergency Manager and takes actions as outlined in the Emergency Plan implementing procedures (ref. 2). If required by the emergency classification or if deemed appropriate by the Emergency Manager, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency.

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

WCGS Basis Reference(s):

1. Wolf Creek Radiological Emergency Response Plan section 5.1 Site Emergency Manager

2. Wolf Creek Radiological Emergency Response Plan section 5. 7 Shift Manager I Page 129 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases 3 .. NEI 99-01 HA?

Page 130 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Manager Judgment Initiating Condition: Other conditions existing that in the judgment of the Emergency Manager warrant declaration of a Site Area Emergency EAL:

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

All Definition(s):

HOSTILE ACTION -An act toward WCGS 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 WCGS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area).

Basis:

The Emergency Manager is the designated onsite individual having the responsibility and authority for implementing the Wolf Creek Radiological Emergency Response Plan (ref. 1).

The Operations Shift Manager (SM) initially acts in the capacity of the Emergency Manager and takes actions as outlined in the Emergency Plan implementing procedures (ref. 2). If required by the emergency classification or if deemed appropriate by the Emergency Manager, emergency response personnel are notified and-instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency.

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 Manager to fall under the emergency classification level description for a Site Area Emergency.

WCGS Basis Reference{s):

Page 131 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases

1. Wolf Creek Radiological Emergency Response Plan section 5.1 Site Emergency Manager

2. Wolf Creek Radiological Emergency Response Plan section 5. 7 Shift Manager

3. NEI 99-01 HS?

Page 132 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Manager Judgment Initiating Condition: Other conditions exist which in the judgment of the Emergency Manager warrant declaration of a General Emergency EAL:

HG7 .1 General Emergency Other conditions exist which in the judgment of the Emergency Manager indicate that events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area

• Mode Applicability:

All Definition(s ):

. HOSTILE ACTION - An act toward WCGS 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 WCGS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area).

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

Basis:

The Emergency Manager is the designated onsite individual having the responsibility and authority for implementing the Wolf Creek Radiological Emergency Response Plan (ref. 1).

The Operations Shift Manager (SM) initially acts in the capacity of the Emergency Manager and takes actions as outlined in the Emergency Plan implementing procedures (ref. 2). If required by the emergency Classification or if deemed appropriate by the Emergency Manager,

.emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency.

Releases can reasonably be expected to exceed EPA PAG plume exposure levels outside the Site Boundary.

Page 133 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Manager to fall under the emergency classification level description for a General Emergency.

WCGS Basis Reference(s):

1. Wolf Creek Radiological Emergency Response Plan section 5.1 Site Emergency Manager

2. Wolf Creek Radiological Emergency Response Plan section 5.7 Shift Manager

3. NEI 99-01 HG?

Page 134 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Category S - System Malfunction EAL Group: Hot Conditions (RCS temperature > 200°F); EALs in this category are applicable only in one or more hot operating modes.

Numerous system-related equipment failure events that warrant emergency classification have been identified in this category. They may pose actual or potential threats to plant safety.

The events of this category pertain to the following subcategories:

1. Loss of Emergency AC Power Loss of emergency electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of onsite and offsite sources for 4.16KV AC emergency 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 voe 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. RTS Failure This subcategory includes events related to failure of the Reactor Trip System (RTS) to initiate and complete reactor trips. In the plant licensing basis, postulated failures of the RTS to complete a reactor trip comprise a specific set of analyzed events referred to as Page 135 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Anticipated Transient Without Scram (ATWS) events; For EAL classification, however, ATWS is intended to mean any trip failure event that does not achieve reactor shutdown. If RTS 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. Affecting Safety Systems Various natural and technological events that result in degraded plant safety system performance or significant VISIBLE DAMAGE warrant emergency classification under this subcategory.

Page 136 of 228 INFORMATION USE

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 emergency buses for 15 minutes or longer EAL:

SU1.1 Unusual Event Loss of all offsite AC power capability, Table S-1, to emergency 4.16KV buses NB01 and NB02 for~ 15 min. (Note 1)

Note 1: The Emergency Manager 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:

• ESF XFMR XNB01

• ESF XFMR XNB02 Onsite:

• EOG NE01

• EOG NE02

• SBO OGs (if already aligned)

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None Basis:

The 4.16KV AC System provides the power requirements for operation and safe shutdown of the plant. The essential switchgear are buses NB01 and NB02 (ref. 1).

Page 137 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases 4.16KV buses NB01 and NB02 are the emergency (essential) buses. NB01 supplies power to Load Group 1 (Red Train) safety related loads and NB02 supplies power to Load Group 2 (Yellow Train) safety related loads. Each bus has two sources of offsite power. One source is from 4.16KV ESF transformer XNB01 and the other source is from the ESF transformer XNB02. Transformer XNB01 is the normal supply to bus NB01 and the alternate supply to bus NB02; XNB02 is the normal supply to bus NB02 and the alternate supply to bus NB01 (ref. 1, 2).

In addition, NB01 and NB02 each have an emergency diesel generator which supply electrical power to the bus automatically in the event that the preferred source becomes unavailable (ref.

2).

An additional source of power are the SBO diesel generators SBO DGs (ref. 3, 4). Credit can be taken for this source only if they are already aligned because they cannot be aligned within 15 minutes.

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 SA 1.

WCGS Basis Reference(s):

1. USAR figure 8.2-5 Electrical one line diagram of Wolf Creek 345 KV switchyard

2. USAR Section 8.3

3. OFN NB-030 Loss of AC Emergency Bus NB01 (NB02)

4. SYS KU-121 (122) Energizing NB01 (NB02) From Station Blackout Diesel Generators

5. NEI 99-01 SU1 Page 138 of 228 INFORMATION USE

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 emergency buses for 15 minutes or longer EAL:

SA1.1 Alert AC power capability, Table S-1, to emergency 4.16KV buses NB01 and NB02 reduced to a single power source for~ 15 min. (Note 1)

AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS Note 1: The Emergency Manager 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:

• ESF XFMR XNB01

• ESF XFMR XNB02 Onsite:

• EDG NE01

• EDG NE02

• SBO DGs (if already aligned)

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.

Page 139 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Basis:

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 4.16KV AC System provides the power requirements for operation and safe shutdown of the plant. The essential switchgear are buses NB01 and NB02 (ref. 1).

4.16KV buses NB01 and NB02 are the emergency (essential) buses. NB01 supplies power to Load Group 1 (Red Train) safety related loads and NB02 supplies power to Load Group 2 (Yellow Train) safety related loads. Each bus has two sources of offsite power. One source is from 4.16KV ESF transformer XNB01 and the other source is from the ESF transformer XNB02. Transformer XNB01 is the normal supply to bus NB01 and the alternate supply to bus NB02; XNB02 is the normal supply to bus NB02 and the alternate supply to bus NB01 (ref. 1, 2).

In addition, NB01 and NB02 each have an emergency diesel generator which supply electrical power to the bus automatically in the event that the preferred source becomes unavailable (ref.

2).

An additional source of power are the SBO diesel generators SBO DGs (ref. 3, 4 ). Credit can be taken for this source only if they are already aligned because they cannot be aligned within 15 minutes.

This hot condition EAL is equivalent to the cold condition EALCU2.1.

If the capability of a second source of emergency bus power is not restored to at least one emergency bus within 15 minutes, an Alert is declared under this EAL.

This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to Safety Systems. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment. This IC provides an escalation path from IC SU1.

An "AC power source" is a source recognized in OFNs and EMGs, and capable of supplying

• required power to an emergency bus. Some examples of this condition are presented below.

• A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).

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

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

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

Escalation of the emergency classification level would be via IC SS1 ..

Page 140 of 228 INFORMATION USE

ATJ"ACHMENT 1 EAL Bases WCGS Basis Reference(s):

1. USAR figure 8.2-5 Electrical one line diagram of Wolf Creek 345 KV switchyard

2. USAR Section 8.3

3. OFN NB-030 Loss of AC Emergency Bus NB01 (NB02)

4. SYS KU-121(122) Energizing NB01(NB02) From Station Blackout Diesel Generators

5. NEI 99-01 SA 1

• 1~~~~~~~~~~~~~P_ag_e~14_1_o_f_2_2_8~~~~~~~1N_F_O_R_M_A_T_l_O_N_U_S_E__,

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 1 - Loss of Emergency AC Power Initiating Condition: Loss of all offsite power and all onsite AC power to emergency buses for 15 minutes or longer EAL:

SS1 .1 Site Area Emergency Loss of all offsite and all onsite AC power to emergency 4.16KV buses NB01 and NB02 for~ 15 min. (Note 1)

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

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None Basis:

The 4.16KV AC System provides the power requirements for operation and safe shutdown of the plant. The essential switchgear are buses NB01 and NB02 (ref. 1).

4.16KV buses NB01 and NB02 are the emergency (essential) buses. NB01 supplies power to Load Group 1 (Red Train) safety related loads and NB02 supplies power to Load Group 2 (Yellow Train) safety related loads. Each bus has two sources of offsite power. One source is from 4.16KV ESF transformer XNB01 and the other source is from the ESF transformer XNB02. TransformerXNB01 is the normal supply to bus NB01 and the alternate supply to bus NB02; XNB02 is the normal supply to bus NB02 and the alternate supply to bus NB01 (ref. 1, 2).

In addition, NB01 and NB02 each have an emergency diesel generator, which supply electrical power to the bus automatically in the event that the preferred source becomes unavailable (ref.

2).

An additional source of power are the SBO diesel generators SBO DGs (ref. 3, 4). Credit can be taken for this source only if they are already aligned because they cannot be aligned within 15 minutes.

The interval.begins when both offsite and onsite AC power are lost.

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

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.

Page 142 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Escalation of the emergency classification level would be via ICs RG1, FG1 or SG1.

WCGS Basis Reference(s):

1. USAR figure 8.2-5 Electrical one line diagram of Wolf Creek 345 KV switchyard

2. USAR Section 8.3

3. OFN NB-030 Loss of AC Emergency Bus NB01 (NB02)

-4. SYS KU-121 (122) Energizing NB01 (NB02) From Station Blackout Diesel Generators

5. BD-EMG C-0 Loss of All AC Power

6. NEI 99-01 SS1 I Page 143 of 228 INFORMATION USE .I*

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 emergency buses EAL:

SG1 .1 General Emergency Loss of all offsite and all onsite AC power to emergency 4.16KV buses NB01 and NB02 AND EITHER:

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

• CSFST Core Cooling RED Path conditions met Note 1: The Emergency Manager should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None Basis:

This EAL is indicated* by the extended loss of all offsite and on site AC power to 4.16KV emergency buses NB01 and NB02 either for greater then the WCGS Station Blackout (SBO) coping analysis time (4 hrs.) (ref. 1) 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. 2).

The 4.16KV AC System provides the power requirements for operation and safe shutdown of the plant. The essential switchgear are buses NB01 and NB02 (ref. 3).

• 4.16KV buses NB01 and NB02 are the emergency (essential) buses. NB01 supplies power to Load Group 1 (Red Train) safety related loads and NB02 supplies power to Load Group 2 (Yellow Train) safety related loads. Each bus has two sources of offsite power. One source is from 4.16KV ESF transformer XNB01 and the other source is from the ESF transformer

• XNB02. Transformer XNB01 is the normal supply to bus NB01 and the alternate supply to bus NB02; XNB02 is the normal supply to bus NB02 and the alternate supply to bus NB01 (ref. 3, 4). .

In addition, NB01 and NB02 each have an emergency diesel generator which supply electrical power to the bus automatically in the event that the preferred source becomes* unavailable (ref.

4). .

An additional source of power are the SBO diesel generators SBO DGs (ref. 5) .. Credit can be taken for this source only if they can be aligned within the 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> period.

l Page 144 of 228 INFORMATION USE.

ATTACHMENT 1 EAL Bases Indication of continuing core cooling degradation must be based on fission product barrier monitoring with particular emphasis on Emergency Manager 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. Specifically, Core Cooling RED PATH conditions exist if either core exit T/Cs are reading greater than or equal to 1200°F or core exit T/Cs are reading greater than or equal to 712°F with RCS subcooling less than or equal to 30°F [45°F], and RVLIS natural circulation range indication is less than 45% (ref. 2).

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 FG1. This will allow additional time for implementation of offsite protective actions.

Escalation of the emergency classification from Site Area Emergency will occur if it is projected that power cannot be restored to at least one AC 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.

WCGS Basis Reference(s):

1. FSAR Section 8.3A.3

2. CSF F-02 Critical Safety Function Status Trees (CSFST) Figure 2, Core Cooling

3. FSAR figure 8.2-5 Electrical one line diagram of Wolf Creek 345 KV switchyard 4 FSAR Section 8.3

5. SYS KU-121(122) Energizing NB01(NB02) From Station Blackout Diesel Generators

6. BD-EMG C-0 Loss of All AC Power

7. NEI 99-01 SG1 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:

Page 145 of 228 INFORMATION USE I.

ATTACHMENT 1 EAL Bases SG1 .2 General Emergency Loss of all offsite and all onsite AC power to emergency 4.16KV buses NB01 and NB02

• for~ 15 min.

AND Loss of all 125 VDe power based on battery bus voltage indications < 105 VDe on all vital DC buses NK01, NK03 (Division 1) and NK02, NK04 (Division 2) for~ 15 min.

(Note 1)

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

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None Basis:

This EAL is indicated by the loss of all offsite and onsite emergency AC power to 4.16KV emergency buses NB01 and NB02 for greater than 15 minutes in combination with degraded vital DC power voltage. This EAL addresses operating experience from the March 2011 accident at Fukushima Daiichi.

The 4.16KV AC System provides the power requirements for operation and safe shutdown of the plant. The essential switchgear are buses NB01 and NB02 (ref. 1).

4.16KV buses NB01 and NB02 are the emergency (essential) buses. NB01 supplies power to Load Group 1 (Red Train) safety related loads and NB02 supplies power to Load Group 2 (Yellow Train) safety related loads. Each bus has two sources of offsite power. One source is from 4.16KV ESF transformer XNB01 and the other source is from the 'ESF transformer XNB02. Transformer XNB01 is the normal supply to bus NB01 and the alternate supply to bus NB02; XNB02 is the normal supply to bus NB02 and the alternate supply to bus NB01 (ref. 1, 2).

In addition, NB01 and NB02 each have an emergency diesel generator which supply electrical power to the bus automatically in the event that the preferred source becomes unavailable (ref.

2).

An additional source of power are the SBO diesel generators SBO DGs (ref. 3). Credit can be taken for this source only if they are already aligned because they cannot be aligned within 15 minutes.

The vital DC buses are the following 125 VDC Class 1E buses (ref. 4):

Division 1 (Train A): Division 2 (Train B):

• NK01

• NK02

• NK03

• NK04 There are four, 60 cell, lead-calcium storage batteries (NK11, NK12, NK13 and NK14) that supplement the output of the battery chargers. They supply DC power to the distribution buses I Page 146 of 228 INFORMATION USE

r~--

ATTACHMENT 1 EAL Bases when AC power to the chargers is lost or when transient loads exceed the 300 amp capacity of the battery chargers.

Each battery is designed to have sufficient stored energy to supply the required emergency loads for 240 minutes following a loss of AC power (station blackout) (ref. 4, 5, 6).

Minimum DC bus voltage is 105.0 VDC (ref. 7).

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.

WCGS Basis Reference(s):

1. FSAR figure 8.2-5 Electrical one line diagram of Wolf Creek 345 KV switchyard

2. FSAR Section 8.3

3. SYS KU-121 (122) Energizing NB01 (NB02) From Station Blackout Diesel Generators

4. OFN NK-020 Loss of Vital 125 VDC Bus NK01, NK02, NK03, and NK04

5. FSAR Tables 8.3-1, -2, -3

6. FSAR Section 8.3.2

7. Calculation NK-E-001 125 VDC Class 1 E Battery System Sizing, Voltage Drop and Short Circuit Studies

8. NEI 99-01 SGS Page 147 of 228 INFORMATION USE

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 Loss of all 125 VDC power based on battery bus voltage indications < 105 VDC on all vital DC buses NK01, NK03 (Division 1) and NK02, NK04 (Division 2) for~ 15 min. (Note 1)

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

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None Basis:

The vital DC buses are the following 125 VDC Class 1E buses (ref. 1):

Division 1 (Train A): Division 2 (Train B):

• NK01

• NK02

• NK03

• NK04 There are four, 60 cell, lead-calcium storage batteries (NK11, NK12, NK13 and NK14) that supplement the output of the battery chargers. They supply DC power to the distribution buses when AC power to the chargers is lost or when transient loads exceed the 300 amp capacity of the battery chargers.

Each battery is designed to have sufficient stored energy to supply the required emergency loads for 240 minutes following a loss of AC power (station blackout) (ref. 2, 3, 4).

Minimum DC bus voltage is 105.0 VDC (ref. 4).

This IC addresses a loss of vital DC power which compromises the ability to monitor and control SAFETY SYSTEMS. In modes above Cold Shutdown, this condition involves a major failure of plant functions needed for the protection of the public.

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

Escalation of the emergency classification level would be via ICs RG1, FG1 or SG1.

Page 148 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases WCGS Basis Reference(s}:

1. OFN NK-020 Loss of Vital 125 VDC Bus NK01, NK02, NK03, and NK04

2. FSAR Tables 8.3-1, -2, -3

3. FSAR Section 8.3.2

4. Calculation NK-E-001 125 VDC Class 1 E Battery System Sizing, Voltage Drop and Short Circuit Studies

5. NEI 99-01 SSS Page 149 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 3 - Loss of Control Room Indications Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer EAL:

SU3.1 Unusual Event An UNPLANNED event results in the inability to monitor one or more Table S-2 parameters from within the Control Room for~ 15 min. (Note 1)

Note 1: The Emergency Manager 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 TIC temperature

• Level in at least one SIG

• Auxiliary or emergency feed flow in at least one SIG Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

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

Basis:

Safety System parameters listed in Table S-2 are monitored in the Control Room through a combination of hard control paner indicators as well as computer based information systems.

The NPIS, which displays SPDS required information, serves as a redundant compensatory indicator which may be utilized in lieu of normal Control Room indicators (ref. 1, 2).

This IC addresses the difficulty associated with monitoring normal plant conditions Without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition is a precursor to a more significant event and represents a potential degradation in the level of safety of the plant.

As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room.

I Page 150 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.

This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling and 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 deterryiine the values of other Safety System parameters may be impacted as well. For example, if the value for reactor vessel level cannot be determined from the indications and recorders on a main control board, the SPDS or NPIS, the availability of other parameter values may be compromised as well.

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

Escalation of the emergency classification level would be via IC SA3.

WCGS Basis Reference(s):

1. USAR Section 7.5 Safety-RelatedDisplay Instrumentation

2. OFN RJ-023 NPIS Malfunctions

3. NEI 99-01 SU2 Page 151 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 3 - Loss of Control Room Indications Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress EAL:

SA3.1 Alert An UNPLANNED event results in the inability to monitor one or more Table S-2 parameters from within the Control Room for~ 15 min. (Note 1)

AND Any significant transient is in progress, Table S-3 Note 1: The Emergency Manager 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 SIG

• Auxiliary or emergency feed flow in at least one SIG Table S-3 Significant Transients

• Reactor trip

• Runback ~ 25% thermal power

• Electrical load rejection > 25% full electrical load

• ECCS actuation Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

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

Pa.ge 152 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases 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 NPIS, which displays SPDS required information, serves as a redundant compensatory indicator which may be utilized in lieu of normal Control Room indicators (ref. 1, 2).

Significant transients are listed in Table S-3 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.

This IC addresses the difficulty associated with monitoring rapidly changing plant conditions during a transient without the ability to obtain Safety System parameters from within the Control Room. During this condition, the margin to a potential fission product barrier challenge is reduced. It thus represents a potential substantial degradation in the level of safety of the plant.

As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room.

An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.

This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling and 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 cannot be determined from the indications and recorders on a main control board, the SPDS or NPIS, the availability of other parameter values may* be compromised as well.

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

Escalation of the emergency classification level would be via ICs FS1 or IC RS1 WCGS Basis Reference(s):

1. USAR Section 7.5 Safety-Related Display Instrumentation

2. OFN RJ-023 NPIS Malfunctions

3. NEI 99-01 SA2 Category: S - System Malfunction Page 153 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Subcategory: 4 - RCS Activity Initiating Condition: Reactor coolant activity greater than Technical Specification allowable limits EAL:

SU4.1 Unusual Event Sample analysis indicates RCS activity > Technical Specification Section 3.4.16 limits Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None Basis:

The specific iodine activity is limited to either::;; 60 µCi/gm Dose Equivalent 1-131 or::;; 1.0

µCi/gm Dose Equivalent 1-131 for a > 48 hr continuous period. The specific Xe-133 activity is limited to ::;; 500 µCi/gm Dose Equivalent Xe-133 (ref 1, 2).

This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant.

Escalation of the emergency classification level would be via ICs FA1 or the Recognition Category R ICs.

• WCGS Basis Reference(s):

1. Wolf Creek Technical Specifications section 3.4.16 RCS Specific Activity

2. OFN BB-006 High Reactor Coolant Activity

3. NEI 99-01 SU3 Page 154 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 5 - RCS Leakage Initiating Condition: RCS leakage for 15 minutes or longer EAL:

SUS.1 Unusual Event RCS unidentified or pressure boundary leakage > 10 gpm for~ 15 min.

OR RCS identified leakage > 25 gpm for~ 15 min.

OR Leakage from the RCS to a location outside containment > 25 gpm for ~ 15 min.

(Note 1)

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

Mode Applicability:*

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None Basis:

Manual or computer-based methods of performing an RCS inventory balance are normally used to determine RCS leakage. The NPIS Computer is preferred method of calculating RCS leak rate. When the NPIS Computer is not available, procedural guidance is available to perform the manual RCS inventory balance (ref. 1, 2).

Identified leakage includes

• Leakage such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank, or

• Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary leakage, or

• RCS leakage through a steam generator to the secondary system (ref. 3).

Unidentified leakage is all leakage (except RCP seal water injection or leakoff) that is not identified leakage (ref. 3).

Pressure Boundary leakage is leakage (except primary to secondary leakage) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall (ref. 3)

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 Page 155 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases such as Chemical & Volume Control System, Safety Injection, Nuclear Sampling system and Residual Heat Removal system (when in the shutdown cooling mode) (ref. 4, 5)

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.

Thresholds #1 and #2 are focused on a loss of mass from the RCS due to "unidentified leakage", "pressure boundary leakage" or "identified leakage" (as these leakage types are defined in the plant Technical Specifications). The third threshold addresses a RCS mass loss caused by an unisolable leak through an interfacing system. These EALs thus apply to leakage into the containment, a secondary-side system (e.g:, steam generator tube leakage) or a location outside of containment.

The leak rate values for each EAL were selected because they are usually observable with normal Control Room indications. Lesser values typically require time-consuming calculations to determine (e.g., a mass balance calculation). The first. threshold 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. An emergency classification would be required if a mass loss is caused by a relief valve that is not functioning as designed/expected (e.g., a relief valve sticks open and the line flow cannot be isolated).

The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible.

WCGS Basis Reference(s):

1. STS BB-006 RCS Water Inventory Balance Using the NPIS Computer

2. STS BB-004 RCS Water Inventory Balance

3. Wolf Creek Technical Specifications Definitions section 1.1

4. USAR Section 5.2.5.2.1 lntersystem Leakage

5. OFN BB-007 RCS Leakage High

6. NEI 99-01 SU4 Page 156 of 228

• INFORMATION USE

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 6 - RTS Failure Initiating Condition: Automatic or manual trip fails to shut down the reactor EAL:

SU6.1 Unusual Event An automatic trip did not shut down the reactor as indicated by reactor power ~ 5% after any RTS setpoint is exceeded AND A subsequent automatic trip or manual trip action taken at the reactor control consoles (SB HS-1 or SB HS-42) is successful in shutting down the reactor as indicated by reactor power< 5% (Note 8)

Note 8: A manual trip action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods, tripping rod drive power or implementation of boron injection strategies.

Mode Applicability:

1 - Power Operation Definition(s):

None Basis:

The first condition of this EAL identifies the need to cease critical reactor operations by actuation of the automatic Reactor Trip System (RTS) trip function. A reactor trip is automatically initiated by the RTS 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 RTS to bring the reactor power below the immedi~te shutdown decay heat level of 5% (ref. 2, 3, 4).

For the purposes of emergency classification, successful manual trip actions are those which can be quickly performed from the reactor control console; SB HS-1 on Panel RL003 or SB HS-42 on Panel RL006. Reactor shutdown achieved by use of other trip actions specified in EMG FR-S1 Response to ATWS (such as opening PG HIS-16 and PG HIS-18 supply breakers, depressing manual pushbutton on turbine control panel, emergency boration or manually driving control rods) do not constitute a successful manual trip (ref. 4).

Following any automatic RTS trip signal, EMG E-0 (ref. 2) and EMG FR-S1 (ref. 4) prescribe insertion of redundant manual trip signals to back up the automatic RTS trip function and Page 157 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases 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. 6).

A reactor trip resulting from actuation of the ATWS Mitigation System Actuation Circuitry (AMSAC) logic that results in full insertion of control rods and diminishing neutron flux is considered a successful reactor trip. AMSAC automatically initiates auxiliary feedwater and a turbine trip under conditions indicative of an Anticipated Transient Without Scram (ATWS) event (ref. 5).

In the event that the operator identifies a reactor trip is imminent and initiates a successful manual reactor trip before the automatic RTS trip setpoint is reached, no declaration is required. The successful manual trip of the reactor before it reaches its automatic trip setpoint.

or reactor trip signals caused by instrumentation channel failures do not lead to a potential fission product barrier loss. However, if subsequent manual reactor trip actions fail to reduce reactor power below 5%, the event escalates to the Alert under EAL SA6.1.

If by procedure, operator actions include the initiation of an immediate manual trip following receipt of an automatic trip signal and there are no clear indications that the automatic trip failed (such as a time delay following ind.ications 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 potentlal damage, and the reporting requirements of 50. 72 should be considered for the transient event.

This IC addresses a failure of the RTS to initiate or* complete an automatic reactor trip that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic trip is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.

Following the failure on an automatic reactor trip, operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip. If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.

A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip). This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles".

The plant response to the failure of an automatic reactor trip will vary based upori 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 SA5. Depending upon the plant response, escalation is also possible via IC Page 158 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases FA 1. Absent the plant conditions needed to meet either IC SA6 or FA 1, an Unusual Event declaration is appropriate for this event.

A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.

Should a reactor trip signal be generated as a result of plant work (e.g., RTS setpoint testing),

the following classification guidance should be applied.

• If the signal causes a plant transient that should have included an automatic reactor trip and the RTS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and should be evaluated.

• If the signal does not cause a plant transient and the trip failure is determined through other means (e.g., assessment of test results), then this IC and the EALs are not applicable and no classification is warranted.

WCGS Basis Reference(s):

1. Wolf Creek Technical Specifications section 3.3.1 Reactor Trip System {RTS)

Instrumentation

2. EMG E-0 Reactor Trip or Safety Injection

3. EMG F-0 Critical Safety Function Status Trees - Subcriticality

4. EMG FR-S1 Response to Nuclear Power Generation/ATWS

5. FSAR Section 7.7.1 6 NEI 99-01 SU5 Page 159 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 6 - RTS Failure Initiating Condition: Automatic or manual trip fails to shut down the reactor EAL:

SU6.2 Unusual Event A manual trip did not shut down the reactor as indicated by reactor power~ 5% after any manual trip action was initiated AND A subsequent automatic trip or manual trip action taken at the reactor control console (SB HS-1 or SB HS-42) is successful in shutting down the reactor as indicated by reactor power< 5% (Note 8)

Note 8: A manual trip action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods, tripping rod drive power or implementation of boron injection strategies.

• Mode Applicability:

1 - Power Operation Definition(s):

None 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< 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 RTS to bring the reactor power below the immediate shutdown decay heat level of 5% .(ref. 2, 3, 4 ).

For the purposes of emergency classification, successful manual trip actions are those which can be quickly performed from the reactor control console; SB HS-1 on Panel RL003 or SB HS-42 on Panel RL006. Reactor shutdown achieved by use of other trip actions specified in EMG FR-S1 Response to ATWS (such as opening PG HIS-16 and PG HIS-18 supply breakers, depressing manual pushbutton on turbine control panel, emergency boration or manually driving control rods) do not constitute a successful manual trip (ref. 4).

Following the failure of any manual trip signal, EMG E-0 (ref. 2) and EMG FR-S1 (ref. 4) prescribe insertion of redundant manual trip signals to back up the RTS trip function and Page 160 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases 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. 6).

• A reactor trip resulting from actuation of the ATWS Mitigation System Actuation Circuitry (AMSAC) logic that results in full insertion of control rods and diminishing neutron flux is considered a successful reactor trip. AMSAC automatically initiates auxiliary feedwater and a turbine trip under conditions indicative of an Anticipated Transient Without Scram (ATWS) event (ref. 5).

If both subsequent automatic and subsequent manual reactor trip actions in the Control Room fail to reduce reactor power below the power associated with the safety system design(< 5%)

following a failure of an initial manual trip, the event escalates to an Alert under EAL SA6.1 This IC addresses a failure of the RTS to initiate or complete a manual reactor trip that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic trip is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.

If an initial manual reactor trip is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip using a different switch). Depending upon several factors, the initial or subsequent effort to manually trip the reactor, or a concurrent plar:it condition, may lead to the generation of an automatic reactor trip signal. If a subsequent manual or automatic trip is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.

A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip). This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles".

The plant response to the failure of a manual reactor trip will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC SA6.

Depending upon the plant response, escalation is also possible via IC FA 1. Absent the plant conditions needed to meet either IC SA6 or FA 1, an Unusual Event declaration is appropriate for this event.

A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.

Page 161 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Should a reactor trip signal be generated as a result of plant work (e.g., RTS setpoint testing),

the following classification guidance should be applied.

• If the signal causes a plant transient that should have included an automatic reactor trip and the RTS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and should be evaluated.

• If the signal does not cause a plant transient and the trip failure is determined through other means (e.g., assessment of test results), then this IC and the EALs are not applicable and no classification is warranted.

WCGS Basis Reference(s):

1. Wolf Creek Technical Specifications section 3.3.1 Reactor Trip System (RTS)

Instrumentation

2. EMG E-0 Reactor Trip or Safety Injection

3. EMG F-0 Critical Safety Function Status Trees - Subcdticality

4. EMG FR-S1 Response to Nuclear Power Generation/ATWS

5. FSAR Section 7. 7 .1

6. NEI 99-01 SU5 Page 162 of 228 INFORMATION USE I*

ATTACHMENT 1 EAL Bases Category: S- System MalfunGtion Subcategory: 2 - RTS 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

53 AND Manual trip actions taken at the reactor control console (SB HS-1 or SB HS-42) are not successful in shutting down the reactor as indicated by reactor power;;

5% (Note 8)

Note 8: A manual trip action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods, tripping rod drive power or implementation of boron injection strategies.

Mode Applicability:

1 - Power Operation Definition(s ):

None Basis:

This EAL addresses any automatic or manual reactor trip signal that fails to shut down the reactor (reactor power< 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; SB HS-1 on Panel RL003 or SB HS-42 on Panel RL006. Reactor shutdown achieved by use of other trip actions specified in EMG FR-S1 Response to ATWS (such as opening PG HIS-16 and PG HIS-18 supply breakers, depressing manual pushbutton on turbine control panel, emergency boration or manually driving control rods) do not constitute a successful manual trip (ref. 4).

A reactor trip. resulting from actuation of the A TWS Mitigation System Actuation Circuitry (AMSAC) logic that results in full insertion of control rods and diminishing neutron flux is i

considered a successful reactor trip. AMSAC automatically initiates auxiliary feedwater and a turbine trip under conditions indicative of an Anticipated Transient Without Scram (ATWS) event (ref. 5).

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 Page 163 of 228 INFORMATION USE

ATTACHMENT 1

. EAL Bases normal shutdown. Nuclear instrumentation can be used to determine if reactor power is greater than 5 % power (ref. 3, 4 ).

Escalation of this event to a Site Area Emergency would be under EAL SS6.1 or Emergency Manager judgment.

This IC addresses a failure of the RTS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and subsequent operator manual actions taken at the reactor control consoles to shutdown the reactor are also unsuccessful. This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if the reactor is subsequently shutdown by an action taken away from the reactor control consoles since this event entails a significant failure of the RTS.

A manual action at the reactor control 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 rip.

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".

The plant response to the failure of an automatic or manual reactor will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If the failure to shutdown the reactor is prolonged enough to cause a challenge to the core cooling or RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC SS6. Depending upon plant responses and symptoms, escalation is also possible via IC FS 1. Absent the plant conditions needed to meet either IC SS6 or FS1, an Alert declaration is appropriate for this event.

It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F ICs; however, this IC and EAL are included to ensure a timely emergency declaration.

A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.

WCGS Basis Reference(s):

1. Wolf Creek Technical Specifications section 3.3.1 Reactor Trip System (RTS)

Instrumentation

2. EMG E-0 Reactor Trip or Safety Injection

3. EMG F-0 Critical Safety Function Status Trees - Subcriticality

4. EMG FR-S1 Response to Nuclear Power.Generation/ATWS

5. FSAR Section 7.7.1

6. NEI 99-01 SA5 Page 164 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 2 - RTS 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

5%

AND All actions to shut down the reactor are not successful as indicated by reactor power

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 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 EMG FR-S1 Response to Nuclear Power Generation/ATWS (such as opening PG HIS-16 and PG HIS-18 supply breakers, depressing manual pushbutton on turbine control panel, emergency boration or manually driving control rods) are also credited as a successful manual trip provided reactor power can be reduced below 5% before indications of an extreme challenge to either core cooling or heat removal exist (ref. 1, 4).

5% rated power is a minimum reading on the power range scale that indicates continued power production. It also approximates the decay heat which the shutdown systems were designed to remove and is indicative of a condition requiring immediate response to prevent Page 165 of 228 . INFORMATION USE

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, 4 ).

Indication of continuing core cooling degradation is manifested by CSFST Core Cooling RED PATH conditions being met (ref. 2). Specifically, Core Cooling RED PATH conditions exist if either core exit T/Cs are reading greater than or equal to 1200°F or core exit T/Cs are reading greater than or equal to 712°F with RCS subcooling less than or equal to 30°F [45°F], and RVLIS natural circulation range indication is less than 45% (ref. 2).

Indication of inability to adequately remove heat from the RCS is manifested by CSFST Heat Sink RED PATH conditions being met (ref. 2). Specifically, Heat Sink RED PATH conditions exist if narrow range level in at least on steam generator is not greater than or equal to 6%

[29%] and total feedwater flow to the steam generators is less than or equal to 270,000 lbm/hr.

(ref. 3).

This IC addresses a failure of the RTS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, all subsequent operator actions to manually shutdown the reactor are unsuccessful, and continued power generation is challenging the capability to adequately remove heat from the core and/or the 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 shutdown the reactor. The inclusion of this IC and EAL ensures the timely declaration of a Site Area Emergency in response to prolonged failure to shutdown the reactor.

A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.

Escalation of the emergency classification level would be via IC RG1 or FG1.

WCGS Basis Reference(s):

1. EMG F-0 Critical Safety Function Status Trees - Figure 1 Subcriticality

2. EMG F-0 Critical Safety Function Status Tress - Figure 2 Core Cooling

3. EMG F-0 Critical Safety Function Status Tress - Figure 3 Heat Sink

4. EMG FR-S1 Response to Nuclear Power Generation/ATWS

5. NEI 99.;01 SS5 Page *166 of 228 INFORMATION USE

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 NRG communication methods Table S-4 Communication Methods System Onsite* Offsite NRC PA system x Plant Radios x x Site Telephone System x x x Local Telephone Company Direct Lines x x x ENS Line x Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None Basis:

Onsite/offsite/NRC communications include one or more of the systems listed in Table S-4 (ref.

1, 2).

1. Public Address (PA) system The system provides five separate independent Communication lines and one general page channel. Communication between parties in the plant and the Control Room can be easily and quickly established using the Control Room page channel (channel 1).

Communication between parties within the plant can be easily and quickly established by using the general page channel. The party line channel is normally used after the page call Page 167 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases is completed. As many as five party lines may communicate simultaneously.

2. Plant Radios The Plant Radio System consists of two repeater sites: the Turbine Building and the Meteorological Tower. The sites have multiple repeaters to support communications inside the Plant structures as well as in a 5 mile radius of the Plant. The system provides two-way portable and mobile communications for normal and emergency Plant operation. Portable and mobile users have communications capability with fixed operator positions in the Control Room, TSC and EOF, and with security radio consoles, if desired.

3. Site Telephone System The Touchtone Telephone System consists of telephone stations located throughout the Power Block, in the main Control Room, Security Building, Administration Building and various other buildings around the site. The system has diverse routing to provide multiple routes for both inward and outward dialing. The telephone system is powered through a battery backup system, which can provide about 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of service after loss of offsite power.

4. Local Telephone Company Direct Lines In the event of a total system failure there are multiple direct telephone lines from the local telephone company which remain operational for access to the local Burlington exchange ..

5. ENS line The NRC Emergency Notification System (ENS) is a Federal Telephone System (FTS) telephone used for official communications with NRC Headquarters. The NRC Headquarters has the capability to patch into the NRC Regional offices. The primary purpose of this phone is to provide a reliable method for the initial notification of the NRC and to maintain continuous communications with the NRC after initial notification. ENS telephones are located in the Control Room, TSC and EOF.

This EAL is the hot condition equivalent of the cold condition EAL CU5.1.

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

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

• The first condition addresses a total loss of the communications methods used in support of routine plant operations.

The second condition addresses a total loss of the communications methods used to notify all offsite organizations of an emergency declaration. The offsite organizations referred to here are the State and Coffey County EOCs.

The third condition addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.

I Page 168 of 228 INFORMATION USE I*

ATTACHMENT 1 EAL Bases

• WCGS Basis Reference(s):

1. Wolf Creek Generating Station Radiological Emergency Response Plan (RERP), Section 6.16.1

2. FSAR Section 9.5.2

3. NEI 99-01 SU6 Page 169 of 228 INFORMATION USE

lI 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:

SUS.1 Unusual Event Any penetration is not isolated within 15 min. of a VALID containment isolation signal OR Containment pressure > 27 psig with < one full train of c_ontainment depressurization equipment operating per d~sign for~ 15 min. (Note 9) *

(Note 1)

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

Note 9: One Containment Spray System train and one Containment Cooling System train comprise one full train of depressurization equipment.

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

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

Basis:

The Containment Spray System consists of two separate trains of equal capacity, each capable of meeting the design bases requirement. Each train includes a containment spray pump, spray headers, nozzles, valves, and piping. The containment spray additive tank supplies 30 weight percent (nominal) sodium hydroxide to both trains The refueling water storage tank (RWST) supplies borated water to the Containment Spray System during the injection phase of operation. In the recirculation mode of operation, Containment Spray pump suction is transferred from the RWST to the Containment sumps (ref. 2).

• The Containment Cooling System consists of two trains of Containment cooling, each of sufficient capacity to supply 100% of the design cooling requirement. Each train of two fan units is supplied with cooling water from a separate train of essential service water (ESW). Air

• is drawn into the coolers through the fan and discharged to the steam generator compartments, pressurizer compartment, and instrument tunnel, and outside the secondary shield wall. The air supplied to each steam generator compartment is drawn upwards through the compartments by the hydrogen mixing fans and discharged into the upper elevations of the containment. During normal operation, all four fan units are normally operating. In post-Page 170 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases accident operation following an actuation signal, the Containment Cooling System fans are designed to start automatically in slow speed if not already running (ref. 3).

The Containment pressure setpoint (27 psig, ref. 4, 5, 6) is the pressure at which the equipment should actuate and begin performing its function. The design basis accident analyses and evaluations assume the loss of one Containment Spray System train and one Containment Cooling System train (ref. 7). Consistent with the design requirement, "one full train of depressurization equipment" is therefore defined to be the availability of one train of each system. If less than this equipment is operating and Containment pressure is above the actuation setpoint, the threshold is met.

This IC addresses a failure of one or more containment penetrations to automatically isolate (close) when required by an actuation signal. It also addresses an event that results in high containment pressure with a concurrent failure of containment pressure control systems.

Absent challenges to another fission product barrier, either condition represents potential degradation of the level of safety of the plant.

For the first threshold, 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 OFNs and EMGs. The 15-minute criterion is included to allow operators time to manually isolate the required penetrations, if possible.

The second threshold 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.

WCGS Basis Reference(s):

1. FSAR Section 6.2.2

2. FSAR Section 6.2.2.1.2.1

3. FSAR Section 6.2.2.2.2

4. EMG F-0 Critical Safety Function Status Trees. (CSFST) Fig Lire 6, Containment

5. EMG FR-Z1 Response to High Containment Pressure

6. Technical Specifications Table 3.3.2-1

7. Technical Specifications 83.6.6

8. NEI 99-01 SU?

Page 171 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 9 - Hazardous Event Affecting Safety Systems Initiating Condition: Hazardous event affecting SAFETY SYSTEMS needed for the current operating mode EAL:

SA9.1 Alert The occurrence of any Table S-5 hazardous event AND Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode AND EITHER:

• Event damage has caused indications of degraded performance to a second train of a SAFETY SYSTEM needed for the current operating mode

• Event damage has resulted in VISIBLE DAMAGE to a second train of a SAFETY SYSTEM needed for the current operating mode (Notes 11, 12)

Note 11: If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then emergency classification is not warranted.

Note 12: If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted.

Table 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 Manager 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, l Page 172 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases 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 evidence of heat are observed.

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

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

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

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

VISIBLE DAMAGE - Damage to a SAFETY SYSTEM train that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected SAFETY SYSTEM train ..

Basis:

This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the current operating mode. In order to provide the appropriate context for consideration of an ALERT classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train, and there must be either indications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance, and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE such that the potential exists for performance issues. Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has indications of degraded performance; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events.

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

VISIBLE DAMAGE addresses damage to a SAFETY SYSTEM train that is not in service/operation and that potentially could cause performance issues. Operators will make this determination based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantificati.on of the Page 173 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases damage. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.

Escalation of the emergency classification level would be via IC FS 1 or RS 1.

WCGS Basis Reference(s):

1. NEI 99-01 SA9 Page 174 of 228 INFORMATION USE

ATTACHMENT 1 EAL Bases Category F - Fission Product Barrier Degradation EAL Group: Hot Conditions (RCS temperature > 200°F); EALs in this category are applicable only in one or more hot operating modes.

EALs in this category represent threats to the defense in depth design concept that precludes the release of highly radioactive fission products to the environment. This concept relies on multiple physical barriers any one of which, if maintained intact, precludes the release of signific.ant 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 (CMT): The Containment Barrier includes the containment building and connections up to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the containment building up to and including the outermost secondary side isolation valve.

Containment Barrier thresholds are used as criteria for escalation of the ECL from Alert to a Site Area Emergency or a General Emergency.

The EALs in this category require evaluation of the loss and potential loss thresholds listed in the fission product barrier matrix of Table F-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 potential loss of either Fuel Clad or RCS Site Area Emergency:

Loss or potential loss of any two barriers General Emergency:

Loss of any*two barriers and loss or potential loss of third barrier The logic used for emergency classification based on fission product barrier monitoring should reflect the following considerations:

• The Fuel Clad Barrier and the RCS Barrier are weighted more heavily than the Containment Barrier.

• Unusual Event ICs associated with RCS and Fuel Clad Barriers are addressed under System Malfunction ICs.

Page 175 of 228 INFORMATION USE

ATTACHMENT 1 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 WCGS 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 consid~red 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 Manager would have more assurance that there was no immediate need to escalate to a General Emergency.

Page 176 of 228 INFORMATION USE

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:

FA1.1 Alert Any loss or any potential loss of EITHER Fuel Clad or RCS (Table F-1)

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

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 FS1 .1 WCGS Basis Reference(s):

1. NEI 99-01 FA1 Page 177 of 228 INFORMATION US.E

ATTACHMENT 1 EAL Bases Category: Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss or potential loss of any two barriers EAL:

FS1 .1 Site Area Emergency Loss or potential loss of any two barriers (Table F-1)

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s ):

None 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 Manager would have greater assurance that escalation to a General Emergency is less imminent.

WCGS Basis Reference(s):

1. NEI 99-01 FS1 Page 178 of 228 INFORMATION USE

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:

FG1.1 General Emergency Loss of any two barriers AND Loss or potential loss of third barrier (Table F-1)

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None 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 WCGS Basis Reference(s):

1. NEI 99-01 FG1 Page 179 of 228 INFORMATION USE

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. GMT Radiation I RCS Activity D. GMT Integrity or Bypass E. Emergency Manager 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.1," the third Containment barrier Potential Loss in Category C would be assigned "GMT 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-1, locates the likely category and then reads across the fission product barrier Loss and Potential Loss thresholds in that category to determine if a threshold has been exceeded. If a threshold has not been exceeded, the EAL-user proceeds to the next likely category and continues review of the thresholds in the new category If the EAL-user determines that any threshold has been exceeded, by definition, the barrier is lost or potentially lost - even if multiple thresholds in the same barrier column are exceeded, only that one barrier is lost or potentially lost. The EAL-user must examine each of the three fission product barriers to determine if other barrier thresholds in the category are lost or 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 180 of 228 INFORMATION USE

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, FS1 .1, and FA 1.1 to determine the appropriate emergency classification.

In the remainder of this Attachment, the Fuel Clad barrier threshold bases appear first, followed by the RCS barrier and finally the Containment barrier threshold bases. In each barrier, the bases are given according category Loss followed by category Potential Loss beginning with Category A, then B, C, D, E.

Page 181 of 228 INFORMATION USE

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 (CMT) Barrier Category Loss Potential Loss Loss Potential Loss Loss Potential Loss

1. RCS leakage > 50 gpm with

1. An automatic or manual letdown isolated due to A ECCS (SI) actuation required by EITHER:

EITHER:

• UNISOLABLE RCS 1. A leaking or RUPTURED SG RCS or None None leakage is FAULTED outside of None

• UNISOLABLE RCS SG Tube

• SG tube leakage containment leakage Leakage

• SG tube RUPTURE 2. CSFST Integrity-RED Path conditions met

1. CSFST Core Cooling-ORANGE Path conditions 1. CSFST Core Cooling-RED B met 1. CSFST Heat Sink-RED Path conditions met Path conditions met Inadequate

1. CSFST Core Cooling- 2. CSFST Heat Sink-RED Path AND None None RED Path conditions met conditions met AND Restoration procedures not Heat Removal AND Heat sink is required effective within 15 min.

(Note 1)

Heat sink is required

1. Containment radiation c > 600 R/hr on GT RE-59 or 1. Containment radiation 1. Containment radiation CMT > 60 R/hr on* > 6,000 R/hr on GT RE-60 None None None Radiation GT RE-59 or GT RE-59 or GT RE-60

/RCS 2. Dose equivalent 1-131 GT RE-60 Activity coolant activity > 300

µCi/gm

1. Containment isolation is required AND EITHER: 1. CSFST Containment-RED Path conditions met

• Containment integrity

2. Containment hydrogen D has been lost based on concentration o:: 4%

Emergency Manager CMT None None None None 3. Containment pressure > 27 judgment Integrity psig with < one full train of or Bypass

• UNISOLABLE pathway from depressurization equipment Containment to the operating per design for environment exists

> 15 min. (Note 1, 9)

2. Indications of RCS leakage outside of Containment

1. Any condition in the opinion 1. Any condition in the opinion 1. Any condition in the opinion 1. Any condition in the opinion 1. Any condition in the opinion 1. Any condition in the opinion E of the Emergency Manager of the Emergency Manager of the Emergency Manager of the Emergency Manager of the Emergency Manager of the Emergency Manager EC that indicates loss of the that indicates potential loss that indicates loss of the that indicates potential loss of that indicates loss of the that indicates potential loss of Judgment fuel clad barrier of the fuel clad barrier RCS barrier the RCS barrier Containment barrier the Containment barrier Page 182 of 228 INFORMATION USE

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 Page 183 of 228 INFORMATION USE*

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:

None Page 184 of 228 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: B. Inadequate Heat Removal Degradation Threat: Loss Threshold:

1. CSFST Core Cooling-RED Path conditions met Definition(s):

None Basis:

Critical Safety Function Status Tree (CSFST) Core Cooling-RED path indicates significant core exit superheating and core uncovery. The CSFSTs can be monitored using the SPDS display on the NPIS Computer (ref. 1).

This reading indicates temperatures within the core are sufficient to cause significant superheating of reactor coolant.

WCGS Basis Reference(s):

1. EMG F-0 Critical Safety Function Status Trees

2. NEI 99-01 Inadequate Heat Removal Fuel Clad Loss 2.A Page 185 of 228 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: B. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold:

1. CSFST Core Cooling-ORANGE Path conditions met*

Definition(s):

None Basis:

Critical Safety Function Status Tree (CSFST) Core Cooling-ORANGE path indicates subcooling has been lost and that some fuel clad damage may potentially occur. The CSFSTs can be monitored using the SPDS display on the NPIS Computer (ref. 1).

This reading indicates a reduction in reactor vessel water level sufficient to allow the onset of heat-induced cladding damage.

WCGS Basis Reference(s):

1. EMG F-0 Critical Safety Function Status Trees

2. NEI 99-01 Inadequate Heat Removal Fuel Clad Loss 2.A Page 186 of 228 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: B. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold:

2. CSFST Heat Sink-RED Path conditions met AND Heat sink is required Definition(s):

None Basis:

In combination with 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 can be monitored using the SPDS display on the NPIS 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 EOP. For example, EMG FR-H1 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. 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. 2).

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.

• Page 187 of 228 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases WCGS Basis Reference(s):

1. EMG F-0 Critical Safety Function Status Trees Figure 3 Heat Sink

2. EMG FR-H1 Response to Loss of Secondary Heat Sink

3. NEI 99-01 Inadequate Heat Removal Fuel Clad Loss 2.B Page 188 of 228 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: C. CMT Radiation I RCS Activity Degradation Threat: Loss Threshold:

1. Containment radiation > 600 R/hr on GT RE-59 or GT RE-60 Definition(s):

None Basis:

Containment radiation monitor readings greater than 600 R/hr (ref. 1) indicate the release of reactor coolant, with elevated activity(> 300 µCi/gm dose equivalent 1-131) indicative of fuel damage, into the Containment (ref. 1).

Monitors used for this fission product barrier loss threshold are the Containment High Range Radiation Monitors GT RE-59 and GT RE-60 (ref.1 ).

The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals 300 µCi/gm dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to 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 C.1 since it indicates a loss of both the Fuel Clad Barrier and the RCS Barrier.

Note that a combination of the two monitor readings appropriately escalates the emergency classification level to a Site Area Emergency.

WCGS Basis Reference(s):

1. EP-CALC-WCNOC-1602 Containment Radiation EAL Threshold Values

2. NEI 99-01 CMT Radiation I RCS Activity Fuel Clad Loss 3.A Page 189 of 228 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: C. CMT Radiation I RCS Activity Degradation Threat: Loss Threshold:

I

2. Dose equivalent 1-131 coolant activity> 300 µCi/gm Definition(s):

None Basis:

Dose Equivalent Iodine (DEi) is determined by Chemistry procedure CHA RC-004 Gamma Isotopic, Total Curie Content and Dose Equivalent Iodine Determination (ref. 1).

This threshold indicates that RCS radioactivity concentration is greater than 300 µCi/gm dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.

It is recognized that sample collection and analysis of reactor coolant with highly elevated activity levels could require several hours to complete. Nonetheless, a sample-related threshold is included as a backup to other indications.

There is no Potential Loss threshold associated with RCS Activity I Containment Radiation.

WCGS Basis Reference(s):

1. CHA RC-004 Gamma Isotopic, Total Curie Content and Dose Equivalent Iodine Determination

2. NEI 99-01 CMT Radiation I RCS Activity Fuel Clad Loss 3.B Page 190 of 228 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: C. CMT Radiation I RCS Activity Degradation Threat: Potential Loss Threshold:

None Page 191 of 228 INFORMATION USE

l ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: D. CMT Integrity or Bypass Degradation Threat: Loss Threshold:

Page 192 of 228 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: D. GMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:

[None Page 193 of 228 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: E. Emergency Manager Judgment Degradation Threat: Loss Threshold:

1. Any condition in the opinion of the Emergency Manager that indicates loss of the Fuel Clad barrier Definition(s):

None Basis:

This threshold addresses any other factors that are to be used by the Emergency Manager in determining whether the Fuel Clad barrier is lost.

WCGS Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment Fuel Clad Loss 6.A Page 194 of 228 INFORMATION USE ,.

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: E. Emergency Manager Judgment Degradation Threat: Potential Loss Threshold:

1. Any condition in the opinion of the Emergency Manager that indicates potential loss of the Fuel Clad barrier Basis:

This threshold addresses any other factors that are to be used by the Emergency Manager in determining whether the Fuel Clad barrier is potentially lost. The Emergency Manager should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

WCGS Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment Potential Fuel Clad Loss 6.A Page 195 of 228 INFORMATION USE

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

• SGTR Definition(s):

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

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

Basis:

ECCS (SI) actuation is caused by (ref. 1):

• Pressurizer low pressure < 1830 psig

• Steamline low pressure < 615 psig

• Containment high pressure > 3.5 psig 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 RCS Barrier.

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 A.1 will also be met.

WCGS Basis Reference(s):

1. EMG E-0 Reactor Trip or Safety Injection

2. EMG E-3 Steam Generator Tube Rupture

3. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Loss 1.A Page 196 of 228 INFORMATION USE

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. RCS leakage > 50 gpm with letdown isolated due to EITHER:

• UNISOLABLE RCS leakage

• SG tube leakage Definition(s):

None Basis:

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 A.1 will also be met.

WCGS Basis Reference(s):

1. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Potential Loss 1.A Page 197 of 228 INFORMATION USE

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:

The "Potential Loss" threshold is defined by the CSFST Integrity - RED path. CSFST Integrity -

Red Path plant conditions and associated PTS Limit Curve A indicates an extreme challenge to the safety function when plant parameters are to the left of the limit curve following excessive RCS cooldown under pressure (ref. 1, 2).

This condition indicates an extreme challenge to the integrity of the RCS pressure boundary 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).

• WCGS Basis Reference(s):

1. EMG F-0 Critical Safety Function Status Trees

2. EMG FR-P1 Response to Imminent Pressurized Thermal Shock

3. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Potential Loss 1.B Page 198 of 228 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: B. Inadequate Heat Removal Degradation Threat: Loss Threshold:

I None Page 199 of 228 INFORMATION USE

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:

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 can be monitored using the SPDS display on the NPIS 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 EOP. For example, EMG FR-H1 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. 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. 2).

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 B.2; both will be met. This condition warrants a 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.

WCGS Basis Reference(s):

1. EMG F-0 Critical Safety Function Status Trees I Page 200 of 228 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases

2. EMG FR-H1 Response to Loss of Secondary Heat Sink I

3. NEI 99-01 Inadequate Heat Removal RCS Loss 2.B II Page 201 of 228 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: C. CMT Radiation/ RCS Activity Degradation Threat: Loss Threshold:

1. Containment radiation =::- 60 R/hr on GT RE-59 or GT RE-60 Definition(s):

N/A Basis:

Containment radiation monitor readings greater than 60 R/hr (ref. 1) indicate the release of reactor coolant, with Technical Specification allowed spiked coolant activity, into the Containment.

Monitors used for this fission product barrier loss threshold are the Containment High Range Radiation Monitors GT RE-59 and GT RE-60 (ref.1 ).

The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for Fuel Clad Barrier Loss threshold C.1 since it indicates a loss of the RCS Barrier only.

There is no Potential Loss threshold associated with RCS Activity I Containment Radiation.

WCGS Basis Reference(s):

1. EP-CALC-WCNOC-1602 Containment Radiation EAL Threshold Values

2. NEI 99-01 CMT Radiation I RCS Activity RCS Loss 3.A Page 202 of 228 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: C. CMT Radiation/ RCS Activity Degradation Threat: Potential Loss Threshold:

None Page 203 of 228 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: D. GMT Integrity or Bypass Degradation Threat: Loss Threshold:

I None Page 204 of 228 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: D. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:

I None Page 205 of 228 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Pdtential Loss Matrix and Bases Barrier: Reactor Coolant System Category: E. Emergency Manager Judgment Degradation Threat: Loss Threshold:

1. Any condition in the opinion of the Emergency Manager that indicates loss of the RCS barrier Definition(s):

None Basis:

This threshold addresses any other factors that may be used by the Emergency Manager in determining whether the RCS Barrier is lost.

WCGS Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment RCS Loss 6.A Page 206 of 228 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: E. Emergency Manager Judgment Degradation Threat: Potential Loss Threshold:

1. Any condition in the opinion of the Emergency Manager that indicates potential loss of the RCS barrier Definition(s):

None Basis:

This threshold addresses any other factors that may be used by the Emergency Manager in determining whether the RCS Barrier is potentially lost. The Emergency Manager should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

WCGS Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment RCS Potential Loss 6.A

-~I~~~~~~~~~~~-P_a_ge~20_7_o_f_2_2_8~~~~~~-l_N_F_O_R_M_A_T_IO_N~U_S_E_.

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 RUPTURED SG is FAULTED outside of containment Definition(s):

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

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

Basis:

This threshold addresses a RUPTURED Steam Generator (SG) that is also FAULTED outside of containment. The condition of the SG is determined in accordance with the threshold for RCS Loss A.1. This condition represents a bypass of the containment barrier.

FAUL TED is a defined term within the NEI 99-01 methodology; this determination is not necessarily dependent upon entry into, or diagnostic steps within, an EOP. For example, if the pressure in a steam generator is decreasing uncontrollably (part of the FAULTED definition) and the FAULTED steam generator isolation procedure is not entered because EOP user rules are dictating implementation of another procedure to address a higher priority condition, the steam generator is still considered FAULTED for emergency classification purposes.

The FAULTED criterion establishes an appropriate lower bound on the size of a steam release that may require an emergency classification. Steam releases of this size are readily observable with normal Control Room indications. The lower bound for this aspect of the containment barrier is analogous to the lower bound criteria specified in IC SU4 for the fuel clad barrier (i.e., RCS activity values) and IC SUS 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 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 I Page 208 of 228 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases 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 rupture, there may be minor radiological releases through a secondary-side system component (e.g., air ejectors, glad seal exhausters, valve packing, etc.). These types of releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R ICs.

The emergency classification levels resulting from primary-to-secondary leakage, with or without a steam release from the FAULTED SG, are summarized below.

Affected SG is FAULTED Outside of Containment?

P-to-5 Leak Rate Yes No Less than or equal to 25 gpm No classification No classification Greater than 25 gpm Unusual Event per SU5.1 Unusual Event per SU5.1 Requires an automatic or manual Site Area Emergency per ECCS (SI) actuation (RCS Barrier Alert per FA1 .1 FS1.1 Loss)

There is no Potential Loss threshold associated with RCS or SG Tube Leakage.

WCGS Basis Reference(s):

1. EMG E-2 Faulted Steam Generator Isolation

2. EMG E-3 Steam Generator Tube Rupture

3. NEI 99-01 RCS or SG Tube Leakage Containment Loss 1.A Page 209 of 228 INFORMATION USE

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:

I None Page 210 of 228 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: B. Inadequate heat Removal Degradation Threat: Loss Threshold:

I None Page 211 of 228 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Pbtential 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 procedure_s not effective within 15 min. (Note 1)

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

Definition(s ):

None Basis:

Critical Safety Function Status Tree (CSFST) Core Cooling-RED path indicates significant core exit superheating and core uncovery. The CSFSTs can be monitored using the SPDS display on the NPIS 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, 2, 3).

This condition represents an IMMINENT core melt sequence which, if not corrected, could lead to vessel failure and an increased potential for containment failure. For this condition to occur, there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. If implementation of a procedure(s) to restore adequate core cooling is not effective (successful) within 15 minutes, it is assumed that the event trajectory will likely lead to core melting and a subsequent challenge of the Containment Barrier.

The restoration procedure is considered "effective" if core exit thermocouple readings are decreasing and/or if reactor vessel level is increasing. Whether or not the procedure(s) will be effective should be apparent within 15 minutes. The Emergency Manager should escalate the emergency classification level as soon as it is determined that the procedure(s) will not be effective.

Severe accident analyses (e.g., NUREG-1150) have concluded that function restoration procedures can arrest core degradation in .a significant fraction of core damage scenarios, and that the likelihood of containment failure is very small in these events. Given this, it is appropriate to provide 15 minutes beyond the required entry point to determine if procedural actions can reverse the core melt sequence.

WCGS Basis Reference(s):

1. EMG F-0 Critical Safety Function Status Trees Page 212 of 228 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases

2. EMG FR-C1 Response to Inadequate Core Cooling

3. EMG FR-C.2 Response to Degraded Core Cooling

4. NEI 99-01 Inadequate Heat Removal Containment Potential Loss 2.A Page 213 of 228 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: C. CMT Radiation/RCS Activity Degradation Threat: Loss Threshold:

None Page 214 of 228 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: C. CMT Radiation/RCS Activity Degradation Threat: Potential Loss Threshold:

1. Containment radiation > 6,000 R/hr on GT RE-59 or GT RE-60 Definition(s):

None Basis:

Containment radiation monitor readings greater than 6,000 R/hr (ref. 1) indicate the release of reactor coolant, with coolant activity corresponding to 20% clad failure, into the Containment.

Monitors used for this fission product barrier loss threshold are the Containment High Range Radiation Monitors GT RE-59 and GT RE-60 (ref. 1).

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. 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.

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 ldss of containment which would then escalate the emergency classification to a General Emergency.

WCGS Basis Reference(s):

1. EP-CALC-WCNOC-1602 Containment Radiation EAL Threshold Values

2. NEI 99-01 CMT Radiation I RCS Activity Containment Potential Loss 3.A Page 215 of 228 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Pptential Loss Matrix and Bases Barrier: Containment Category: D. CMT Integrity or Bypass Degradation Threat: Loss Threshold:

1. Containment isolation is required AND EITHER:

• Containment integrity has been lost based on Emergency Manager judgment

• UNISOLABLE pathway from containment to the environment exists I*..

Definition(s ):

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

Basis:

The status of the containment barrier during an event involving steam generator tube leakage is assessed using Loss Threshold A.1.

These thresholds address a situation where containment isolation is required and one of two conditions exists as discussed below. Users are reminded that there may be accident and release conditions that simultaneously meet both bulleted thresholds.

First Bulleted 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 Manager will assess this threshold using judgment, and with due consideration given to current plant conditions, and available operational and radiological data (e.g., containment pressure, readings on radiation monitors outside containment, operating status of containment pressure control equipment, etc.).

Refer to the middle piping run of Figure 1. Two simplified examples are provided. One is leakage from a penetration and the other is leakage from an in-service system valve.

Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure.

Another example would be a loss or potential loss of the 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 I Page 216 of 228 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases or potential loss of containment but should be evaluated using the Recognition Category R ICs.

Second Bulleted Threshold - Conditions are such that there is an UNISOLABLE pathway for the migration of radioactive material from the containment atmosphere to the environment. As used here, the term "environment" includes the atmosphere of a room or area, outside the containment, that may, in turn, communicate with the outside-the-plant atmosphere (e.g.,

through discharge of a ventilation system or atmospheric leakage). Depending upon a variety of factors, this condition may or may not be accompanied by a noticeable drop in containment pressure.

Refer to the top piping run of Figure 1. In this simplified example, the inboard and outboard isolation valves remained open after a containment isolation was required (i.e., containment isolation was not successful). There is now an UNISOLABLE pathway from the containment to the environment.

The existence of a filter is not considered in the threshold assessment. Filters do not remove fission product noble gases. In addition, a filter could become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream.

Leakage between two interfacing liquid systems, by itself, does not meet this threshold.

Refer to the bottom piping run of Figure 1. In this simplified example, leakage in an RCP seal cooler is allowing radioactive material to enter the Auxiliary Building. The radioactivity would be detected by the Process Monitor. If there is no leakage from the closed water cooling system to the Auxiliary Building, then no threshold has been met. If the pump developed a leak that allowed steam/water to enter the Auxiliary Building, then the second threshold would be met. Depending upon radiation monitor locations and sensitivities, this leakage could be detected by any of the four monitors depicted in the figure and cause the first threshold to be met as well.

Following the leakage of 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 R ICs.

WCGS Basis Reference(s):

1. NEI 99-01 GMT Integrity or Bypass Containment Loss 4.A Page 217 of 228 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Pbtential Loss Matrix and Bases Barrier: Containment Category: D. CMT Integrity or Bypass Degradation Threat: Loss Threshold:

2. Indications of RCS leakage outside of containment Definition(s):

None Basis:

The status of the containment barrier during an event involving steam generator tube leakage is asses~ed using Loss Threshold A.1.

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 threshold A.1 to be met.

EMG C-12 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, 2, 3):

• Residual Heat Removal

• Safety Injection

• Chemical & Volume Control

• RCP seals

• RCS sample lines 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 1. In this simplified example, a leak has occurred at a reducer on a pipe carrying reactor coolant in the Auxiliary Building. Depending upon radiation monitor locations and sensitivities; the leakage could be detected by any of the four monitors depicted in the figure and cause threshold D.1 to be met as well.

WCGS Basis Reference(s):

Page 218 of 228. INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases

1. EMG C-12 LOCA Outside Containment

2. EMG E-1 Loss of Reactor or Secondary Coolant

3. USAR Section 5.2.5.2 lntersystem Leakage

4. NEI 99-01 GMT Integrity or Bypass Containment Loss Page 219 of 228 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Figure 1: Containment Integrity or Bypass Examples

• * * *2nd * * * ::::::::

. ::: Threshold- :::::: :-

Auxiliary Building Inside Reactor Building Damper Open valve Damper)f t RCP Seal Cooling Page 220 of 228 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: D. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:

1. CSFST Containment-RED path conditions met Definition(s ):

None Basis:

Critical Safety Function Status Tree (CSFST) Containment-RED path is entered if containment pressure is greater than or equal to 60 psig and represents an extreme challenge to the containment barrier. The CSFSTs can be monitored using the SPDS display on the NPIS Computer (ref. 1).

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.

• WCGS Basis Reference(s):

1. BD-EMG F-0 Critical Safety Function Status Trees

2. NEI 99-01 CMT Integrity or Bypass Containment Potential Loss 4.A Page 221 of 228 INFORMATION USE*

ATTACHMENT 2 Fission Product Barrier Loss/Ptltential Loss Matrix and Bases Barrier: Containment Category: D. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:

2. Containment hydrogen concentration ~ 4%

Definition(s):

None Basis:

Following a design basis accident, hydrogen gas may be generated inside the containment by reactions such as zirconium metal with water, corrosion* of materials of construction and radiolysis of aqueous solution in the core and sump. (ref. 1).

WCGS is equipped with a Hydrogen Control System (HCS) which serves to limit or reduce combustible gas concentrations in the Containment. The HCS is an engineered safety feature with redundant hydrogen recombiners, hydrogen mixing system, hydrogen monitoring subsystem, and a backup hydrogen purge subsystem. The HCS is designed to maintain the Containment hydrogen concentration below 4% by volume (ref. 1).

HCS operation is prescribed by EMGs if Containment hydrogen concentration should reach 0.5% by volume (ref. 2). If the Potential Loss threshold is reached or exceeded, the primary means of controlling Containment hydrogen concentration must have failed to perform its design function or has otherwise been inadequate in mitigating the hydrogen generation rate.

For either case, continued hydrogen production may yield a flammable hydrogen concentration and a consequent threat to Containment integrity.

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.

• Two Containment hydrogen monitors (GS Al-10 and GS Al-19) with a range of 0% to 10%

provide indication on Control Room Panel RL020 and NPIS (ref. 1, 3).

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.

• WCGS Basis Reference(s):

1. USAR Section 6.2.5 Combustible Gas Control in Containment

2. EMG FR-C1 Response to Inadequate Core Cooling

3. USAR Section 7.5 Safety-Related Display Instrumentation

4. NEI 99-01 CMT Integrity or Bypass Containment Potential Loss 4.B Page 222 of 228 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: D. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:

3. Containment pressure > 27 psig with < one full train of containment depressurization equipment operating per design for~ 15 min. (Note 1, 9)

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

Note 9: One Containment Spray System train and one Containment Cooling System train comprise one full train of depressurization equipment.

Definition(s):

None Basis:

The Containment Spray System consists of two separate trains of equal capacity, each capable of meeting the design bases requirement. Each train includes a containment spray pump, spray headers, nozzles, valves, and piping. The containment spray additive tank supplies 30 weight percent (nominal) sodium hydroxide to both trains. The refueling water storage tank (RWST) supplies borated water to the Containment Spray System during the injection phase of operation. In the recirculation mode of operation, Containment Spray pump suction is transferred from the RWST to the Containment sumps (ref. 2).

The Containment Cooling System consists of two trains of Containment cooling, each of sufficient capacity to supply 100% of the design cooling requirement. Each train of two fan units is supplied with cooling water from a separate train of essential service water (ESW). Air is drawn into the coolers through the fan and discharged to the steam generator compartments, pressurizer compartment, and instrument tunnel, and outside the secondary shield wall. The air supplied to each steam generator compartment is drawn upwards through the compartments by the hydrogen mixing fans and discharged into the upper elevations of the containment. During normal operation, all four fan units are normally operating. In post-accident operation following an actuation signal, the Containment Cooling System fans are designed to start automatically in slow speed if not already running (ref. 3).

• The Containment pressure setpoint (27 psig, ref. 4, 5, 6) is the pressure at which the equipment should actuate and begin performing its function. The design basis accident analyses and evaluations assume the loss of one Containment Spray System train and one Containment Cooling System train (ref. 7). Consistent with the design requirement, "one full train of depressurization equipment" is therefore defined to be the availability of one train of each system. If less than this equipment is operating and Containment pressure is above the actuation setpoint, the threshold is met.

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 I Page 223 of 228 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases 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.

WCGS Basis Reference(s):

1. USAR Section 6.2.2

2. USAR Section 6.2.2.1.2.1

3. USAR Section 6.2.2.2.2

4. EMG F-0 Critical Safety Function Status Trees (CSFST)

5. EMG FR-Z1 Response to High Containment Pressure

6. Technical Specifications Table 3.3.2-1

7. Technical Specifications B3.6.6

8. NEI 99-01 GMT Integrity or Bypass Containment Potential Loss 4.C Page 224 of 228 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: E. Emergency Manager Judgment Degradation Threat: Loss Threshold:

1. Any condition in the opinion of the Emergency Manager that indicates loss of the Containment barrier Definition(s):

None Basis:

This threshold addresses any other factors that may be used by the Emergency Manager in determining whether the Containment Barrier is lost.

WCGS Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment PC Loss 6.A Page 225 of 228 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/P6tential Loss Matrix and Bases Barrier: Containment Category: E. Emergency Manager Judgment Degradation Threat: Potential Loss Threshold:

1. Any condition in the opinion of the Emergency Manager that indicates potential loss of

.the Containment barrier

• Definition(s):

None Basis:

This threshold addresses any other factors that may be used by the Emergency Manager in determining whether the Containment Barrier is lost. The Emergency Manager should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

WCGS Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment PC Potential Loss 6.A Page 226 of 228 INFORMATION USE

ATTACHMENT 3 Safe Operation & Shutdown Areas Tables R-3 & H-2 Bases

Background

NEI 99-01 Revision 6 ICs AA3 and HAS prescribe declaration of an Alert based on impeded access to rooms or areas (due to either area radiation levels or hazardous gas concentrations) where equipment necessary for normal plant operations, cooldown or shutdown is located.

These areas are intended to be plant operating mode dependent. Specifically the Developers Notes for AA3 and HAS states:

• The "site-specific list of plant rooms or areas with entry-related mode applicability identified" should specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, coo/down and shutdown. Do not include rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations). In addition, the list should specify the plant mode(s) during which entry would be required for each room or area.

The list should not include rooms or areas for which entry is required solely to perform actions of an administrative or record keeping nature (e.g., normal rounds or routine inspections).

Further, as specified in IC HAS:

The list need.not include the Control Room if adequate engineered safety/design features are in place to preclude a Control Room evacuation due to the release of a hazardous gas.

Such features may include, but are not limited to, capability to draw air from multiple air intakes at different and separate locations, inner and outer atmospheric boundaries, or the capability to acquire and maintain positive pressure within the Control Room envelope.

Page 227 of228 INFORMATION USE 1*

ATTACHMENT 3 Safe Operation & Shutdown Areas Tables R-3 & H-2 Bases WCGS 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:

~r Required for~horma1f::

• ~!

WCGS

~.rocedure and

• iS:tep Action Byilding/Elevatioo/Roorri.0.

' . plant operations,

<:, '; M~re ..~

• cooldown or**~.

~*. Step

>~' :§:/' , \, c:,__,-<

shutdown?

GEN 00-005 Chemistry directed to Aux/2000/Sampling Room 3,4,5 Yes - Chemistry Step 6.8.1, 6.25 obtain boron sample sampling requires access and 6.11 to sampling panel GEN 00-006 Isolate Accumulators Aux/2026/ Electrical Pen 3 Yes - for breaker Step 6.18.1 and Rooms operation in electrical Attachment J pen rooms GEN 00-006 Make SI pumps and Control/2000/ESF 4 Yes - NB Breakers must Step 6.22.3 one CCP incapable Switchgear Rooms be racked down in of injection switchgear rooms GEN 00-006 Place RHR in service Aux/2000/Heat Exchanger 4,5 Yes - for breaker Step 6.22.4 and using SYS EJ-120 Rooms operation and low 6.33.2 Aux/2026/Electrical Pen pressure letdown Rooms Table R-3 & H-2 Results Table R-3/H-2 Safe Operation & Shutdown Rooms/Areas Room/Area Mode Applicability North Electrical Pen. Room A 3,4 South Electrical Pen. Room B 3,4 ESF SWGR Room No. 1 (TRN A) 4 ESF SWGR Room No. 2 (TRN B) 4 Auxiliary Building/West Hall Elev 2000 3,4,5 Plant Operating Procedures Reviewed

1. GEN 00-004 - Power Operation

2. GEN 00-005 - Minimum Load to Hot Standby

3. GEN 00-006 - Hot Standby to Cold Shutdown

4. OFN MA-038 - Rapid Plant Shutdown Page 228 of 228 INFORMATION USE