NEI 99-01, Wolf Creek Generating Station Proposed Revised EAL Technical Basis Document

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Wolf Creek Generating Station Proposed Revised EAL Technical Basis Document
ML16279A327
Person / Time
Site: Wolf Creek Wolf Creek Nuclear Operating Corporation icon.png
Issue date: 09/12/2016
From: McCain S
Wolf Creek
To:
Office of Nuclear Reactor Regulation
References
NEI 99-01, Rev 6, WO 16-0045
Download: ML16279A327 (390)
v • d • e


Text

Enclosure II Wolf Creek Generating Station Proposed Revised EAL Technical Basis Document (226 Pages)

APF [XX-XXX-XX]

Revision xxx Draft DS 9/12/16 Page 1of226 INFORMATION USE TABLE OF CONTENTS SECTION PAGE 1.0 PURPOSE ..............................................................................................................................

3 2.0 DISCUSSION

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3 2.1 Background

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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

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6 2.6 Operating Mode Applicability

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8 3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS

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9 3.1 General Considerations

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9 3.2 Classification Methodology

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1 O

4.0 REFERENCES

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13 4.1 Developmental

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13 4.2 Implementing

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13 5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS

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14 6.0 WCGS TO NEI 99-01 Rev. 6 EAL CROSS-REFERENCE

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21 7.0 ATTACHMENTS

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25 1 Emergency Action Level Technical Bases ................................................................

25 . Category R Abnormal Rad Release I Rad Effluent..

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26 Category C Cold Shutdown I Refueling System Malfunction

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60 Category H Hazards ......................................................................................

101 Category S System Malfunction

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135 Category F Fission Product Barrier Degradation

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17 4 2 Fission Product Barrier Loss I Potential Loss Matrix and Bases ....................................................................................................

179 3 Safe Operation

& Shutdown Areas Tables R-3 & H-2 Bases .................................

225 Page 2 of 226 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). 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 making (e.g., the Emergency Manager refers to it during an event), the NRC staff expects that changes to the basis document will be evaluated in accordance with the provisions of 10 CFR 50.54(q).

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 ML 12326A805) (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 226 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 (ECL) from Alert to a Site Area Emergency or a General Emergency 2.3 Fission Product Barrier Classification Criteria The following criteria are the bases for event classification related to fission product barrier loss or potential loss: Alert: Any loss or any potential loss of either Fuel Clad or RCS barrier Site Area 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 226 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 226 INFORMATION USE EAL Groups, Categories and Subcategories EAL Group/Category Any Operating Mode: R -Abnormal Rad Levels I Rad Effluent H -Hazards and Other Conditions Affecting Plant Safety Hot Conditions:

S -System Malfunction I EAL Subcategory 1 -Radiological Effluent 2 -Irradiated Fuel Event 3 -Area Radiation Levels 1 -Security 2 -Seismic Event 3 -Natural or Technological Hazard 4-Fire 5 -Hazardous Gases 6 -Control Room Evacuation 7 -Emergency Manager Judgment 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 Malfunction 1-RCS Level 2 -Loss of Emergency AC Power 3 -RCS Temperature 4 -Loss of Vital DC Power 5 -Loss of Communications 6 -Hazardous Event Affecting Safety Systems 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 226 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. §econd 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. Page 7 of 226 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 Kett;;:: 0.99 and rated thermal power> 5% 2 Startup Kett;;:: 0.99 and rated thermal powers 5% 3 Hot Standby Kett< 0.99 and average reactor coolant temperature;;::

350°F 4 Hot Shutdown Kett < 0.99 and average reactor coolant temperature 350°F > T avg > 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." 5 Cold Shutdown Kett< 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. Page 8 of 226 INFORMATION USE 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 NRC staff has provided guidance on implementing this requirement in NSIR/DPR-ISG-01, "Interim Staff Guidance, Emergency Planning for Nuclear Power Plants" (ref. 4.1.9). 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 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 verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

3.1.3 Imminent Conditions For ICs and EALs that have a stipulated time duration (e.g., 15 minutes, 30 minutes, etc.), the Emergency 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 Page 9 of 226 INFORMATION USE I '-----------------------

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 1 OCFR 50. 72 (ref. 4.1.4). 3.1.5 Classification Based on Analysis The assessment of some EALs is based on the results of analyses that are necessary to ascertain whether a specific EAL threshold has been exceeded (e.g., dose assessments, chemistry sampling, RCS leak rate calculation, etc.). For these EALs, the EAL wording or the associated basis discussion will identify the necessary analysis.

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

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

3.1.6 Emergency 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 For a full discussion of this timing requirement, refer to NSIR/DPR-ISG-01 (ref. 4.1.9). 3.2.1 Classification of Multiple Events and Conditions When multiple emergency events or conditions are present, the user will identify all met or exceeded EALs. The highest applicable ECL identified during this review is declared.

For example: -* If an Alert EAL and a Site Area Emergency EAL are met, a Site Area Emergency should be declared.

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

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

Page 10 of 226 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 I Cs 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 I Cs and/or EALs employ time-based criteria.

These criteria will require that the IC/EAL conditions be present for a defined period of time before an emerg.ency 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 226 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 A TWS 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 the classification should be based on the A TWS only. It is important to stress that the 15-minute emergency classification assessment period (process clock) is not a "grace period" during which a classification may be delayed to allow the performance of a corrective action that would obviate the need to classify the event. Emergency classification assessments must be deliberate and timely, with no undue delays. The provision discussed above addresses only those rapidly evolving situations when an operator is able to take a successful corrective action prior to the Emergency 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 NRC is discussed in NUREG-1022 (ref. 4.1.3). Page 12 of 226 INFORMATION USE

4.0 REFERENCES

4.1 Developmental 4.1.1 NEI 99-01 Revision 6, Methodology for the Development of Emergency Action Levels for Non-Passive Reactors, ADAMS Accession Number ML 12326A805 4.1.2 RIS 2007-02 Clarification of NRG 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 4.1.10 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 226 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 process, or have occurred, which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of hostile action. Any releases are expected to be small fractions of the EPA Protective Action Guideline exposure levels. Containment Closure The procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.

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.

Page 14 of 226 INFORMATION USE

  • I 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 process or have occurred which involve actual or imminent substantial core degradation or melting with potential for loss of containment integrity or hostile actions that result in an actual loss of physical control of the facility.

Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area. Hostage A person(s) held as leverage against the station to ensure that demands will be met by the station. Hostile Action An act toward 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. 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.

Page 15 of 226 INFORMATION USE 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. Safety System A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as related (as defined in 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.

Page 16 of 226 INFORMATION USE Security Condition Any security event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A security condition does not involve a hostile action. Site Area Emergency Events are in process or have occurred which involve actual or likely 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 process or have occurred which indicate a potential degradation in the level of safety of the plant or indicate a security threat to facility protection has been initiated.

No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs. Valid An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

Visible Damage Damage to a component or structure that is readily observable without measurements, testing, or analysis.

The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure.

Damage resulting from an equipment failure and limited to the failed component (i.e., the failure did not cause damage to a structure or any other equipment) is not visible damage. Page 17 of 226 INFORMATION USE 5.2 Abbreviations/Acronyms

°F .......................................................................................................

Degrees Fahrenheit 0 ***************************************************************************************************************************

Degrees AC .......................................................................................................

Alternating Current A lWS ......................................................................

Anticipated Transient Without Scram COE .......................................................................................

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

Code of Federal Regulations CMT ...............................................................................................................

Containment CSFST .......................................................................

Critical Safety Function Status Tree OBA ...............................................................................................

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

Direct Current EAL .............................................................................................

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

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

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 MPC ...................................

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 18 of 226 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 19 of 226 INFORMATION USE 6.0 WCGS-TO-NEI 99-01 Rev. 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 EAL IC Example 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 20 of 226 INFORMATION USE WCGS NEI 99-01 Rev. 6 EAL IC Example 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 21 of 226 INFORMATION USE WCGS NEI 99-01 Rev. 6 EAL IC Example EAL HU4.1 HU4 1 HU4.2 HU4 2 HU4.3 HU4 3 HU4.4 HU4 4 HU7.1 HU7 1 HA1.1 HA1 1, 2 HA5.1 HA5 1 HA6.1 HA6 1 HA7.1 HA7 1 HS1.1 HS1 1 HS6.1 HS6 1 HS7.1 HS7 1 HG7.1 HG7 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 SU7.1 SU6 1, 2, 3 SU8.1 SU7 1, 2 SA1.1 SA1 1 SA3.1 SA2 1 SA6.1 SA5 1 SA9.1 SA9 1 Page 22 of 226 INFORMATION USE WCGS NEI 99-01 Rev. 6 EAL IC Example EAL SS1.1 SS1 1 SS2.1 SSS 1 SS6.1 SSS 1 SG1.1 SG1 1 SG1.2 SGS 1 Page 23 of 226 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 24 of 226 INFORMATION USE 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 25 of 226 INFORMATION USE UI ::i 0 CD UI C'CI (!) "C *3 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 Release Point Monitor GE SAE Alert UE 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 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 Waste Water Treatment O-HF-RE-95 2.09E-3 µCi/ml ------------er System :J Liquid Radwaste Discharge Secondary Liquid Waste System Mode Applicability:

All Definition(s):

None O-HB-RE-18 O-HF-RE-45

1.00E-2 µCi/ml ------------1.00E-2 µCi/ml Page 26 of 226 INFORMATION USE Basis: ATTACHMENT 1 EAL Bases 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 level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release).

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

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

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

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

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

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

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

Releases should not be prorated or averaged.

For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL. This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways.

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 associ.ated with planned batch releases from non-continuous release pathways (e.g., radwaste, waste gas). Escalation of the emergency classification level would be via IC RA1. 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 AU 1 Page 27 of 226 INFORMATION USE Category:

Subcategory:

ATIACHMENT 1 EAL Bases R -Abnormal Rad Levels I Rad Effluent 1 -Radiological Effluent Initiating Condition:

Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer. EAL: RU1.2 Unusual Event Sample analysis for a gaseous or liquid release indicates a concentration or release rate > 2 x ODCM limits 60 min. (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):

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

It includes any gaseous or liquid radiological 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. Page 28 of 226 INFORMATION USE WCGS Basis Reference(s):

ATTACHMENT 1 EAL Bases 1. AP 078-003 Offsite Dose Calculation Manual Section 2. NEI 99-01 AU1 Page 29 of 226 INFORMATION USE Ill :::I 0 QI Ill I'll (!) "C *3 ATIACHMENT 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 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 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 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 Waste Water Treatment O-HF-RE-95 2.09E-3 µCi/ml ------------er System :J Liquid Radwaste Discharge Secondary Liquid Waste System Mode Applicability:

All Definition(s):

None O-HB-RE-18 O-HF-RE-45

1.00E-2 µCi/ml ------------1.00E-2 µCi/ml Page 30 of 226 INFORMATION USE

.

Basis: ATIACHMENT 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:

  • 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 31 of 226 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 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 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. Escalation of the emergency classification level would be via IC RS1. WCGS Basis Reference(s):

1. NEI 99-01 AA 1 Page 32 of 226 INFORMATION USE ATIACHMENT 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: RA1.3 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.

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. Escalation of the emergency classification level would be via IC RS1. Page 33 of 226 INFORMATION USE WCGS Basis Reference(s):

ATTACHMENT 1 EAL Bases 1. AP 078-003 Wolf Creek Offsite Dose Calculation Manual Section 2. NEI 99-01 AA 1 Page 34 of 226 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 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. I Page 35 of 226 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 AA 1 Page 36 of 226 INFORMATION USE Ill :J 0 Ill Ill RI C) "C *s 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: RS1.1 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 RA1 .1, RS1 .1 and RG1 .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

  • Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE Unit Vent (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 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 Waste Water Treatment O-HF-RE-95 2.09E-3 µCi/ml ------------er System :J Liquid Radwaste Discharge Secondary Liquid Waste System Mode Applicability:

All Definition(s):

None O-HB-RE-18 O-HF-RE-45

1.00E-2 µCi/ml ------------1.00E-2 µCi/ml Page 37 of 226 INFORMATION USE Basis: 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:

  • 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 38 of 226 INFORMATION USE ATIACHMENT 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 React9r. 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 39 of 226 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 40 of 226 INFORMATION USE WCGS Basis Reference(s):

1. EPP 06-011, Team Formation
2. NEI 99-01 AS1 ATTACHMENT 1 EAL Bases Page 41 of 226 INFORMATION USE UI :I 0 GI UI ns (!) "C *s 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 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 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 Waste Water Treatment O-HF-RE-95 2.09E-3 µCi/ml ------------C" System :i Liquid Radwaste Discharge Secondary Liquid Waste System Mode Applicability:

All Definition(s):

None O-HB-RE-18

1.00E-2 µCi/ml O-HF-RE-45

1.00E-2 µCi/ml Page 42 of 226 INFORMATION USE Basis: 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 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 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 TEDE and thyroid COE. Classification based on effluent monitor readings assumes that a release path to the environment is established.

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

WCGS Basis Reference(s):

1. EP-CALC-WCNOC-1601 Radiological Effluent EAL Values 2. NEI 99-01 AG1 Page 43 of 226 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: RG1.2 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 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 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 atthe 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 44 of 226 INFORMATION USE ATIACHMENT 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 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 45 of 226 INFORMATION USE WCGS Basis Reference(s):

1. EPP 06-011, Team Formation
2. NEI 99-01 AG1 ATTACHMENT 1 EAL Bases Page 46 of 226 INFORMATION USE ATTACHMENT 1 EAL Bases Category: R -Abnormal Rad Levels I Rad Effluent Subcategory: 2 -Irradiated Fuel Event Initiating Condition:

Unplanned loss of water level above irradiated fuel EAL: RU2.1 Unusual Event UNPLANNED water level drop in the REFUELING PATHWAY as indicated by low water level alarm or indication (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 TableR-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 4 7 of 226 INFORMATION USE Basis: ATTACHMENT 1 EAL Bases 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 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 classification level would be via IC RA2. Page 48 of 226 INFORMATION USE WCGS Basis Reference(s):

1. ALR 00-0760 SFP LEV HI LO ATTACHMENT 1 EAL Bases 2. Technical Specification Section 3.7.15 Fuel Storage Pool Water Level 3. Technical Specification Section 3.9.7 Refueling Pool Water Level 4. OFN KE-018 Fuel Handling Accident 5. NEI 99-01 AU2 Page 49 of 226 INFORMATION USE Category:

Subcategory:

ATTACHMENT 1 EAL Bases R -Abnormal Rad Levels I Rad Effluent 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. This IC applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the Confinement Boundary is classified in accordance with IC EU1. Escalation of the emergency would be based on either Recognition Category R or C I Cs. 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 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. Page 50 of 226 INFORMATION USE WCGS Basis Reference(s):

1. ALR 00-0760 SFP LEV HI LO ATTACHMENT 1 EAL Bases 2. OFN KE-018 Fuel Handling Accident 3. NEI 99-01 AA2 Page 51 of 226 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 I Cs. Page 52 of 226 INFORMATION USE WCGS Basis Reference(s):

ATIACHMENT 1 EAL Bases 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 226 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 I Cs 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 226 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 Ep-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. 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 AS2 Page 55 of 226 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. 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 AG2 Page 56 of 226 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 R, C or F I Cs. WCGS Basis Reference(s):

1. USAR Section 12.3 Table 12.3-2 Area Radiation Monitors 2. NEI 99-01 AA3 Page 57 of 226 INFORMATION USE Category:

Subcategory:

ATTACHMENT 1 EAL Bases R -Abnormal Rad Levels I Rad Effluent 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: 3 -Hot Standby, 4 -Hot Shutdown, 5 -Cold 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).

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 226 INFORMATION USE I ATIACHMENT 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 I Cs. 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 226 INFORMATION USE ATIACHMENT 1 EAL Bases Category C -Cold Shutdown I Refueling System Malfunction EAL Group: Cold Conditions (RCS temperature

5 200°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 226 INFORMATION USE

5. Loss of Communications ATTACHMENT 1 EAL Bases 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 226 INFORMATION USE ATIACHMENT 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 fop 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 226 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 CA1 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 226 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 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
  • Auxiliary Building Sump
  • Liquid Waste Holdup Tank
  • Recycle Holdup Tank
  • CCW Surge Tank Mode Applicability: 5 -Cold Shutdown, 6-Refueling Definition(s):

UNISOLABLE -Ari 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 226 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 CA1 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 226 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: CA1.1 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 CA1 Page 66 of 226 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
  • Auxiliary Building Sump
  • Liquid Waste Holdup Tank
  • Recycle Holdup Tank
  • CCW Surge Tank Mode Applicability: 5 -Cold Shutdown, 6 -Refueling Definition(s):

UNJSOLABLE -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 226 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 CA1 Page 68 of 226 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 226 INFORMATION USE WCGS Basis Reference(s):

ATTACHMENT 1 EAL Bases 1. WCOP-24 EMG/OFN Setpoints

-Setpoint F.36 and F.37 2. NEI 99-01 CS1 Page 70 of 226 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. Page 71 of 226 INFORMATION USE WCGS Basis Reference(s):

ATTACHMENT 1 EAL Bases 1. WCOP-24 EMG/OFN Setpoints

-Setpoint F.37 2. NEI 99-01 CS1 Page 72 of 226 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 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 Tank:;"" "

  • Containment Normal Sump
  • 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 226 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 o_f 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. Page 7 4 of 226 INFORMATION USE WCGS Basis Reference(s):

1. OFN BB-007 RCS Leakage High ATTACHMENT 1 EAL Bases 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 226 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%

30 min. (Note 1) AND Any Containment Ch_allenge 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.

Mode Applicability: 5 -Cold Shutdown Definition(s):

Table C-2 Containment Challenge Indications

  • CONTAINMENT CLOSURE not established (Note 6)
  • UNPLANNED rise in Containment pressure 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.

Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area. Page 76 of 226 INFORMATION USE ATTACHMENT 1 EAL Bases 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 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 226 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.

I Table C-1 Sumps I Tanks

  • Containment Normal Sump
  • 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)

4%

  • UNPLANNED rise in Containment pressure Page 78 of 226 INFORMATION USE Mode Applicability: 5 -Cold Shutdown, 6 -Refueling Definition(s):

ATTACHMENT 1 EAL Bases 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: 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 determinatio.ns (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 226 INFORMATION USE ATIACHMENT 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 cannot be restored, fuel damage is probable.

With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment.

If CONTAINMENT CLOSURE is established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.

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

It therefore represents a challenge to Containment integrity.

In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment.

If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged.

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 226 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 226 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.

Mode Applicability:

Table C-3 AC Power Sources Offsite:

  • ESF XFMR XNB02 Onsite:
  • EOG NE01
  • EOG NE02 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 1 OCFR50.2):

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

Page 82 of 226 INFORMATION USE Basis: ATTACHMENT 1 EAL Bases For emergency classification purposes, "capability" means that an AC power source(s) is available to the emergency buses, whether or not the buses are powered from it. The condition indicated by this EAL is the degradation of the offsite and onsite power sources such that any additional single failure would result in a loss of all AC power to the emergency buses. 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 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. Page 83 of 226 INFORMATION USE ATTACHMENT 1 EAL Bases Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power. The subsequent loss of the remaining single power source would escalate the event to an Alert in accordance with IC CA2. 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 226 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 off site and all on site AC power capability to emergency 4.16KV buses N BO 1 and NB02 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: For emergency classification purposes, "capability" means that an AC power source(s) is available to the emergency buses, whether or not the buses are powered from it. 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. Page 85 of 226 INFORMATION USE ATTACHMENT 1 EAL Bases When in the cold shutdown, refueling, or defueled mode, this condition is not classified as a Site Area Emergency because of the increased time available to restore an emergency bus to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition represents an actual or potential substantial degradation of the level of safety of the plant. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via IC CS 1 or RS 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 CA2 Page 86 of 226 INFORMATION USE AITACHMENT 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 226 INFORMATION USE 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 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. GEN 00-002 Cold Shutdown to Hot Standby 3. USAR Section 7.2.2.3.2
4. NEI 99-01 CU3 Page 88 of 226 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 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
  • 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 Wide Range 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 226 INFORMATION USE ...._ ____________________

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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 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. 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 226 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 not apply during 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 RCS Status CONTAINMENT 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 226 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. Page 92 of 226 INFORMATION USE WCGS Basis ATTACHMENT 1 EAL Bases 1. Wolf Creek Technical Specifications Table 1.1-1 2. FSAR Section 7.2.2.3.2

3. GEN 00-006 Hot Standby to Cold Shutdown 4. NEI 99-01 CA3 Page 93 of 226 INFORMATION USE ATTACHMENT 1 EAL Bases 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): 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). 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 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. Page 94 of 226 INFORMATION USE ATTACHMENT 1 EAL Bases 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 95 of 226 INFORMATION USE ATIACHMENT 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: CU5.1 Unusual Event Loss of all Table C-5 onsite communication methods OR Loss of all Table C-5 offsite communication methods OR Loss of all Table C-5 NRG communication methods Table C-5 Communication Methods System PA system Plant Radios Site Telephone System Local Telephone Company Direct Lines ENS Line -Mode Applicability: 5 -Cold Shutdown, 6 -Refueling, D -Defueled Definition(s):

  • None Basis: Onsite x x x x Offsite x x x NRC x x x. Onsite/offsite/NRC communications include one 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 96 of 226 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 SU7.1. This IC addresses a significant loss of on-site or offsite communications capabilities.

While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to 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 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 97 of 226 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 98 of 226 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 a SAFETY SYSTEM needed for the current operating mode EAL: CA6.1 Alert The occurrence of any Table C-6 hazardous event AND EITHER:

  • Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode
  • The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode Table C-6 Hazardous Events
  • Internal or external FLOODING event
  • FIRE
  • EXPLOSION
  • Other events with similar hazard characteristics as determined by the Emergency Manager Mode Applicability: 5 -Cold Shutdown, 6 -Refueling Definition(s):

EXPLOSION

-A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization.

A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion.

Such events require a 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.

Page 99 of 226 INFORMATION USE

,ATTACHMENT 1 EAL Bases FLOODING -A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area. SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 1 OCFR50.2):

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

VISIBLE DAMAGE -Damage to a component or structure that is readily observable without measurements, testing, or analysis.

The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure.

Damage resulting from an equipment failure and limited to the failed component (i.e., the failure did not cause damage to a structure or any other equipment) is not visible damage. Basis: This IC addresses a hazardous event that causes damage to a Safety System, or a structure containing Safety System components, needed for the current operating mode. This condition significantly reduces the margin to a loss or potential loss of a fission product barrier, and therefore represents an actual or potential substantial degradation of the level of safety of the plant. The first condition addresses damage to a Safety System train that is in service/operation since indications for it will be readily available.

The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the Safety System train. With respect to hazards caused by an equipment failure (e.g., an electrical breaker failure leading to an explosion), no emergency declaration is warranted if the hazard did not cause any damage to another safety system, or another train of the affected safety system. If the hazard resulting from an equipment failure causes damage to another safety system, or another train of the affected safety system (i.e., a system or train that was not the source of the initiating equipment failure), then an emergency declaration is required per this EAL. The second condition addresses damage to a Safety System component that is not in service/operation or readily apparent through indications alone, or to a structure containing Safety System components.

Operators will make this determination based on the totality of available event and damage report information.

This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. Escalation of the emergency classification level would be via IC CS 1 or RS 1. WCGS Basis Reference(s):

1. NEI 99-01 CA6 Page 100 of 226 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 5. 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 226 INFORMATION USE

7. Emergency Manager Judgment ATTACHMENT 1 EAL Bases 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.

Page 1 02 of 226 INFORMATION USE Category:

H -Hazards Subcategory:

1 -Security ATTACHMENT 1 EAL Bases 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 NRC providing information of an aircraft threat Mode Applicability:

All Definition(s):

SECURITY CONDITION

-Any security event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A security condition does not involve a hostile action. HOSTILE ACTION -An act toward 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 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 226 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 226 INFORMATION USE Category:

H -Hazards Subcategory:

1 -Security ATTACHMENT 1 EAL Bases 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 226 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 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 site-specific procedures.

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 Emergency/Rescue, and Spills 4. NEI 99-01 HA1 Page 106 of 226 INFORMATION USE Category:

H -Hazards Subcategory:

1 -Security ATTACHMENT 1 EAL Bases 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 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.

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.

I Page 107of226 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 226 INFORMATION USE ATIACHMENT 1 EAL Bases Category:

H -Hazards and Other Conditions Affecting Plant Safety Subcategory: 2 -Seismic Event Initiating Condition:

Seismic event greater than QBE level EAL: HU2.1 Unusual Event Seismic event > QBE 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, QBE will illuminate if the seismic instrument detects ground motion in excess of the QBE threshold (ref. 2). OFN SG-003, Natural Events provides the guidance for determining if the QBE 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 QBE 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 (QBE). An earthquake greater than an QBE but less than a Safe Shutdown Earthquake (SSE) should have no significant impact on safety-related systems, structures and components; however, some time may be required for the plant staff to ascertain the actual post-event condition of the plant (e.g., performs downs and post-event inspections).

Given the time necessary to perform walk-downs and inspections, and fully understand any impacts, this event represents a potential degradation of the level of safety of the plant. Event verification with external sources should not be necessary during or following an QBE. 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 226 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 226 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 under EAL CA6.1 or SA9.1. A tornado striking (touching down) within the Protected Area warrants declaration of an Unusual Event regardless of the measured wind speed at the meteorological tower. A tornado is defined as a violently rotating column of air in contact with the ground and extending from the base of a thunderstorm.

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

  • Escalation of the emergency classification level would be based on I Cs in Recognition Categories R, F, Sor C. WCGS Basis Reference(s):
1. OFN SG-003 Natural Events 2. NEI 99-01 HU3 Page 111 of 226 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 E'.CCS. These are typically systems classified as safety-related (as defined in 1 OCFR50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to preve!:lt 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 I Cs in Recognition Categories R, F, Sor C. Page 112 of 226 INFORMATION USE

WCGS Basis Reference(s):

1. NEI 99-01 HU3 ATTACHMENT 1 EAL Bases Page 113 of 226 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):

IMPEDE(D)

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

PROTECTED AREA -An area encompassed by physical barriers and to which access is controlled.

The Protected Area refers to the designated security area around the process buildings and is depicted in 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 I Cs in Recognition Categories R, F, Sor C. WCGS Basis Reference(s):

1. NEI 99-01 HU3 Page 114 of 226 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, Sor C.

  • WCGS Basis Reference(s):
1. NEI 99-01 HU3 Page 115 of 226 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 in the same fire area
  • 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.

Mode Applicability:

All Definition(s):

Table H-1 Fire Areas

  • Auxiliary Building
  • Reactor Building
  • Control Building
  • Fuel Building
  • Refueling Water Storage Tank Valve Room 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 in the same fire area 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). I Page 116 of 226 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 of a 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 226 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.

Mode Applicability:

All Definition(s):

Table H-1 Fire Areas

  • Auxiliary Building
  • Reactor Building
  • Control Building
  • Fuel Building
  • Refueling Water Storage Tank Valve Room 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 226 INFORMATION USE ATIACHMENT 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 functioFl 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. Page 119 of 226 INFORMATION USE WCGS Basis Reference(s):

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 226 INFORMATION USE ATIACHMENT 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 226 INFORMATION USE ATIACHMENT 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 226 INFORMATION USE ATTACHMENT 1 EAL Bases Category:

Subcategory: H -Hazards and Other Conditions Affecting Plant Safety 5 -Hazardous Gases Initiating Condition:

Gaseous release IMPEDING access to equipment necessary for normal plant operations, cooldown or shutdown EAL: HAS.1 Alert Release of a toxic, corrosive, asphyxiant or flammable gas 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 11--------

North Electrical Pen. Room A 3, 4 South Electrical Pen. Room B ESF SWGR Room No. 1 (TRN A) ESF SWGR Room No. 2 (TRN B) Auxiliary Building/West Hall Elev 2000 Mode Applicability: . 3 -Hot Standby, 4 -Hot Shutdown, 5 -Cold Shutdown Definition(s):

3,4 4 4 3,4,5 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 necessary 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 226 INFORMATION USE 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 I Cs. 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 99-01 AA3 Page 124 of 226 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 Control 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 a plant shutdown from outside the Control Room, in addition to responding to the event that required the evacuation of the Control Room, will present challenges to plant operators and other on-shift personnel.

Activation of the ERO and emergency response facilities will assist in responding to these challenges.

Escalation of the emergency classification level would be via IC HS6. 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 226 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 226 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 226 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 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 226 INFORMATION USE ATIACHMENT 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. Page 129 of 226 INFORMATION USE WCGS Basis Reference(s):

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 HA7 Page 130 of 226 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.

Page 131 of 226 INFORMATION USE WCGS Basis Reference(s):

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 HS7 Page 132 of 226 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 226 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 HG7 Page 134 of 226 INFORMATION USE Category S -System Malfunction ATTACHMENT 1 EAL Bases 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 anal zed events referred to as Page 135 of 226 INFORMATION USE ATTACHMENT 1 EAL Bases Anticipated Transient Without Scram (ATWS) events. For EAL classification, however, A TWS 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 226 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 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:

Table S-1 AC Power Sour Offsite:

  • ESF XFMR XNB02 Onsite:
  • EOG NE01
  • EOG NE02 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 NBd2 (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 so.urce 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 OGs (ref. 3, 4). Credit can be taken for this source only if they are already aligned because they cannot be aligned within 15 minutes. Page 137 of 226 INFORMATION USE ATIACHMENT 1 EAL Bases 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 226 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.

Mode Applicability:

Table S-1 AC Power Sou Offsite:

  • ESF XFMR XNB02 Onsite:
  • EOG NE01
  • EDG NE02 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 1 OCFR50.2):

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

Page 139 of 226 INFORMATION USE Basis: ATIACHMENT 1 EAL Bases For emergency classification purposes, "capability" means that an AC power source is available to the emergency buses, whether or not the buses are powered from it. The 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 essentialswitchgear 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 die.sel 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 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 226 INFORMATION USE WCGS Basis Reference(s):

ATTACHMENT 1 EAL Bases 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 SA1 Page 141 of 226 INFORMATION USE 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 capability 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: For emergency classification purposes, "capability" means that an AC power source is available to the emergency buses, whether or not the buses are powered from it. The 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. The interval begins when both offsite and onsite AC power capability are lost. This IC addresses a total loss of AC power that compromises the performance of all Safety Systems requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. In addition, fission product barrier monitoring capabilities may be degraded under these Page 142 of 226 INFORMATION USE ATTACHMENT 1 EAL Bases conditions.

This IC represents a condition that involves actual or likely major failures of plant functions needed for the protection of the public. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via I Cs 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 Page 143 of 226 INFORMATION USE ATIACHMENT 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 capability 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 Manage r shou l d declare the event promptly upon determ i n i ng that t i me l i mit has been exceeded , or will l i kely 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 onsite AC power capability 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). For emergency classification purposes , " capability" means that an AC power source is available to the emergency buses , whether or not the buses are powered from it. The 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). Page 144 of 226 INFORMATION USE ATIACHMENT 1 EAL Bases 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. 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 Page 145 of 226 INFORMATION USE ATIACHMENT 1 EAL Bases Category:

S -System Malfunction Subcategory: 1 -Loss of Emergency AC Power Initiating Condition:

Loss of all AC and vital DC power sources for 15 minutes or longer EAL: SG1 .2 General Emergency Loss of all offsite and all onsite AC power capability to emergency 4.16KV buses NB01 and NB02 15 min. AND 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) 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 capability 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. For emergency classification purposes, "capability" means that an AC power source is available to the emergency buses, whether or not buses are powered from it. 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). Page 146 of 226 INFORMATION USE ATIACHMENT 1 EAL Bases 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 1 E 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 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 SG8 Page 147 of 226 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 1 E 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 226 INFORMATION USE WCGS Basis Reference(s):

ATTACHMENT 1 EAL Bases 1. OFN NK-020 Loss of Vital 125 VOC 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 VOC Class 1 E Battery System Sizing, Voltage Drop and Short Circuit Studies 5. NEI 99-01 SSS Page 149 of 226 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 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:

Table S-2 Safety System Parameters

  • Reactor power
  • Core Exit TIC temperature
  • Level in at least one SIG
  • Auxiliary or emergency feed flow in at least one SIG 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 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). 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 226 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 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 IC SA3. WCGS Basis Reference(s):

1. USAR Section 7.5 Safety-Related Display Instrumentation
2. OFN RJ-023 NPIS Malfunctions
3. NEI 99-01 SU2 Page 151 of 226 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.

Mode Applicability:

Table S-2 Safety System Paramete

  • Reactor power
  • Core Exit TIC temperature
  • Level in at least one SIG
  • Auxiliary or emergency feed flow in at least one SIG Table S-3 Significant Transients
  • Runback ;::: 25% thermal power
  • Electrical load rejection

> 25% electrical load

  • ECCS actuation 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. Page 152 of 226 INFORMATION USE Basis: ATTACHMENT 1 EAL Bases 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 canriot 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 Page 153 of 226 INFORMATION USE ATTACHMENT 1 EAL Bases Category:

S -System Malfunction Subcategory:

4 -RCS Activity Initiating Condition:

Reactor coolant activity greater than Technical Specification allowable limits EAL: SU4.1 Unusual Event 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 s 60 µCi/gm Dose Equivalent 1-131 ors 1.0 µCi/gm Dose Equivalent 1-131 for a > 48 hr continuous period. The specific Xe-133 activity is limited to s 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 I Cs FA 1 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 226 INFORMATION USE ATTACHMENT 1 EAL Bases Category:

S -System Malfunction Subcategory: 5 -RCS Leakage Initiating Condition:

RCS leakage for 15 minutes or longer EAL: SU5.1 Unusual Event RCS unidentified or pressure boundary leakage > 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 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 226 INFORMATION USE I AITACHMENT 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. 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 226 INFORMATION USE Category:

Subcategory:

ATTACHMENT 1 EAL Bases S -System Malfunction 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 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 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 226 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 indications that a trip setpoint was exceeded), it may be difficult to determine if the reactor was shut down because of automatic trip or manual actions. If a subsequent review of the trip actuation indications reveals that the automatic trip did not cause the reactor to be shut down, then consideration should be given to evaluating the fuel for potential damage, and the reporting requirements of 50. 72 should be considered for the transient event. This IC addresses a failure of the 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 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 SA5. Depending upon the plant response, escalation is also possible via IC Page 158 of 226 INFORMATION USE ATTACHMENT 1 EAL Bases FA1. Absent the plant conditions needed to meet either IC SA6 or FA1, an Unusual Event declaration is appropriate for this event. A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.

Should a reactor trip 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 226 INFORMATION USE I Category:

Subcategory:

ATIACHMENT 1 EAL Bases S -System Malfunction 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 226 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 plant condition, may lead to the generation of an automatic reactor trip signal. If a subsequent manual or automatic trip is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems. A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip). This action does not include manually driving in control rods or implementation of boron injection strategies.

Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles".

The plant response to the failure of 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 FA1, 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 226 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 -Subcriticality
4. EMG FR-S1 Response to Nuclear Power Generation/ATWS
5. FSAR Section 7.7.1 6. NEI 99-01 SU5 Page 162 of 226 INFORMATION USE Category:

Subcategory:

ATTACHMENT 1 EAL Bases S -System Malfunction 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 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 considered a successful reactor trip. AMSAC automatically initiates auxiliary feedwater and a turbine trip under conditions indicative of an Anticipated Transient Without Scram (A TWS) 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 226 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 s)nce 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 FS1. Absent the plant conditions needed to meet either IC SS6 or FS1, an Alert declaration is appropriate for this event. It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F ICs; however, this IC and EAL are 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 226 INFORMATION USE ATIACHMENT 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 AND All actions to shut down the reactor are not successful as indicated by reactor power 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-S 1 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 226 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-S 1 Response to Nuclear Power Generation/A TWS 5. NEI 99-01 SS5 Page 166 of 226 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 PA system x Plant Radios x Site Telephone System x Local Telephone Company Direct Lines x ENS Line Mode Applicability: 1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown Definition(s):

None Basis: Offsite x x x NRC x x x 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 226 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 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 226 INFORMATION USE WCGS Basis Reference(s):

ATTACHMENT 1 EAL Bases 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 226 INFORMATION USE ATTACHMENT 1 EAL Bases Category:

S -System Malfunction Subcategory: 8 -Containment Failure Initiating Condition:

Failure to isolate containment or loss of containment pressure control. EAL: SU8.1 Unusual Event Any penetration is not isolated within 15 min. of a VALID containment isolation signal OR Containment pressure > 27 psig with < one full train of containment depressurization equipment operating per design for==: 15 min. (Note 9) (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.

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 226 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) Figure 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 SU7 Page 171 of 226 INFORMATION USE ATTACHMENT 1 EAL Bases Category: S -System Malfunction Subcategory: 9 -Hazardous Event Affecting Safety Systems Initiating Condition:

Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode EAL: SA9.1 Alert The occurrence of any Table S-5 hazardous event AND EITHER:

  • Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode
  • The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode Mode Applicability:

Table S-5 Hazardous Events

  • Internal or external FLOODING event
  • FIRE
  • EXPLOSION
  • Other events with similar hazard characteristics as determined by the Emergency Manager 1 -Power Operation, 2 -Startup, 3,. Hot Standby, 4 -Hot Shutdown Definition(s):

EXPLOSION -A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization.

A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion.

Such events require a 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 I Page 172 of 226 INFORMATION USE ATTACHMENT 1 EAL Bases classified as safety-related (as defined in 1 OCFR50.2):

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

VISIBLE DAMAGE -Damage to a component or structure that is readily observable without measurements, testing, or analysis.

The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure.

Damage resulting from an equipment failure and limited to the failed component (i.e., the failure did not cause damage to a structure or any other equipment) is not visible damage. Basis: This IC addresses a hazardous event that causes damage to a Safety System, or a structure containing Safety System components, needed for the current operating mode. This condition significantly reduces the margin to a loss or potential loss of a fission product barrier, and therefore represents an actual or potential substantial degradation of the level of safety of the plant. The first condition addresses damage to a Safety System train that is in service/operation since indications for it will be readily available.

The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the Safety System train. With respect to hazards caused by an equipment failure (e.g., an electrical breaker failure leading to an explosion), no emergency declaration is warranted if the hazard did not cause any damage to another safety system, or another train of the affected safety system. If the hazard resulting from an equipment failure causes damage to another safety system, or another train of the affected safety system (i.e., a system or train that was not the source of the initiating equipment failure), then an emergency declaration is required per this EAL. The second condition addresses damage to a Safety System component that is not in service/operation or readily apparent through indications alone, or to a structure containing safety system components.

Operators will make this determination based on the totality of available event and damage report information.

This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. Escalation of the emergency classification level would be via IC FS1 or RS1. WCGS Basis Reference(s):

1. NEI 99-01 SA9 Page 173 of 226 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 significant amounts of radioactive fission products to the environment.

The primary fission product barriers are: A. Fuel Clad (FC): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets. B. Reactor Coolant System (RCS): The RCS Barrier includes the RCS primary side and its connections up to and including the pressurizer safety and relief valves, and other connections up to and including the primary isolation valves. C. Containment (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 174 of 226 INFORMATION USE ATIACHMENT 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 designed/expected operation of a relief valve is not considered to be RCS leakage.

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

For example, if the Fuel Clad and RCS fission product barriers were both lost, then there should be frequent assessments of containment radioactive inventory and integrity.

Alternatively, if both the Fuel Clad and RCS fission product barriers were potentially lost, the Emergency Manager would have more assurance that there was no immediate need to escalate to a General Emergency.

Page 175 of 226 INFORMATION USE I Category:

Subcategory:

ATTACHMENT 1 EAL Bases Fission Product Barrier Degradation N/A Initiating Condition:

Any loss or any potential loss of either Fuel Clad or RCS EAL: FA1.1 Alert Any loss or any potential loss of EITHER Fuel Clad or RCS (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 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 176 of 226 INFORMATION USE 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 177 of 226 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 178 of 226 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. CMT Radiation I RCS Activity D. CMT 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 "CMT 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 179 of 226 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 180 of 226 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. An automatic or manual A ECCS (SI) actuation required A leaking or RUPTURED SG by EITHER: 1. CSFST I ntegr i ty-RED Path 1. RCS or None None is FAUL TED outside of None SG Tube

  • UNISOLABLE RCS conditions met containment Leakage leakage
  • SG tube RUPTURE 1. CSFST Core Cooling-CSFST Core Cool i ng-RED ORANGE Path conditions
1. B met 1. CSFST Heat Sin k-RED Path Path conditions met 1. CSFST Core Cooling-2. CSFST Heat S i nk-RED Path conditions met AND Inadequate None None Heat RED Path cond i tions met conditions met AND Restoration procedures not Removal AND Heat sink is required effective within 15 min. Heat sink is required (Note 1) 1. Conta i nment radiation c > 600 R/hr on Containment radiation
1. Containment radiation GT RE-59 or 1. CMT GT RE-60 None > 60 R/hr on None None > 6 , 000 R l hr on 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 1. CSFST Conta i nment-RED Path AND EITHER: conditions met D
  • Containment integrity
2. Containment hydrogen has been lost based on Emergency Manager concentration 2' 4% CMT None None None None judgment 3. Containment pressure > 27 Integrity
  • UNISOLABLE pathway from psig with < one full train of or Bypass Containment to the depressurization equipment environment exists operating per des i gn for 2. Indications of RCS leakage > 15 min. (Note 1 , 9) outside of Containment E 1. Any condition i n the opin i on 1. Any condition in the opinion 1. Any condit i on in the opinion 1. Any condit i on in the opinion 1. Any condition in the opinion 1. Any condition i n the opinion 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 c l ad barrier of the fuel clad barr i er RCS barr i er the RCS barrier Conta i nment barrier the Containment barrier Page 181 of 226 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:

I None Page 182 of 226 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:

I None Page 183 of 226 INFORMATION USE ATIACHMENT2 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 184 of 226 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 185 of 226 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 186 of 226 INFORMATION USE ATIACHMENT2 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 187 of 226 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 188 of 226 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:

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. 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 189 of 226 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 190 of 226 INFORMATION USE 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 191 of 226 INFORMATION USE 1 __

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category:

D. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:

Page 192 of 226 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: The Emergency Manager judgment threshold addresses any other factors relevant to determining if the Fuel Clad barrier is lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

  • Imminent barrier degradation exists if the degradation will likely occur within relatively short period of time based on a projection of current safety system performance.

The term "imminent" refers to recognition of the inability to reach safety acceptance criteria before completion of all checks.

  • Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators.

This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.

  • Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Manager should be mindful of the Loss of AC power (Station Blackout) and A TVVS EALs to assure timely emergency classification declarations.

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 193 of 226 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: The Emergency Manager judgment threshold addresses any other factors relevant to determining if the Fuel Clad barrier is potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.
  • Imminent barrier degradation exists if the degradation will likely occur within relatively short period of time based on a projection of current safety system performance.

The term "imminent" refers to recognition of the inability to reach safety acceptance criteria before completion of all checks.

  • Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators.

This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.

  • Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Manager should be mindful of the Loss of AC power (Station Blackout) and A TWS EALs to assure timely emergency classification declarations.

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 194 of 226 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

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

UN/SOLABLE

-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 195 of 226 INFORMATION USE I_ ATIACHMENT2 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. CSFST Integrity-RED Path conditions met Definition(s):

None Basis: The "Potential Loss" threshold is defined by the CSFST Reactor Coolant Integrity

-RED path. CSFST RCS 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 196 of 226 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 197 of 226 INFORMATION USE _I 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. Page 198 of 226 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 2. EMG FR-H1 Response to Loss of Secondary Heat Sink 3. NEI 99-01 Inadequate Heat Removal RCS Loss 2.B Page 199 of 226 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 200 of 226 INFORMATION USE I_ 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 201 of 226 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: Loss Threshold:

Page 202 of 226 INFORMATION USE J 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:

Page 203 of 226 INFORMATION USE '----------------------------------

... ..-J ATIACHMENT2 Fission Product Barrier Loss/Potential 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: ' The Emergency Manager judgment threshold addresses any other factors relevant to determining if the RCS barrier is lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

  • Imminent barrier degradation exists if the degradation will likely occur within relatively short period of time based on a projection of current safety system performance.

The term "imminent" refers to recognition of the inability to reach safety acceptance criteria before completion of all checks.

  • Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators.

This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.

  • Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Manager should be mindful of the Loss of AC power (Station Blackout) and A TWS EALs to assure timely emergency classification declarations.

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 204 of 226 INFORMATION USE J 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: The Emergency Manager judgment threshold addresses any other factors relevant to determining if the RCS barrier is potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

  • Imminent barrier degradation exists if the degradation will likely occur within relatively short period of time based on a projection of current safety system performance.

The term "imminent" refers to recognition of the inability to reach safety acceptance criteria before completion of all checks.

  • Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators.

This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.

  • Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Manager should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.

This threshold addresses any other factors that may be used by the Emergency 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 Page 205 of 226 INFORMATION USE 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 FAUL TED 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. FAULTED is a defined term within the NEI 99-01 methodology; this determination is not necessarily dependent upon entry into, or diagnostic steps within, an EOP. For example, if the pressure in a steam generator is decreasing uncontrollably (part of the FAUL TED 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 FAUL TED 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 FAUL TED 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 206 of 226 INFORMATION USE _J ATIACHMENT2 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 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 FAUL TED SG, are summarized below. Affected SG is FAUL TED Outside of Containment?

P-to-S Leak Rate Less than or equal to 25 gpm Greater than 25 gpm Requires an automatic or manual ECCS (SI) actuation (RCS Barrier Loss) Yes No classification Unusual Event per SU5.1 Site Area Emergency per FS1.1 No No classification Unusual Event per SU5.1 Alert per FA1 .1 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 207 of 226 INFORMATION USE _J 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 208 of 226 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 209 of 226 INFORMATION USE ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category:

B. Inadequate heat Removal Degradation Threat: Potential Loss Threshold:

1. CSFST Core Cooling-RED path conditions met AND Restoration procedures not effective within 15 min. (Note 1) Note 1: The Emergency Manager should declare the event promptly upon determ i n i ng that t i me lim i t has been exceeded , or w i ll likely be e x ceeded. 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 operat i ng 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 i ncreasing (ref. 1 , 2 , 3). This condition represents an IMMINENT core melt sequence which, i f not corrected , could lead to vessel failure and an increased potent i al for containment fai l ure. For this condition to occu r, 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 Barr i e r. 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 with i n 15 m i nutes. 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 restorat i on procedures can arrest core degradat i on in a significant fraction of core damage scenarios , and that the likelihood of containment failure is very sma ll i n 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. Page 210 of 226 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 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 211 of 226 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:

I None Page 212 of 226 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 loss 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 213 of 226 INFORMATION USE ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category:

D. CMT Integrity or Bypass Degradation Threat: Loss Threshold:

1. Containment isolation is required AND EITHER: *.
  • Containment integrity has been lost based on Emergency Manager judgment
  • UNISOLABLE pathway from containment to the environment exists Definition(s):

UNISOLABLE -An open or breached system line that cannot be isolated, remotely or locally. Basis: 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 214 of 226 INFORMATION USE

ATIACHMENT2 Fission Product Barrier Loss/Potential Loss Matrix and Bases or potential loss of containment but should be evaluated using the Recognition Category R I Cs. 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 CMT Integrity or Bypass Containment Loss 4.A Page 215 of 226 INFORMATION USE ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category:

D. CMT Integrity br 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 assessed 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):

  • Safety Injection
  • Chemical & Volume Control
  • 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. Page 216 of 226 INFORMATION USE ATIACHMENT2 Fission Product Barrier Loss/Potential Loss Matrix and Bases WCGS Basis Reference(s): 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 217 of 226 INFORMATION USE Inside Reactor Building Damper RCP Seal Cooling ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Figure 1: Containment Integrity or Bypass Examples Auxiliary Building Page 218 of 226 * * * :2n d. * * * * * . * .
  • Threshold-. *. * .
  • Airborne 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 GMT Integrity or Bypass Containment Potential Loss 4.A Page 219 of 226 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:
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 220 of 226 INFORMATION USE ATIACHMENT2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category:

D. GMT 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 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 221 of 226 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 CMT Integrity or Bypass Containment Potential Loss 4.C Page 222 of 226 INFORMATION USE

, ATIACHMENT2 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: The Emergency Manager judgment threshold addresses any other factors relevant to determining if the Primary Containment barrier is lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

  • Imminent barrier degradation exists if the degradation will likely occur within relatively short period of time based on a projection of current safety system performance.

The term "imminent" refers to recognition of the inability to reach safety acceptance criteria before completion of all checks.

  • Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators.

This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results. *

  • Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Manager should be mindful of the Loss of AC power (Station Blackout) and A TWS EALs to assure timely emergency classification declarations.

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 223 of 226 INFORMATION USE ' J ATIACHMENT2 Fission Product Barrier Loss/Potential 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: The Emergency Manager judgment threshold addresses any other factors relevant to determining if the Primary Containment barrier is potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

  • Imminent barrier degradation exists if the degradation will likely occur within relatively short period of time based on a projection of current safety system performance.

The term "imminent" refers to recognition of the inability to reach safety acceptance criteria before completion of all checks.

  • Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators.

This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.

  • Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Manager should be mindful of the Loss of AC power (Station Blackout) and A TWS EALs to assure timely emergency classification declarations.

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 Potential Loss 6.A Page 224 of 226 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 225 of 226 INFORMATION USE 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:

WCGS Required for normal Procedure and Step Action Building/Elevation/Room Mode(s) plant operations, cooldown or Step 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 226 of 226 INFORMATION USE Enclosure Ill Wolf Creek Generating Station EAL Comparison Matrix (120 Pages)

Wolf Creek Generating Station NEI 99-01 Revision 6 EAL Comparison Matrix [Draft 05 9/13/16]

EAL Comparison Matrix Table of Contents Section Introduction

1 Com pa ri son Matrix Form at -------------------------------------------------------------------------------------------------------------------------------------1 EAL Wording-------------------------------------------------------------------------------------------------------------------------------------------------------1 EAL Emphasis Tech n i q u es-------------------------------------------------------------------------------------------------------------------------------------1 Global Differences------------------------------------------------------------------------------------------------------------------------------------------------2 Differences and Deviations-------------------------------------------------------------------------------------------------------------------------------------3 Category A-Abnormal Rad Levels I Rad Effluents

13 Category C -Cold Shutdown I Refueling System Malfunction

33 Category D -Permanently Defueled Station Malfunction

54 Category E -Events Related to I ndependerit Spent Fuel Storage Installations

56 Category F -Fission Product Barrier Degradation

58 Category H -Hazards and Other Conditions Affecting Plant Safety-------------------------------------------------------------------------------

71 Category S -System Malfunction

91 Table 1 -WCGS EAL Categories/Subcategories

5 Table 2 -NEI I WCGS EAL Identification Cross-Reference

6 Table 3 -Summary of Deviations i of i WCGS EAL Comparison Matrix Introduction This document provides a line-by-line comparison of the Initiating Conditions (ICs), Mode Applicability and Emergency Action Levels (EALs) in NEI 99-01 Rev. 6 Final, Development of Emergency Action Levels for Non-Passive Reactors, ADAMS Accession Number ML 12326A805, and the Wolf Creek Generating Station (WCGS) ICs, Mode Applicability and EALs. This document provides a means of assessing WCGS differences and deviations from the NRC endorsed guidance given in NEI 99-01. Discussion of WCGS EAL bases and lists of source document references are given in the EAL Technical Bases Document.

It is, therefore, advisable to reference the EAL Technical Bases Document for background information while using this document.

Comparison Matrix Format The ICs and EALs discussed in this document are grouped according to NEI 99-01 Recognition Categories.

Within each Recognition Category, the ICs and EALs are listed in tabular format according to the order in which they are given in NEI 99-01. Generally, each row of the comparison matrix provides the following information:

  • NEI EAL/IC identifier
  • NEI EAL/IC wording
  • WCGS EAL/IC identifier
  • Description of any differences or deviations EAL Wording In Section 4.1, NEI recommends the following: "The guidance in NEI 99-01 is not intended to be applied to plants "as-is"; however, developers should attempt to keep their site-specific schemes as close to the generic guidance as possible.

The goal is to meet the intent of the generic Initiating Conditions (I Cs) and Emergency Action Levels (EALs) within the context of site-specific characteristics

-locale, plant design, operating features, terminology, etc. Meeting this goal will result in a shorter and less cumbersome NRC review and approval process, closer alignment with the schemes of other nuclear 1 of 118 WCGS power plant sites and better positioning to adopt future industry-wide scheme enhancements" To assist the Emergency Manager (EM), the WCGS EALs have been written in a clear and concise style (to the extent that the differences from the NEI EAL wording could be reasonably documented and justified).

This supports timely and accurate classification in the tense atmosphere of an emergency event. The EAL differences introduced to reduce reading burden comprise almost all of the differences justified in this document.

EAL Emphasis Techniques Due to the width of the table columns and table formatting constraints in this document, line breaks and indentation may differ slightly from the appearance of comparable wording in the source documents.

NEI 99-01 is the source document for the NEI EALs; the WCGS EAL Technical Bases Document for the WCGS EALs. Development of the WCGS IC/EAL wording has attempted to minimize inconsistencies and apply sound human factors principles.

As a result, differences occur between NEI and WCGS ICs/EALs for these reasons alone. When such difference may-infer a technical difference in the associated NEI IC/EAL, the difference is identified and a justification provided.

The print and paragraph formatting conventions summarized below guide presentation of the WCGS EALs in accordance with the EAL writing criteria.

Space restrictions in the EAL table of this document sometimes override this criteria in cases when following the criteria would introduce undesirable complications in the EAL layout.

  • Upper case-bold print is used for the logic terms AND, OR and EITHER.
  • Bold font is used for certain logic terms, negative terms (not, cannot, etc.), any, all.
  • Upper case print is reserved for defined terms, acronyms, system abbreviations, logic terms (and, or, etc. when not used as a conjunction), annunciator window engravings.
  • Three or more items in a list are normally introduced with "Any of the following

... " or "All of the following

... " Items of the list begin with bullets when a priority or sequence is not inferred.

EAL Comparison Matrix

  • The use of AND/OR logic within the same EAL has been avoided when possible.

When such logic cannot be avoided, indentation and separation of subordinate contingent phrases is employed.

Global Differences The differences listed below generally apply throughout the set of EALs and are not repeated in the Justification sections of this document.

The global differences do not decrease the effectiveness of the intent of NEI 99-01. 1. The NEI phrase "Notification of Unusual Event" has been changed to "Unusual Event" or abbreviated "UE" to reduce EAL-user reading burden. 2. NEI 99-01 IC Example EALs are implemented in separate plant EALs to improve clarity and readability.

For example, NEI lists all IC HU3 Example EALs under one IC. The corresponding WCGS EALs appear as unique EALs (e.g., HU3.1 through HU3.4). 3. Mode applicability identifiers (numbers/letter) modify the NEI 99-01 mode applicability names as follows: 1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown, 5 -Cold Shutdown, 6 -Refueling, D -Defueled.

NEI 99-01 defines Defueled as follows: "Reactor Vessel contains no irradiated fuel (full core off-load during refueling or extended outage)." 4. NEI 99-01 uses the terms greater than, less than, greater than or equal to, etc. in the wording of some example EALs. For consistency and reduce EAL-user reading burden, WCGS has adopted use of boolean symbols in place of the NEI 99-01 text modifiers within the EAL wording. 5. "min." is the standard abbreviation for "minutes" and is used to reduce EAL user reading burden. 6. The term "Emergency Director" has been replaced by "Emergency Manager" consistent with site-specific nomenclature.

7. Wherever the generic bracketed PWR term "reactor vessel/RCS" is provided, WCGS uses the term "RCS" as the site-specific nomenclature.
8. IC/EAL identification:

2 of 118 WCGS

  • NEI Recognition Category A "Abnormal Radiation Levels/ Radiological Effluents" has been changed to Category R "Abnormal Rad Levels I Rad Effluents." The designator "R" is more intuitively associated with radiation (rad) or radiological events. NEI IC designators beginning with "A" have likewise been changed to "R."
  • NEI 99-01 defines the thresholds requiring emergency classification (example EALs) and assigns them to ICs which, in turn, are grouped in "Recognition Categories." WCGS endeavors to optimize the NEI EAL organization and identification scheme to enhance usability of the plant-specific EAL set. To this end, the WCGS IC/EAL scheme includes the following features:
a. Division of the NEI EAL set into three groups: o EALs applicable under all plant operating modes -This group would be reviewed by the EAL-user any time emergency classification is considered.

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

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

b. Within each of the above three groups, assignment of EALs to categories/subcategories

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

Subcategories are used as necessary to further divide the EAL Comparison Matrix EALs of a category into logical sets of possible emergency classification thresholds.

The WCGS EAL categories/subcategories and their relationship to NEI Recognition Categories are listed in Table 1. c. . Unique identification of each EAL -Four characters comprise the EAL identifier as illustrated in Figure 1. Figure 1 -EAL Identifier EAL Identifier xxx.x Category (R, H, S, F, C) 111 Sequential number within category/classification Emergency Classification (G, S, A, U) Subcategory number (1 if no subcategory)

The first character is a letter associated with the category in which the EAL is located. The second character is a letter associated with the emergency classification level (G for General Emergency, S for Site Area Emergency, A for Alert, and U for Notification of Unusual Event). The third character is a number associated with one or more subcategories within a given category.

Subcategories are sequentially numbered beginning with the number "1". If a category does not have a subcategory, this character is assigned the number "1". The fourth character is a number preceded by a period for each EAL within a subcategory.

EALs are sequentially numbered within the emergency classification level of a subcategory beginning with the number "1". The EAL identifier is designed to fulfill the following objectives:

o Uniqueness

-The EAL identifier ensures that there can be no confusion over which EAL is driving the need for emergency classification.

o Speed in locating the EAL of concern -When the EALs are displayed in a matrix format, knowledge of the EAL identifier alone can lead the EAL-user to the location of the EAL within the classification matrix. The identifier conveys the category, 3of118 WCGS subcategory and classification level. This assists ERO responders (who may not be in the same facility as the EM) to find the EAL of concern in a timely manner without the need for a word description of the classification threshold.

o Possible classification upgrade -The category/subcategory/identifier scheme helps the EAL-user find higher emergency classification EALs that may become active if plant conditions worsen. Table 2 lists the WCGS ICs and EALs that correspond to the NEI !Cs/Example EALs when the above EAUIC organization and identification scheme is implemented.

Differences and Deviations In accordance NRC Regulatory Issue Summary (RIS) 2003-18 "Use of Nuclear Energy Institute (NEI) 99-01, Methodology for Development of Emergency Action Levels" Supplements 1 and 2, a difference is an EAL change in which the basis scheme guidance differs in wording but agrees in meaning and intent, such that classification of an event would be the same, whether using the basis scheme guidance or the WCGS EAL. A deviation is an EAL change in which the basis scheme guidance differs in wording and is altered in meaning or intent, such that classification of the event could be different between the basis scheme guidance and the WCGS proposed EAL. Administrative changes that do not actually change the textual content are neither differences nor deviations.

Likewise, any format change that does not alter the wording of the IC or EAL is considered neither a difference nor a deviation.

The following are examples of differences:

  • Choosing the applicable EAL based upon plant type (i.e., BWR vs. PWR).
  • Using a numbering scheme other than that provided in NEI 99-01 that does not change the intent of the overall scheme.
  • Where the NEI 99-01 guidance specifically provides an option to not include an EAL if equipment for the EAL does 11ot exist at WCGS (e.g., automatic real-time dose assessment capability).

EAL Comparison Matrix

  • Pulling information from the bases section up to the actual EAL that does not change the intent of the EAL.
  • Choosing to state ALL Operating Modes are applicable instead of stating N/A, or listing each mode individually under the Abnormal Rad Level/Radiological Effluent and Hazard and Other Conditions Affecting Plant Safety sections.
  • Using synonymous wording (e.g., greater than or equal to vs. at or above, less than or equal vs. at or below, greater than or less than vs. above or below, etc.)
  • Adding WCGS equipment/instrument identification and/or noun names to EALs.
  • Combining like I Cs that are exactly the same but have different operating modes as long as the intent of each IC is maintained and the overall progression of the EAL scheme is not affected.
  • Any change to the IC and/or EAL, and/or basis wording, as stated in NEI 99-01, that does not alter the intent of the IC and/or EAL, i.e., the IC EAL continues to: o Classify at the correct classification level. o Logically integrate with other EALs in the EAL scheme. o Ensure that the resulting EAL scheme is complete (i.e., classifies all potential emergency conditions).

The following are examples of deviations:

  • Use of altered mode applicability.
  • Altering key words or time limits.
  • Changing words of physical reference (protected area, safety-related equipment, etc.).
  • Eliminating an IC. This includes the removal of an IC from the Fission Product Barrier Degradation category as this impacts the logic of Fission Product Barrier ICs.
  • Changing a Fission Product Barrier from a Loss to a Potential Loss or vice-versa.
  • Not using NEI 99-01 definitions as the intent is for all NEI 99-01 users to have a standard set of defined terms as defined in NEI 99-01. 4 of 118 WCGS Differences due to plant types are permissible (BWR or PWR). Verbatim compliance to the wording in NEI 99-01 is not necessary as long as the intent of the defined word is maintained.

Use of the wording provided in NEI 99-01 is encouraged since the intent is for all users to have a standard set of defined terms as defined in NEI 99-01.

  • Any change to the IC and/or EAL, and/or basis wording as stated in NEI 99-01 that does alter the intent of the IC and/or EAL, i.e., the IC and/or EAL: o Does not classify at the classification level consistent with NEI 99-01. o Is not logically integrated with other EALs in the EAL scheme. o Results in an incomplete EAL scheme (i.e., does not classify all potential emergency conditions).

The "Difference/Deviation Justification" columns in the remaining sections of this document identify each difference between the NEI 99-01 IC/EAL wording and the WCGS IC/EAL wording. An explanation that justifies the reason for each difference is then provided.

If the difference is determined to be a deviation, a statement is made to that affect and explanation is given that states why classification may be different from the NEI 99-01 IC/EAL and the reason for its acceptability.

In all cases, however, the differences and deviations do not decrease the effectiveness of the intent of NEI 99-01. A summary list of WCGS EAL deviations from NEI 99-01 is given in Table 3.

EAL Comparison Matrix Table 1 -WCGS EAL Categories/Subcategories Category Group: Any Operating Mode: R -Abnormal Rad Levels/Rad Effluent H -Hazards and Other Conditions Affecting Plant Safety Group: Hot Conditions:

S -System Malfunction F -Fission Product Barrier Group: Cold Conditions:

C -Cold Shutdown/Refueling System Malfunction WCGS EALs I Subcategory 1 -Radiological Effluent 2 -Irradiated Fuel Event 3 -Area Radiation Levels 1 -Security 2 -Seismic Event 3 -Natural or Technological Hazard 4-Fire 5 -Hazardous Gases 6 -Control Room Evacuation 7 -Emergency Manager Judgment 1 -Loss of Emergency AC Power 2 -Loss of Vital DC Power 3 -Loss of Control Room Indications 4-RCS Activity 5-RCS Leakage 6 -RPS Failure 7 -Loss of Communications 8 -Containment Failure 9 -Hazardous Event Affecting Safety Systems None 1 -RCS Level 2 -Loss of Emergency AC Power 3 -RCS Temperature 4 -Loss of Vital DC Power 5 -Loss of Communications 6 -Hazardous Event Affecting Safety Systems 5 of 118 NEI Recognition Category Abnormal Rad Levels/Radiological Effluent ICs/EALs Hazards and Other Conditions Affecting Plant Safety ICs/EALs System Malfunction ICs/EALs Fission Product Barrier ICs/EALs Cold Shutdown./

Refueling System Malfunction ICs/EALs WCGS EAL Comparison Matrix WCGS Table 2 -NEI I WCGS EAL Identification Cross-Reference NEI WCGS IC Example Category and Subcategory EAL EAL AU1 1 R -Abnormal Rad Levels I Rad Effluent, 1 -Radiological Effluent RU1.1 AU1 2 R -Abnormal Rad Levels I Rad Effluent, 1 -Radiological Effluent RU1.1 AU1 3 R -Abnormal Rad Levels I Rad Effluent, 1 -Radiological Effluent RU1.2 AU2 1 R -Abnormal Rad Levels I Rad Effluent, 2 -Irradiated Fuel Event RU2.1 AA1 1 R -Abnormal Rad Levels I Rad Effluent, 1 -Radiological Effluent RA1.1 AA1 2 R -Abnormal Rad Levels I Rad Effluent, 1 -Radiological Effluent RA1.2 AA1 3 R -Abnormal Rad Levels I Rad Effluent, 1 -Radiological Effluent RA1.3 AA1 4 R -Abnormal Rad Levels I Rad Effluent, 1 -Radiological Effluent RA1.4 AA2 1 R -Abnormal Rad Levels I Rad Effluent, 2 -Irradiated Fuel Event RA2.1 AA2 2 R -Abnormal Rad Levels I Rad Effluent, 2 -Irradiated Fuel Event RA2.2 AA2 3 R -Abnormal Rad Levels I Rad Effluent, 2 -Irradiated Fuel Event RA2.3 AA3 1 R -Abnormal Rad Levels I Rad Effluent, 3 -Area Radiation Levels RA3.1 AA3 2 R -Abnormal Rad Levels I Rad Effluent, 3 -Area Radiation Levels RA3.2 AS1 1 R -Abnormal Rad Levels I Rad Effluent, 1 -Radiological Effluent RS1.1 AS1 2

  • R -Abnormal Rad Levels I Rad Effluent, 1 -Radiological Effluent RS1.2 AS1 3 R -Abnormal Rad Levels I Rad Effluent, 1 -Radiological Effluent RS1.3 6 of 118 EAL Comparison Matrix WCGS NEI WCGS IC Example Category and Subcategory EAL EAL AS2 1 R -Abnormal Rad Levels I Rad Effluent, 2 -Irradiated Fuel Event RS2.1 AG1 1 R -Abnormal Rad Levels I Rad Effluent, 1 -Radiological Effluent RG1.1 AG1 2 R -Abnormal Rad Levels I Rad Effluent, 1 -Radiological Effluent RG1.2 AG1 3 R -Abnormal Rad Levels I Rad Effluent, 1 -Radiological Effluent RG1.3 AG2 1 R -Abnormal Rad Levels I Rad Effluent, 2 -Irradiated Fuel Event RG2.1 CU1 1 C -Cold SD/ Refueling System Malfunction, 1 -RCS Level CU1.1 CU1 2 C -Cold SD/ Refueling System Malfunction, 1 -RCS Level CU1.2 CU2 1 C -Cold SD/ Refueling System Malfunction, 2 -Loss of ESF AC Power CU2.1 CU3 1 C -Cold SD/ Refueling System Malfunction, 3 -RCS Temperature CU3.1 CU3 2 C -Cold SD/ Refueling System Malfunction, 3 -RCS Temperature CU3.2 CU4 1 C -Cold SD/ Refueling System Malfunction, 4 -Loss of Vital DC Power CU4.1 CU5 1, 2, 3 C -Cold SD/ Refueling System Malfunction, 5 -Loss of Communications CU5.1 CA1 1 C -Cold SD/ Refueling System Malfunction, 1 -RCS Level CA1.1 CA1 2 C -Cold SD/ Refueling System Malfunction, 1 -RCS Level CA1.2 CA2 1 C -Cold SD/ Refueling System Malfunction, 1 -Loss of ESF AC Power CA2.1 CA3 1, 2 C -Cold SD/ Refueling System Malfunction, 3 -RCS Temperature CA3.1 CA6 1 C -Cold SD/ Refueling System Malfunction, 6 -Hazardous Event Affecting Safety Systems CA6.1 CS1 1 C -Cold SD/ Refueling System Malfunction, 1 -RCS Level CS1.1 7of118 EAL Comparison Matrix WCGS NEI WCGS IC Example Category and Subcategory EAL EAL CS1 2 C -Cold SD/ Refueling System Malfunction, 1 -RCS Level CS1.2 CS1 3 C -Cold SD/ Refueling System Malfunction, 1 -RCS Level CS1.3 CG1 1 C -Cold SD/ Refueling System Malfunction, 1 -RCS Level CG1.1 CG1 2 C -Cold SD/ Refueling System Malfunction, 1 -RCS Level CG1.2 E-HU1 1 NIA NIA FA1 1 F -Fission Product Barrier Degradation FA1.1 FS1 1 F -Fission Product Barrier Degradation FS1.1 FG1 1 F -Fission Product Barrier Degradation FG1.1 HU1 1, 2, 3 H -Hazards and Other Conditions Affecting Plant Safety, 1 -Security HU1.1 HU2 1 H -Hazards and Other Conditions Affecting Plant Safety, 2 -Seismic Event HU2.1 HU3 1 H -Hazards and Other Conditions Affecting Plant Safety, 3 -Natural or Technological Hazard HU3.1 HU3 2 H -Hazards and Other Conditions Affecting Plant Safety, 3 -Natural or Technological Hazard HU3.2 HU3 3 H -Hazards and Other Conditions Affecting Plant Safety, 3 -Natural or Technological Hazard HU3.3 HU3 4 H -Hazards and Other Conditions Affecting Plant Safety, 3 -Natural or Technological Hazard HU3.4 HU3 5 NIA NIA HU4 1 H -Hazards and Other Conditions Affecting Plant Safety, 4 -Fire or Explosion HU4.1 HU4 2 H -Hazards and Other Conditions Affecting Plant Safety, 4 -Fire or Explosion HU4.2 HU4 3 H -Hazards and Other Conditions Affecting Plant Safety, 4 -Fire or Explosion HU4.3 8 of 118 EAL Comparison Matrix WCGS NEI WCGS IC Example Category and Subcategory EAL EAL HU4 4 H -Hazards and Other Conditions Affecting Plant Safety, 4 -Fire or Explosion HU4.4 HU? 1 H -Hazards and Other Conditions Affecting Plant Safety, 7 -Judgment HU7.1 HA1 1, 2 H -Hazards and Other Conditions Affecting Plant Safety, 1 -Security HA1.1 HAS 1 H -Hazards and Other Conditions Affecting Plant Safety, S -Hazardous Gases HAS.1 HA6 1 H -Hazards and Other Conditions Affecting Plant Safety, 6 -Control Room Evacuation HA6.1 HA? 1 H -Hazards and Other Conditions Affecting Plant Safety, 7 -Judgment HA7.1 HS1 1 H -Hazards and Other Conditions Affecting Plant Safety, 1 -Security HS1.1 HS6 1 H -Hazards and Other Conditions Affecting Plant Safety, 6 -Control Room Evacuation HS6.1 HS? 1 H -Hazards and Other Conditions Affecting Plant Safety, 7 -Judgment HS7.1 HG1 1 N/A N/A HG? 2 H -Hazards and Other Conditions Affecting Plant Safety, 7 -Judgment HG7.1 SU1 1 S -System Malfunction, 1 -Loss of Emergency AC Power SU1.1 SU2 1 S -System Malfunction, 3 -Loss of Control Room Indications SU3.1 SU3 1 N/A N/A SU3 2 S -System Malfunction, 4 -RCS Activity SU4.1 SU4 1, 2, 3 S -System Malfunction, S -RCS Leakage SUS.1 SUS 1 S -System Malfunction, 6 -RPS Failure SU6.1 SUS 2 S -System Malfunction, 6 -RPS Failure SU6.2 9 of 118 EAL Comparison Matrix WCGS NEI WCGS IC Example Category and Subcategory EAL EAL SU6 1,2, 3 S -System Malfunction, .7 -Loss of Communications SU7.1 SU? 1, 2 S -System Malfunction, S -Containment Failure SUS.1
  • SA1 1 S -System Malfunction, 1 -Loss of Emergency AC Power SA1.1 SA2 1 S -System Malfunction, 3 -Loss of Control Room Indications SA3.1 SA5 1 S -System Malfunction; 6 -RPS Failure SA6.1 SA9 1 S -Hazardous Event Affecting Safety Systems SA9.1 SS1 1 S -System Malfunction, 1 -Loss of Emergency AC Power SS1.1 SS5 1 S -System Malfunction, 6 -RPS Failure SS6.1 SSS 1 S -System Malfunction, 2 -Loss of Vital DC Power SS2.1 SG1 1 S -System Malfunction, 1 -Loss of Emergency AC Power SG1.1 SGS 2 S -System Malfunction, 1 -Loss of Emergency AC Power SG1.2 10of11 S EAL Comparison Matrix WCGS Table 3 -Summary of Deviations NEI WCGS EAL Description IC Example EAL HG1 1 N/A IC HG1 and associated example EAL are not implemented in the WCGS scheme. There are several other I Cs that are redundant with this IC, and are better suited to ensure timely and effective emergency declarations.

In addition, the development of new spent fuel pool level EALs, as a result of NRC Order EA-12-051, clarified the intended emergency classification level for spent fuel pool level events. This deviation is justified because: 1. Hostile Action in the Protected Area is bounded by ICs HS1 and HS7. Hostile Action resulting in a loss of physical control is bound by EAL HG7, as well as any event that may lead to radiological releases to the public in excess of Environmental Protection Agency (EPA) Protective Action Guides (PAGs). a. If, for whatever reason, the Control Room must be evacuated, and control of safety functions (e.g., reactivity control, core cooling, and RCS heat removal) cannot be reestablished, then IC HS6 would apply, as well as IC HS7 if desired by the EAL decision-maker.

b. Also, as stated above, any event (including Hostile Action) that could reasonably be expected to have a release exceeding EPA PAGs would be bound by IC HG?. . C. From a Hostile Action perspective, I Cs HS1, HS? and HG? are appropriate, and therefore, make this part of HG1 redundant and unnecessary.
d. From a loss of physical control perspective, ICs HS6, HS7 and HG7 are appropriate, and therefore, make this part of HG1 redundant and unnecessary.
2. Any event which causes a loss of spent fuel pool level will be bounded by I Cs AA2, AS2 and AG2, regardless of whether it was based upon a Hostile Action or not, thus making this part of HG1 redundant and unnecessary.
a. An event that leads to a radiological release will be bounded by I Cs AU1, AA 1, AS 1 and AG 1. Events that lead to radiological releases in excess of 11 of 118 EAL Comparison Matrix WCGS NEI WCGS EAL Description IC Example EAL EPA PAGs will be bounded by EALs AG1 and HG7, thus making this part of HG1 redundant and unnecessary.

I Cs AA2, AS2, AG2, AS1, AG1, HS1, HS6, HS7 and HG7 have been implemented consistent with NEI 99-01 Revision 6 and thus HG1 is adequately bounded as described above. This is an acceptable deviation from the generic NEI 99-01 Revision 6 guidance.

HS6 1 HS6.1 Deleted defueled mode applicability.

Control of the cited safety functions are not critical for a defueled reactor as there is no energy source in the reactor vessel or RCS. The Mode applicability for the reactivity control safety function has been limited to Modes 1, 2, and 3 (hot operating conditions).

In the cold operating modes adequate shutdown margin exists under all conditions.

This is an acceptable deviation from the generic NEI 99-01 Revision 6 guidance.

12 of 118 EAL Comparison Matrix WCGS Category A Abnormal Rad Levels I Radiological Effluent 13 of 118 EAL Comparison Matrix WCGS NEI IC# NEI IC Wording and Mode WCGS WCGS IC Wording and Mode Difference/Deviation Justification Applicability IC#(s) Applicability AU1 Release of gaseous or liquid RU1 Release of gaseous or liquid The WCGS ODCM is the site-specific effluent release radioactivity greater than 2 times radioactivity greater than 2 times the controlling document.

the (site-specific effluent release ODCM limits for 60 minutes or longer controlling document) limits for MODE: All 60 minutes or longer. MODE: All NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Reading on ANY effluent Reading on any Table R-1 effluent Example EALs #1 and #2 have been combined into a single radiation monitor greater than 2 radiation monitor> column "UE" for EAL to simplify presentation.

times the (site-specific effluent <::60 min. The NEI phrase" ... effluent radiation monitor greater than 2 release controlling document) (Notes 1, 2, 3) times the (site-specific effluent release controlling limits for 60 minutes or longer: document)" and "effluent radiation monitor greater than 2 (site-specific monitor list and times the alarm setpoint established by a current radioactivity threshold values corresponding discharge permit" have been replaced with " ... any Table R-1 to 2 times the controlling effluent radiation monitor> column "UE". document limits) UE thresholds for all WCGS continuously monitored gaseous RU1.1 2 Reading on ANY effluent release pathways are listed in Table R-1 to consolidate the radiation monitor greater than 2 information in a single location and, thereby, simplify times the alarm setpoint identification of the thresholds by the EAL user. The values established by a current shown in Table R-1 column "UE", consistent with the NEI radioactivity discharge permit for bases, represent two times the ODCM release limits for both 60 minutes or longer. liquid and gaseous release. EP-CALC-WCNOC-1601

-The UE gaseous and liquid thresholds are derived from the station ODCM setpoint calculations with assumptions and inputs from the ODCM. 3 Sample analysis for a gaseous or RU1.2 Sample analysis for a gaseous or The WCGS ODCM is the site-specific effluent release liquid release indicates a liquid release indicates a concentration controlling document.

concentration or release rate or release rate > 2 x ODCM limits for<:: 14 of 118 l EAL Comparison Matrix WCGS NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# greater than 2 times the (site-60 min. specific effluent release (Notes 1, 2) controlling document) limits for 60 minutes or longer. Notes

  • The Emergency Director N/A Note 1: The Emergency Manager The classification timeliness note has been standardized should declare the Unusual should declare the event across the WCGS EAL scheme by referencing the "time limit" Event promptly upon promptly upon determining specified within the EAL wording. determining that 60 minutes that time limit has been has been exceeded, or will exceeded, or will likely be likely be exceeded.

exceeded.

  • If an ongoing release is Note 2: If an ongoing release is The classification timeliness note has been standardized detected and the release detected and the release across the WCGS EAL scheme by referencing the "time limit" start time is unknown, start time is unknown, specified within the EAL wording. assume that the release assume that the release duration has exceeded 60 duration has exceeded the minutes. specified time limit.
  • If the effluent flow past an Note 3: If the effluent flow past an effluent monitor is known to effluent radiation monitor is None have stopped due to actions known to have stopped, to isolate the release path, indicating that the release then the effluent monitor path is isolated, the effluent reading is no longer valid for monitor reading is no longer classification purposes.

VALi D for classification purposes.

15 of 118 EAL Comparison Matrix WCGS Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE fl) 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 ::l 0 cu fl) ca 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 Cl 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

  • O-HF-RE-95 2.09E-3 µCi/ml ':i System ------------er :J Liquid Radwaste Discharge O-HB-RE-18

1.00E-2 µCi/ml Secondary Liquid Waste O-HF-RE-45

1.00E-2 µCi/ml System 16 of 118 EAL Comparison Matrix WCGS NEI IC# NEI IC Wording and Mode WCGS WCGS IC Wording and Mode Difference/Deviation Justification Applicability IC#(s) Applicability AU2 UNPLANNED loss of water level RU2 Unplanned loss of water level above None above irradiated fuel. irradiated fuel MODE: All MODE: All NEI Ex. NEI EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 a. UNPLANNED water level RU2.1 UNPLANNED water level drop in the Site-specific area radiation monitors are listed in Table R-2. drop in the REFUELING REFUELING PATHWAY as indicated PATHWAY as indicated by by low water level alarm or indication ANY of the following: (site-specific level (EC Ll-39A, EC Ll-398, EC LIT-39, indications).

local observation of SFP level) AND AND b. UNPLANNED rise in area UNPLANNED rise in corresponding radiation levels as indicated by ANY of the following area radiation levels as indicated by radiation monitors.

any Table R-2 radiation monitors (site-specific list of area radiation monitors) 17 of 118 EAL Comparison Matrix WCGS 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 18 of 118 EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) AA1 Release of gaseous or liquid RA1 Release of gaseous or liquid None radioactivity resulting in offsite radioactivity resulting in offsite dose dose greater than 10 mrem TEDE greater than 10 mrem TEDE or 50 or 50 mrem thyroid COE. mrem thyroid COE MODE: All MODE: All NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Reading on ANY of the following RA1.1 Reading on any Table R-1 effluent The WCGS radiation monitors that detect radioactivity radiation monitors greater than radiation monitor > column "ALERT" effluent release to the environment are listed in Table R-1. the reading shown for 15 for ;:: 15 min. (Notes 1, 2, 3, 4) UE, Alert, SAE and GE thresholds for all WCGS minutes or longer: continuously monitored gaseous and liquid release pathways (site-specific monitor list and are listed in Table R-1 to consolidate the information in a single location and, thereby, simplify identification of the threshold values) thresholds by the EAL-user.

EP-CALC-WCNOC-1601

-The GE, SAE and Alert gaseous thresholds are derived from the EDCP dose assessment model with assumptions and inputs from the USAR. 2 Dose assessment using actual RA1.2 Dose assessment using actual The site boundary is the site-specific receptor point. meteorology indicates doses meteorology indicates doses > 10 greater than 10 mrem TEDE or mrem TEDE or 50 mrem thyroid COE 50 mrem thyroid COE at or at or beyond the SITE BOUNDARY beyond (site-specific dose (Notes 4) receptor point). 3 Analysis of a liquid effluent RA1.3 Analysis of a liquid effluent sample The site boundary is the site-specific receptor point. sample indicates a concentration indicates a concentration or release or release rate that would result rate that would result in doses > 1 O in doses greater than 1 O mrem mrem TEDE or 50 mrem thyroid COE TEDE or 50 mrem thyroid COE at or beyond the SITE BOUNDARY at or beyond (site-specific dose for 60 min. of exposure (Notes 1, 2) receptor point) for one hour of 19of118 EAL Comparison Matrix WCGS exposure.

4 Field survey results indicate RA1.4 Field survey results indicate EITHER The site boundary is the site-specific receptor point. EITHER of the following at or beyond (site-specific dose of the following at or beyond the SITE receptor point): BOUNDARY:

  • Closed window dose rates
  • Closed window dose rates > 1 O greater than 10 mR/hr mR/hr expected to continue for expected to continue for 60 ;:::50 min. minutes or longer.
  • Analyses of field survey
  • Analyses of field survey samples indicate thyroid samples indicate thyroid COE COE greater than 50 mrem > 50 mrem for 60 min. of for one hour of inhalation.

inhalation. (Notes 1, 2) Notes

  • The Emergency Director N/A Note 1: The Emergency Manager The classification timeliness note has been standardized should declare the Alert v should declare the event across the WCGS EAL scheme by referencing the "time promptly upon determining promptly upon determining limit" specified within the EAL wording. that the applicable time has that time limit has been been exceeded, or will likely exceeded, or will likely be be exceeded.

exceeded.

  • If an ongoing release is Note 2: If an ongoing release is The classification timeliness note has been standardized detected and the release detected and the release across the WCGS EAL scheme by referencing the "time start time is unknown, start time is unknown, limit" specified within the EAL wording. assume that the release assume that the release duration has exceeded 15 duration has exceeded the minutes. specified time limit.
  • If the effluent flow past an Note 3: If the effluent flow past an effluent monitor is known to effluent radiation monitor is None have stopped due to actions known to have stopped, to isolate the release path, indicating that the release then the effluent monitor path is isolated, the effluent reading is no longer valid for monitor reading is no longer classification purposes.

VALID for classification purposes.

  • The pre-calculated effluent Note 4 The pre-calculated effluent Incorporated site-specific EAL numbers associated with monitor values presented in monitor values presented in generic EAL#1. EAL #1 should be used for EALs RA1.1, RS1.1 and 20 of 118 EAL Comparison Matrix WCGS emergency classification RG1 .1 should be used for assessments until the results emergency classification from a dose assessment assessments until the using actual meteorology are results from a dose available.

assessment using actual meteorology are available.

21 of 118 EAL Comparison Matrix NEI IC# NEI IC Wording AA2 Significant lowering of water level above, or damage to, irradiated fuel. MODE: All NEI Ex. NEI Example EAL Wording EAL# 1 Uncovery of irradiated fuel in the REFUELING PATHWAY. 2 Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by ANY of the following radiation monitors: (site-specific listing of radiation monitors, and the associated readings, setpoints and/or alarms) 3 Lowering of spent fuel pool level to (site-specific Level 2 value). [See Developer Notes] WCGS IC#(s) RA2 WCGS EAL# RA2.1 RA2.2 RA2.3 WCGS IC Wording Significant lowering of water level above, or damage to, irradiated fuel MODE: All WCGS EAL Wording Uncovery of irradiated fuel in the REFUELING PATHWAY 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) Lowering of spent fuel pool level to 120 in. on EC-Ll-0059 or 0060 (Leve12) 22of118 WCGS Difference/Deviation Justification None Difference/Deviation Justification None Added the word "Mechanical" to EAL statement consistent with the generic bases that states: "This EAL addresses a release of radioactive material caused by mechanical damage to irradiated fuef' Site-specific list of radiation monitors was bulletized.

The bases for the ventilation radiation and Fuel Building and Containment area radiation hi-hi alarms are a spent fuel handling accident.

Post-Fukushima order EA-12-051 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

.

EAL Comparison Matrix WCGS (Level 2) and SFP level at the top of the fuel racks (Level 3). For Wolf Creek SFP Level 2 is a reading of 120 in. on EC-LI-0059 or 0060 (1 O ft. above the top of the spent fuel racks). 23 of 118 EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) AA3 Radiation levels that impede RA3 Radiation levels that IMPEDE access Mode applicability of RA3.2 limited to the modes determined access to equipment necessary to equipment necessary for normal to be applicable per Table R-3. See RA3.2 bases and for normal plant operations, plant operations, cooldown or Attachment 3 of the Technical Bases Document.

cooldown or shutdown shutdown MODE: All MODE: All except for RA3.2 applicable in only modes 1, 2, 3 and 4. NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Dose rate greater than 15 mR/hr RA3.1 Dose rate > 15 mR/hr in EITHER of No other site-specific areas requiring continuous occupancy in ANY of the following areas: the following areas: exist at WCGS.

  • Control Room
  • Control Room (SD-RE-33)

SD-RE-33 is the installed CR ARM.

  • Central Alarm Station
  • Central Alarm Station (by The CAS does not have installed area radiation monitoring
  • (other site-specific survey) and thus must be determined by survey. areas/rooms) 2 An UNPLANNED event results RA3.2 An UNPLANNED event results in Analysis by Operations Subject Matter Experts determined in radiation levels that prohibit or radiation levels that prohibit or areas external to the Main Control Room that require access impede access to any of the IMPEDE access to any Table R-3 in any operating mode to maintain safe plant operations or to following plant rooms or areas: rooms or areas (Note 5) perform a normal plant shutdown and cooldown to Cold (site-specific list of plant rooms Shutdown conditions.

Table R-3 and Attachment 3 of the or areas with entry-related mode Technical Bases Document.

applicability identified)

Note If the equipment in the listed N/A Note 5: If the equipment in the listed None room or area was already room or area was already inoperable or out-of-service inoperable or out-of-service before the event occurred, then before the event occurred, no emergency classification is then no emergency warranted.

classification is warranted.

24 of 118 EAL Comparison Matrix WCGS Table R-3 Safe Operation

& Shutdown Rooms/Areas 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 Auxiliar Buildin /West Hall Elev 2000 3,4,5 25of118 EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) AS1 Release of gaseous radioactivity RS1 Release of gaseous radioactivity None resulting in offsite dose greater resulting in offsite dose greater than than 100 mrem TEDE or 500 100 mrem TEDE or 500 mrem thyroid mrem thyroid COE COE MODE: All MODE: All MODE: All NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Reading on ANY of the following RS1.1 Reading on any Table R-1 effluent The WCGS radiation monitors that detect radioactivity effluent radiation monitors greater than radiation monitor > column "SAE" for release to the environment are listed in Table R-1. UE, Alert, the reading shown for 15 2! 15 min. SAE and GE thresholds for all WCGS continuously monitored minutes or longer: (Notes 1, 2, 3, 4)

  • gaseous and liquid release pathways are listed in Table R-1 to (site-specific monitor list and consolidate the information in a single location and, thereby, threshold values) simplify identification of the thresholds by the EAL-user.

EP-CALC-WCNOC-1601 -The GE, SAE and Alert gaseous thresholds are derived from the EDCP dose assessment model with assumptions and inputs from the USAR. 2 Dose assessment using actual RS1.2 Dose assessment using actual The site boundary is the site-specific receptor point. meteorology indicates doses meteorology indicates doses > 100 greater than 100 mrem TEDE or mrem TEDE or 500 mrem thyroid COE 500 mrem thyroid COE at or at or beyond the SITE BOUNDARY beyond (site-specific dose (Notes 4) receptor point) 3 Field survey results indicate RS1.3 Field survey results indicate EITHER The site boundary is the site-specific receptor point. EITHER of the following at or of the following at or beyond the SITE beyond (site-specific dose BOUNDARY:

receptor point):

  • Closed window dose rates>
  • Closed window dose rates 100 mR/hr expected to continue greater than 100 mR/hr for 2! 60 min. expected to continue for 60 26 of 118 EAL Comparison Matrix WCGS minutes or longer.
  • Analyses of field survey
  • Analyses of field survey samples indicate thyroid COE > samples indicate thyroid 500 mrem for 60 min. of COE greater than 500 inhalation.

mrem for one hour of (Notes 1, 2) inhalation.

Notes

  • The Emergency Director Note 1: The Emergency Manager The classification timeliness note has been standardized should declare the Site Area Emergency promptly upon should declare the event across the WCGS EAL scheme by referencing the "time limit" determining that the promptly upon determining specified within the EAL wording. applicable time has been that time limit has been exceeded, or will likely be exceeded, or will likely be exceeded.

exceeded.

  • If an ongoing release is Note 2: If an ongoing release is The classification timeliness note has been standardized detected and the release start detected and the release across the WCGS EAL scheme by referencing the "time limit" time is unknown, assume that start time is unknown, specified within the EAL wording. -the release duration has assume that the release exceeded 15 minutes. duration has exceeded the
  • If the effluent flow past an specified time limit. effluent monitor is known to have stopped due to actions Note 3: If the effluent flow past an to isolate the release path, effluent radiation monitor is None then the effluent monitor known to have stopped, reading is no longer valid for indicating that the release classification purposes.

path is isolated, the effluent monitor reading is no longer

  • The pre-calculated effluent VALi D for classification monitor values presented in purposes.

EAL #1 should be used for Note 5 The pre-calculated effluent Incorporated site-specific EAL numbers associated with emergency classification monitor values presented in generic EAL#1. assessments until the results EALs RA1.1, RS1.1 and from a dose assessment RG1 .1 should be used for using actual meteorology are emergency classification available.

assessments until the results from a dose assessment using actual meteorology are available.

27 of 118 EAL Comparison Matrix NEI IC# NEI IC Wording AS2 Spent fuel pool level at specific Level 3 description)

MODE: All NEI Ex. NEI Example EAL Wording EAL# 1 Lowering of spent fuel pool level to (site-specific Level 3 value) WCGS IC#(s) RS2 WCGS EAL# RS2.1 WCGS IC Wording Spent fuel pool level at the top of the fuel racks WCGS EAL Wording Lowering of spent fuel pool level to 15 in. on EC-Ll-0059 or 0060 (Level 3) 28 of 118 l WCGS Difference/Deviation Justification Top of the fuel racks is the site-specific Level 3 description.

Difference/Deviation Justification Post-Fukushima order EA-12-051 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 Wolf Creek SFP Level 3 is a reading of 15 in. on EC-LI-0059 or 0060.

EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) AG1 Release of gaseous radioactivity RG1 Release of gaseous radioactivity None resulting in offsite dose greater resulting in offsite dose greater than 1,000 mrem TEDE or than 1,000 mrem TEDE or 5,000 5,000 mrem thyroid COE. mrem thyroid COE MODE: All MODE: All NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Reading on ANY of the following RG1.1 Reading on any Table R-1 effluent The WCGS radiation monitors that detect radioactivity effluent radiation monitors greater than radiation monitor> column "GE" release to the environment are listed in Table R-1. UE, Alert, SAE the reading shown for 15 for ::::15min.

and GE thresholds for all WCGS continuously monitored gaseous or minutes or longer: (Notes 1, 2, 3, 4) liquid release pathways are listed in Table R-1 to consolidate the (site-specific monitor list and information in a single location and, thereby, simplify identification of threshold values) the thresholds by the EAL-user.

EP-CALC-WCNOC-1601

-The GE, SAE and Alert gaseous thresholds are derived from the EDCP dose assessment model with assumptions and inputs from the USAR. 2 Dose assessment using actual RG1.2 Dose assessment using actual The site boundary is the site-specific receptor point. meteorology indicates doses meteorology indicates doses > greater than 1,000 mrem TEDE 1000 mrem TEDE or or 5,000 mrem thyroid COE at 5000 mrem thyroid COE at or or beyond (site-specific dose beyond the SITE BOUNDARY receptor point). (Note 4) 3 Field survey results indicate RG1.3 Field survey results indicate The site boundary is the site-specific receptor point. EITHER of the following at or EITHER of the following at or beyond (site-specific dose beyond the SITE BOUNDARY:

receptor point):

  • Closed window dose rates >
  • Closed window dose rates 1000 mR/hr expected to greater than 1,000 mR/hr continue for:::: 60 min. expected to continue for 60
  • Analyses of field survey minutes or lonoer. 29of118 EAL Comparison Matrix WCGS *Analyses of field survey samples indicate thyroid CDE samples indicate thyroid CDE > 5000 mrem for 60 min. of greater than 5,000 mrem for inhalation.

one hour of inhalation. (Notes 1, 2) Notes

  • The Emergency Director Note 1: The Emergency The classification timeliness note has been standardized across the should declare the Site Area Manager should declare WCGS EAL scheme by referencing the "time limit" specified within Emergency promptly upon the event promptly upon the EAL wording. determining that the determining that time applicable time has been limit has been exceeded, or will likely be exceeded, or will likely exceeded.

be exceeded.

  • If an ongoing release is Note 2: If an ongoing release is The classification timeliness note has been standardized across the detected and the release detected and the release start time is unknown, start time is unknown, WCGS EAL scheme by referencing the "time limit" specified within assume that the release assume that the release the EAL wording. duration has exceeded 15 duration has exceeded minutes. the specified time limit.
  • If the effluent flow past an Note 3: If the effluent flow past effluent monitor is known to an effluent radiation have stopped due to actions monitor is known to None to isolate the release path, have stopped, indicating then the effluent monitor that the release path is reading is no longer valid for isolated, the effluent classification purposes.

monitor reading is no

  • The pre-calculated effluent longer VALID for monitor values presented in classification purposes.

EAL #1 should be used for Note 5 The pre-calculated emergency classification effluent monitor values assessments until the results presented in EALs Incorporated site-specific EAL numbers associated with generic from a dose assessment RA1.1, RS1.1 and EAL#1. using actual meteorology are RG1 .1 should be used available.

for emergency classification assessments until the results from a dose assessment using actual meteoroloov are 30 of 118 L EAL Comparison Matrix WCGS available.

31 of 118 EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) AG2 Spent fuel pool level cannot be RG2 Spent fuel pool level cannot be Top of the fuel racks is the site-specific Level 3 description.

restored to at least (site-specific restored to at least the top of the fuel Level 3 description) for 60 racks for 60 minutes or longer minutes or longer MODE: All MODE: All NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Spent fuel pool level cannot be RG2.1 Spent fuel pool level cannot be Post-Fukushima order EA-12-051 required the installation of restored to at least (site-specific restored to at least 15 in. (Level 3) on reliable SFP level indication capable of identifying normal level Level 3 value) for 60 minutes or EC-Ll-0059 or 0060 for;:::: 60 min. (Level 1 ), SFP level 10 ft. above the top of the fuel racks longer (Note 1) (Level 2) and SFP level at the top of the fuel racks (Level 3). For Wolf Creek SFP Level 3 is a reading of 15 in. on EC-LI-0059 or 0060. Note The Emergency Director should N/A Note 1: The Emergency Manager The classification timeliness note has been standardized declare the General Emergency should declare the event across the WCGS EAL scheme by referencing the "time limit" promptly upon determining that promptly upon determining specified within the EAL wording. 60 minutes has been exceeded, that time limit has been or will likely be exceeded.

exceeded, or will likely be exceeded.

32of118 L EAL Comparison Matrix WCGS Category C Cold Shutdown I Refueling System Malfunction 33of118 EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording J Difference/Deviation Justification IC#(s) CU1 UNPLANNED loss of (reactor CU1 UNPLANNED loss of RCS Deleted " ... for 15 minutes or longer" as the 15 minute criteria only vessel/RCS

[PWR] or RWC inventory applies to EAL threshold

  1. 1. [BWR]) inventory for 15 minutes MODE: 5 -Cold Shutdown, or longer. Refueling MODE: Cold Shutdown, Refueling NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 UNPLANNED loss of reactor CU1.1 UNPLANNED loss of reactor None coolant results in (reactor coolant results in RCS water c vessel/RCS

[PWR] or RWC level less than a required lower [BWR]) level less than a limit for 2!: 15 min. (Note 1) required lower limit for 15 minutes or longer. 2 a. (Reactor vessel/RCS

[PWR] CU1.2 RCS water level cannot be Table C-1 provides a tabularized list of site-specific applicable or RWC [BWR]) level cannot monitored sumps and tanks. be monitored.

AND EITHER Added bulleted criteria "Visual observation of UNISOLABLE RCS AND

  • UNPLANNED increase in leakage" to include direct observation of RCS leakage. any Table C-1 sump/tank Added " ... due to loss of RCS inventory" consistent with the IC and b. UNPLANNED increase in level due to loss of RCS generic bases. (site-specific sump and/or inventory tank) levels.
  • Visual observation of UNISOLABLE RCS leakage Note The Emergency Director N/A Note 1: The Emergency The classification timeliness note has been standardized across the should declare the Unusual Manager should WCGS EAL scheme by referencing the "time limit" specified within Event promptly upon declare the event the EAL wording. determining that 15 minutes promptly upon has been exceeded, or will determining that time likely be exceeded.

limit has been exceeded, or will likely 34 of 118 EAL Comparison Matrix WCGS be exceeded.

Table C-1 Sumps I Tanks

  • Containment Normal Sump
  • Auxiliary Building Sump
  • Liquid Waste Holdup Tank
  • Recycle Holdup Tank
  • CCW Surge Tank 35of118

--EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) CU2 Loss of all but one AC power CU2 Loss of all but one AC power None source to emergency buses for source to emergency buses for 15 minutes or longer. 15 minutes or longer. MODE: Cold Shutdown, MODE: 6 -Cold Shutdown, Refueling, Defueled Refueling, Defueled NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 a. AC power capability to (site-CU2.1 AC power capability, Table C-3, 4.16KV buses NB01 and NB02 are the emergency buses. specific emergency buses) is to emergency 4.16KV buses Site-specific AC power sources are tabularized in Table C-3. reduced to a single power NB01 and NB02 reduced to a source for 15 minutes or single power source for ?: 15 longer. min. (Note 1) AND AND b. Any additional single power Any additional single power source failure will result in source failure will result in loss of loss of all AC power to all AC power to SAFETY SAFETY SYSTEMS. SYSTEMS Note The Emergency Director should N/A Note 1: The Emergency The classification timeliness note has been standardized across the declare the Unusual Event Manager should WCGS EAL scheme by referencing the "time limit" specified within promptly upon determining that declare the event the EAL wording. 15 minutes has been exceeded, promptly upon or will likely be exceeded.

determining that time limit has been exceeded, or will likely be exceeded.

36 of 118 EAL Comparison Matrix Table C-3 AC Power Sources Off site:

  • ESF XFMR XNB02 Onsite:

-I EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording IC#(s) Difference/Deviation Justification CU3 UNPLANNED increase in RCS CU3 UNPLANNED increase in RCS None temperature temperature MODE: Cold Shutdown, MODE: Cold Shutdown, Refueling Refuel in NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 UNPLANNED increase in RCS CU3.1 UNPLANNED increase in RCS 200°F is the site-specific Tech. Spec. cold shutdown temperature temperature to greater than (site-temperature to > 200°F (Note 10) limit. specific Technical Specification Cited Note 1 O below. cold shutdown temperature limit) 2 Loss of ALL RCS temperature CU3.2 Loss of all RCS temperature and None and (reactor vessel/RCS

[PWR] RCS level indication 15 min. or RWC [BWR]) level indication (Note 1) for 15 minutes or longer. Note The Emergency Director should N/A Note 1: The Emergency The classification timeliness note has been standardized across the declare the Unusual Event Manager should declare WCGS EAL scheme by referencing the "time limit" specified within promptly upon determining that the event promptly upon the EAL wording. 15 minutes has been exceeded, determining that time or will likely be exceeded limit has been exceeded, or will likely be exceeded.

N/A N/A N/A Note 10: Begin monitoring hot Added note to remind end-user that the hot condition EALs become condition EALs applicable once operating mode changes to hot conditions.

concurrently for any new event or condition not related to the loss of decay heat removal. 38 of 118 EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) CU4 Loss of Vital DC power for 15 CU4 Loss of Vital DC power for 15 None minutes or longer. minutes or longer. MODE: Cold Shutdown, MODE: 5 -Cold Shutdown, 6 -Refueling Refueling NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Indicated voltage is less than CU4.1 < 1 05 VDC bus voltage 105 VDC is the site-specific minimum vital DC bus voltage. (site-specific bus voltage value) indications on Technical DC operability requirements are specified in Technical on required Vital DC buses for 15 Specification required 125 VDC Specifications.

minutes or longer. buses for 15 min. (Note 1) Note The Emergency Director should N/A Note 1: The Emergency Manager The classification timeliness note has been standardized across the declare the Unusual Event should declare the event WCGS EAL scheme by referencing the "time limit" specified within promptly upon determining that promptly upon determining the EAL wording. 15 minutes has been exceeded, that time limit has been or will likely be exceeded.

exceeded, or will likely be exceeded.

-39 of 118 EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) CU5 Loss of all onsite or offsite CU5 Loss of all onsite or offsite None communications capabilities.

communications capabilities.

MODE: Cold Shutdown, MODE: Cold Shutdown, Refueling, Defueled Refueling, Defueled NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Loss of ALL of the following CU5.1 Loss of all Table C-5 onsite Example EALs #1, 2 and 3 have been combined into a single communication methods onsite communication methods: OR EAL for simplification of presentation. (site specific list of Loss of all Table C-5 offsite Table C-5 provides a site-specific list of onsite, ORO and NRC communications methods) communication methods communications methods. OR Changes the acronym "ORO" to "offsite" consistent with WC 2 Loss of ALL of the following ORO Loss of all Table C-5 NRC communications methods: communication methods usage. (site specific list of communications methods) 3 Loss of ALL of the following NRC communications methods: (site specific list of communications methods) 40 of 118

....... EAL Comparison Matrix WCGS 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 41 of 118 EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) CA1 Loss of (reactor vessel/RCS CA1 Loss of RCS inventory , None [PWR] or RWC [BWR]) MODE: Cold Shutdown, inventory Refueling MODE: Cold Shutdown, Refueling NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Loss of (reactor vessel/RCS CA1.1 Loss of RCS inventory as 12 in. is the lowest RCS water level for unrestricted operation of an [PWR] or RWC [BWR]) inventory indicated by RCS level< 12 in. RHR pump in the shutdown cooling mode. as indicated by level less than -(site-specific level). 2 a. (Reactor vessel/RCS

[PWR] CA1.2 RCS water level cannot be Table C-1 provides a tabularized list of site-specific applicable sumps or RWC [BWR]) level cannot monitored 15 min. (Note 1) and tanks. be monitored for 15 minutes AND EITHER Added bulleted criteria "Visual observation of UNISOLABLE RCS or longer leakage" to include direct observation of RCS leakage. AND

  • UNPLANNED increase in any Table C-1 Sump I Tank b. UNPLANNED increase in level (site-specific sump and/or
  • Visual observation of tank) levels due to a loss of (reactor vessel/RCS

[PWR] UNISOLABLE RCS leakage or RWC [BWR]) inventory.

Note The Emergency Director should N/A Note 1 :The Emergency The classification timeliness note has been standardized across the declare the Alert promptly upon Manager should declare WCGS EAL scheme by referencing the "time limit" specified within determining that 15 minutes has the event promptly upon the EAL wording. been exceeded, or will likely be determining that time exceeded limit has been exceeded, or will likely be exceeded.

42of118 EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) CA2 Loss of all offsite and all onsite CA2 Loss of all offsite and all onsite None AC power to emergency buses AC power to emergency buses for 15 minutes or longer for 15 minutes or longer. MODE: Cold Shutdown, MODE: Cold Shutdown, Refueling, Defueled Refueling, Defueled NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Loss of ALL offsite and ALL CA2.1 Loss of all offsite and all onsite 4.16KV buses NB01 and NB01 are the site-specific emergency onsite AC Power to (site-specific AC power capability to buses. emergency buses) for 15 emergency 4.16KV buses NB01 minutes or longer. and NB02 for;;:: 15 min. (Note 1) Note The Emergency Director should N/A Note 1: The Emergency The classification timeliness note has been standardized across the declare the Unusual Event Manager should WCGS EAL scheme by referencing the "time limit" specified within promptly upon determining that declare the event the EAL wording. 15 minutes has been exceeded, promptly upon or will likely be exceeded.

determining that time limit has been exceeded, or will likely be exceeded.

43of118 L EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) CA3 Inability to maintain the plant in CA3 Inability to maintain the plant in None cold shutdown.

cold shutdown.

MODE: Cold Shutdown, MODE: Cold Shutdown, Refueling Refueling NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 UNPLANNED increase in RCS UNPLANNED increase in RCS Example EALs #1 and #2 have been combined into a single EAL temperature to greater than temperature to > 200°F for as EAL #2 is the alternative threshold based on a loss of RCS (site-specific Technical

> Table C-4 duration temperature indication.

Specification cold shutdown (Note1,10) 200°F is the site-specific Tech. Spec. cold shutdown temperature temperature limit) for greater OR limit. than the duration specified in the following table. CA3.1 UNPLANNED RCS pressure Table C-3 is the site-specific implementation of the generic RCS increase > 10 psig (This EAL Heat-up Duration Threshold table. 2 UNPLANNED RCS pressure does not apply during water-10 psig is the site-specific pressure increase readable by Control increase greater than (site-solid plant conditions)

Room indications.

specific pressure reading). (This EAL does not apply during Cited Note 10 below. water-solid plant conditions.

[PWR]) Note The Emergency Director should N/A Note 1: The Emergency The classification timeliness note has been standardized across the declare the Unusual Event Manager should WCGS EAL scheme by referencing the "time limit" specified within promptly upon determining that declare the event the EAL wording. 15 minutes has been exceeded, promptly upon or will likely be exceeded.

determining that time limit has been exceeded, or will likely be exceeded.

44of118 EAL Comparison Matrix WCGS N/A N/A N/A Note 10: Begin monitoring hot Added note to remind end-user that the hot condition EALs become condition EALs applicable once operating mode changes to hot conditions.

concurrently for any new event or condition not related to the loss of decay heat removal. Table: RCS Heat-up Duration Thresholds RCS Status Containment Closure Status Heat-up Duration Intact (but not at reduced Not applicable 60 minutes* inventory

[PWR]) Not intact (or at reduced Established 20 minutes* inventory

[PWR]) Not Established O minutes

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

Table C-4: RCS Heat-up Duration Thresholds RCS Status CONTAINMENT CLOSURE Heat-up Duration Status Intact (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.

45 of 118 EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) CA6 Hazardous event affecting a CA6 Hazardous event affecting a None SAFETY SYSTEM needed for SAFETY SYSTEM needed for the current operating mode. the current operating mode. MODE: Cold Shutdown, MODE: Cold Shutdown, Refueling Refueling 46 of 118 EAL Comparison Matrix WCGS NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 a. The occurrence of ANY of CA6.1 The occurrence of any Table The hazardous events have been tabularized in Table C-6. the following hazardous C-6 hazardous event Replaced "Shift Manager" with "Emergency Manager" as the EC events:

  • Event damage has caused
  • Internal or external flooding event indications of degraded
  • High winds or tornado performance in at least one strike train of a SAFETY SYSTEM
  • FIRE needed for the current
  • EXPLOSION operating mode * (site-specific hazards)
  • The event has caused
  • Other events with similar VISIBLE DAMAGE to a hazard characteristics as SAFETY SYSTEM determined by the Shift component or structure Manager needed for the current AND operating mode b. EITHER of the following:
1. Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode. OR 2. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode. 47of118 EAL Comparison Matrix L Table C-6 Hazardous.

Events

  • Internal or external FLOODING event
  • FIRE
  • EXPLOSION
  • Other events with similar hazard characteristics as determined by the Emergency Manager 48 of 118 WCGS EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) CS1 Loss of (reactor vessel/RCS CS1 Loss of RCS inventory affecting RVLIS is the only vessel level instrumentation capable of measuring

[PWR] or RWC [BWR]) core decay heat removal reactor vessel water level below the bottom of the RCS hot leg. inventory affecting core decay capability RVLIS is disconnected when in the Refueling mode and therefore heat removal capability.

MODE: CS1 .1 and CS1 .2 -EAL CS1 .1 and CS1 .1 are only applicable in Cold Shutdown.

MODE: Cold Shutdown, Cold Shutdown only Refueling CS1 .3 -Cold Shutdown, Refueling NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 a. CONTAINMENT CLOSURE CS1.1 With CONTAINMENT 72% on RVLIS natural circulation range corresponds to 6" below not established.

CLOSURE not established, the bottom ID of the RCS loop. AND RVLIS natural circulation range <72% b. (Reactor vessel/RCS

[PWR] or RWC [BWR]) level less than (site-specific level). 2 a. CONTAINMENT CLOSURE CS1.2 With CONTAINMENT 66% on RVLIS natural circulation range corresponds to the top of established.

CLOSURE established, RVLIS reactor fuel in the reactor vessel. AND natural circulation range < 66% b. (Reactor vessel/RCS

[PWR] or RWC [BWR]) level less than (site-specific level). 3 a. (Reactor vessel/RCS

[PWR] CS1.3 RCS water level cannot be Table C-1 provides a tabularized list of site-specific applicable or RWC [BWR]) level cannot monitored for 2! 30 min. (Note 1) sumps and tanks. be monitored for 30 minutes AND Manipulator crane radiation monitor SD RE-41 > Hi-Hi alarm would or longer. Core uncovery is indicated by any of the following:

be indicative of possible core uncovery in the Refueling mode. The AND

  • UNPLANNED increase in dose rate due to core shine when the top of the core becomes b. Core uncovery is indicated by any Table C-1 sump/tank uncovered should result in upscale dose rates. 49 of 118 EAL Comparison Matrix WCGS ANY of the following:

level of sufficient magnitude

  • (Site-specific radiation to indicate core uncovery monitor) reading greater
  • Manipulator bridge crane than (site-specific value) radiation monitor SD RE-41
  • Erratic source range Hi-Hi alarm monitor indication

[PWR]

  • Erratic Source Range
  • UNPLANNED increase in Monitor indication (site-specific sump and/or tank) levels of sufficient magnitude to indicate core uncovery * (Other site-specific indications)

Note The Emergency Director should N/A Note 1: The Emergency The classification timeliness note has been standardized across the declare the Site Area Manager should declare WCGS EAL scheme by referencing the "time limit" specified within Emergency promptly upon the event promptly upon the EAL wording. determining that 30 minutes has determining that time been exceeded, or will likely be limit has been exceeded exceeded, or will likely be exceeded.

50 of 118 EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) CG1 Loss of(reactor vessel/RCS CG1 Loss of RCS inventory affecting RVLIS is the only vessel level instrumentation capable of [PWR] or RWC [BWR]) inventory fuel clad integrity with measuring reactor vessel water level below the bottom of the RCS affecting fuel clad integrity with containment challenged hot leg. RVLIS is disconnected when "in the Refueling mode and containment challenged MODE: CG1 .1 -Cold Shutdown therefore EAL CG1 .1 is only applicable in Cold Shutdown.

MODE: Cold Shutdown, only Refueling CG1 .2 -Cold Shutdown, Refueling NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 a. (Reactor vessel/RCS

[PWR] CG1.1 RVLIS natural circulation range 66% on RVLIS natural circulation range corresponds to the top of or RWC [BWR]) level less < 66%

30 min. (Note 1) reactor fuel in the reactor vessel. than (site-specific level) for 30 minutes or longer. AND Table C-2 provides a tabularized list of containment challenge AND Any Containment Challenge indications.

b. ANY indication from the indication, Table C-2 4% hydrogen concentration in the presence of oxygen represents Containment Challenge Table (see below). an explosive mixture in containment.

2 a. (Reactor vessel/RCS

[PWR] CG1.2 RCS water level cannot be Table C-1 provides a tabularized list of site-specific applicable or RWC [BWR]) level cannot monitored for 30 min. (Note 1) sumps and tanks. be monitored for 30 minutes AND Manipulator crane radiation monitor SD RE-41 > Hi-Hi alarm or longer. Core uncovery is indicated by any of the following:

would be indicative of possible core uncovery in the Refueling AND mode. The dose rate due to core shine when the top of the core b. Core uncovery is indicated by

  • UNPLANNED increase in any becomes uncovered should result in upscale dose rates. ANY of the following:

Table C-1 sump/tank level Table C-2 provides a tabularized list of containment challenge (Site-specific radiation

  • Manipulator bridge crane indications.
  • radiation monitor SD RE-41 monitor) reading greater 4% hydrogen concentration in the presence of oxygen represents
  • than (site-specific value) Hi-Hi alarm an explosive mixture in containment.
  • Erratic source range
  • Erratic Source Range Monitor monitor indication

[PWR] indication

  • UNPLANNED increase in 51 of 118 EAL Comparison Matrix WCGS (site-specific sump and/or AND tank) levels of sufficient Any Containment Challenge magnitude to indicate core indication, Table C-2 uncovery * (Other site-specific indications)

AND c. ANY indication from the Containment Challenge Table (see below). Note The Emergency Director should N/A Note 1: The Emergency The classification timeliness note has been standardized across declare the General Emergency Manager should declare the WCGS EAL scheme by referencing the "time limit" specified promptly upon determining that the event promptly upon within the EAL wording. 30 minutes has been exceeded, determining that time or will likely be exceeded.

limit has been exceeded, or will likely be exceeded.

Note 6: If CONTAINMENT Note 6 implements the asterisked note associated with the N/A CLOSURE is re-established prior to generic Containment Challenge table. exceeding the 30-minute time limit, declaration of a General Emergency is not required.

Containment Challenge Table

  • CONTAINMENT CLOSURE not established*
  • (Explosive mixture) exists inside containment
  • UNPLANNED increase in containment pressure
  • Secondary containment radiation monitor reading above (site-specific value) [BWR] *If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.

52 of 118 EAL Comparison Matrix WCGS Table C-2 Containment Challenge Indications

  • CONTAINMENT CLOSURE not established (Note 6)
  • Unplanned rise in Containment pressure 53 of 118 EAL Comparison Matrix WCGS Category D Permanently Defueled Station 54 of 118 EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) PD-AU1 Recognition Category D N/A NIA NEI Recognition Category PD ICs and EALs are applicable only to PD-AU2 Permanently Defueled Station permanently defueled stations.

WCGS is not a defueled station. PD-SU1 PD-HU1 PD-HU2 PD-HU3 PD-AA1 PD-AA2 PD-HA1 PD-HA3 55 of 118 EAL Comparison Matrix WCGS Category E Independent Spent Fuel Storage Installation 56of118 EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) E-HU1 Damage to a loaded cask N/A N/A NEI Recognition Category E ICs and EALs are applicable only to CONFINEMENT BOUNDARY stations with ISFSl's. WCGS does not have an ISFSI. MODE: All NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording EAL# EAL# Difference/Deviation Justification 1 Damage to a loaded cask N/A N/A NEI Recognition Category E ICs and EALs are applicable only to CONFINEMENT BOUNDARY as stations with ISFSl's. WCGS does not have an ISFSI. indicated by an on-contact radiation reading greater than (2 times the site-specific cask specific technical specification allowable radiation level) on the surface of the spent fuel cask. 57 of 118 EAL Comparison Matrix WCGS Category F Fission Product Barrier Degradation 58 of 118 EAL Comparison Matrix NEI IC# FA1 NEI Ex. EAL# 1 NEI IC Wording Any Loss or any Potential Loss of either the Fuel Clad or RCS barrier. MODE: Power Operation, Hot Standby, Startup, Hot Shutdown NEI Example EAL Wording Any Loss or any Potential Loss of either the Fuel Clad or RCS barrier. WCGS IC#(s) FA1 WCGS EAL# FA1.1 WCGS IC Wording Any loss or any potential loss of either Fuel Clad or RCS MODE: 1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown WCGS EAL Wording Any loss or any potential loss of EITHER Fuel Clad or RCS (Table F-1) 59 of 118 Difference/Deviation Justification None Difference/Deviation Justification Table F-1 provides the fission product barrier loss and potential loss thresholds.

EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) FS1 Loss or Potential Loss of any two FS1 Loss or potential loss of any two None barriers barriers MODE: Power Operation, Hot MODE: 1 -Power Operation, 2 -Standby, Startup, Hot Shutdown Startup, 3 -Hot Standby, 4 -Hot Shutdown NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Loss or Potential Loss of any two FS1 .1 Loss or potential loss of any two Table F-1 provides the fission product barrier loss and potential loss barriers barriers (Table F-1) thresholds.

60 of 118 EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) FG1 Loss of any two barriers and FG1 Loss of any two barriers and loss None Loss or Potential Loss of third or potential loss of the third barrier barrier MODE: Power Operation, Hot MODE: 1 -Power Operation, 2 -Standby, Startup, Hot Shutdown Startup, 3 -Hot Standby, 4 -Hot Shutdown NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Loss of any two barriers and FG1.1 Loss of any two barriers Table F-1 provides the fission product barrier loss and potential loss Loss or Potential Loss of third AND thresholds.

barrier Loss or potential loss of the third barrier (Table F-1) 61 of 118 EAL Comparison Matrix WCGS PWR Fuel Clad Fission Product Barrier Degradation Thresholds NEI NEI Threshold Wording WCGS WCGS FPS Wording Differen c e/Deviation Justification FPS# FPS #(s) FC Loss RCS or SG Tube Leakage N/A N/A N/A 1 Not Appl i cable FC Loss Inadequate Heat Removal FC Loss CSFST Core Cooling-RED Consistent with the gener i c developers note options CSFST Core 2 A. Core exit thermocouple S.1 Pat h conditions me t Cooling Red Path is used i n lieu of CET temperatu r es. readings greater than (s i te-specific temperature value). FC Loss RCS Activity/CMNT Rad FC Loss Containment radiation GT RE-59 and GT RE-60 are the s i te-specific conta i nment high > 600 R/hr on range radiation monitors. The specified monitors and values are 3 A. Conta i nment radiat i on C.1 GT RE-59 or containment radiation monitor readings correspond i ng to 300 monitor reading greater t han G T RE-60 µCi/gm. (s i te-spec ifi c value) EP-CALC-WCNOC-1602 -The conta i nment pot enti a l fa i lure and OR fuel clad fa i lure fission product barrier thresholds a r e derived from (Site-specific indicat i ons the station core damage assessment procedure , wh i ch itself is B. based on the Westinghouse Owner's Group Co r e Damage that reactor coolant activ i ty is Assessment Guidance WCAP-14696-A , with assump t ions and greater than 300 µCi/gm dose inputs from the USAR and other guidance documents calcs and equivalent 1-131) other guidance documents. FC Loss Dose equivalent 1-1 31 coolant Site-spec i fic units for DEi is µCi/gm. C.2 act i v i ty > 300 µCi/gm FC Loss CNMT Integrity or Bypass N/A N/A N/A 4 Not Applicable FC Loss Other Indications N/A N/A No other s i te-specific Fuel Clad Loss i ndication has been ident i f i ed 5 A. (site-specific as applicable) forWCGS. 6 2 of 1 18 EAL Comparison Matrix WCGS NEI NEI Threshold Wording WCGS WCGS FPB Wording Difference/Deviation Justification FPB# FPB #(s) FC Loss ED Judgment FC Loss Any condition in the opinion of None 6 A ANY condition in the E.1 the Emergency Manager that opinion of the Emergency indicates loss of the fuel clad Director that indicates Loss of barr i e r the Fuel Clad Barrier. FC RCS or SG Tube Leakage N/A N/A See FC P-Loss B.1. The RCS l evel threshold is imp l emented as P-Loss A RCS/reactor vessel l evel CSFST Core Cooling Orange Path conditions me t. 1 less than (site-specific level) FC Inadequate Heat Removal FC CSFST Core Cooling-Consistent with the generic developers note options C SFST Core P-Loss A Core exit thermocouple P-Loss ORANGE Path conditions met Cooling Orange Path is used in lieu of CET temperatures. 2 readings greater than (site-B.1 specific temperature value) OR F C CSFST Heat S i nk-RED Path Consistent with the generic developers note options CSFST Heat B. Inadequate RCS heat P-Loss cond iti ons met Sink Red Path is used. r emoval capability via steam B.2 AND generators as indicated by The phrase " and heat s i nk required" was added to preclude the Heat s i nk is required need for classification for conditions in which RCS pressure is less (site-speci fi c i ndications

). than SG pressure or Heat Sink-RED path entry was created through operator action directed by an EOP. FC RCS Activity/CMNT Rad N/A N/A N/A P-Loss Not Appl i cable 3 FC CNMT Integrity or Bypass N/A N/A N/A P-Loss Not Applicable 4 FC Other Indications N/A N/A No other s i te-specific Fuel Clad Potential Loss indica t ion has been P-Loss A (site-s pecific as applicable) identified for WCGS. 5 63 of 118 EAL Comparison Matrix WCGS 'NEI NEI Threshold Wording WCGS WCGS FPB Wording Difference/Deviation Justification FPB# FPB #(s) FC Emergency Director FC Any condition in the opinion of None P-Loss Judgment P-Loss the Emergency Manager that 6 A. Any condition in the opinion E.1 indicates potential loss of the of the Emergency Director that fuel clad barrier indicates Potential Loss of the Fuel Clad Barrier. 64of118 EAL Comparison Matrix WCGS PWR RCS Fission Prod,uct Barrier Degradation Thresholds NEI ! WCGS FPS# NEI 'le Wording FPS #(s) WCGS FPS Wording Difference/Deviation Justification I\ RCS RCS or SC?tfube Leakage RCS Loss An automatic or manual None Loss '!-. A. An aut!iiatic or manual A.1 ECCS (SI) actuation required 1 ECCS (SI) actuation is by EITHER: required by EITHER of the

  • UNISOLASLE RCS following:

leakage 1. UNISOLASLE RCS

  • SGTR leakage OR 2. SG tube RUPTURE. RCS Inadequate Heat Removal N/A N/A N/A Loss Not Applicable 2 RCS RCS Activity/CMNT Rad RCS Loss Containment radiation GT RE-59 and GT RE-60 are the site-specific containment high Loss > 60 R/hr on range radiation monitors.

The specified monitors and values are 3 A. Containment radiation C.1 GT RE-59 or containment radiation monitor readings corresponding to T.S. monitor reading greater than GT RE-60 spiked coolant activity release to containment. (site-specific value). EP-CALC-WCNOC-1602

-The RCS failure fission product barrier threshold is derived from NUREG-1940 with assumptions and inputs from the USAR, station engineering calcs and other guidance documents.

RCS CNMT Integrity or Bypass N/A N/A N/A Loss Not Applicable 4 65of118 EAL Comparison Matrix WCGS NEI NEI , IC Wording WCGS WCGS FPB Wording Difference/Deviation Justification FPB# FPB #(s) RCS Other Indications N/A N/A No other site-specific RCS Loss indication has been identified for Loss A. (si te-spe:,cific as applicable)

WCGS. 5 RCS Emergency Director Judgment RCS Loss Any condition in the opinion None Loss A. ANY condition in the opinion E.1 of the Emergency Manager that indicates loss of the RCS 6 of the Emergency Director that barr i er indicates Loss of the RCS Barrier. RCS RCS or SG Tube Leakage N/A N/A Wolf Creek operating procedures that respond to RCS leaks or P-L oss 1 A. Operation of a standby Steam Generator Tube Leakage direct actuation of Safety Injection if an RCS leak exceeds the capac i ty of the currently charging (makeup) pump is running charging pump. Therefore any leak of a magnitude of the required by EITHER of the capacity of a running charging pump would , by procedure , exceed following: RCS Loss A.1. Refer to OFN BB-007 RCS Leakage High and 1. UNISOLABLE RCS OFN BB-007 A Steam Generator Tube Leakage. Therefore RCS leakage P-L oss 1.A is not i mplemented. OR RCS CSFST Inte grity-RED Path Consistent with the generic developers note options CSFST 2. SG tube leakage. P-Loss A.1 conditions met Integrity Red Path is used. OR B. RCS cooldown rate greater th an (site-specific pressurized thermal shock cr i teria/limits defined by site-specific indications). RCS Inadequate Heat Removal RCS CSFST Heat Sink-RED Path Consistent with the generic developers note options CSFST Heat P-Loss B.1 conditions met Sink Red Path is used. P-Loss 2 A. Inadequate RCS heat AND remova l capability via steam The phrase " and heat sink required" was added to preclude the generators as indicated by Heat sink is required need for classification for conditions in which RCS pressure is less (s it e-specific indications).

than SG pressure or Heat Sink-RED path entry was created through operator action directed by an EOP. 66 of 118 EAL Comparison Matrix WCGS NEI NEI IC Wording WCGS WCGS FPB Wording Difference/Deviation Justification FPB# FPB #(s) RCS CS Activity/CMNT Rad N/A N/A N/A P-Loss 3 Not Applicable RCS CNMT Integrity or Bypass N/A N/A N/A P-Loss 4 Not Applicable RCS Other Indications N/A N/A No other site-specific RCS Potential Loss indication has been P-Loss 5 A. (site-specific as applicable) identified for WCGS. RCS Emergency Director Judgment RCS Any condition in the opinion of None P-Loss 6 A. ANY condition in the opinion P-Loss E.1 the Emergency Manager that indicates potential loss of the of the Emergency Director that RCS barrier indicates Potential Loss of the RCS Barrier. 67 of 118 EAL Comparison Matrix WCGS PWR Containment Fission Product Barrier Degradation Thresholds NEI NEI IC Wording WCGS WCGS FPB Wording Difference/Deviation Justification FPB# FPB #(s) CNMT RCS or SG Tube Leakage CNMT A RUPTURED SG is FAUL TED outside Deleted the term "leaking." Consistent with the Loss A. A leaking or RUPTURED SG is Loss of containment generic bases, the condition of the SG whether leaking or ruptured is determined in accordance with 1 FAUL TED outside of containment.

A.1 the thresholds for RCS barrier potential loss 1.A or Loss 1.A. However, since RCS potential loss 1.A [A.1] is not implemented in the WC scheme (see RCS potential loss 1.A justification), Containment loss 1.A [A.1] is bounded by the ruptured SG only. CNMT Inadequate Heat Removal N/A N/A N/A Loss Not Applicable 2 CNMT RCS Activity/CMNT Rad N/A N/A N/A Loss Not applicable 3 CNMT CNMT Integrity or Bypass CNMT Containment isolation is required None Loss A. Containment isolation is required Loss AND EITHER: 4 AND D.1

  • Containment integrity has been EITHER of the following:

lost based on Emergency Manager judgment 1. Containment integrity has been lost based on

  • UNISOLABLE pathway from Emergency Director judgment.

Containment to the environment exists 68 of 118 EAL Comparison Matrix WCGS NEI NEI IC Wording WCGS WCGS FPB Wording Difference/Deviation Justification FPB# FPB #(s) OR CNMT Indications of RCS leakage outside of None 2. UNISOLABLE pathway from Loss containment the containment to the D.2 environment exists. OR B. Indications of RCS leakage outside of containment.

CNMT Other Indications N/A N/A No other site-specific Containment Loss indication Loss A. (site-specific as applicable) has been identified for WCGS. 5 CNMT Emergency Director Judgment CNMT Any condition in the opinion of the None Loss ANY condition in the opinion of the Loss Emergency Manager that indicates loss 6 Emergency Director that indicates E.1 of the containment barrier Loss of the Containment Barrier. CNMT RCS or SG Tube Leakage N/A N/A N/A P-L oss Not Applicable 1 CNMT Inadequate Heat Removal CNMT CSFST Core Cooling-RED Path Consistent with the generic developers note options P-Loss P-Loss conditions met CSFST Core Cooling Red Path is used in lieu of GET A. 1. (Site-specific criteria for entry AND temperatures and RCS levels. 2 in to core cooling restoration 8.1 procedure)

Restoration procedures not effective Added Note 1 consistent with other thresholds with a AND within 15 min. (Note 1) timing component.

2. Restoration procedure not effective within 15 minutes. 69 of 118 I EAL Comparison Matr i x WCGS NEI NEI IC Wording WCGS WCGS FPB Word i ng Difference/Deviation Just ifi cat i on FPB# FPB #(s) CNMT RCS Activity/CMNT Rad CNMT Containment radiation GT RE-59 and GT RE-60 are the site-specific P-Loss P-Loss > 6 , 000 R/hr on containment high range radiation monitors.

The A. Containment radiation monitor GT RE-59 or specified monitors and val u es are containment 3 reading greater than (site-specific C.1 GT RE-60 radiation monitor readings corresponding to 20% clad value). failure. EP-CALC-WCNOC-1602

-The containment potential failure and fue l clad failure fission product barrier thresholds are derived from the station core damage assessment procedure , which itself i s based on the Westinghouse Owner's Group Core Damage Assessment Guidance WCAP-14696-A , with assumptions and inputs from the USAR and other guidance documents. CNMT CNMT Integrity or Bypass CNMT CSFST Containment-RED Path Consistent with the generic developers note options P-Loss A. Containment pressure greater P-Loss conditions met CSFST Containment Red Path is used in l ieu of containment pressure. For WCGS , the Containment 4 than (site-specific value) D.1 CSFST Red Path is defined by containment pressure OR exceeding containment design pressure of 60 psig. B. Explosive mixture exists inside CNMT Containment hydrogen concentration 4% hydrogen concentration in the presence of containment P-Loss 4% oxygen represents an explosive mixture in OR D.2 containment.

C. 1. Containment pressure greater than (site-s pecific pressure setpoint)

AND CNMT Containment pressure > 27 psig with The Containment pressure setpoint (27 psig) is the 2. Less than one full train of P-Loss < one full train of Containment pressure at which the Containment Spray System (sit e-specific system or D.3 depressurizat i on equipment operating should actuate and begin performing its function. equipment) is operating per per design for 15 min. (Note 1 , 9) Added Note 1 consistent with other thresholds with a design for 15 minutes or timing component.

longer. Added new Note 9 to define what constitutes one full train of depressurization equipmen t. 70of118 EAL Comparison Matrix WCGS NEI NEI IC Wording WCGS WCGS FPB Wording Difference/Deviation Justification FPB# FPB #(s) CNMT Other Indications N/A N/A No other site-specific Containment Potential Loss P-Loss A. (site-specific as applicable) indication has been identified for WCGS. 5 CNMT Emergency Director Judgment CNMT Any condition in the opinion of the None P-Loss A. ANY condition in the opinion of the P-Loss Emergency Manager that indicates potential loss of the containment 6 Emergency Director that indicates E.1 barrier Potential Loss of the Containment Barrier. 71 of118 EAL Comparison Matrix WCGS Category H Hazards and Other Conditions Affecting Plant Safety 72 of 118 EAL Comparison Matrix NEI IC# NEI IC Wording HU1 Confirmed SECURITY CONDITION or threat MODE: All NEI Ex. NEI Example EAL Wording EAL# 1 A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the (site-specific security shift supervision).

2 Notification of a credible security threat directed at the site. 3 A validated notification from the NRG providing information of an aircraft threat. WCGS IC#(s) HU1 WCGS EAL# HU1.1 WCGS IC Wording Confirmed SECURITY CONDITION or threat. MODE: All WCGS EAL Wording 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 73 of 118 WCGS Difference/Deviation Justification None Difference/Deviation Justification Example EALs #1, 2 and 3 have been combined into a single EAL for ease of presentation and use. The security shift supervision is defined as the Security Shift Lieutenant EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) HU2 Seismic event greater than OBE HU2 Seismic event greater than OBE None level level MODE: All MODE: All NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Seismic event greater than HU2.1 Seismic event > OBE as The WCGS OBE indicator is Seismic Activity, Annunciator 00-098D. Operating Basis Earthquake indicated by Seismic Activity (OBE) as indicated by: Annunciator 00-098D (site-specific indication that a seismic event met or exceeded OBE limits) 74 of 118 EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) HU3 Hazardous event. HU3 Hazardous event None MODE: All MODE: All NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 A tornado strike within the HU3.1 A tornado strike within the None PROTECTED AREA. PROTECTED AREA 2 Internal room or area flooding of a HU3.2 Internal room or area FLOODING None magnitude sufficient to require of a magnitude sufficient to manual or automatic electrical require manual or automatic isolation of a SAFETY SYSTEM electrical isolation of a SAFETY component needed for the current SYSTEM component needed for operating mode. the current operating mode 3 Movement of personnel within the HU3.3 Movement of personnel within the None PROTECTED AREA is impeded PROTECTED AREA is due to an offsite event involving IMPEDED due to an offsite event hazardous materials (e.g., an involving hazardous materials offsite chemical spill or toxic gas (e.g., an offsite chemical spill or release).

toxic gas release) 4 A hazardous event that results in HU3.4 A hazardous event that results in Added reference to Note 7. on-site conditions sufficient to on-site conditions sufficient to prohibit the plant staff from prohibit the plant staff from accessing the site via personal accessing the site via personal vehicles.

vehicles (Note 7) 5 (Site-specific list of natural or N/A N/A No other site-specific hazard has been identified for WCGS. technological hazard events) Note EAL #3 does not apply to routine N/A Note 7: This EAL does not This note, designated Note #7, is intended to apply to generic traffic impediments such as fog, apply to routine traffic example EAL #4, not #3 as specified in the generic guidance.

75of118 EAL Comparison Matrix WCGS snow, ice, or vehicle breakdowns impediments such as or accidents.

fog, snow, ice, or vehicle breakdowns or accidents.

76 of 118 EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) HU4 FIRE potentially degrading the HU4 FIRE potentially degrading the None level of safety of the plant. level of safety of the plant MODE: All MODE: All NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 a. A FIRE is NOT extinguished HU4.1 A FIRE is not extinguished Added " ... in the same fire area" to the second bullet. Single file within 15-minutes of ANY of the within 15 min. of any of the alarms in different fire areas is not indicative of a higher likelihood of following FIRE detection following FIRE detection a fire in either area. indications:

indications (Note 1 ): Table H-1 provides a tabularized list of site-specific fire areas.

  • Report from the field (i.e.,
  • Report from the field (i.e., visual observation) visual observation)
  • Receipt of multiple (more
  • Receipt of multiple (more than 1) fire alarms or than 1) fire alarms or indications indications in the same fire Field verification of a single area
  • fire alarm
  • Field verification of a single AND fire alarm b. The FIRE is located within AND ANY of the following plant rooms The FIRE is located within any or areas: Table H-1 area (site-specific list of plant rooms or areas) 2 a. Receipt of a single fire alarm HU4.2 Receipt of a single fire alarm Table H-1 provides a tabularized list of site-specific fire areas. (i.e., no other indications of a (i.e., no other indications of a FIRE). FIRE) AND AND b. The FIRE is located within The fire alarm is indicating a 77 of 118 EAL Comparison Matrix WCGS ANY of the following plant rooms FIRE within any Table H-1 area or areas: AND (site-specific list of plant rooms or The existence of a FIRE is not areas) verified within 30 min. of alarm AND receipt (Note 1) C. The existence of a FIRE is not verified within 30-minutes of alarm receipt. 3 A FIRE within the plant or ISFSI HU4.3 A FIRE within the plant WCGS does not have an ISFSI. [for plants with an ISFSI outside PROTECTED AREA not the plant Protected Area] extinguished within 60 min. of PROTECTED AREA not the initial report, alarm or extinguished within 60-minutes of indication (Note 1) the initial report, alarm or indication.

4 A FIRE within the plant or ISFSI HU4.4 A FIRE within the plant WCGS does not have an ISFSI. [for plants with an ISFSI outside PROTECTED AREA that the plant Protected Area] requires firefighting support by PROTECTED AREA that requires an offsite fire response agency to firefighting support by an offsite extinguish fire response agency to extinguish.

Note Note: The Emergency Director N/A Note 1: The Emergency Manager The classification timeliness note has been standardized across the should declare the Unusual Event should declare the event WCGS EAL scheme by referencing the "time limit" specified within promptly upon determining that promptly upon the EAL wording. the applicable time has been determining that time exceeded, or will likely be limit has been exceeded, exceeded.

or will likely be exceeded.

78 of 118 EAL Comparison Matrix WCGS Table H-1 Fire Areas

  • Auxiliary Building
  • Reactor Building
  • Control Building
  • Fuel Building
  • Refueling Water Storage Tank Valve Room
  • Refueling Water Storage Tank (TBN01) 79 of 118 1 EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) HU7 Other conditions exist which in the HU6 Other conditions existing that in None judgment of the Emergency the judgment of the Emergency Director warrant declaration of a Manager warrant declaration of a (NO)UE UE MODE: All MODE: All NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Other conditions exist which in the HU6.1 Other conditions exist which in Emergency Manager is the site-specific title for Emergency director.

judgment of the Emergency the judgment of the Emergency Director indicate that events are in Manager indicate that events are progress or have occurred which in progress or have occurred indicate a potential degradation of which indicate a potential the level of safety of the plant or degradation of the level of safety indicate a security threat to facility of the plant or indicate a security protection has been initiated.

No threat to facility protection has releases of radioactive material been initiated.

No releases of requiring offsite response or radioactive material requiring monitoring are expected unless offsite response or monitoring further degradation of safety are expected unless further systems occurs. degradation of SAFETY SYSTEMS occurs. 80 of 118 EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording IC#(s) Difference/Deviation Justification HA1 HOSTILE ACTION within the HA1 HOSTILE ACTION within the None OWNER CONTROLLED AREA or OWNER CONTROLLED AREA airborne attack threat within 30 or airborne attack threat within 30 minutes. minutes MODE: All MODE: All NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 A HOSTILE ACTION is occurring or HA1.1 A HOSTILE ACTION is Example EALs #1 and #2 have been combined into a single EAL has occurred within the OWNER occurring or has occurred within for ease of use. CONTROLLED AREA as reported the OWNER CONTROLLED The security shift supervision is defined as the Security Shift by the (site-specific security shift AREA as reported by the Lieutenant supervision).

Security Shift Lieutenant 2 A validated notification from NRG of OR an aircraft attack threat within 30 A validated notification from minutes of the site. NRG of an aircraft attack threat within 30 min. of the site 81 of 118 EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) HA5 Gaseous release impeding HA5 Gaseous release impeding Mode applicability of HA5 limited to the modes determined to be access to equipment necessary access to equipment necessary applicable per Table H-2. See bases note and Attachment 3 of the for normal plant operations, for normal plant operations, Technical Bases Document.

cooldown or shutdown.

cooldown or shutdown.

MODE: All MODE: 1 -Power Operation, 2 -Startup, 3 -Hot Standby, 4 -Hot Shutdown NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 a. Release of a toxic, HA5.1 Release of a toxic, corrosive, Analysis by Operations Subject Matter Experts determined areas corrosive, asphyxiant or asphyxiant or flammable gas that external to the Main Control Room that require access in any flammable gas into any of the prohibits or IMPEDES access to operating mode to maintain safe plant operations or to perform a following plant rooms or areas: any Table H-2 rooms or areas normal plant shutdown and cooldown to Cold Shutdown conditions. (site-specific list of plant rooms (Note 5) Table H-2 and Attachment 3 of the Technical Bases Document.

or areas with entry-related mode applicability identified)

AND b. Entry into the room or area is prohibited or impeded. Note Note: If the equipment in the N/A Note 5: If the equipment in the None listed room or area was already listed room or area was inoperable or out-of-service already inoperable or before the event occurred, then out-of-service before the no emergency classification is event occurred, then no warranted.

emergency classification is warranted.

82of118 EAL Comparison Matrix WCGS Table H-2 Safe Operation

& Shutdown Rooms/Areas Room/Area Mode Applicability North Electrical Pen. Room A 3,4 South Electrical Pen. Room 8 3,4 ESF SWGR Room No. 1 (TRN A) 4 ESF SWGR Room No. 2 (TRN 8) 4 Auxiliary BuildinQ/West Hall Elev 2000 3,4,5 83of118 EAL Comparison Matrix NEI IC# NEI IC Wording WCGS IC#(s) HA6 Control Room evacuation HA6 resulting in transfer of plant control to alternate locations.

MODE: All NEI Ex. NEI Example EAL Wording WCGS EAL# EAL# 1 An event has resulted in plant HA6.1 control being transferred from the Control Room to (site-specific remote shutdown panels and local control stations).

WCGS IC Wording Control Room evacuation resulting in transfer of plant control to alternate locations MODE: All WCGS EAL Wording An event has resulted in pla nt control being transferred fro m the ry Control Room to the Auxilia Shutdown Panel (ASP) -84 of 11 8 Difference/Deviation Justification None Difference/Deviation Justification Auxiliary Shutdown Panel (ASP) is the site-specific remote shutdown panels/local control stations.

WCGS EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) HA7 Other conditions exist which in the HA7 Other conditions exist that in the None judgment of the Emergency Director judgment of the Emergency Manager warrant declaration of an Alert. warrant declaration of an Alert MODE: All MODE: All NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Other conditions exist which, in the HA7.1 Other conditions exist which, in the Emergency Manager is the site-specific title for judgment of the Emergency Director, judgment of the Emergency Manager, Emergency director.

indicate that events are in progress or indicate that events are in progress or have occurred which involve an actual or have occurred which involve an actual or potential substantial degradation of the potential substantial degradation of the level of safety of the plant or a security level of safety of the plant or a security event that involves probable life event that involves probable life threatening risk to site personnel or threatening risk to site personnel or damage to site equipment because of damage to site equipment because of HOSTILE ACTION. Any releases are HOSTILE ACTION. Any releases are expected to be limited to small fractions expected to be limited to small fractions of the EPA Protective Action Guideline of the EPA Protective Action Guideline exposure levels. exposure levels. 85 of 118 EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) HS1 HOSTILE ACTION within the HS1 HOSTILE ACTION within the None PROTECTED AREA PROTECTED AREA MODE: All MODE: All NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 A HOSTILE ACTION is occurring HS1.1 A HOSTILE ACTION is occurring or has The security shift supervision is defined as the Security Shift or has occurred within the occurred within the PROTECTED AREA Lieutenant PROTECTED AREA as reported as reported by the Security Shift by the (site-specific security shift Lieutenant supervision).

86of118 EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) HS6 Inability to control a key safety HS6 Inability to control a key safety function Deleted defueled mode applicability.

Control of the cited function from outside the Control from outside the Control Room safety functions are not critical for a defueled reactor as there Room. MODE: 1 -Power Operation, 2 -Startup, is no energy source in the reactor vessel or RCS. MODE: All 3 -Hot Standby, 4 -Hot Shutdown, 5 -This is an acceptable deviation from the generic NEI 99-Cold Shutdown, 6 -Refueling 01 Revision 6 guidance.

NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 a. An event has resulted in plant HS6.1 An event has resulted in plant control Auxiliary Shutdown Panel (ASP) is the site-specific remote control being transferred from the being transferred from the Control Room shutdown panels/local control stations.

Control Room to (site-specific to the Auxiliary Shutdown Panel (ASP) The Mode applicability for the reactivity control safety remote shutdown panels and local control stations).

AND function has been limited to Modes 1, 2, and 3 (hot operating Control of any of the following key safety conditions).

In the cold operating modes adequate shutdown AND functions is not re-established within 15 margin exists under all conditions.

b. Control of ANY of the min. (Note 1 ): This is an acceptable deviation from the generic NEI 99-following key safety functions is
  • Reactivity (Modes 1, 2 and 3 01 Revision 6 guidance.

not reestablished within (site-specific number of minutes).

only) Reactivity control

  • Core cooling *
  • Core cooling [PWR] I RWC
  • RCS heat removal water level [BWR]
  • RCS heat removal 87 of 118 EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) HS? Other conditions exist which in HS? Other conditions existing that in the None the judgment of the Emergency judgment of the Emergency Manager Director warrant declaration of a warrant declaration of a Site Area Site Area Emergency.

Emergency MODE: All MODE: All NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Other conditions exist which in HS7.1 Other conditions exist which in the Emergency Manager is the site-specific title for Emergency the judgment of the Emergency judgment of the Emergency Manager director.

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

result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the site boundary.

88 of 118 EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) HG1 HOSTILE ACTION resulting in N/A N/A IC HG1 and associated example EAL are not implemented in loss of physical control of the the WCGS scheme. facility.

There are several other ICs that are redundant with this IC, MODE: All and are better suited to ensure timely and effective emergency declarations.

In addition, the development of new spent fuel pool level EALs, as a result of NRG Order EA-12-051, clarified the intended emergency classification level for spent fuel pool level events. This is an acceptable deviation from the generic NEI 99-01 Revision 6 guidance.

NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 a. A HOSTILE ACTION is NIA N/A IC HG1 and associated example EAL are not implemented occurring or has occurred within in the WCGS scheme. the PROTECTED AREA as There are several other I Cs that are redundant with this IC, reported by the (site-specific and are better suited to ensure timely and effective security shift supervision).

emergency declarations.

In addition, the development of new AND spent fuel pool level EALs, as a result of NRG Order EA-12-b. EITHER of the following has 051, clarified the intended emergency classification level for spent fuel pool level events. This deviation is justified occurred:

because: 1. ANY of the following safety 1. Hostile Action in the Protected Area is bounded by ICs functions cannot be HS1 and HS7. Hostile Action resulting in a loss of controlled or maintained.

physical control is bound by EAL HG7, as well as any

  • Protective Action Guides (PAGs). [PWR]/RWC water level [BWR] a. If, for whatever reason, the Control Room must be RCS heat removal evacuated, and control of safety functions (e.g.,
  • reactivity control, core cooling, and RCS heat 89 of 118 EAL Comparison Matrix WCGS OR removal) cannot be reestablished, then IC HS6 would 2. Damage to spent fuel has apply, as well as IC HS7 if desired by the EAL occurred or is IMMINENT.

decision-maker.

b. Also, as stated above, any event (including Hostile Action) that could reasonably be expected to have a release exceeding EPA PAGs would be bound by IC HG7. C. From a Hostile Action perspective, ICs HS1, HS7 and -. HG7 are appropriate, and therefore, make this part of HG1 redundant and unnecessary.
d. From a loss of physical control perspective, ICs HS6, HS7 and HG7 are appropriate, and therefore, make this part of HG1 redundant and unnecessary.
2. Any event which causes a loss of spent fuel pool level will be bounded by ICs AA2, AS2 and AG2, regardless of whether it was based upon a Hostile Action or not, thus making this part of HG1 redundant and unnecessary.
a. An event that leads to a radiological release will be bounded by ICs AU1, AA1, AS1 and AG1. Events that lead to radiological releases in excess of EPA PAGs will be bounded by EALs AG1 and HG7, thus making this part of HG1 redundant and unnecessary.

ICs AA2, AS2, AG2, AS1, AG1, HS1, HS6, HS7 and HG7 have been implemented consistent with NEI 99-01 Revision 6 and thus HG1 is adequately bounded as described above. This is an acceptable deviation from the generic NEI 99-01 Revision 6 guidance.

90of118 EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) HG? Other conditions exist which in HG6 Other conditions exist which in the None the judgment of the Emergency judgment of the Emergency Manager Director warrant declaration of a warrant declaration of a General General Emergency Emergency MODE: All MODE: All NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Other conditions exist which in HG6.1 Other conditions exist which in the Emergency Manager is the site-specific title for Emergency the judgment of the Emergency judgment of the Emergency Manager director.

Director indicate that events are indicate that events are in progress or in progress or have occurred have occurred which involve actual or which involve actual or IMMINENT substantial core degradation IMMINENT substantial core or melting with potential for loss of . degradation or melting with containment integrity or HOSTILE potential for loss of containment ACTION that results in an actual loss of integrity or HOSTILE ACTION physical control of the facility.

Releases that results in an actual loss of can be reasonably expected to exceed physical control of the facility.

EPA Protective Action Guideline Releases can be reasonably exposure levels offsite for more than the expected to exceed EPA immediate site area. Protective Action Guideline exposure levels offsite for more than the immediate site area. 91of118 EAL Comparison Matrix Category S System Malfunction 92 of 118 WCGS EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) SU1 Loss of all offsite AC power SU1 Loss of all offsite AC power None capability to emergency buses for capability to emergency buses for 15 minutes or longer. 15 minutes or longer MODE: Power Operation, Startup, MODE: 1 -Power Operation, 2 -Hot Standby, Hot Shutdown Startup, 3 -Hot Standby, 4 -Hot Shutdown NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Loss of ALL offsite AC power SU1.1 Loss of all offsite AC power 4.16KV buses N BO 1 and N 802 are the site-specific emergency capability to (site-specific capability, Table S-1, to buses. emergency buses) for 15 minutes emergency 4.16KV buses NB01 Site-specific AC power sources are tabularized in Table S-1. or longer. and NB02 15 min. (Note 1) Note The Emergency Director should N/A Note 1: The Emergency The classification timeliness note has been standardized across the declare the Unusual Event Manager should declare WCGS EAL scheme by referencing the "time limit" specified within promptly upon determining that 15 the event promptly upon the EAL wording. minutes has been exceeded, or determining that time will likely be exceeded.

limit has been exceeded, or will likely be exceeded.

93 of 118 EAL Comparison Matrix Table 5-1 AC Power Sources Offsite:

  • ESF XFMR XNB02 Onsite:
  • EOG NE01
  • EDG NE02 94 of 118 EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) SU2 UNPLANNED loss of Control SU3 UNPLANNED loss of Control None Room indications for 15 minutes Room indications for 15 minutes or longer. or longer. MODE: Power Operation, MODE: 1 -Power Operation, 2 -Startup, Hot Standby, Hot Startup, 3 -Hot Standby, 4 -Hot Shutdown Shutdown NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 An UNPLANNED event results in SU3.1 An UNPLANNED event results in The site-specific Safety System Parameter list is tabulated in Table the inability to monitor one or the inability to monitor one or S-2. more of the following parameters more Table S-2 parameters from Added the words "to at least one SIG" to Auxiliary or emergency from within the Control Room for within the Control Room for;::: 15 feedwater flow. This is consistent with Level in at least on SIG. 15 minutes or longer. min. (Note 1) Note The Emergency Director should N/A Note 1: The Emergency The classification timeliness note has been standardized across the declare the Unusual Event Manager should declare WCGS EAL scheme by referencing the "time limit" specified within promptly upon determining that the event promptly upon the EAL wording. 15 minutes has been exceeded, determining that time or will likely be exceeded.

limit has been exceeded, or will likely be exceeded.

95 of 118 EAL Comparison Matrix [BWR parameter list] [PWR parameter list] Reactor Power Reactor Power RWC Water Level RCS Level RWC Pressure RCS Pressure Primarv Containment Pressure In-Core/Core Exit Temperature Suppression Pool Level Levels in at least (site-specific number) steam generators Suppression Pool Temperature Steam Generator Auxiliary or Emergency Feed Water Flow Table S-2 Safety System Parameters

  • Reactor power
  • Core Exit TIC temperature
  • Level in at least one SIG
  • Auxiliary or emergency feed flow in at least one SIG 96 of 118 WCGS EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) SU3 Reactor coolant activity greater SU4 Reactor coolant activity greater None than Technical Specification than Technical Specification allowable limits. allowable limits MODE: Power Operation, Startup, MODE: 1 -Power Operation, 2 -Hot Standby, Hot Shutdown Startup, 3 -Hot Standby, 4 -Hot Shutdown NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 (Site-specific radiation monitor) N/A N/A WCGS does not have any site-specific radiation monitor correlation reading greater than (site-specific to TS coolant activity limits. value). 2 Sample analysis indicates that a SU4.1 Sample analysis indicates RCS Changed 'reactor coolant activity" to "RCS" to conform to site reactor coolant activity value is activity>

Technical Specification specific terminology.

greater than an allowable limit Section 3.4.16 limits WCGS T.S. Section 3.4.16 provides the TS allowable coolant activity specified in Technical limits. Specifications.

97 of 118 EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) SU4 RCS leakage for 15 minutes or SUS RCS leakage for 15 minutes or None longer. longer MODE: Power Operation, MODE: 1 -Power Operation, 2 -Startup, Hot Standby, Hot Startup, 3 -Hot Standby, 4 -Hot Shutdown Shutdown NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 RCS unidentified or pressure SU5.1 RCS unidentified or pressure Example EALs #1, 2 and 3 have been combined into a single EAL boundary leakage greater than boundary leakage > 10 gpm for;::: for usability. (site-specific value) for 15 15 min. minutes or longer. OR 2 RCS identified leakage greater RCS identified leakage > 25 gpm than (site-specific value) for 15 for;::: 15 min. minutes or longer. OR 3 Leakage from the RCS to a Leakage from the RCS to a location outside containment location outside containment

> 25 greater than 25 gpm for 15 gpm for;::: 15 min. minutes or longer. (Note 1) Note The Emergency Director should N/A Note 1: The Emergency The classification timeliness note has been standardized across the declare the Unusual Event Manager should declare WCGS EAL scheme by referencing the "time limit" specified within promptly upon determining that the event promptly upon the EAL wording. 15 minutes has been exceeded, determining that time or will likely be exceeded.

limit has been exceeded, or will likely be exceeded.

98of118 EAL Comparison Matrix NEI IC# SU5 NEI Ex. EAL# 1 2 NEI IC Wording Automatic or manual (trip [PWR] I scram [BWR]) fails to shutdown the reactor. MODE: Power Operation NEI Example EAL Wording a. An automatic (trip [PWR] I scram [BWR]j did not shutdown the reactor. AND b. A subsequent manual action taken at the reactor control consoles is successful in shutting down the reactor. a. A manual trip ([PWR] I scram [BWR]) did not shutdown the reactor. AND b. EITHER of the following:

1. A subsequent manual action taken at the reactor control consoles is successful in shutting down the reactor. WCGS IC#(s) SU6 WCGS EAL# SU6.1 SU6.2 WCGS IC Wording Automatic or manual trip fails to shut down the reactor MODE: 1 -Power Operation WCGS EAL Wording 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) 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 99of118 WCGS Difference/Deviation Justification None Difference/Deviation Justification As specified in the generic developers guidance "Developers may include site-specific EOP criteria indicative of a successful reactor shutdown in an EAL statement, the Basis or both (e.g., a reactor power level)." Reactor power< 5% is the site-specific indication of a successful reactor trip. Added the words" ... as indicated by reactor power;;::

5% after any RTS setpoint is exceeded" to clarify that it is a failure of the automatic trip when a valid trip signal has been exceed. SB HS-1 (Panel RL003) or SB HS-42 (Panel RL006) are the site-specific reactor control console trip switches credited for a successful manual trip. As specified in the generic developers guidance "Developers may include site-specific EOP criteria indicative of a successful reactor shutdown in an EAL statement, the Basis or both (e.g., a reactor power level)." Reactor power < 5% is the site-specific indication of a successful reactor trip. Added the words" ... as indicated by reactor power;;::

5% after any manual trip action was initiated" to clarify that it is a failure of any manual trip when an actual manual trip signal has been inserted.

Combined conditions b.1 and b.2 into a single statement to simplify the presentation.

EAL Comparison Matrix WCGS OR power < 5% (Note 8) SB HS-1 (Panel RL003) or SB HS-42 (Panel RL006) are the site-2 A subsequent automatic specific reactor control console trip switches credited for a (trip [PWR] I scram successful manual trip. [BWR]) is successful in shutting down the reactor. Notes Note: A manual action is any operator action, or set of actions, N/A Note 8: A manual action is any " ... tripping rod drive power ... " as a site-specific, manual trip which causes the control rods to operator action, or set of action taken away from the reactor control consoles.

be rapidly inserted into the core, actions, which causes and does not include manually the control rods to be driving in control rods or rapidly inserted into the implementation of boron core, and does not injection strategies.

include manually driving in control rods, tripping rod drive power or implementation of boron injection strategies.

100of118 EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) SU6 Loss of all onsite or offsite SU7 Loss of all onsite or offsite None communications capabilities.

communications capabilities.

MODE: Power Operation, MODE: 1 -Power Operation, 2 -Startup, Hot Standby, Hot Startup, 3 -Hot Standby, 4 -Hot Shutdown Shutdown NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Loss of ALL of the following SU7.1 Loss of all Table S-4 onsite Example EALs #1, 2 and 3 have been combined into a single EAL onsite communication methods: communication methods for simplification of presentation. (site-specific list of OR Table S-4 provides a site-specific list of onsite, ORO and NRG communications methods) Loss of all Table S-4 offsite communications methods. 2 Loss of ALL of the following communication methods Changes the acronym "ORO" to "offsite" consistent with WC usage. ORO communications methods: OR (site-specific list of Loss of all Table S-4 NRG communications methods) communication methods 3 Loss of ALL of the following NRG communications methods: (site-specific list of communications methods) 101 of 118 EAL Comparison Matrix WCGS Table S-4 Communication Methods System On site 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 102 of 118 EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) SU7 Failure to isolate containment or SUS Failure to isolate containment or None loss of containment pressure loss of containment pressure control. [PWR] control MODE: Power Operation, MODE: 1 -Power Operation, 2 -Startup, Hot Standby, Hot Startup, 3 -Hot Standby, 4 -Hot Shutdown Shutdown NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 a. Failure of containment to Any penetration is not isolated Reworded EAL to better describe the intent. Penetrations cannot isolate when required by an within 15 min. of a VALID close, but they can be isolated by closure of one or more isolation containment isolation signal actuation signal. OR valves associated with that penetration.

The revised wording AND Containment pressure > 27 psig maintains the generic example EAL intent while more clearly with < one full train of describing failure to isolate threshold.

b. ALL required penetrations containment depressurization The containment pressure setpoint (27 psig) is the pressure at are not closed within 15 minutes of the actuation signal. equipment operating per design which the containment depressurization equipment should actuate 15 min. (Note 9) and begin performing its function.

One train of containment 2 Containment pressure SUS.1 (Note 1) depressurization equipment is defined in Note 9. a. greater than (site-specific pressure).

AND b. Less than one full train of (site-specific system or equipment) is operating per design for 15 minutes or longer. N/A N/A N/A Note 1: The Emergency Manager Added Note 1 to be consistent in its use for EAL thresholds with a should declare the event timing component.

promptly upon determining that time 103 of 118 EAL Comparison Matrix WCGS limit has been exceeded, or will likely be exceeded. N/A N/A N/A Note 9: One Containment Spray Added new Note 9 to provide clarification regarding what System train and one constitutes on full train of a containment depressurization system. Containment Cooling System train comprise one full train of depressurization equipment 104of118 EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) SA1 Loss of all but one AC power SA1 Loss of all but one AC power None source to emergency buses for source to emergency buses for 15 minutes or longer. 15 minutes or longer. MODE: Power Operation, MODE: 1 -Power Operation, 2 -Startup, Hot Standby, Hot Startup, 3 -Hot Standby, 4 -Hot Shutdown Shutdown NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 a. AC power capability to (site-SA1.1 AC power capability, Table S-1, 4.16KV buses NB01 and NB02 are the site-specific emergency specific emergency buses) is to emergency 4.16KV buses buses. reduced to a single power source NB01 and NB02 reduced to a Site-specific AC power sources are tabularized in Table S-1. for 15 minutes or longer. single power source 15 min. AND (Note 1) b. Any additional single power AND source failure will result in a loss Any additional single power of all AC power to SAFETY source failure will result in loss of SYSTEMS. all AC power to SAFETY SYSTEMS Note The Emergency Director should N/A Note 1: The Emergency The classification timeliness note has been standardized across the declare the Alert promptly upon Manager should declare WCGS EAL scheme by referencing the "time limit" specified within determining that 15 minutes has the event promptly upon the EAL wording. been exceeded, or will likely be determining that time exceeded.

limit has been exceeded, or will likely be exceeded.

105 of 118 EAL Comparison Matrix WCGS Table S-1 AC Power Sources Offsite:

  • ESF XFMR XNB02 Onsite:
  • EOG NE01
  • EOG NE02 106 of 118 EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) SA2 UNPLANNED loss of Control SA3 UNPLANNED loss of Control None Room indications for 15 minutes Room indications for 15 minutes or longer with a significant or longer with a significant transient in progress.

fransient in progress.

MODE: Power Operation, MODE: 1 -Power Operation, 2 -Startup, Hot Standby, Hot Startup, 3 -Hot Standby, 4 -Hot Shutdown Shutdown NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 An UNPLANNED event results in SA3.1 An UNPLANNED event results in The site-specific Safety System Parameter list to tabulated in Table the inability to monitor one or the inability to monitor one or S-2. more of the following parameters more Table S-2 parameters from The site-specific significant transients list to tabulated in Table S-3. from within the Control Room for within the Control Room for 2': 15 15 minutes or longer. min. (Note 1) WCGS is a PWR and thus does not include thermal power AND AND oscillations

> 10%. ANY of the following transient Any significant transient is in events in progress.

progress, Table S-3

  • Automatic or manual run back greater than 25% thermal reactor power
  • Electrical load rejection greater than 25% full electrical load
  • Reactor scram [BWR] I trip [PWR]
  • Thermal power oscillations greater than 10% [BWR] 107 of 118 EAL Comparison Matrix Note The Emergency Director should NIA declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.

fBWR 1Jarameter lisfl Reactor Power RWC Water Level RWC Pressure Primary Containment Pressure Suppression Pool Level Suppression Pool Temperature

WCGS Note 1: The Emergency The classification timeliness note h Manager should declare as been standardized across the the "time limit" specified within WCGS EAL scheme by referencing the event promptly upon the EAL wording. determining that time limit has been exceeded, or will likely be exceeded.

[PWR f]arameter list] Reactor Power RCS Level RCS Pressure In-Core/Core Exit Temperature Levels in at least (site-specific number) steam oenerators Steam Generator Auxiliary or Emergency Feed Water Flow Table S-2 Safety System Parameters

  • Reactor power
  • Core Exit TIC temperature
  • Level in at least one SIG
  • Auxiliary or emergency feed flow in at least one SIG Table S-3 Significant Transients
  • Runback ;:: 25% thermal power
  • Electrical load rejection

> 25% electrical load

  • ECCS actuation 108 of 118 EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) SA5 Automatic or manual (trip [PWR] SA6 Automatic or manual trip fails to None I scram [BWR]) fails to shutdown shut down the reactor and the reactor, and subsequent subsequent manual actions manual actions taken at the taken at the reactor control reactor control consoles are not consoles are not successful in successful in shutting down the shutting down the reactor reactor. MODE: 1 -Power Operation MODE: Power Operation NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 a. An automatic or manual (trip SA6.1 An automatic or manual trip fails As specified in the generic developers guidance "Developers may [PWR] I scram [BWR]) did not to shut down the reactor as include site-specific EOP criteria indicative of a successful reactor shutdown the reactor. indicated by reactor power shutdown in an EAL statement, the Basis or both (e.g., a reactor AND ::::5% power level)." Reactor power< 5% is the site-specific indication of a AND successful reactor trip. b. Manual actions taken at the SB-HS-1 (Panel RL003) .or SB-HS-42 (Panel RL006) are the site-reactor control consoles are not Manual trip actions taken at the successful in shutting down the reactor control console (SB-HS-1 specific reactor control console trip switches credited for a reactor. or SB-HS-42) are not successful successful manual trip. in shutting down the reactor as indicated by reactor power :::: 5% (Note 8) Notes Note: A manual action is any N/A Note 8: A manual action is any Added " ... tripping rod drive power. .. " as a site-specific, manual trip operator action, or set of actions, which causes the control rods to operator action, or set of action taken away from the reactor control consoles.

be rapidly inserted into the core, actions, which causes the control rods to be and does not include manually rapidly inserted into the driving in control rods or implementation of boron injection core, and does not strategies.

include manually driving in control rods, tripping rod drive power or 109of118 l EAL Comparison Matrix WCGS implementation of boron injection strategies.

11Oof118 EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) SA9 Hazardous event affecting a SA9 Hazardous event affecting a None SAFETY SYSTEM needed for SAFETY SYSTEM needed for the current operating mode. the current operating mode MODE: Power Operation, MODE: 1 -Power Operation, 2 -Startup, Hot Standby, Hot Startup, 3 -Hot Standby, 4 -Hot Shutdown Shutdown 111of118 EAL Comparison Matrix WCGS NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 a. The occurrence of ANY of SA9.1 The occurrence of any Table S-The hazardous events have been tabularized in Table S-5. the following hazardous events: 5 hazardous event Replaced "Shift Manager" with "Emergency Manager" as the EC

AND EITHER: can be either the SM or augmented ERO EC.

  • Internal or external flooding
  • Event damage has caused event indications of degraded
  • High winds or tornado strike performance in at least one train of a SAFETY SYSTEM
  • FIRE needed for the current
  • EXPLOSION operating mode * (site-specific hazards)
  • The event has caused VISIBLE DAMAGE to a
  • Other events with similar SAFETY SYSTEM hazard characteristics as component or structure determined by the Shift needed for the current Manager operating mode AND b. EITHER of the following:
1. Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode. OR 2. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode. 112 of 118 EAL Comparison Matrix Table S-5 Hazardous Events
  • Internal or external FLOODING event
  • FIRE
  • EXPLOSION
  • Other events with similar hazard characteristics as determined by the Emergency Manager 113of118 WCGS EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) SS1 Loss of all offsite and all onsite SS1 Loss of all offsite and all onsite None AC power to emergency buses AC power to emergency buses for 15 minutes or longer. for 15 minutes or longer MODE: Power Operation, MODE: 1 -Power Operation, 2 -Startup, Hot Standby, Hot Startup, 3 -Hot Standby, 4 -Hot Shutdown Shutdown NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Loss of ALL offsite and ALL SS1.1 Loss of all offsite and all onsite 4.16KV buses NB01 and NB02 are the site-specific emergency onsite AC power to (site-specific AC power capability to buses. emergency buses) for 15 minutes emergency 4.16KV buses NB01 or longer. and NB02 15 min. (Note 1) Note The Emergency Director should N/A Note 1: The Emergency The classification timeliness note has been standardized across the declare the Unusual Event Manager should WCGS EAL scheme by referencing the "time limit" specified within promptly upon determining that declare the event the EAL wording. 15 minutes has been exceeded, promptly upon or will likely be exceeded.

determining that time limit has been exceeded, or will likely be exceeded.

114 of 118 EAL Comparison Matrix NEI IC# NEI IC Wording SS5 Inab i lity to shutdown the reactor caus i ng a challenge to (core cool i ng [PWR] I RWC water l evel [BWR]) or RCS heat removal. MODE: Power Operation NEI Ex. NE I Example EAL Wording EAL# 1 a. An automatic or manual (trip [PWR] I scram [BWR]) did not shutdown the reactor. AND b. All manual actions to shutdown the reactor have been unsuccessful.

AND c. EITHER of the following cond i t i ons exist: * (Site-specific indication of an inability to adequately remove heat from the core) * (Site-specific indication o f an inability to adequately remove heat from the RCS) WCGS IC#(s) SS6 WCGS EAL# SS6.1 WCGS IC Wording Inabi l ity to shut down the reactor causing a challenge to core cooling or RCS heat removal MODE: 1 -Power Operation WCGS EAL Wording An automatic or manual trip fails to shut down the reactor as indicated by reactor power AND All actions to shut down the reactor are not successful as i ndicated by reactor power AND EITHER:

  • CSFST Core Cool i ng RED Path condit i ons met
  • CSFST Heat Sink RED Path conditions met 115 of 118 WCGS Difference/Deviation Just i fication None Difference/Dev i ation Justificat i on As specified in the generic developers guidance " Developers may include site-specific EOP criteria indicative of a successful reactor shutdown in an EAL statement , the Basis or both (e.g., a reactor power level)." Reactor power< 5% is the site-specific ind i cation of a successful reactor trip. Indication that core cooling is extremely challenged is manifested by CSFST Core Cooling RED Path conditions met. Ind i cation that heat removal is extremely challenged is manifested by CSFST Heat Sink RED Path conditions me t.

EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) SSS Loss of all Vital DC power for 15 SS2 Loss of all vital DC power for 15 None minutes or longer. minutes or longer. MODE: Power Operation, MODE: 1 -Power Operation, 2 -Startup, Hot Standby, Hot Startup, 3 -Hot Standby, 4 -Hot Shutdown Shutdown NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 Indicated voltage is less than SS2.1 Loss of all 125 VDC power 105 VDC is the site-specific minimum vital DC bus voltage. (site-specific bus voltage value) based on battery bus voltage DC buses NK01, NK03 (Division

1) and NK02, NK04 (Division
2) are on ALL (site-specific Vital DC indications

< 105 voe on all the site-specific vital DC buses. busses) for 15 minutes or longer. vital DC buses NK01, NK03 (Division

1) and NK02, NK04 (Division
2) 15 min. (Note 1) Note The Emergency Director should N/A Note 1: The Emergency The classification timeliness note has been standardized across the declare the Unusual Event Manager should declare the WCGS EAL scheme by referencing the "time limit" specified within promptly upon determining that event promptly upon the EAL wording. 15 minutes has been exceeded, determining that time limit has or will likely be exceeded.

been exceeded, or will likely be exceeded.

116 of 118 EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviat i on Justificat i on IC#(s) SG1 Prolonged loss of all offsite and SG1a Prolonged loss of all offsite and None all onsite AC power to all onsite AC power to emergency buses. emergency buses MODE: Power Operat i on , MODE: 1 -Power Operation , 2 -Startup , Hot Standby , Hot Startup , 3 -Hot Standby , 4 -Hot Shutdown Shutdown NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording D i fference/Deviat i on Justificat i on EAL# EAL# 1 a. Loss of ALL offsite and ALL SG1.1 Loss of all offs i te and all ons i te 4.16KV buses NB01 and NB02 are the site-spec i fic emergency onsite AC power to (site-specific AC power capab i lity to buses. emergen cy buses). emergency 4.16KV buses NB01 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> i s the site-specific SBO coping analys i s t i me. and NB02 AND CSFST Core Cooling RED Path conditions met i ndicates significant

b. EITHER of the follow i ng: AND EITHER: core ex i t superheating and core uncovery. Restoration of at least
  • Restoration of at least one
  • emergency bus i n < 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> one AC emergency bus i n less than (site-specific i s not likely (Note 1) hours) is not likely.
  • CSFST Core Cooling RED Path cond i tions met * (S i te-specific indication of an i nability to adequately remove heat from the core) Note The Emergency Director should N/A Note 1: The Emergency The classification timeliness note has been standardized across the declare the General Emergency Manager should WCGS EAL scheme by referencing the "time l i m it" specified within promptly upon determin i ng that declare the event the EAL wording. (site-specific hours) has been promptly upon exceeded , or will likely be determining that time exceeded. limit has been exceeded , or will likely be exceeded. 1 17 of 118 l EAL Comparison Matrix WCGS NEI IC# NEI IC Wording WCGS WCGS IC Wording Difference/Deviation Justification IC#(s) SGS Loss of all AC and Vital DC SG1b Loss of all AC and vital DC None power sources for 15 minutes or power sources for 15 minutes or longer. longer MODE: Power Operation, MODE: 1 -Power Operation, 2 -Startup, Heit Standby, Hot Startup, 3 -Hot Standby, 4 -Hot Shutdown Shutdown NEI Ex. NEI Example EAL Wording WCGS WCGS EAL Wording Difference/Deviation Justification EAL# EAL# 1 a. Loss of ALL offsite and ALL SG1.2 Loss of all offsite and all onsite 4.16KV buses NB01 and NB02 are the site-specific emergency onsite AC power to (site-specific AC power capability to buses. emergency buses) for 15 minutes emergency 4.16KV buses NB01 105 VDC is the site-specific minimum vital DC bus voltage. or longer. and NB02 for;:: 15 min. AND AND NK01, NK03 (Division
1) and NK02, NK04 (Division
2) are the site-specific vital DC buses. b. Indicated voltage is less than Loss of all 125 VDC power (site-specific bus voltage value) based on battery bus voltage on ALL (site-specific Vital DC indications

< 105 voe on all busses) for 15 minutes or longer. vital DC buses NK01, NK03 (Division

1) and NK02, NK04 (Division
2) for;:: 15 min. (Note 1) Note The Emergency Director should N/A Note 1: The Emergency The classification timeliness note has been standardized across the declare the Unusual Event Manager should declare the WCGS EAL scheme by referencing the "time limit" specified within promptly upon determining that 15 event promptly upon the EAL wording. minutes has been exceeded, or determining that time limit has will likely be exceeded.

been exceeded, or will likely be exceeded.

118 of 118 Enclosure IV Wolf Creek Nuclear Operating Corporation EP-CALC-WCNOC-1601 Revision 0 "Radiological Effluent EAL Values" (23 Pages)

Wolf Creek Nuclear Operating Corporation r-----**-* -------------------.

I I ; -:,-. --.t(\5 ___ ! Wolf Creek Nuclear

  • ' "' Operating Corporation (WC NOC) Radiological Effluent EAL Values EP-CALC-WCNOC-1601 REVISION 0 Document Author: Scott McCain Chemistry Reviewer:

Kurtis Mitchell EPS Approval:

Tim East Document Author: Chemistry Reviewer: 07/27/16 07/27/16 EPM Approval:

WCNOC EAL Technical Bases Calculations

-Rx1 Effluent Series Table of Contents 1. PURPOSE ............................................................................................................................

3 2. DEVELOPMENT METHODS AND BASES ...........................................................................

3 2.1. Threshold Limits ...........................................................................................................

3 2.2. Effluent Release Points ................................................................................................

6 2.3. Source Term .................................................................................................................

7 2.4. Release Duration ..........................................................................................................

8 2.5. Meteorological lnputs ....................................................................................................

9 3. DESIGN INPUTS ................................................................................................................

10 3.1. General Constants and Conversion Factors ...............................................................

10 3.2. Liquid Effluent ....................................

,. ........................................................................

10 3.3. Gaseous Effluent ........................................................................................................

11 4. CALCULATIONS

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

12 4.1 . RU 1.1 Liquid Release .................................................................................................

12 4.2. RU1 .1 Gaseous Release ............................................................................................

13 4.3. RA1.1, RS1.1 and RG1.1 Gaseous Release ..............................................................

13 5. CONCLUSIONS

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

14 6. REFERENCES

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

15 ATTACHMENTS Attachment 1, RU 1.1 Liquid Effluent EAL Calculations

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

16 Attachment 2, RU1 .1 Gaseous Effluent EAL Calculations

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

17 Attachment 3, RA 1.1, RS 1.1 and RG 1.1 EDCP Dose Assessments

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

18 EP-CALC-WCNOC-1601 Page 2 of 23 Revision 0 WCNOC EAL Technical Bases Calculations

-Rx1 Effluent Series 1. PURPOSE The Wolf Creek Nuclear Operating Corporation (WCNOC) Emergency Action Level (EAL) Technical Bases Manual contains background information, event declaration thresholds, bases and references for the EAL and Fission Product Barrier (FPB) values used to implement the Nuclear Energy Institute (NEI) 99-01 Rev. 6 EAL guidance. This calculation document provides additional technical detail specific to the derivation of the gaseous and liquid radiological effluent EAL values developed in accordance with the guidance in NEI 99-01 Rev. 6. Documentation of the assumptions, calculations and results are provided for the WCNOC Rx1 series EAL effluent monitor values associated the NEI 99-01 Rev 6 EALs listed below.

  • NEI EAL AU1 .1 (gaseous and liquid)
  • NEI EAL M1 .1 (gaseous and liquid)
  • NEI EAL AS1 .1 (gaseous)
  • NEI EAL AG1 .1 (gaseous)
2. DEVELOPMENT METHODS AND BASES 2.1. Threshold Limits 2.1.1. RU1 .1 Liquid Threshold Limits Guidance Criteria RU 1. 1 for WC NOC RA 1.1 for WCNOC RS 1.1 for WC NOC RG 1.1 for WC NOC The RU1 Initiating Condition (IC) a release of gaseous or liquid radioactivity greater than 2 times the Offsite Dose Calculation Manual (ODCM) limits for 60 minutes or longer. WCNOC Bases AP O?B-003 Attachment A Section 2.1 states that the concentration of radioactive material released in liquid effluents from the site to unrestricted areas shall be limited as follows:
  • 10 times the limit specified in 10 CFR Part 20, Appendix B, Table 2, Column 2,for radionuclides other than dissolved or entrained noble gases
  • 2.0E-04 µCi/ml total activity for dissolved and entrained noble gases The site specific RU1 .1 liquid effluent EAL threshold values will equate to 2 times the ODCM limit. EP-CALC-WCNOC-1601 Page 3of23 Revision 0 I I J WCNOC EAL Technical Bases Calculations

-Rx1 Effluent Series 2.1.2. RU1 .1 Gaseous Threshold Limits Guidance Criteria The RU1 Initiating Condition (IC) addresses a release of gaseous or liquid radioactivity greater than 2 times the Offsite Dose Calculation Manual limits for 60 minutes or longer. WCNOC Bases AP 078-003 Attachment A Section 3.1 states that the dose rate due to radioactive materials released in gaseous effluents from the site to areas at or beyond the site boundary shall be limited to the following:

  • Less than or equal to 500 mrem/yr to the whole body (Noble Gasses)
  • Less than or equal to 3000 mrem/yr to the skin (Noble Gasses)
  • Less than or equal to 1500 mrem/yr to any organ (1-131, 1-133, tritium, and particulate with half-lives greater than 8 days) The inhalation (internal organ) ODCM dose limit is not applicable for the Unusual Event EAL threshold since; (1) individual uptake dose assessment is highly variable and requires real time representative samples, and (2) a predominant Noble Gas source term (no core damage) is assumed for the unusual event accident level. The site specific RU1 .1 gaseous effluent EAL threshold values will equate to 2 times the ODCM limit for the lesser of the whole body or skin exposure pathways.

2.1.3. RA 1.1 Liquid Threshold Limits Guidance Criteria The RA1 Initiating Condition (IC) addresses a release of radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid COE. This is based on values at 1 % of the EPA Protective Action Guides (PAGs). Per NEI 99-01, the effluent monitor readings should correspond to the above dose limits at the "site-specific dose receptor point" (consistent with the calculation methods employed) for one hour of exposure.

WCNOC Bases Liquid effluent limits are based on water concentration values given in 10 CFR 20 Appendix 8 Table 2 Column 2 (see Section 2.1.1 above). Those values are equivalent to the radionuclide concentrations which, if ingested continuously over the course of a year, would produce a total effective dose equivalent of 0.05 rem (50 millirem).

The EPA PAGs are based on a TEDE dose from atmospheric immersion, inhalation and ground deposition.

The 10 CFR 20 limits and the EPA limits do not represent the same type of exposure and thus cannot be compared on a one to one basis. Thus, the site specific EALs will not contain an RA 1.1 liquid effluent monitor threshold value that equates to 1 % of the EPA PAG. However, EALs RA 1.3 and RA 1.4 will remain applicable for liquid effluent releases that exceed the threshold based upon sample and field survey results. EP-CALC-WCNOC-1601 Page 4 of 23 Revision 0 WCNOC EAL Technical Bases Calculations

-Rx1 Effluent Series 2.1.4. RA 1.1 Gaseous Threshold Limits Guidance Criteria The RA 1 IC addresses a release of radioactivity resulting in offsite dose greater than 10 mrem TEOE or 50 mrem thyroid COE. Per NEI 99-01, the effluent monitor readings are based on values at 1 % of the EPA Protective Action Guides (PAGs) at the "site-specific dose receptor point" (consistent with the calculation methods employed) for one hour of exposure.

WCNOC Bases The gaseous effluent limits for RA 1.1 are based on values that equate to an offsite dose greater than 10 mrem TEOE or 50 mrem COE thyroid, which are 1 % of the EPA PAGs. 2.1.5. RS1 .1 Gaseous Threshold Limits Guidance Criteria The RS1 IC addresses a release of radioactivity resulting in offsite dose greater than 100 mrem TEOE or 500 mrem thyroid COE. This is based on values at 10% of the EPA Protective Action Guides (PAGs) at the specific dose receptor point" (consistent with the calculation methods employed) for one hour of exposure.

WCNOC Bases The gaseous effluent limits for RS1 .1 are based on values that equate to an offsite dose greater than 100 mrem TEOE or 500 mrem COE thyroid, which are 10% of the EPA PAGs. 2.1.6. RG1 .1 Gaseous Threshold Limits Guidance Criteria The RG1 IC addresses a release of radioactivity resulting in offsite dose greater than 1,000 mrem TEOE or 5,000 mrem thyroid COE. This is based on values at 100% of the EPA Protective Action Guides (PAGs) at the "site-specific dose receptor point" (consistent with the calculation methods employed) for one hour of exposure.

WCNOC Bases The gaseous effluent limits for RG1 .1 are based on values that equate to an offsite dose greater than 1,000 mrem TEOE or 5,000 mrem COE thyroid, which are 100% of the EPA PAGs. EP-CALC-WCNOC-1601 Page 5 of 23 Revision 0 WCNOC EAL Technical Bases Calculations

-Rx.1 Effluent Series 2.2. Effluent Release Points Note -All effluent release points assume a background reading of zero to conservatively account for all modes of operation applicable to the EALs. 2.2.1. Liquid Release Points Guidance Criteria Per NEI 99-01, the RU1 IC addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways (NEI AU1 EAL #1) and planned batch releases from non-continuous release pathways (NEI AU1 EAL #2). Per NEI 99-01, the RA 1 IC includes events or conditions involving a radiological release, whether gaseous or liquid, monitored or un-monitored.

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.

The "site-specific monitor list 'and threshold values" should be determined with consideration of the selection of the appropriate installed gaseous and liquid effluent monitors.

WCNOC Bases There are three monitors associated with continuous liquid releases that discharge to the environment at WCNOC (ODCM Appendix A Section 2.4.4.1) Monitor ID Description O-BM-RE-52 Steam Generator Slowdown Discharge Monitor O-LE-RE-59 Turbine Building Drain Monitor O-HF-RE-95 Waste Water Treatment System Influent Monitor There are two monitors associated with batch liquid releases that discharge to the environment at WCNOC (ODCM Appendix A Section 2.4.4.2) Monitor ID Description O-HB-RE-18 Liquid Radwaste Discharge Monitor O-HF-RE-45 Secondary Liquid Waste System Monitor 2.2.2. Gaseous Release Points Guidance Criteria Per NEI 99-01, the RU1 IC addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways (EAL #1) and planned batch releases from non-continuous release pathways (EAL #2). EP-CALC-WCNOC-1601 Page 6 of 23 Revision 0 WCNOC EAL Technical Bases Calculations

-Rx.1 Effluent Series Per NEI 99-01, the RA1 IC includes events or conditions involving a radiological release, whether gaseous or liquid, monitored or un-monitored.

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.

Per NEI 99-01, the RS1 and RG1 ICs address monitored and un-monitored releases of gaseous radioactivity.

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.

The "site-specific monitor list and threshold values" should include the effluent monitors described in emergency plan and emergency dose assessment procedures.

WCNOC Bases Gaseous effluent releases from the Unit Vent and Radwaste Building Vent are monitored continuously.

The Unit Vent is the release point for the Fuel/Auxiliary Building, access control area, containment purge, and condenser air discharge.

The Radwaste Building Vent is the release point for Waste Gas Decay Tanks and the Radwaste Building Ventilation System. Waste Gas Decay Tank releases and Containment Building releases are treated as batch releases.

Containment Building releases (purges) are monitored by the Plant Unit Vent Monitor. Waste Gas Decay Tank releases are monitored by the Radwaste Building Exhaust Monitor. Monitor ID O-GT-RE-21 A/B O-GH-RE-10 A/B 2.3. Source Term Description Unit Vent Radwaste Building Vent 2.3.1. RU1 .1 Liquid Source Term Guidance Criteria NEI 99-01 does not provide source term guidance for the RU1 .1 EAL liquid source term assumptions.

WCNOC Bases Per ODCM Attachment A Section 2.4.4.1, monitor setpoints for liquid effluent pathways will be conservatively based on 1-131, which the most restrictive isotope expected to be present. The 10 CFR 20 Appendix B Table 2 Column 2 1-131 limit is 1 E-6 µCi/ml. EP-CALC-WCNOC-1601 Page 7 of 23 Revision 0 WCNOC EAL Technical Bases Calculations

-Rx1 Effluent Series 2.3.2. RU1 .1 Gaseous Source Term Guidance Criteria NEI 99-01 does not provide source term guidance for the RU1 .1 gaseous source term assumptions.

WCNOC Bases ODCM setpoints for gaseous monitors are based on the source activity taken from USAR Table 11.1-1 (as referenced from ODCM Attachment A Section 3.4.4.2.1

). 2.3.3. RA1.1. RS1.1 and RG1.1 Gaseous Source Terms Guidance Criteria NEI 99-01 specifies that the calculation of monitor readings will require use of an assumed release isotopic mix; the selected mix should be the same for ICs AA1, AS1 andAG1. WCNOC Bases The source term utilized for the Unit and Radwaste vent EAL thresholds is provided by the Emergency Dose Calculation Program (EDCP) dose model with the following settings:

  • LOCA event with clad damage
  • Particulate filtration on A zero (0) time after shutdown (TAS) was chosen for the source decay period based on the assumption that the event, trip and release occur simultaneously.

2.4. Release Duration Guidance Criteria Per NEI 99-01, the effluent monitor readings for RA1 .1, RS1 .1 and RG1 .1 gaseous EAL threshold values should correspond to a dose at the "site-specific dose receptor point" (consistent with the calculation methods employed) for one hour of exposure.

WCNOC Bases The effluent monitor readings for RA1 .1, RS1 .1 and RG1 .1 gaseous EAL threshold values are calculated for a release duration of one hour. EP-CALC-WCNOC-1601 Page 8 of 23 Revision 0 WCNOC EAL Technical Bases Calculations

-Rx1 Effluent Series 2.5. Meteorological Inputs Guidance Criteria The effluent monitor readings should correspond to the applicable dose limit at the specific dose receptor point." The "site-specific dose receptor point" is the distance(s) and/or locations used by the licensee to distinguish between on-site and offsite doses. The selected distance(s) and/or locations should reflect the content of the emergency plan, and the procedural methods used to determine offsite doses and protective action recommendations.

This is typically the boundary of the Owner Controlled Area. Monitor readings will be calculated using a set of assumed meteorological data or atmospheric dispersion factors; the data or factors selected for use should be the same for ICs M1, AS1 and AG1. WCNOC Bases The site specific meteorology used for selections and inputs in the EDCP dose assessment model is based upon the USAR as documented below. 2.5.1. Limiting Site Boundary Direction The WCNOC site boundary is equivalent to the Exclusion areas boundary described in USAR Section 2.1.2.1. It is a 1,200 meter radius circle centered on the reactor. Thus, there is no limiting site boundary direction that would result in a more limiting X/Q dispersion value. A wind direction from 180° is utilized for the EDCP dose assessment cases. 2.5.2. Stability The predominant stability class based on annual percent occurrence is Class D from the total % occurrences given in USAR Section 2.3.2.1.5.

2.5.3. Wind Speed The mean wind speed for a D stability class is 13.1 knots (15 mph) given in USAR Section 2.3.2.1.5.

2.5.4. Dispersion Inputs Note -All releases from the plant are considered as ground releases (USAR Section 2.3.5.1.2.1

). RU1 .1 site boundary atmospheric dispersion X/Q value of 2.2E-06 sec/m 3 is obtained from ODCM Attachment A Section 3.1.2. RA1 .1, RS1 .1 and RG1 .1 site boundary atmospheric dispersion X/Q values were developed from the EDCP dose assessment model using the USAR mean wind speed and most predominant stability as inputs. 2.5.5. Other Parameters No precipitation is assumed to occur for the duration of the release and plume transport throughout the EPZ. EP-CALC-WCNOC-1601 Page 9 of 23 Revision 0 WCNOC EAL Technical Bases Calculations

-Rx1 Effluent Series 3. DESIGN INPUTS 3.1. General Constants and Conversion Factors 3.1.1. 10 6 µCi per Ci 3.2. Liquid Effluent 3.2.1. Liquid Effluent Monitor Ranges (UFSAR Table 11.5-2) 1) Steam Generator Slowdown Discharge Monitor (O-BM-RE-52)

..... 1 E-7 to 1 E-2 µCi/cc 2) Turbine Building Drain Monitor (O-LE-RE-59)

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

1 E-7 to 1 E-2 µCi/cc 3) Waste Water Treatment System Influent Monitor (O-HF-RE-95)

... 1 E-7 to 1 E-2 µCi/cc 4) Liquid Radwaste Discharge Monitor (O-HB-RE-18)

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

1 E-7 to 1 E-2 µCi/cc 5) Secondary Liquid Waste System Monitor (O-HF-RE-45)

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

1 E-7 to 1 E-2 µCi/cc 3.2.2. Liquid Effluent Dilution Flow (A Normal Situations Note -a single circ pump is assumed for dilution on the applicable discharge points. 1) Turbine Building Sump (Al 078-019 6.2.15.2)

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

0 gpm 2) All Other Liquid Discharge Points (Al 078-019 6.2.15.1)

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

1.24E+5 gpm 3.2.3. Liquid Effluent Source Flow (fl 1) SG Slowdown (Al 078-036 Section 6.1.5) .....................................................

360 gpm 2) Turbine Building Drain (Al 078-036 Section 6.1.5) .........................................

500 gpm 3) Waste Water Treatment System (Al 078-036 Section 6.1.5) ......................

1,200 gpm 4) Liquid Radwaste Discharge (Al 078-036 Section 6.1.5) .................................

100 gpm 5) Secondary Liquid Waste System (Al 078-036 Section 6.1.5) .........................

200 gpm 3.2.4. Source Term Limit (C;) 10 CFR 20 Appendix B, Table 2, Column 21-131 Limit. ...............................

.. 1E-6 µCi/ml ODCM 1-131 Limit.. .........................................................................................

1E-5 µCi/ml EP-CALC-WCNOC-1601 Page 10 of 23 Revision 0 WCNOC EAL Technical Bases Calculations

-Rx1 Effluent Series 3.3. Gaseous Effluent 3.3.1. Vent Flow (USAR Table 11.5-4) 1) Unit Vent. ...................................................................................................

66,000 cfm 2) Radwaste Building Vent.. ...........................................................................

12,000 cfm 3.3.2. Gaseous Effluent Monitor Ranges (UFSAR Table 11.5-2) 1) Unit Vent (O-GT-RE-21 B) ............................................................

1 E-7 to 1 E+5 µCi/cc 2) Radwaste Building Vent (O-GH-RE-10B)

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

1E-7 to 1E+5 µCi/cc 3.3.3. RU1 .1 Dispersion Factor (X/Q) Dispersion Factor (ODCM Attachment A Section 3.1.2) ............................

2.2E-06 sec/m 3 3.3.4. RU1 .1 Source Term Fraction (Si) Source activity taken from USAR Table 11.1-1 (as referenced from ODCM Attachment A Section 3.4.4.2.1

). RCS Activity (µCi/g) Source Term Fraction -Si Kr-83m 6.93E-02 1.69E-03 Kr-85m 2.83E-01 6.89E-03 Kr-85 1.18E+OO 2.87E-02 Kr-87 1.84E-01 4.48E-03 Kr-88 5.33E-01 1.30E-02 Kr-89 1.51E-02 3.67E-04 Xe-131m 4.26E-01 1.04E-02 Xe-133m 6.71E-01 1.63E-02

  • Xe-133 3.63E+01 8.83E-01 Xe-135m 7.55E-02 1.84E-03 Xe-135 1.23E+OO 2.99E-02 Xe-137 2.SOE-02 6.81 E-04 Xe-138 1.02E-01 2.48E-03 4.11 E+01 1.00E+OO EP-CALC-WCNOC-1601 Page 11 of 23 Revision 0 WCNOC EAL Technical Bases Calculations

-Rx1 Effluent Series 3.3.5. ODCM Dose Factors (ODCM Attachment A Table A.1-2) Total Body OF Skin Beta OF Gamma Air OF Ki Li Mi (mrem/yr per i.iCi/m 3) (mrem/yr per i.iCi/m 3) (mrad/yr per i.iCi/m 3) Kr-83m 7.56E-02 O.OOE+OO 1.93E+01 Kr-85m 1.17E+03 1.46E+03 1.23E+03 Kr-85 1.61E+01 1.34E+03 1.72E+01 Kr-87 5.92E+03 9.73E+03 6.17E+03 Kr-88 1.47E+04 2.37E+03 1.52E+04 Kr-89 1.66E+04 1.01 E+04 1.73E+04 Xe-131m 9.15E+01 4.76E+02 1.56E+02 Xe-133m 2.51E+02 9.94E+02 3.27E+02 Xe-133 2.94E+02 3.06E+02 3.53E+02 Xe-135m 3.12E+03 7.11E+02 3.36E+03 Xe-135 1.81 E+03 1.86E+03 1.92E+03 Xe-137 1.42E+03 1.22E+04 1.51E+03 Xe-138 8.83E+03 4.13E+03 9.21E+03 4. CALCULATIONS 4.1. RU1 .1 Liquid Release 4.1.1. ODCM Liquid Release Limit (ODCM Appendix A Section 2.4.4.1) . F(gpm) + f (gpm) S(µCi/ml)

= ECL 1_131 f (gpm) x AF x SF Where: S radiation monitor setpoint (µCi/ml) ECL1-131 effluent concentration limit of 1-131, (µCi/ml) F dilution flow (gpm) f effluent (waste water) flow rate (gpm) AF allocation fraction (Not used for EAL calculations)

SF safety factor (Not used for EAL calculations) 4.1.2. RU1 .1 Liquid Release EAL Threshold RU1 .1 liquid is two times the calculated ODCM limit setpoint.

See Attachment 1 for the spreadsheet calculations that develop the RU1 .1 liquid effluent EAL threshold values for each applicable monitor. EP-CALC-WCNOC-1601 Page 12 of 23 Revision O WCNOC EAL Technical Bases Calculations

-Rx.1 Effluent Series 4.2. RU1 .1 Gaseous Release 4.2.1. ODCM Gaseous Total Body Dose Limit (ODCM Appendix A Section 3.4.4.1.1)

Where: total body setpoint for the given ODCM limit (µCi/sec)

ODCM whole body limit (mrem/yr) allocation fraction (not used for EAL calculations) safety factor (not used for EAL calculations)

Stb 500 AF SF XIQRP Si highest annual average dispersion factor for the release point (sec/m 3) activity released (fraction

-unit less) Ki total body dose correction factor (mrem/yr per µCi/m 3) 4.2.2. ODCM Gaseous Skin Dose Limit (ODCM Appendix A Section 3.4.4.1.2) 3000 x AF x SF Sskin = X IQ x i:csi x (Li + i.1Mi)) Where: Sskin RP skin setpoint for the given ODCM limit (µCi/sec)

ODCM skin limit (mrem/yr) allocation fraction (not used for EAL calculations) safety factor (not used for EAL calculations) 3000 AF SF XIQRP highest annual average dispersion factor for the release point (sec/m 3) Si activity released (fraction

-unit less) Li + 1.1 Mi skin dose correction factor (mrem/yr per µCi/m 3) 4.2.3. RU1 .1 Gaseous Release EAL Threshold RU 1.1 gaseous is two times the lesser of the calculated total body or skin ODCM limit setpoint.

See Attachment 2 for the spreadsheet calculations that develop the RU1 .1 gaseous effluent EAL threshold values for each applicable monitor. 4.3. RA1.1. RS1.1 and RG1.1 Gaseous Release The RA1.1, RS1.1 and RG1.1 gaseous release EAL thresholds are developed using the site specific EDCP dose assessment model with the inputs described in Sections 2.3.3, 2.5 and 3.3. Refer to Attachment 3 for the results of the EDCP gaseous effluent EAL threshold calculations.

EP-CALC-WCNOC-1601 Page 13 of 23 Revision 0 WCNOC EAL Technical Bases Calculations

-Rx1 Effluent Series 5. CONCLUSIONS Monitor GE SAE Alert OE Unit Vent (EFF) O-GT-RE-21 B 4.45E+8 4.45E+7 4.45E+6 7.85E+5 : (µCi/sec)

(µCi/sec)

(µCi/sec)

(µCi/sec) 0 l CD ; I/) Radwaste Vent O-GH-RE-1 OB 4.45E+8 4.45E+7 4.45E+6 7.85E+5 (EFF) (µCi/sec)

(µCi/sec)

(µCi/sec)

(µCi/sec)

' Steam Generator 6.91 E-3 Slowdown O-BM-RE-52 N/A N/A N/A (µCi/ml) Discharge Monitor ' Turbine Building O-LE-RE-59 N/A N/A N/A 2.00E-5 1 Drain Monitor (µCi/ml) ftS * '* Waste Water 2.09E-3 O
Treatment System O-HF-RE-95 N/A N/A N/A (µCi/ml) . Influent Monitor er :J Liquid Radwaste O-HB-RE-18 N/A N/A N/A 1.00E-2 Discharge Monitor (µCi/ml) Secondary Liquid 1.00E-2 Waste System O-HF-RE-45 NIA N/A N/A (µCi/ml) Monitor Currently, the highest procedural alarm setpoint for the Unit Vent is 3.15E+5 µCi/sec (Al 078-026 Section 6.6.6). The UE threshold for the Unit Vent is sufficiently above the highest alarm setpoint for that monitor. Currently, the highest procedural alarm setpoint for the Radwaste Vent is 6.23E+4 µCi/sec (Al 078-026 Section 6.6.6). The UE threshold for the Radwaste Vent is sufficiently above the highest alarm setpoint for that monitor. Currently, the highest procedural alarm setpoint for the Steam Generator Slowdown Discharge Monitor is 8.64E-4 µCi/ml based on the ECL for 1-131 and an allocation factor of 0.25 (Al 078-036 Section 6.1.17). The UE threshold for the Steam Generator Slowdown Discharge Monitor.is sufficiently above the highest alarm setpoint for that monitor. Currently, the highest procedural alarm setpoint for the Turbine Building Drain Monitor is 1.00E-5 µCi/ml based on the ECL for 1-131 (Al 078-036 Section 6.1.17). The UE threshold for the Turbine Building Drain Monitor is sufficiently above the highest alarm setpoint for that monitor. Currently, the highest procedural alarm setpoint for the Waste Water Treatment System Influent Monitor is 2.61 E-4 µCi/ml based on the ECL for 1-131 and an allocation factor of 0.25 (Al 078-036 Section 6.1.17). The UE threshold for the Waste Water Treatment System Influent Monitor is sufficiently above the highest alarm setpoint for that monitor. EP-CALC-WCNOC-1601 Page 14 of 23 Revision O WCNOC EAL Technical Bases Calculations

-Rx1 Effluent Series Currently, the highest procedural alarm setpoint for the Liquid Radwaste Discharge and the Secondary Liquid Waste System monitors is 1.0E-2 µCi/ml (Al 078-019 Section 6.4.1.6.f.2.c).

The UE threshold for the Liquid Radwaste Discharge and the Secondary Liquid Waste System monitors are lowered to the maximum readable value to ensure the ability to identify the threshold is established.

6. REFERENCES 6.1. 10 CFR Part 20, Appendix 8, Table 2, Column 2 6.2. NEI 99-01 R6, Methodology for Development of Emergency Action Levels, November 2012 6.3. AP 078-003, Offsite Dose Calculation Manual, Revision 8 6.4. AP 078-004, Offsite Dose Calculation Manual (Radiological Environmental Monitoring Program), Revision 21 6.5. Al 078-019, Instructions For Liquid Release Permits, Revision 20 6.6. Al 078-026, Instructions for Unit Vent and Radwaste Vent Permits, Revision 12 6.7. Al 078-036, Liquid Release Permits Using RADEAS, Revision 2 6.8. Wolf Creek Final Safety Analysis Report (USAR)
  • Chapter 2.0, Site Characteristics

-0.125% Fuel Defects

  • Table 11.5-2, Liquid Effluent Radioactivity Monitors
  • Table 11.5-4, Airborne Effluent Radioactivity Monitors EP-CALC-WCNOC-1601 Page 15 of 23 Revision 0 Attachment 1 .. s '2 0 :::& O-BM-RE-52 O-LE-RE-59 O-HF-RE-95 O-HB-RE-18 O-HF-RE-45 EP-CALC-WCNOC-1601 RU1 .1 Liquid Effluent EAL Calculations
  • :II ii E > :::::i "Cl .. cs "' s: LL. .... Q I I c i: i: :s s: 0 0 "' .... u::: u::: * ..J a: -c( -c .. .. -w-0.-c S E .... .e *-:c:i E :::i E *--c *-..: u .2 a. E a. 00 =>,:. i5 s wS :::& :S: a: -1.24E+05 360 3.45E-03 6.91E-03 O.OOE+OO 500 1.00E-05 2.00E-05 1.24E+05 1200 1.04E-03 2.09E-03 1.24E+05 100 1.24E-02 2.48E-02 1.24E+05 200 6.21 E-03 1.24E-02 1-131 10 CFR 20 Limit (µCi/ml): 1.00E-06 1-131 ODCM Limit -ECL (µCi/ml): 1.00E-05 Page 16 of 23 Revision 0 Attachment 2 I .. ca E IL ::: GI 0 (I) :::i. 8 >-Q. 'g .. i8 f j§ i! o E ... _ Kr-83m 7.56E-02 Kr-85rn 1.17E+03 Kr-85 1.61 E+01 Kr-87 5.92E+03 Kr-88 1.47E+04 Kr-89 1.66E+04 Xe-131m 9.15E+01 Xe-133m 2.51E+02 Xe-133 2.94E+02 Xe-135m 3.12E+03 Xe-135 1.81E+03 Xe-137 1.42E+03 Xe-138 8.83E+03 EP-CALC-WCNOC-1601 RU1 .1 Gaseous Effluent EAL Calculations i :J I I .. .. -s_ SM UM u E E ca ::: -IL 0 GI *-(I) 0 GI :::i. 8 = (I) .. 8 & ._ GI *-Q. J! !;. < .. ill e E -g .5 i! E ca E U) -C> -U) I c 0 :Cl u l'!! IL e 'i:' >--*-E lll:: GI :I )( a:: 0 *-E U) U) -O.OOE+OO 1.93E+01 1.69E-03 1.27E-04 1.46E+03 1.23E+03 6.89E-03 8.06E+OO 1.34E+03 1.72E+01 2.87E-02 4.62E-01 9.73E+03 6.17E+03 4.48E-03 2.65E+01 2.37E+03 1.52E+04 1.30E-02 1.91E+02 1.01E+04 1.73E+04 3.67E-04 6.10E+OO 4.76E+02 1.56E+02 1.04E-02 9.48E-01 9.94E+02 3.27E+02 1.63E-02 4.10E+OO 3.06E+02 3.53E+02 8.83E-01 2.60E+02 7.11E+02 3.36E+03 1.84E-03 5.73E+OO 1.86E+03 1.92E+03 2.99E-02 5.42E+01 1.22E+04 1.51 E+03 6.81 E-04 9.67E-01 4.13E+03 9.21E+03 2.48E-03 2.19E+01 1.00E+OO 5.79E+02 Calculation Constants Total Body Do Skin Do se Rate Limit (mRem/yr): se Rate Limit (mRem/yr): XJQ (sec/m3): DCF (mRad to mRem): Calculation Results t for Total Body (µCi/sec): ODCM Limi ODCM Limit for Skin (µCi/sec): 2x ODCM Limit (µCi/sec): Page 17 of 23 -i """: ... 'i:' + >--:J E -GI )( a:: *-E U) -3.58E-02 1.94E+01 3.90E+01 7.40E+01 2.48E+02 1.07E+01 6.71E+OO 2.21E+01 6.13E+02 8.10E+OO 1.19E+02 9.44E+OO 3.54E+01 1.20E+03 500 3000 2.20E-06 1.1 3.92E+05 1.13E+06 7.85E+05 Revision 0 Attachment 3 RA1.1, RS1.1 and RG1.1 EDCP Dose Assessments O-GT-RE-21 B Unit Vent -Alert EOCP
  • V ersio n S.2 *R a d iation Mo nitorin g S ystem Wind Speed Stabi l ity C ass (A-G) Wind D ir ect ion F ro m 18 0 t o 0 Pro je cted Rel e as e Dura ion ( rs) 1 3 D 1 Gaseous Activity (uCi/cc) l odi Obie Gas Ra 0 Particu ate Activity ( C i/cc) Ven Fl ow (cfm) .43E-0 1 5.75E-03 2 87E-0 3 6.60E+04 Ti m e S in ce RX T ri p I RCS Sa m p le (hrs) 0 Start I 1 me o t t<e ease: Re arks Pe orm edB y _________ _ Approved By:----------Ve f1 B Y...,...--------

--Tlme C a lc u l a t ons P erf o nn&d: Oa1 Time: __ _ Thurs d a y , June 2 3 1 201612: 06:17 PM R ele aH Rat H (Cl/u c): Part: 2.87E-0 3 Noble Gas: 4.45E+OO I odine: 2.56 E-02 PAR's Do ses ng P A R s Inclu de S u mmed n d P Ojec t e d Do s . *cc L *JRR Eva cu ate Pr o iect ed Ce D i stance EAB 2MI 5M I 10 M I e r hne Dose Seg m nt T EDE;(RE M} 233E 840E

  • J RR a d CC L are r ecom m en d ed for evacua *on pon declaration of an SAE o above. Dose Rate R/H Distance TE D E Dose Ra t e R/H r Thyroid EAB 1.95E-0 2 2 M l 4. SE-03 5 M l 1.0SE-03 10 Ml 3.79E-04 Doses a bute d to su b zones , bot T EDE Thy Dose SBZ (REM) (R E M) CTR 1 COE 2 1. 32c 2 CCL 1 OOE-2 '-32E 2 JRR O.OOE 0 OOOE+O N 1 1.6 6E 3 7 1 E-3 N E 1 OOE*O 0 0[* E1 000[ 0 OOOE+O SE1 0.0 E+O 0 OOEt-0 S1 OOOE+O coo + 4.5 E-0 2 9 65E-03 2.4 3E-0 3 8.76 E-04 SBZ SW1 W 1 NW 1 N2 NE2 NE3 E2 I 1od 1ne F dtrat l on S ta lu >= O N (90%)11 P art icul a Calcu la: ion W<rtSneet.

d ot EP-CALC-WCNOC-1 60 1 Do se R ate Plu m e Arnv al Mi utes 3 9 23 46. -.,SE*S 0 OOOE+ Page 1 8 of 2 3 Es Hou r s Un II Evac u a bo n is Necessary. SBZ SE2 SE3 SE4 sz 5W2 W2 NW2 22.3 0 3 8 41 1.1 142 000 t-0 000 0 CODE 0 DODE 0 0 OE*O 0 OOE*O 0 OEt-0 Rev i s i o n 0 Attachment 3 RA1.1, RS1.1 and RG1.1 EDCP Dose Assessments 0-GT-RE-218 Unit Vent-Site Area Emergency EDCP. Vers i on 5.2 *Rad i ation M on i to ri ng System Win d Speed St a bility C l a s s [A-G) Wind D irec t i on From 1 80 t o O P roject d R l eas Duration ( rs) 13 0 Gaseous Activ i ty ( Ci/cc) lodi e/No bl e Gas R atio P art icu l a e Acbvity (uCi/cc) Ve n t Flow (cfm) 1.43 E+OO 5.75 E-03 2.8 7 E-02 6.60E 04 nme Smee RX Tri p I RCS Sample (hrs) o Start I me of Release: Rema rk s: P e rf ormed B y-------V 1f1 By-.,----,,,--,--

-..,..---Ti me Calcula ti ons P e rformed: Approv e d B y:----------Date Time: __ _ Th ursday , Jun e 23 , 2016 1 2: 06: 01 PM Release Rates (C U see): Part 2.87E-02 o bl e Gas: 4.45E+o 1 lodi e: 2.S&E-01 PAR'* Dose s -Warning PAR"& cfu de S u mmed a nd P l'OJ ect ed Dose P ro i e ct ed Cent e ine Do e Seg m e t Evacuate D stance TE DE(RE M) EAB 1 OOE*O 2 M l 2.1 SE-02 5 M l 542E-03 1 0 M l 1 SSE-03 Th y(REM) 4 32E-O 9.2 5 E-02 2.33 -02 8.40E -03

  • JRR a d CCL a re recomme ed f or e vacuation u po de cl a r atio n of SA E or above. D o se R ate R/Hr D i sta nce T E D E EA B 1.9 5 E-0 1 2 M l 4. 8E*02 5 M l 1.0S E-0 2 10 Ml 3.79 E-03 Dose Ra e R/H r Thyr o i d 4.5 E-0 9.65E-02 2.43E-02 8.76E-03 Dose R a te Plum e Arrival M i utes 3. 9. 23 46. DOSE BY SUBZONE Doses a bu eel to s u bzo es , both summed a n d cu rren t segmen s T EDE Th y D ose TEDE T hy Dose SB Z (REM) (RE M) S B Z (REM) (REM) CT R 1 0 E * .3 2 E SW1 0 OOEt OOOE t O Est Hours Unbl E v acua tion is N ece ssa ry. e con s i de red TED E S BZ (REM) 23 10.5 41.5 1 4.9 Thy Dose (RE M) SE2 O.OOEtO 000 + CC L 1 0 E
  • 32[-W 1 0 00[+0 0 OOE*O S E 3 0 OOE 0 OOOE*O JRR 000[ 0 OO O E 0 NW 1 O.OOE 0 OOOE 0 S E 4 0 OOEtO N 1 1.6 6 E 2 [-2 N2 4 90E 3 2.1 E 2 5 2 O.O OE 0 OOO E*O NE 1 00 [*0 0 0 0[*0 NE2 1.85(-4 *SE-4 SW2 0 O O E10 OOOE* E 1 0 00[+0 0 OOE+O NE3 0 00[+0 0 OOE+O W2 oo oc 0 OOOE t SE1 00 E 0 0 +O E2 0 00[+ 0 00[+0 NW2 9 62E-6 4 4 E-5 S 1 OOOE+O 0 00 + S t a tu = O N (90%)1 1 Con n n t Sp<a y Statu : ON (7 5%)

Calcula t ion W ork.She eld EP-CALC-WCNOC

-1 60 1 Page 19 o f 23 Rev i s i on O Attachment 3 RA1.1 , RS1.1 and RG1.1 EDCP Dose Assessments 0-GT-RE-2 1 B Unit Vent -General Emergency EDCP -V e rs on 5.2

  • Rad i ation Mon i to r i n g Sy s tem Wtn d Speed S ability Class {A*GJ Wind D irection From 180 to O Projecte d Release 0 ration ( rs) 1 3 D 1 Gaseous Activity ( Ci/cc) lodinel oble Gas Ra o Pa rt icu a e Acbvrty (uCilcc) Vent Flow (cfm) 1.43 +01 5.75E-03 2.87 -01 8.60 +04 Time Stnoe RX np I RCS Sample (hrs) 0 S t art I 1 me of to(elease: Remarks: P e orme d By _________ _ Approv e d By:----------Ve 'fied By __________

_ Date: TI e* __ _ Ti m e Calculations P e rfonn e d: ThuBday , June 23 , 201612:05: 00 PM Re l ease Rates (C U se c): Part: 2.87E-0 1 o b le Gas: 4.45E+02 lodi e. 2.56E+OO PAR's Doses arning PAR" s i ude Summed and PrOjected Doses. *ccL *J RR CTR Evacuate Proj ed C nte ine Dose Segmen D i stance TEDE(REM) EAB 1 OOEi 2MI 5MI 10MI Thy(REM) 4.32E+OO 9 251:: 2 33E-01 840E 2 "JRR a d CCL a r e recomme ed fo r evacuation upo declara t ion of an SAE or above Dose Rate R/H Distanc e TEDE EAB 1.95E+OO 2 I 4 8 -01 5 Ml 1.05E-01 10 Ml 3.79 -02 Dose Ra e R/H Thyroid 4.5 E+OO 965 -0 2 43E-O 8 76E-02 Dose Rate Plume Arnval Mi utes 3 9 23. 46 DOSE BY SUBZONE Do s attributed to ubzo e , both T EDE Thy Dose SB Z (R EM) (REM) S BZ CTR 1.00E+O 4.32E+O SW1 CCL 1.00E+O 4.32E+O W1 JRR O.OOE+O OOOE NW1 N 1 66E* 7 17E 1 N2 NE 1 00 E*O OOOEt N E2 E 1 OOO E 0 OOOE 0 NE3 SE1 O.OOE 0 0 OE E2 S1 O.OOE+ 0 OOE+O Est HoU/$ Unttl Evacuabon is N ecessary. S BZ SE2 SE3 SE4 S2 SW2. W2. NW2 03 2 45 2.2 Iodine Fii t ra t ion Status .. ON {90% Sta: s

  • ON (90%) Conta en Spra S a t us
  • 0 (75%) E P-CA L C-WCNOC-1 60 1 Page 2 0 o f 23 Re vi s i o n 0 Attachment 3 RA1.1 , RS1.1 and RG1.1 EDCP Dose Assessments O-GH-RE-1 OB Radwaste Vent -Alert EDCP *Version 5.2
  • Ra d i ation M on i t orin g System Win d Spee d S abi ity Cla ss [A-G] Wind D irect i on Fr o m 180 t o O PrOJede d R elease D r ation (hrs) 13 D Gaseous Activrty (uCl/cc) I odine/N o b l e G a s Rati o Part i cu l a e A *
  • y (uC i/cc) Vent Flow (cfm) 7.85E-O 5.75E-0 3 2.86 E-03 1.20E 04 Time S i n ce RX Tnp I RCS Samp (h r s) O Start l me of R elease: Remartcs: Pe ormed B y _________

_ App oved B y:----------V fled By---------Ti me Calcul a tJons Performed: D a te Ti m Thursday , June 23 , 2016 12: 08: 02 PM Release Rat es (Ci/He): P a rt. 2.86 E-0 3 o bl e G a s: 4.45E+OO l odi e: 2.56 E-0 2 P A R's Do ses 'W rn 1 ng PAR'$ l nel ude Summ ed a d P roi c ted Do Projected Ce e r1 1ne Dose Seg en Eva cuate D tance TE D E(REM) *cc L EAB 1 OE 02 " J RR 2M I 2 4E-03 5 I 5 4 E-04 10 M I 1 &SE-04 T hy( R EM) 4 3 E 2 9.23E 3 2 33E-03 8 38E

  • J RR a d CCL are r e commend ed f or e vacu a *on pon d ecla r atio n o f an SA E o abov e. D ose R ate R/H D is ta nce TEDE EAB 1.9 5E-02 2 M 4. 7E-0 3 5 M l 1.05E-03 10MI 3.78E-04 Do se Ra e R/Hr Thy ro id 4.50E-02 9.63E-03 2.43 E-03 8.75E-04 Dose R ate Plume Arnval M i utes 3 9 23. 46. DO SE BY SU BZ O NE Es Hou Until E va c u a n is N ecess a ry. 22.3 04. 411.9 1 44 2 Doses attributed t o subz o es , both sum ed a nd cu rr en t segmen s a r e c o nside r ed. TED E Thy Dose TEDE T h y Dose TEDE T hy Dose SB Z (R EM) (RE M) S B Z (RE M) (RE M) S BZ (REM) {REM) CTR 1 OOE 2 3.E 2 5W1 0 E 0 5 E2 OOOE 0 0 E 0 CC L 1 COE 2 4 3 E 2 W 1 0 OOE+O 5E 3 OOOE+O 0 E 0 J RR 0 OE+O NW 1 0 OE+O S E4 000 +O OOOE 0 N 1 1.66 E 3 7 15[ 3 N 2 4 39 5 2 0 OE+O 0 OOE*O NE 1 O.OOCtO 0 OE*O N E2 84[.5 I SW2. 0 0 ... 10 0 CtO E 1 OOOE 0 0 E+O N E3 000 0 W2. OOOE 0 E 0 SE 1 0 00 0 0 CO +O E2 000 +O NW2 0 0 +O OOOE 0 5 1 0 OOE+O 0 QQC:+Q C alculat io n W ort SheeLdot EP-CALC-WCNOC

-1601 Page 2 1 of 23 Re vi s i on 0 Attachment 3 RA1.1, RS1.1 and RG1.1 EDCP Dose Assessments O-GH-RE-1 OB Radwaste Vent -Site Area Emergency EOCP *Version 5.2

  • Rad i at i on Monitoring System Wind Speed Sta i ity C l ass I A-G) Wi d D ireetio From 180 to 0 Projected Releas Dura i on (h ) 13 0 1 Gaseo s Activity (uCi/cc) I odine/Noble Gas Ra *o Particu l ate Activity ( Ci/cc) Vent ow (cfm) 7.8SE+OO 5.75E-03 2.86E-02 1.20E+o4 Time S r noe RX T p I RCS Sample (hrs) O Start Time ot Re ase: Remarks P rfotmed By----------Approved By:---------Venfi By __________

_ D te Th'lle. __ _ Time CalculatiOM Thuradai.

June 23, 2016 12:07: 48 PM Release Rates lCl/s.c):

Part o le Gas: 4.45E+01 Iodine:

PAR's Doses arni P A s i ncl de Su med and P 01ected Dose . ProJected Ce terl ne Dose Segment Evacuate Dis tance TEDE(REM)

EAB

  • JRR and CCL are recommended or evacuation pon declaration of an SAE or above. Dose Ra e R/Hr O"stance TEDE EAB 1.95E-O 2 M l 4.17E-02 5 M 1.05E-02 10 Ml 3.7BE-03 Dose Rate R/Hr Thyroid 4.50E*01 9.63E*02 2.43E*02 8.75E*03 Dou R a te P ume Arriva I Minutes 3 9 23. 46. DOSE BY SUBZONE Est Hours Un Evacuatio n is Necessary. 2.3 10 5 4 .5 1 5.1 Doses attributed to subzones , bo s mmed and current segments a e considered. TEDE Thy Do e TEOE Thy Doe TEOE Ti'y Dos SBZ (REM) (REM) SBZ (REM) (REM) SBZ (REM) (RE M) CTR 1 OOE 1 4 3 E 1 SW1 0 0 E 0 0 E+O SE2 OOD E 0 OO E 0 CCL 1 OE-1 4 '-11 E-1 W 1 0 OOE*O 000 0 SE3 0 OOE+O 0 OOE+O JRR O.OOE+O OODE 0 tom 0 OOE*O 0 OOE+O SE4 000[ 0 0 OOE+O 1 1.66 -2 1 5E-2 2 4 89E-3 2 10[-2 S2 0 00[+0 000[ 0 E 1 Q QQ IQ OOOC*O NC2 a c-7 4[-4 SW2 oooc 0 0 00(*0 E 1 OOOE 0 0 OOE+O E3 0 OOE+O 0 OOE+O W2 0 +O 000 +O SE1 OOOE 0 000 0 E2 000 *O O OOE+O . 60 -6 4 13 *o S 1 0 OE+O 000 0 l lod 1 trallotl Status= ON (90%)1 1 Partle u late F llration Status= ON (90 Ca 110n WorllShendot EP-CALC-W CNOC-1601 Page 22 of 23 Revision 0 Attachment 3 RA1.1 , RS1.1 and RG1.1 EDCP Dose Assessments O-G H-RE-1 OB Radwaste Vent -General Emergency EDCP *V e rs i on 5.2 *Rad i a ti on M on i tor i ng Sy s t e m Wind Speed Stabi ity C l ass (A*G] Wind D irection From 1 80 to O P r o ected Release 0 tion (h r s} 13 D 1 Gaseous Ac y (uCilcc) Iodine/Noble Gas Ratio Particulate A
  • y (ueilcc) Vent F l ow (cfm) 7.85E+01 5.75E-03 2.86E-O 1.20 E 04 Time Since RX T np I RCS Sample {hrs) 0 8ta I me ot Release: Remarks: P e ormed By _________ _ Approved B y:----------Ve 1ed By __________

_ Oat Time Thu rs day , Ju ne 23 , 2016 1 2: 07: 33 PM T i m e Calcu l a tl on s P e rform e d: Release Rates (C l/He): Part: 2.86E-01 No bl e Gas: 4.45E+02 lodi e: 2.56E+OO PAR'a Do ses '"Warn i ng PAR's i el ude Summed a nd Proj ctcd Doses. *ee L *J R R CTR Evacuate Projected Cente rli ne Do Seg ent 0 stance TE D (REM) EAB 1 OOE+OO 2 Ml 2 4E*O* SMI 54E 2 10 M l 1 95E-02 T hy( REM) 4.3 1 E+OO 9.23 23r 1 8 38E-02

  • JRR and CCL are recommende d fo r evacua on u pon decla r ation o f an SA E or above. Dose Rate RIH Distance TE DE EAB 1.95E+OO 2 M t 4. 7E-O 5 M l 1.05E-O 10 Ml 3.78E-02 Dose Ra e R/Hr Thyroid 4.50E+OO 9.63E-01 2.43E-01 8.75E-02 Dose Rate Plum e Amva l Mi utes 3 9 23 46 DOSE BY SUBZONE Es Hou U n I E vacuation is Necess a ry. 03 2 45 2.2 ed and curre t segments are considered. Thy Dose TEDE Thy Dose TEDE SB Z (REM) S BZ (R EM) (REM) SB Z (R EM) CTR 4.3 E+O SW1 OOOE 0 OOOE 0 S E 2 OC EtO CCL 4.3 1 E+O W 1 OOOE 0 OOOE+O S E 3 OOOE 0 J RR 0 OE+O NW 1 OOOE 0 OOOE t O SE 4 0 OE+O N 1 7 15E 1 N2 489E 2 10E 1 S2 OOOE 0 N E 0 OC10 NE2 84E*3 E-3 SW2. 0 DE*O E 1 00 +O NE3 0 00[+0 OOOE+O W2 OCOE 0 SE1 0 OE+O E2 OOOE 0 0 OOE+O NW2 60E-5 S 1 0 OE+O l od ne F ltratlon Status= O N (9o%) I Pa ieulat F iltration Status= 0 (90%>1 I ContalnŽlnt S ray Status= ON (7 5%) Calcul ation W Shoetdot EP-C ALC-WC N OC-1601 Page 23 of 2 3 Re vi s i o n 0 Enclosure V Wolf Creek Nuclear Operating Corporation EP-CALC-WCNOC-1602 Revision 0 "Containment Radiation EAL Threshold Values" (13 Pages) _J
  • J-*------* ---* 11u-; Wolf Creek Nuclear . -1** -: .. lic':c "-"'*"--i

' l J<F:* -1 Operating Corporation (WC NOC) -.--=_ .. -.*. *. Containment Radiation EAL Threshold Values EP-CALC-WCNOC-1602 Revision 0 Document Author: Scott McCain Technical Reviewer:

William Ketchum EPS Approval:

Tim East Document Author: 07/26/16 Technical Reviewer: ..

-_-*

EPS Approval:

07/27/16 WCNOC EAL Technical Bases Calculations

-CHRRM FPB Series Table of Contents 1. Purpose .................................................................................................................................

3 2. Development Methods and Bases .........................................................................................

3 2.1. Fuel Clad Loss ..............................................................................................................

3 2.2. Reactor Coolant System Loss ......................

., .............................................................

.4 2.3. Containment Potential Loss .........................................................................................

.4 2.4. Source Term .................................................................................................................

5 2.5. Decay Considerations

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

6 3. Design Inputs ........................................................................................................................

7 3.1. Constants and Conversion Factors ...............................................................................

7 3.2. Plant Inputs ..................................................................................................................

7 3.3. Source Term .................................................................................................................

7 4. Calculations

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

8 4.1. Fuel Clad Damage Estimate Based on 300 µCi/g DEl-131 ...............................

...........

8 4.2. GT-RE-59/60 Fission Product Barrier Thresholds

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

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

9 5. Conclusions

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

9 6. References

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

10 Attachments Attachment 1, 300 µCi/g DEl-131 Equivalent Fuel Clad Damage ..............................................

11 Attachment 2, GT-RE-59/60 Fission Product Barrier Threshold Values ....................................

12 Attachment 3, NUREG-1940 Figure 1-1 PWR Containment Monitor Response ........................

13 EP-CALC-WCNOC-1602 Page 2of13 Revision 0 _j WCNOC EAL Technical Bases Calculations

-CHRRM FPB Series 1. PURPOSE The Wolf Creek Nuclear Operating Corporation (WCNOC) Emergency Action Level (EAL) Technical Bases Manual contains background information, event declaration thresholds, bases and references for the EAL and fission product barrier (FPB) values used to implement the Nuclear Energy Institute (NEI) 99-01 Rev. 6 EAL guidance.

This calculation document provides additional technical detail specific to the derivation of the FPB containment high range radiation monitor (CHRRM) readings developed in accordance with the guidance in NEI 99-01 Rev. 6. Documentation of the assumptions, calculations and results are provided for the values associated with the NEI 99-01 Rev. 6 Table 9-F-3, PWR EAL Fission Product Barrier Table, thresholds listed below.

  • NEI Fuel Clad Loss 3.A
  • NEI Containment Potential Loss 3.A 2. DEVELOPMENT METHODS AND BASES 2.1. Fuel Clad Loss Guidance Criteria Per NEI 99-01 Rev. 6, this radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that reactor coolant activity equals 300 µCi/g dose equivalent 1-131 (DEl-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 since it indicates a loss of both the fuel clad and RCS barriers.

The reading should be determined assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory, with RCS radioactivity concentration equal to 300 µCi/g dose equivalent 1-131, into the primary containment atmosphere.

WCNOC Bases The WCNOC fuel clad FPB threshold value is based on an instantaneous release of reactor coolant into the containment at a % fuel clad damage equivalent to 300 µCi/g DEl-131 RCS activity.

That% fuel clad damage value is ratioed to a GT-RE-59/60 reading for 100% fuel clad damage to determine the fuel clad FBP threshold value in R/hr. EP-CALC-WCNOC-1602 Page 3of13 Revision O --------------------------------------

I ----WCNOC EAL Technical Bases Calculations

-CHRRM FPB Series 2.2. Reactor Coolant System Loss Guidance Criteria Per NEI 99-01 Rev. 6, this radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that reactor coolant activity equals Tech Spec allowable limits. This value is lower than that specified for the fuel clad barrier loss threshold since it indicates a loss of the RCS barrier only. The reading should be determined assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory, with RCS activity at Tech Spec allowable limits, into the primary containment atmosphere.

RCS activity at this level will typically result in primary containment radiation levels that can be more readily detected by primary containment radiation monitors, and more readily differentiated from those caused by piping or component "shine" sources. If desired, a plant may use a lesser value of RCS activity for determining this value. In some cases, the site-specific physical location and sensitivity of the containment radiation monitor(s) may be such that radiation from a cloud of released RCS gases cannot be distinguished from radiation emanating from piping and components containing elevated coolant activity.

If so, -determine if an alternate indication is available.

WCNOC Bases WCNOC will adopt NRC guidance as the basis for this EAL threshold.

NUREG-1940 provides estimates for standard plant containment radiation based on spiked RCS activity.

The WCNOC RCS FPB threshold value is based on standard plant containment radiation readings for an instantaneous release of spiked reactor coolant adjusted for the site specific power rating. 2.3. Containment Potential Loss Guidance Criteria Per NEI 99-01 Rev. 6, this radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that 20% of the fuel cladding has failed. This level of fuel clad failure is well above that used to determine the analogous fuel clad and RCS barrier loss thresholds.

NUREG-1228 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 and the fuel clad barriers.

It is therefore prudent to treat this condition as a potential loss of containment which would then escalate the classification level to a General Emergency.

NUREG-1228 provides the basis for using the 20% fuel cladding failure value. Unless there is a site-specific analysis justifying a different value, the reading should be determined assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with 20% fuel clad failure into the primary containment atmosphere.

EP-CALC-WCNOC-1602 Page 4of13 Revision 0 WCNOC EAL Technical Bases Calculations

-CHRRM FPB Series WCNOC Bases The WCNOC containment FPB threshold value is based on an instantaneous release of reactor coolant into the containment at an equivalent of 20% fuel clad damage. The 20% fuel clad damage value is ratioed to a GT-RE-59/60 reading for 100% fuel clad damage to determine the containment FBP threshold value in R/hr. 2.4. Source Term Guidance Criteria NEI 99-01 does not specify a basis for the source term activity or the reduction factors. WCNOC Bases Note -Source term reduction from containment spray is not included as an assumption for these thresholds.

Fuel Clad Damage Equivalent to 300 µCi/g DEl-131 -SAP 99-145, Letter 99-01466, Table 1 core activity was used in conjunction with the NUREG-1465 Table 3.13 halogen release fraction to develop the site specific iodine source term. WCNOC Tech Spec definition of DEl-131 was utilized to determine the iodine conversion factors. The WCNOC Tech Spec definition references EPA-520/1-88-020 (Federal Guidance Report 11) as the basis used to develop the DEl-131 conversion factors. Fuel Clad Loss and Containment Barrier Potential Loss Thresholds

-SAP 99-145, Letter 99-01466, Table 1 core activity was used in conjunction with NUREG-1465 Table 3.13 core to RCS and WCAP-14696-A CDAG CRM3 RCS to containment release fraction assumptions to develop the site specific source term. Per NUREG-1465 Table 3.13, only isotopes from the noble gases, halogens and alkali metals nuclide groups are released from the core to the RCS during a clad failure event. Thus, only the isotopes from those groups, adjusted for decay and abundance considerations, are taken from SAP 99-145, Letter 99-01466, Table 1 for the fuel clad and containment barrier thresholds.

NUREG-1465 Table 3.13 specifies that 3% core activity (noble gases, halogens and alkali metals) may be used as a reduction factor, rather than 5%, if long-term fuel cooling is maintained.

WCAP-14696-A CDAG CRM3 recommends the NUREG-1465 short term fuel clad gap release value of 3% be used. The assumption for this calculation is that long-term fuel cooling is maintained.

WCAP-14696-A CDAG CRM3 specifies that 100% of the noble gas and 50% of the halogens and alkali metals in the RCS are released to containment for the iow RCS pressure case. The assumption for this calculation is that the RCS will depressurize to containment.

RCS Barrier Loss Threshold

-NUREG-1940 Figure 1-1 for spiked coolant is used as a basis to develop this threshold (see Attachment 3). EP-CALC-WCNOC-1602 Page 5 of 13 Revision 0 WCNOC EAL Technical Bases Calculations

-CHRRM FPB Series 2.5. Decay Considerations Guidance Criteria Fission product barrier thresholds and their associated EALs are applicable only when the plant is in Hot Shutdown, Startup, or Power Operation modes (known as the hot operating modes). Per NEI 99-01, the events for these thresholds correspond to an instantaneous release of all reactor coolant mass into the primary containment.

WCNOC Bases This instantaneous release of the RCS to the containment is assumed to occur one hour after the damage event I reactor scram to account for damage progression, dispersion of activity and decay of the very short half-life isotopes.

This is consistent with the WCAP-14696-A and NUREG-1940 assumptions.

EP-CALC-WCNOC-1602 Page 6of13 Revision 0 WCNOC EAL Technical Bases Calculations

-CHRRM FPB Series 3. DESIGN INPUTS 3.1. Constants and Conversion Factors 3.1.1. 453.592 g per lbm water conversion factor 3.2. Plant Inputs 3.2.1. Rated Power

  • Standard Plant (NUREG-1940 Section 1.2.4) .............................................

3,000 MWt

  • Wolf Creek (USAR 1.1.5) ............................................................................

3,565 MWt 3.2.2. RCS Mass at Operating Conditions (USAR Table 6.2.1-5) ........................

5.0452E+5 lbm 3.2.3. GR-RE-59/60 Range (USAR 11.5.2.3.2.4)

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

1 E+O -1 E+8 R/hr 3.3. Source Term 3.3.1. Release Fractions RFcore (NUREG-1465 Table 3.13) RFcore is the fraction of radioactive material released from the core to the fuel cladding (and thus to the RCS) by nuclide groups. The RFcore values are as follows:

  • Noble Gases, Halogens, Alkali Metals -Fuel Clad Damage ........................

0.03 (3%) RFRcs (WCAP-14696-A CDAG CRM3 bases) RF Res is the fraction of radioactive material released from the RCS to the containment by nuclide groups. The RFRcs values are as follows:

  • Noble Gases -Fuel Clad Damage ...............................................................

1 (100%)

  • Non-Noble Gasses -Fuel Clad Damage .....................................................

0.5 (50%) 3.3.2. Source Term Activity (SAP 99-145. Letter 99-01466 Table 1) Core Activity (Ci) 1-131 9.46E+07 1-132 1.37E+08 1-133 1.95E+08 1-134 2.15E+08 1-135 1.83E+08 3.3.3. Iodine Dose Conversion Factors (EPA-520-1-88-020 Table 2.1) Sv/Bct 1-131 2.92E-07 1-132 1.74E-09 1-133 4.86E-08 1-134 2.88E-10 1-135 8.46E-09 EP-CALC-WCNOC-1602 Page 7of13 Revision 0 WCNOC EAL Technical Bases Calculations

-CHRRM FPB Series 4. CALCULATIONS 4.1. Fuel Clad Damage Estimate Based on 300 uCi/g DEl-131 4.1.1. 100% Core Activity Equivalent Reactor Coolant Iodine Concentrations Core RCS Activityi(Ci) x 10 6 100% Core RCS Activityi(µCi/

g) = RCS Mass (g) RCS Activity (uCi/g) 1-131 4.13E+05 1-132 5.99E+05 1-133 8.52E+05 1-134 9.39E+05 1-135 8.00E+OS Total 3.60E+06 4.1.2. 100% Fuel Clad Damage Activity Equivalent Reactor Coolant Iodine Concentrations 100% Clad Damage RCS Activityi(µCi/

g) = 100% Core RCS Activityi(µCi/

g) X RFcore RCS Activity (pCi/g) 1-131 1.24E+04 1-132 1.80E+04 1-133 2.56E+04 1-134 2.82E+04 1-135 2.40E+04 Total 1.08E+05 4.1.3. 100% Fuel Clad Damage Activity Equivalent Reactor Coolant DEl-131 Concentrations 100% DE/ RCS Activity (µCi/ g) = L 100% Clad Damage RCS Activityi(µCi/

g) x DE/ JC Fi The DEl-131 value for each iodine isotope is determined as follows: EPA -520 88 -20 Table 2.1 Iodine DCFi(Sv/Bq) i EPA -520 88 -20 Table 2.1 Iodine DCF 1_131 (Sv/Bq) RCS Activity (uCi/a) 1-131 1.24E+04 1-132 1.07E+02 1-133 4.25E+03 1-134 2.78E+01 1-135 6.95E+02 Total 1.75E+04 EP-CALC-WCNOC-1602 Page 8of13 Revision 0 J WCNOC EAL Technical Bases Calculations

-CHRRM FPB Series 4.1.4. % Fuel Clad Damage Activity Equivalent Reactor Coolant at 300 µCi/g DEl-131 300 µCi/ g % Clad Damage = 100% DE/ RCS Activity (µCi/ g) 300 µCi/g DEl-131 = 1. 72% Fuel Clad Damage See Attachment 1 for the spreadsheet calculations that develop the fuel clad damage source term activity and the % fuel clad damage. 4.2. GT-RE-59/60 Fission Product Barrier Thresholds 4.2.1. Containment Potential Loss (20% Fuel Clad Damage GT-RE-59/60 Reading) GT-RE -59/60 20 0/octaiR/hr)

=GT-RE -59/60 100 0/octad(R/hr)

X 0.20 Containment Potential Loss Threshold GT-RE-59/60 Reading = 6.68E+03 R/hr See Attachment 2 for the spreadsheet calculations that develop the 20% fuel clad damage RM-30/31 reading. 4.2.2. Fuel Clad Loss (300% µCi/g DEl-131 Equivalent Clad Damage GT-RE-59/60 Reading) GT-RE-59/60 1.nO/oclaiR/hr)

=GT-RE -59/60 100 0/oclaiR/hr) x 1.72% Clad Damage Fuel Clad Loss Threshold GT-RE-59/60 Reading = 5. 73E+02 R/hr 4.2.3. RCS Loss (Spiked Coolant GT-RE-59/60 Reading) . MWtwc GT -S9/60spiked (R/hr) =Std Plantspiked (R/hr) x MW tstd Plant RCS Loss Threshold GT-RE-59/60 Reading = 5.94E+01 R/hr 5. CONCLUSIONS 5.1. 300 µCi/g DEl-131 is equivalent to 1.72% fuel clad (gap) damage. 5.2. Calculated containment high range radiation monitor values are as follows: Fuel Clad RCS Containment Loss Loss Potential Loss I GT-RE-59/60

5. 73E+2 R/hr 5.94E+1 R/hr 6.68E+3 R/hr Based on monitor accuracy/readability and human factors, the EAL Fission Product Barrier thresholds are established as follows: Fuel Clad RCS Containment Loss Loss Potential Loss I GT-RE-59/60 600 R/hr 60 R/hr 6,000 R/hr EP-CALC-WCNOC-1602 Page 9 of 13 Revision O WCNOC EAL Technical Bases Calculations

-CHRRM FPB Series 6. REFERENCES 6.1. NEI 99-01 R6, Development of Emergency Action Levels for Non-Passive Reactors, September 2012 6.2. EPA-520/1-88-020 (EPA Federal Guidance Report No. 11 ), Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion, 1988 6.3. WCAP-14696-A, Westinghouse Owners Group Core Damage Assessment Guide, Revision 1 6.4. NUREG-1940, RASCAL 4: Description of Models and Methods, December 2012 6.5. NUREG-1228, Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents, October 1988 6.6. EPP 06-017, Core Damage Assessment Methodology, Revision 4 6.7. AN 98-029, Basis for Wolf Creek Core Damage Assessment Guidance (CDAG), Revision 1 6.8. SAP 99-145, Letter 99-01466, Core Inventory Radiation Sources, Table 1 -Core Fission Product Inventory 6.9. Wolf Creek Final Safety Analysis Report (USAR) 6.10. Technical Specifications, Wolf Creek Generating Station, Unit No. 1 EP-CALC-WCNOC-1602 Page 10of13 Revision O Attachment 1 1-131 9.46E+07 1-132 1.37E+08 300 µCi/g DEl-131 Equivalent Fuel Clad Damage 4.13E+05 1.24E+04 2.92E-07 5.99E+05 1.80E+04 1.74E-09 .... M .... I iii 0 1.00E+OO 5.96E-03 1.24E+04 1.07E+02 1.95E+08 8.52E+05 2.56E+04 4.86E-08 1.66E-01 4.25E+03 2.15E+08 9.39E+05 2.82E+04 2.88E-10 9.86E-04 2.78E+01 1.83E+08 8.00E+05 2.40E+04 8.46E-09 2.90E-02 6.95E+02 .2 E+08 3.60E+06 1.08E+05 1.75E+04 Volume Conversion (g/lbm): 453.592 RCS Mass@ NOT (lbm): 5.05E+05 RCS Liquid Mass@ NOT (g): 2.29E+08 Halogen Core RF (%): 3.0% Target DEl-131 (µCi/g): 3.00E+02 % Clad Damage: EP-CALC-WCNOC-1602 Page 11 of 13 Revision 0 Attachment 2 EP-CALC-WCNOC-1602 GT-RE-59/60 Fission Product Barrier Threshold Values "Cl ftl 0 ';fl. ... a Q .5 I!! -'i u u -qb 0 *s:: 0 :0 M U ... c( a en GI 0 QO:: ftl ... EM i!l ! .! ftl i2 ... LI. ._. .-------b ,,,, w ';fl. c GT -RE-59/60 EPP 06-017 Figure 3 100% Clad Failure (R/hr): 3.34E+04 NUREG-1940 Spiked Coolant (R/hr): 5.00E+0 1 Standard Plant WC Rated Page 1 2 of 13 Rev i sion 0 Attachment 3 NUREG-1940 Figure 1-1 PWR Containment Monitor Response 1 e+6 1 e+5 i:: = I= = ----f--------.c 1 e+4 --0::: t; ;::; t;;; t---t----t--"O -(].) "O 1 e+3 .c I-! (/) c: :::> O> 1 e+2 c "O <O (].) 0::: '-1 e+1 0 -*c: 0 -Key to Spray Status and ->-----'-' -Damage Amount >------c 1 e+O (].) E c = Sprays Sprays E=== == Off On >------>------<O -c 0 1 e-1 () -r 00% >-----§ -= -1 0% 10% >------f--->---------f--->-----1% 1% = 1 e-2 1h 24 1 1h 2 4 h 1 h 24 h 1h 24 h 1 e-3 Normal Cool a n t Sp i ked Coolant Cladding F a i lur e Co re M e l t Damage State and T im e A fte r Rea ctor S hut do w n F i gure 1-1 PWR containment mo n itor r espo n se 1-14 EP-CALC-WCNOC-1 602 Page 13of13 Re v is i on 0 Enclosure VI Wolf Creek Generating Station Proposed EAL Classification Matrix Wall Charts (3 Pages)

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.......... (SORE"'1) rachllon moMGI' (SO AE47 or 31) II.A U I l I I I l I 4 I I I I !Dtt l

°"ECU40Ycw QOe:l(l....i2)

IW.1 I l I i I J I 4 I I I I I Otpl Oowlllt1>

  • ConllolAoom(SO-JtE-331
  • c.nnl Alwm SOii.oft (Dy IUNe)') NolJllA.At.HE01....,,tr..alflrad*tlOrll-1$!Not prolllb( or IMPEDE ICUll ..... .,. hbla R-3 """"'or *re*(Nol*5) ltOSTLt ACTION_ ... CMN U CON'lllOUfO IJtU.* ----*-* llUU I 1 I J I 1 I 4 I I I I I Of,!

>coklrm"l..E"toratOmin(Natw123) 11.UU It I 1 I J Ii I I I I !DC,.

..., (NIIMI l) 11.U2.1!11 1 11 It I I !1 lDVI

  • ... 6c:aledDybw-*lt....i*

..... OflfldiQ!IOn

{ECLI-OQ3M.ECll-OO:l9B ECUT-31

....... UNPLANNED 1111"' oo,._m.g , ,.. ,........., ..... M -..itry 1..,.T-R*2..0.bOftfl'ICWlerl Nor111a.c.w;.iP.,Roorn8 ESFS'NGRRoomNo 1(TANA) ESF SOOA Room No 2 (TAN 8) Ai.lllUlylluoklongM'es!HelBa\12000 HSU I 1 I I I j I 4 I I I I I QE'I HAU I 1 I 2 I j I 4 I 5 I I I Di,. Hiiu I l I J I J I 4 I I I I I Ditl TM Emargancy M*l'llgN IMutl ClldW9 N A HOSTllE ACTION .. OCC\lfmgor occunlCI WC!wl Sll fl lllU!lnllll

..... """"'--l'IH_n_.or ..... w.ly"blft-

i.,.,,;:==::=" ..... 7!."'...:'.:

..

tf\f>a.tft

.. ntflow1191tMef'l

.. ..t.-.. m0fll10f*kl'IOwll1C1n1

... sioppac1 ,..,dlclbngtNo1 tl'll**-111(11

..

<Hdng " -loftglr VALID !or cllMillell.oft

.........

Thapr...,.lclllalld.mu.ntmonitorvat.r..

prN1ntld,.,EAUAAl 1 , RS11-A011

.... _111 ....

..

.... 1110CCUl'r9Cllt\ln 11 0'"'*1111"Ct .. _....*a II cx:>l<lTAINMENT ClOSlRE * *-i.bli.n.d

<11c1 .. 11ono110......Emargia*"V*

-TholEAl.dOH

-9"P1rl011;1W111tnoflll:

.... --

11e .. p1111y..-,.,10lllecora

-4oaa11ot roddnvlpoio111or......,.._DIDOr011 On1Cont--5'nySp1wnua.,1nc1--flll111ond

.. p HMlllZlbon--I No*t1:8ag.womon10mg"'°1condl.OllEAL.1concllflWnlly to51oldK&VnMt_...

°"*-*--..... ,,__ ....... ,_ e..----*-ir...-, H07.l I t I 2 I I I 4 I I I 1 I Di)Pi 01!\e<'COllOCoo,,.u11tlll'lliell"'ir..Jll<lil'"tntol1!1*

E....,IJ'ftC'l'M1"'1119r-111llll-ll'llflptOg1M&or

...

m.or-...... 1 ..

..........

ld\llllDHallllor-icor.1ro1olt111ii.c.lty S-e.anc.iPMRoomA NorltlEIKtnc&IPMRoom8 ESFSWGRAooft'INol(TRNA)

ESF SWOR Room ND 2 (TAN B) AW<.UtyBuldinl1°Wes!t-91Elw2000 3 ,* 5 Hsu I 1 I 1 I J I 1 I I I I I j Iha c-.oi Room., 1\1 AUltMry $11-Pllllll (o'SP) COlltrolof M)l dtlllfolloriMngklys.afll'l'fu""110!tl*

ftOt .. ,5 .... (Noltl) *AlllfMJc-..l(-.12-3

..-i Olhl<COf!dtlCll'lll*lll......m

... in*,IUd{ltNnlolln*

em-g..cy M*l\IQlt<indlc**ll'lat-1.,.nprog<Mt

.. M._.ooc11""'wlldl""'°""'tcftllliori..tir1111JOflt*.rM1of plM t lulldlonl

....... lofpiut-olllltpullkflf

...UC--

oowldlltdlCllll....,,.,.,.,.of.,,

Mt r.11 1 115.,. -tlPICll<I n U PQlllN ..... w111cn .. ca1<1EPAPrutaa

... Ac11on0..-11Ma*P011K*

A'"'°9TILEACTION*0<<11m..io*"'*OCC11"9d-1ll10Nt6 CONTAOltEO AREA 11 *IPll'llCI by !No

  • ...:.":':'::"'

_:cd.., 1rcr1t11tt1c1t TlbteK-1 , ... 11n.. I lhltpdlibtt0<IMPEOES...,...ICI

.., Y l""-H-2,_.,,.

    • -*(Nl>ll5) .... -...........

,. ... _ .. HM.1 I 1 I z I J ! 1 I s 1 1 IOit l lvo-nll\H

..

!ASP) o.harc:<ll'dllOftl H lll...., IClo"'1!\IJl'dgm9n!oftll flnllvooncyM1f11991'lfldicalltn.t__,tl l Nll'IPrOgrHIOf h-OOC11"9<1"""°'.,-..llllC'!Ulolorpcll-l

.-n1.l119gNdilt>onol

... 11r....io1

.. 1e1yof1'11pl11110<

,.......,._tllll

..

... MINf\gnll<

., ... p1._1111.,, .. ,..11"1o_aqu_ti.caustol HOllTilEACTION

!Vty ..... 115.,..eMptiO:ICllobalifl'llllCI 1C1smal"-o;t1DMd!lltEP"P11118ctf**Ac1lllnGwldilltne

  • ICP05W'f

......

fACprowllng llfon.-.t>on HU2.t I t I I I JI* 11 11 !Dipl HUl.1 I l I Z I I Ii 11 I I lDl'I HIJU I 1 I 2 I JI 1 I I I t l p!!!PI lflllf'Mlroomflf1NIA.OOOINGot1.........,.

sullli;Montio._

........ .-... u1 .... 11e:.i.ctnc.1 -t>onol*SN'ETYSYSlEMcomoOflllll"""*'lor ll\9Cllff9nl0111191l"ll-HUl.l t l I # I J I 4 I I I I I Ott I

  • llM'EOEO-toandfsl*.-n\IOIW>g

..

...... 11) HUU I t I 2 I 1 I 1 I I I t I DIP! Anaz.,..,us-1111.-lfloft-141-.-*

1Ulic ... llClll'Olltbllf>&PiMl5lllll'rOft'l-gll'la s.-.... _l_C..("**7)

HlM.1 I l I# I l I 4 11 I I !!l(p! AFlRE11-Ubng....Ma"" t11" 15min o1 ... y of1M loloMnQFlRE<ll111t11onindica110n1{Notel) o

  • HtU.2 I 1 I 2 I l I 1 I S I 1 I I R-ipd*..,gl9ln1ll1TT1(1t 110 otll11tr1<klllon1al Thalr.1IMM*trlClc:lllftlltFIRE-...

1 ft)'T .... H-I Thlf--..otol*FlltE*

-.... r#il<l-30

..... 111 11arm--111(No1t1)

HtU.l I 1i2!lI1i1i1

!p1P I AFIRE""""" NI lilr4 PROTECTEOAAEA 11ot

...

HIJU I l ! I I I I 4 I I I I I @ti A FIRE _,,, ... 1111111 PROTECTED AREA 11.t reqo.W"

_,,,M M"'occu""'""'dl-*POttont*l<ll>gNdllonollle

..... olNM\yol1!141pilol1lOf"'-

' MCUflYll'lfMllO o::;-ru""',,,._

d SNETY Modes: O=:J [TI CD CT] O::J CO I DEF I W$1.F CREEK APf [XX-XXX-U), Addendum 1 R*Y.(llJt)

EAL lrbtrht Pagel ol.S ALL CONDIT I ONS P o-* 0 pet"al ia n Si.rtup Ho t S tand;iy Hot S hutdoYm Co ld Shutdowrl R ttu.U ng Da tu.lt d GENERAL EMERGENCY I SITE AREA EMERGENCY I ALERT I UNUSUAL EVENT 1 Loe.1of 1 ... ,..rw;y .. -*N:.-wlO

... sou I 1 I Z I S I i I

..,....genc:y

  • 11!1(V Du-NBOI ...., N802 AHOEllllEJI:
  • RK10rstiond*leastooe11rnergencybt&

ll'l<-------------i l-<l o llAC..i_oc_

.. _._far15.....,*or

-so u I 1 I Z IS I 4 I lo9lol all al'IWri:l all onM1AC"°"'9cap9bllrlylD

  • fl'IOtfOe<lcy 4 16K'V buMs N B01 ana NBC2 lor l l5 m111 -""'

t_Of .. " "

.. .... 1, ........... IOf l 5,,..._0f.,,....

Ht.t I 1 I :Z I S I i I IA1.1 I 1 I # I S I i I sU1.t Ii I 2 ! l Ii I " " " "

etn*gencv*

1et<Vb!lloKNB01 anclN9021orl15min (Nole1}

Tabll!S.11Dem<<geney41&CV Tabll!S.1 IO bo.*5NB01 em<<gency 4 1&<Vb!-.NB0t andN902fllfl t5m11 lofl15fllln (NOCe l) (Nole\) ...

1c191ol .. AC-IDSAFETYSYSTEWS TallM&-1 ACP'-IYpptiet.

Ol'f5b:

  • ESF XFMR XN002 a.. .. :
  • EDGNEOI
  • EOGNE02 t-* .. -oc_. .. ,,_ * .,......, 2 ss u I 1 I 2 I S I 4 I 3 Lo. of 111 125 VOC baMd on b.nsy bl/1 "'lllage 1051/0Con all ,,.alOCbule1NKD1 NKOO
  • R..c:tor-* RCS ... *RCS....., *
  • L....el1n1111N11oneSIG

\MPt.N.INEO

.... GfCtd,.,......., __ .. """"-U.3.1 I t I 2 I S I 1 I Of rftOfe TIOl9 S-21*-*M fl'om -IM Con!rol Room Arrt 9'Qfllfio;anltr......tl511'11JfOQf9MT!lltlleS-3 SUl.t I 1 I 2 I 3 I 4 I An Ul*'lANNEO 9Y911,....,11'1 lhe.....,.,.

ID monitor oneormor1T.C.S-2pw-MSll'llm....rl'Wl!M Contro1Roomfor115l!WI

{No!9l)

  • at.._on.SIG

=.ci:::oaMrt'***".,r-..--

s 4 "' *-5 . ._ ..

  • Rl.Wllac:k.t25%1hefmlll_.
  • Elednc.911D..irwtedon>

25'11. _ ..... [l ....... ,._ ..... _ ....... o.p_ .... c......., ... "°'-.. -................ ) [l..-..,.--

.... -0., ......... c * ...., .......... 11o1 ____ ,-.1 SlU.1 i 1 I 2 I S! 4 I Sample anal)'5ia indicates RCS KWl)I TecMIClll SU5.t I 1 I 2 IS 1 1 I RCSUllldenttlldorpi""""'boundar)'leakage

> 10gpm lorit15llWI Oft

.. lukagelrom1hl!RCSto1locahonou!l.ldeo::int..nment

> 25gpmlorlt 15mon F 6 7 Lon ot c-8 9 Category A RCSotSG Tube ,_ B c CMTRa:iia1IOlll RCS-D CMT lnlllllfl!yor

  • -E .......

..... lloMn ...

Am"1Ua11t:11011l5anyopenitor10C11011 orlelof rmpodlylM<<tedl'llOtlwetlle and don not

'"l"dlOl1.,.__

Conl-"11 CooW!g S)'llem tr.Ii ClOm?'IM F01.1 I , I 2 I l I 4 I

  • LoMorpg1en!Nllio.&aftnm111amet(TallleF
  • 1l F uel C la d (FC) Barrier ....... -. ... , .... 0' ....... <NI ...........

.-._.,.M:S

__ ,,.... AniaiA.omllll(:ormanualtnplaa1Dtnu1-.U-..rNCtor All KtJoM 10 111111 down me reac101 we not 1UCOKlful

..

  • CSFSl Mlt.iSrik*

R EO Patllcondi!IOl'l&mel

__

...._i_....._.,.1....,.t1M.__, ... **--.u ..........

-.. ... -AnlUIDmll.IC0tm.......,t1¥1r..Jt.to.,,UldoMl!Mre1t:torH ll1Cllc.tedbyreltllOJpowel'15%

l:Mreltllor

  • IMCated byr...:1Dr PCJWef 15" <N'* 8) ...... PA1)'1lem " "
  • Hlgh-moriom.domlke
  • FIRE
  • OlhefeverRWllh11111Mfhazalll I 1 I 2 I i I 4 I Local T elephon* Company OQel Linn ENSC.. Mazarcloul

....ent .rtKtlng 1 SAFET'I' SYSTEM SAl.1 Ii I 2 IS Ii I Theoocwnrnoeof ut)'

Al<<ltfTHVt:

  • e--.i aammge n.u C1UMC1 indocallOM d Mgraliecl pe!foonanc.

in at lea.C one !fMI of

  • SAFET'I' SYSTEMneede-dfOJl:Mcun"entope!1111!1fi1mode
  • The_.. l'IM c:.uMd l/ISJBl..E to 1 SAFETY iltJVC:tl.l"fn.teded FA1.1 I 1 I 2 I 3 ! 4 I Alft lol&ot aft)' PO!enllllbud ETHEi! Futl Clad ot RCS(TalllllF

-1) Table F*1 Fission Product Barrier Matrix Reactor Coolant System (RCS) Barrier A.nautomMICtn(l6od nat trlllldOwnU.1'9:..::IOfM indicated by 1'9:Ktor pcl\"915%

9ftel" 1ny RTSse1parn rn-.. ...

1t11WIUCtOfcomato::in.m.(S8-HS-1or,s&.HS.;12J 11aucc9Ml\111n.,,ul1ngdown1heruc10f*llldlc.!ed by...-e;t0fpowef

<5% (Nol*ll) AmlllU9lfr1Pdld llOt Sl!UllCIOW'nhfuaDI'*

.:r.:pe.,...15%"n.

Ul)'....-fl'IP

... A1ut>teq .....

rNCtJr c 5% (Nole II) sur.1 I 1 I 2 I S Ii I Oft Loud *II T*blt S... olhd* communa!JOn mtlhlldl .. Loud all T..,._ S... NRC com""""IClbOn IMtlW:ICI&

suu 1 1 I z ! S I 4 I AftY pene!f911Cn*llOll&Olm9CIWll'lln15m111 afeVAllO Oft Containmen!

preuure > 27 Plll1Wll'I

< -111111111111 DI de"l!nforl15mon (Note9) (NO!e1) Containment (CMn Barrier Loss I Potential Loss Loss Potential Loss Loss i Potent i al Loss 1 CSFSTCor*CoollniJ-M O 1 CSFSTCoreOlc*lg.QftAt<<JE Pll.li Palhecndiliorllll'IM

..........

'""""""""-'""'""!

onGT-RE.eiotGT-RE-<<1 2

_,,,,.."""" ! 1 I An y oonc1ruonr11heopl!ll0!1atlhe Emefll'MCV Mll\lljlOtf 11111 indlc:Mn Emllfgtney MMaQW tNt buDfthlFi..!Oadllen'*

poWi11911oudttwFutlo.:ttllrnllf 1 MautOl'l'IMeotmaoualEO:S (SIJ1t:t,,..\Jot'lr1JQU11ed1>y ElTHE": *

  • SGTR I 1 CSFST lnl9grty-ltEO Plth QQtldCa. I .. I ! 1 CSFST HNI s.ik-l'lED Path I I :;g---

I Conlalnm..tr-""""" eGR/tvot1 j

I I M&r111Q1<

lhll nc1o;1tn Em*gency Mlnag<< !h.11 ll'ldllcM" klMDf!MRCSt...

I po!enballol&dtN!RCSbl!Tw I A RUPTURED SG 11 FAl..I. TEO outlldecla>nllnnrnenl AHOEI TH EJt;

  • COl'llMl..,.,.l'llfgrtyhMbMfl
  • ----.

I I I 1 CSfST Containmen!-M:O Pali 1--

l ContlUM'llllPf"'"

"27 s-u I

15m1n (Nol* 1 Ill 1 it.ny 0otlditlor1J1thl0pnl0f1atthl 11 Any ccndrllonl'lttlt'OPr\lotlatthl MIM(l'f 11\11. ondlcatH Emelljle!IC)'

MlnlQ9f lt1.c rdiClln IOMDfttwCont111nn1"11t.rrw llOlilnlrllti.clthleataonme"ll

.... Modes: CI:] CD [TI CI] CU CD I DEF I W$1.F CREEi APF [XX*XX X-XX), Actd.ndum

1. Rev.(u] EAL Claslifleatlon Matm Paoe2of3 HOT CONDITIONS RCS> 200'F) P OW9< Opera t ion S tan u p H ot H o t Shutd own Cold Shutdown R efueli ng De l\Jeled I c .,... ... --*-GENERAL EMERGENCY I SITE AREA EMERGENCY ALERT I UNUSUAL EVENT 1 (Noc.1) AHO TlltlleC-2 cou Ci ::::i=J::::I=::JIC:

ID l:::i i[II ::::J RCSle...i carinot t1eMOn1CQC1klflJ0"11" (Nolel) ANO

  • UNPl.ANNEOn:rnMlll
  • "Y l ....

1.....i alwfliC9111'19Qlltudela!l'dlclleOOle

.-y cs1.1 C:::::C::::I=C:

IJCI=C::J WtnCClNTAINMENl ClOISt.RE MC...,..,_ RI/US 1*inlorcwil0ollnnuec72'fl RVl.15 nMLDI IW09 c en. c1u I I i It I I RCS_...., cannot lMl!IOnllor9dlor2lO"" (Nole1) .. , CoreW'CO'WY*lllCIQ!tdtl)'

MV dll'llflolowil'lg CA1.1 Ci =ic::::c::I=10::, :r1:::;*a1

i I cu1.1 Ci =i=i::::c=i1:;1a1
l*::CI =i CA1.J I I S I f I RCS,,,_..,.

unno1.11emm1a1ec1b1\!5""" (Nl:M 'I AHDlrTHEll

  • l.NPlANl-EO io. d ...aa ClllCllMI

,_.. 111 RCS ""9-. ,_,

  • 1!5IVI !Nd*ll cuu Ct ::::i=J::::I::::Jl
JICI! ::l*::I i ::::J ACSwM.-llvel u_be_.,,.ed .t.NOmn.9':

-.pl 1'ink ..... Cluelo1io.cr1RCS-.tory

  • Vl5Ueloi-vnoncrlUNISOL.A9l.ERCSINQroe
  • M.-ciui.t0tbndgolioraner.a.MH1nmot'*>l'S().A£-41 "
  • 2 -* 3 .... ,_ 4 ...... ... oc T.-C-2 ... T.W.C.1 Su,...naM1
  • Containn!lnlNof!Ml,S.,.,.,p
  • Ai.Hy8UlliingSUmp
  • LIQIMIW...HoblpTMk
  • CONS..V.Tri
  • Ena!acScuc:.R..,gieMonllDl"ftloaltion
  • Unpllnned Me'" CorUsvMnl ""'""
  • ESf XFMR XN801
  • ESF XFMR XN802 '""
  • EOGNBl1 .... ..:.-.-.., ...........

lllCS

... o.i.. ...... .....--REDUCED ltNENTOR'I')

....... "" REDJCED INVENTORY

""'-'

_,, CA>.1 Ci ::::i=J::::I=::JIC:*Di

i*::I=::J
10) UNPLANNED RCS pi-.i1* nereaM > 10119'8 (ltw EAL C U 2.t Cl ::::i=J::::C=::Jl:J ICi l::i l::IJI t..nNB01-NB02otdllOt<llO*

.. nglitl)(JW9l'llOUlmlol' A#IW.,..,,.. .... ....... ,_."'b91 CUJ.1 Ci ::::i=::c:::I=::JIC:

IDl:::i*CII

J UNP\.ANl'EOlllOHMll'IRCStempe..tlnlo>200F (NGl101 cuu Cl =::J=::C:::I=::JIC:

ID l:::i*CII ::::J ..

RCS lrJel ll'ldleiltion for :t15 m1n(Nole1)

CLM.t Ci =i=::c:::I=::JIC:

IDl:J*CI::::J so.ob!):)n r.qulrecl 125VOCMMlor:t t5lflll'I (NO:e 1) cus.t IC::::C::I::JC:::Il:JIO:I

il::Ill D!ll!t!JI 5 PAtytt"" --t.o.d al TICll:C-!5on.<<i*oommunora11Dnm<<hodl

.. SUT..,,....5)11

....

o*

6 --.. .... .,._,

.....

I CONTAINMENT 0.0SURE * ,,._...,_, pnOf Ill eCMdinQ en. 31).fnnM '-.....

...... Note10:

lol.ny,_...,llfllortol'ldllionl'IClllfNl9dlOll\e lotl.dc1tetyn

.. 1remov.I

..... l -*

  • lntemtl0tl'l!ltmtiFLOODl<<>.--
  • Hvti.W.0.0tlor-lllQ
  • FORE
  • Olhel-W!lhllfftlllr'-ZMI CNl'IC1ant11e1*1delerminedbyll'I*

Erne1geric:yM""""'r .... ---CAU Ci ::::i=J::::I=::JIC:

IDI ::Jl::II ::::J The-d anv T .. C6hN.o&:lu5....,.

  • NO..,,... .

pm!formaneer1Mi1Uone1rM1crl*SA.FETI' SYSTEM !-*I for the curren!

o Tl'* ....... c.UMO VISIBLE DAMAGE ID* SAFETY SYSTEM eomponene or IUllCllH nwa.d fOfll'l*eunencoper.ung mode DEFINITIONS

.... MMe!Mnb..-.

..

.-.U.CAPS) 11-ConttlNNntao....ni -l:Ol'l\l)Ol'llnK*

l'unaloMltlelNllo6-P'OOldrtlffM...no..,.._eondtb::N

.....

...... l-TPWI R-=tor\l .... Fllnge()pellbllnl..-.m<<

..........

A grounding 91C11111 Sud'l_.,,reQYftlpotl.-,,--.ID de! ............

Mll'lbut*crl.,,uplclllon..-.Pf-.1 , ...... Tiie F1n 00---U!*firn Obtavnoncrlflltnt

  • prdtn<<lbut*

not rtQUHClfllrg.equ-dlf!IOke*ndht*-

-Floodln11

.....itingin*,..d

........

Hos1111Actlon tak9hl* .. 9D'Of Hllllllt WdlO-. NorH__..MCIE.\Llll'O.llODl:llllllllDaCIOf-MOlllctM!IHr1e tlWlm.yn:t.ICllt"'lllllnl**be!wMn ll'dMCllMlllll'lllW_eoroolllll_)

--ClneOflTICn

........

INl!lll'CI

...

rMlgMIOnOfllOl'rec:tto/elCtJON

.........

MQ!M'"'lli..dpn:ittdr¥eeQUICllftent.111t11*9CSAI ._,

Owner c.rtrollld Al'H (QC.A) b!ilotNtel!ler'ClorWdlOeekGener1itlf'lll$tM>llnlarwlw:ll pubkllCC*ll*lf!tllld

'roJtdll orPtlWMlll u"'y '""*dM,.,...

, ,,.,

..

"'"'"'""

....

modecrlgplfM>lln(t11

-ftteie ..... ornaultd8111&1 RHc;t1;11VftM1"9r1Qt(c&c 11n)wdltuel1nb.....i l?!nlluelllOcrl-IUlll..,_,*mn.ICllf'll)f

..

--

..... dllllllelltnlmlgllt\IOllD,.qust*

... f!!V -l ...

... IONICIPIQbilllrrllb Satetyapt.,,

indudolng IM eccs Th .. -svwi-* dMl'lt<t M .. tety.i.i.ted

{M aefined"'

10CfR50 21 !:::-: 1.,.iemsll'ldcomponent11ti.1.,*rehedupOl'llO....,N1f\lrv:t1on.illll'inglllllklllowlngllellfilllb9lll i11 Tl'*

rHC!Dleoolantpieuureboundlly (2)TnetaplbjlytolhutoownNl'NClor-mM11..,*1n*..-1..,_.downoondttlon .S.Clll'llyCOfldidoft

'*'°"Y f\feM/1111CID1feP91.anne!

..... datf!!yd!M

...

EaciuMDnM*Blllundlt)'**SYJIOfl'l'TICIUlemllorSH&oundilry

.. dfll'nedMl'le-lllll

__

Rmaot Tllenct..w:wl**boundwflllcl1111nDClll\Cldl!IWll'IU....n:e.dare*bOIASlr)'

CotCrc:ldto::a&IO-*

MopenottwHehtetl)'ltemllnetMt unn<<belMllMedrtm0t9fyorloollly u..-..nlled

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