ML23278A095

EIP-ZZ-00101 Addendum 2 - Emergency Action Level Technical Bases Document Administrative Correction Revision 021

ML23278A095
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Site:	Callaway
Issue date:	10/05/2023
From:	Ameren Missouri
To:	Office of Nuclear Reactor Regulation, Document Control Desk
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EIP-ZZ-00101 ADDENDUM 2 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT ADMINISTRATIVE CORRECTION Revision 021 Page 1 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT TABLE OF CONTENTS Section Page Number 1.0 PURPOSE ...................................................................................................................................... 6 2.0 DISCUSSION ................................................................................................................................ 6 2.1. Background.......................................................................................................................... 6 2.2. Fission Product Barriers ....................................................................................................... 7 2.3. Fission Product Barrier Classification Criteria...................................................................... 7 2.4. EAL Organization ................................................................................................................ 8 2.5. Technical Bases Information .............................................................................................. 10 2.6. Operating MODE Applicability (ref. 4.1.8) ........................................................................ 11 3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS .......................................... 12 3.1. General Considerations ...................................................................................................... 12 3.1.1. Classification Timeliness ........................................................................................ 12 3.1.2. Valid Indications .................................................................................................... 12 3.1.3. Imminent Conditions .............................................................................................. 13 3.1.4. Planned vs. Unplanned Events ................................................................................ 13 3.1.5. Classification Based on Analysis ............................................................................ 13 3.1.6. Emergency Coordinator Judgment .......................................................................... 13 3.2. Classification Methodology ............................................................................................... 14 3.2.1. Classification of Multiple Events and Conditions ................................................... 14 3.2.2. Consideration of MODE Changes During Classification......................................... 14 3.2.3. Classification of Imminent Conditions .................................................................... 15 3.2.4. Emergency Classification Level Upgrading and Downgrading ............................... 15 3.2.5. Classification of Short-Lived Events ...................................................................... 15 3.2.6. Classification of Transient Conditions .................................................................... 16 3.2.7. After-the-Fact Discovery of an Emergency Event or Condition .............................. 17 3.2.8. Retraction of an Emergency Declaration................................................................. 17

4.0 REFERENCES

............................................................................................................................ 18 4.1. Developmental ................................................................................................................... 18 4.2. Implementing ..................................................................................................................... 18 5.0 DEFINITIONS, ACRONYMS, & ABBREVIATIONS ............................................................. 19 5.1. Definitions (ref. 4.1.1 except as noted) ............................................................................... 19 5.2. Abbreviations/Acronyms ................................................................................................... 23 6.0 CALLAWAY-TO-NEI 99-01 REV. 6 EAL CROSS-REFERENCE.......................................... 25 7.0 ATTACHMENTS ........................................................................................................................ 26 Page 2 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT TABLE OF CONTENTS Section Page Number 8.0

SUMMARY

OF CHANGES ....................................................................................................... 26 ATTACHMENT 1................................................................................................................................... 31 Category R - Abnormal Rad Release / Rad Effluent ...................................................................... 31 RU1.1 Unusual Event ....................................................................................................... 32 RU1.2 Unusual Event ....................................................................................................... 34 RA1.1 Alert ...................................................................................................................... 36 RA1.2 Alert ...................................................................................................................... 38 RA1.3 Alert ...................................................................................................................... 39 RA1.4 Alert ...................................................................................................................... 41 RS1.1 Site Area Emergency ............................................................................................. 43 RS1.2 Site Area Emergency ............................................................................................. 45 RS1.3 Site Area Emergency ............................................................................................. 46 RG1.1 General Emergency ............................................................................................... 48 RG1.2 General Emergency ............................................................................................... 50 RG1.3 General Emergency ............................................................................................... 51 RU2.1 Unusual Event ....................................................................................................... 53 RA2.1 Alert ...................................................................................................................... 55 RA2.2 Alert ...................................................................................................................... 56 RA2.3 Alert ...................................................................................................................... 57 RS2.1 Site Area Emergency ............................................................................................. 58 RG2.1 General Emergency ............................................................................................... 59 RA3.1 Alert ...................................................................................................................... 60 RA3.2 Alert ...................................................................................................................... 61 Category E - Independent Spent Fuel Storage Installation (ISFSI)................................................. 63 EU1.1 Unusual Event ....................................................................................................... 64 Category C - Cold Shutdown / Refueling System Malfunction ...................................................... 66 CU1.1 Unusual Event ....................................................................................................... 67 CU1.2 Unusual Event ....................................................................................................... 69 CA1.1 Alert ...................................................................................................................... 71 CA1.2 Alert ...................................................................................................................... 73 CS1.1 Site Area Emergency ............................................................................................. 75 CS1.2 Site Area Emergency ............................................................................................. 77 CS1.3 Site Area Emergency ............................................................................................. 79 CG1.1 General Emergency ............................................................................................... 81 CG1.2 General Emergency ............................................................................................... 85 CU2.1 Unusual Event ....................................................................................................... 89 CA2.1 Alert ...................................................................................................................... 92 CU3.1 Unusual Event ....................................................................................................... 94 Page 3 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT TABLE OF CONTENTS Section Page Number CU3.2 Unusual Event ....................................................................................................... 96 CA3.1 Alert ...................................................................................................................... 98 CU4.1 Unusual Event ..................................................................................................... 100 CU5.1 Unusual Event ..................................................................................................... 102 CA6.1 Alert .................................................................................................................... 105 Category H - Hazards and Other Conditions Affecting Plant Safety ............................................ 108 HU1.1 Unusual Event ..................................................................................................... 109 HA1.1 Alert .................................................................................................................... 111 HS1.1 Site Area Emergency ........................................................................................... 113 HU2.1 Unusual Event ..................................................................................................... 115 HU3.1 Unusual Event ..................................................................................................... 117 HU3.2 Unusual Event ..................................................................................................... 118 HU3.3 Unusual Event ..................................................................................................... 119 HU3.4 Unusual Event ..................................................................................................... 120 HU4.1 Unusual Event ..................................................................................................... 121 HU4.2 Unusual Event ..................................................................................................... 124 HU4.3 Unusual Event ..................................................................................................... 127 HU4.4 Unusual Event ..................................................................................................... 128 HA5.1 Alert .................................................................................................................... 129 HA6.1 Alert .................................................................................................................... 131 HS6.1 Site Area Emergency ........................................................................................... 132 HU7.1 Unusual Event ..................................................................................................... 134 HA7.1 Alert .................................................................................................................... 135 HS7.1 Site Area Emergency ........................................................................................... 136 HG7.1 General Emergency ............................................................................................. 137 Category S - System Malfunction................................................................................................ 138 SU1.1 Unusual Event .................................................................................................... 140 SA1.1 Alert .................................................................................................................... 142 SS1.1 Site Area Emergency ........................................................................................... 145 SG1.1 General Emergency ............................................................................................. 147 SG1.2 General Emergency ............................................................................................. 149 SS2.1 Site Area Emergency ........................................................................................... 151 SU3.1 Unusual Event ..................................................................................................... 153 SA3.1 Alert .................................................................................................................... 155 SU4.1 Unusual Event ..................................................................................................... 157 SU5.1 Unusual Event ..................................................................................................... 158 SU6.1 Unusual Event ..................................................................................................... 160 SU6.2 Unusual Event ..................................................................................................... 163 SA6.1 Alert .................................................................................................................... 166 Page 4 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT TABLE OF CONTENTS Section Page Number SS6.1 Site Area Emergency ........................................................................................... 168 SU7.1 Unusual Event ..................................................................................................... 170 SU8.1 Unusual Event ..................................................................................................... 173 SA9.1 Alert .................................................................................................................... 175 Category F - Fission Product Barrier Degradation ....................................................................... 178 FA1.1 Alert .................................................................................................................... 180 FS1.1 Site Area Emergency ........................................................................................... 181 FG1.1 General Emergency ............................................................................................. 182 ATTACHMENT 2, Fission Product Barrier Loss/Potential Loss Matrix And Bases.183 Page 5 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 1.0 PURPOSE This document provides an explanation and rationale for each Emergency Action Level (EAL).

Decision-makers responsible for implementation of EIP-ZZ-00101, Classification of Emergencies, should use this document as a technical reference in support of EAL interpretation. This information may assist the Emergency Coordinator in making classifications, particularly those involving judgment or multiple events. The basis information may also be useful in training and for explaining event classifications to offsite officials.

The expectation is that emergency classifications are to be made as soon as conditions are present and recognizable for the classification, but within 15 minutes or less in all cases of conditions present.

Use of this document for assistance is not intended to delay the emergency classification.

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

Additionally, changes to plant AOPs and EOPs 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 Callaway Plant Radiological Emergency Response Plan (RERP).

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EIP-ZZ-00101 ADDENDUM 2 Rev. 021 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 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 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 7 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 2.4. EAL Organization The Callaway EAL scheme includes the following features:

• Division of the EAL set into three broad groups:

• EALs applicable under all conditions - This group would be reviewed by the EAL-user any time emergency classification is considered.

• 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.

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

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EIP-ZZ-00101 ADDENDUM 2 Rev. 021 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT EAL Groups, Categories and Subcategories EAL Group / Category EAL Subcategory All Conditions:

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

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

C - Cold Shutdown / Refueling System 1 - RCS Level Malfunction 2 - Loss of Emergency AC Power 3 - RCS Temperature 4 - Loss of Vital DC Power 5 - Loss of Communications 6 - Hazardous Event Affecting Safety Systems The primary tool for determining the emergency classification level is the EAL Classification Matrix. The user of the EAL Classification Matrix may (but is not required to) consult the EAL Technical Bases Document in order to obtain additional information concerning the EALs under classification consideration. The user should consult Section 3.0 and Attachments 1 & 2 of this document for such information.

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EIP-ZZ-00101 ADDENDUM 2 Rev. 021 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 2.5. Technical Bases Information EAL technical bases are provided in Attachment 1 for each EAL according to EAL group (All, Hot, Cold), EAL category (R, C, H, S, E and F) and EAL subcategory. A summary explanation of each category and subcategory is given at the beginning of the technical bases discussions of the EALs included in the category. For each EAL, the following information is provided:

Category Letter & Title 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, E or F)

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

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

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

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

Classification (enclosed in rectangle):

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

EAL (enclosed in rectangle)

Exact wording of the EAL as it appears in the EAL Classification Matrix.

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.

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EIP-ZZ-00101 ADDENDUM 2 Rev. 021 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Step 2.5 Cont'd Basis:

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

Callaway Basis Reference(s):

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

2.6. Operating MODE Applicability (ref. 4.1.8) 1 Power Operation Keff 0.99 and reactor thermal power > 5%.

2 Startup Keff 0.99 and reactor thermal power 5%.

3 Hot Standby Keff < 0.99 and average coolant temperature 350°F.

4 Hot Shutdown Keff < 0.99 and average coolant temperature 350°F > Tavg > 200°F and at least 53 of 54 reactor vessel head closure bolts fully tensioned.

5 Cold Shutdown Keff < 0.99 and average coolant temperature 200°F.

6 Refueling Two or more reactor vessel head closure bolts are less than fully tensioned.

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

The MODE in effect at the time that an event or condition occurred, and prior to any plant or operator response, is the MODE that determines whether or not an IC is applicable. If an event or condition occurs, and results in a MODE change before the emergency is declared, the emergency classification level is still based on the MODE that existed at the time that the event or condition was initiated (and not when it was declared). Once a different MODE is reached, any new event or condition, not related to the original event or condition, requiring emergency classification should be evaluated against the ICs and EALs applicable to the operating MODE at the time of the new event or condition. For events that occur in Cold Shutdown or Refueling, escalation is via EALs that are applicable in the Cold Shutdown or Refueling MODES, even if Hot Shutdown (or a higher MODE) is entered during the subsequent plant response. In particular, the fission product barrier EALs are applicable only to events that initiate in the Hot Shutdown MODE or higher.

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EIP-ZZ-00101 ADDENDUM 2 Rev. 021 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS 3.1. General Considerations When making an emergency classification, the Emergency Coordinator must consider all information having a bearing on the proper assessment of an Initiating Condition (IC). This includes the Emergency Action Level (EAL) plus the associated Operating MODE Applicability, Notes, and the informing basis information. In the Recognition Category F matrices, EALs are based on loss or potential loss of Fission Product Barrier Thresholds.

3.1.1. Classification Timeliness NRC regulations require the licensee to establish and maintain the capability to assess, classify, and declare an emergency condition within 15 minutes after the availability of indications to plant operators that an emergency action level has been exceeded and to promptly declare the emergency condition as soon as possible following identification of the appropriate emergency classification level. The NRC staff has provided guidance on implementing this requirement in NSIR/DPR-ISG-01, "Interim Staff Guidance, Emergency Planning for Nuclear Power Plants" (ref. 4.1.11).

Time based EALs should be evaluated upon first indication of the conditions. If someone is working to mitigate the condition in less than the time required, the declaration can wait to see if they are successful within the time constraints. If there is indication that the threshold will be exceeded for the time period, the declaration should immediately be declared, regardless of the time remaining. In the case of leaks, the exceeded threshold will take some additional period of time to lower and must be taken into account.

When assessing an EAL that specifies a time duration for the off-normal condition, the clock for the EAL time duration runs concurrently with the emergency classification process clock.

3.1.2. Valid Indications All emergency classification assessments shall be based upon valid indications, reports or conditions. A valid indication, report, or condition is one that has been verified through appropriate means such that there is no doubt regarding the indicators operability, the conditions existence, or the reports 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 indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.

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EIP-ZZ-00101 ADDENDUM 2 Rev. 021 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 3.1.3. Imminent Conditions For ICs and EALs that have a stipulated time duration (e.g., 15 minutes, 30 minutes, etc.), the Emergency Coordinator should not wait until the applicable time has elapsed but should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time. If an ongoing radiological release is detected and the release start time is unknown, it should be assumed that the release duration specified in the IC/EAL has been exceeded, absent data to the contrary.

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

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

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

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EIP-ZZ-00101 ADDENDUM 2 Rev. 021 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 3.2. Classification Methodology To make an emergency classification, the user will compare an event or condition (i.e., the relevant plant indications and reports) to an EAL(s) and determine if the EAL has been met or exceeded. The evaluation of an EAL must be consistent with the related Operating MODE Applicability and Notes. If an EAL has been met or exceeded, the associated IC is likewise met, the emergency classification process clock starts, and the ECL must be declared in accordance with plant procedures no later than 15 minutes after the process clock started.

When assessing an EAL that specifies a time duration for the off-normal condition, the clock for the EAL time duration runs concurrently with the emergency classification process clock. For a full discussion of this timing requirement, refer to NSIR/DPR-ISG-01 (ref. 4.1.11).

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

• If an Alert EAL and a Site Area Emergency EAL are met, a Site Area Emergency should be declared, and the Alert noted in facility logs.

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

• If two Alert EALs are met, one of the Alerts should be declared and the other Alert should be noted in the facility logs.

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 emergency classification level is still based on the MODE that existed at the time that the event or condition was initiated (and not when it was declared). Once a different MODE is reached, any new event or condition, not related to the original event or condition, requiring emergency classification should be evaluated against the ICs and EALs applicable to the operating MODE at the time of the new event or condition.

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

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EIP-ZZ-00101 ADDENDUM 2 Rev. 021 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 3.2.3. Classification of Imminent Conditions Although EALs provide specific thresholds, the Emergency Coordinator must remain alert to events or conditions that could lead to meeting or exceeding an EAL within a relatively short period of time (i.e., a change in the ECL is IMMINENT).

If, in the judgment of the Emergency Coordinator, meeting an EAL is IMMINENT, the emergency classification should be made as if the EAL has been met. While applicable to all emergency classification levels, this approach is particularly important at the higher emergency classification levels since it provides additional time for implementation of protective measures.

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

As noted above, guidance concerning classification of rapidly escalating events or conditions is provided in RIS 2007-02 (ref. 4.1.2).

3.2.5. Classification of Short-Lived Events Event-based ICs and EALs define a variety of specific occurrences that have potential or actual safety significance. By their nature, some of these events may be short-lived and, thus, over before the emergency classification assessment can be completed. If an event occurs that meets or exceeds an EAL, the associated ECL must be declared regardless of its continued presence at the time of declaration. Examples of such events include an earthquake or a failure of the reactor protection system to automatically trip the reactor followed by a successful manual trip.

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EIP-ZZ-00101 ADDENDUM 2 Rev. 021 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 3.2.6. Classification of Transient Conditions Many of the ICs and/or EALs employ time-based criteria. These criteria will require that the IC/EAL conditions be present for a defined period of time before an emergency declaration is warranted. In cases where no time-based criterion is specified, it is recognized that some transient conditions may cause an EAL to be met for a brief period of time (e.g., a few seconds to a few minutes). The following guidance should be applied to the classification of these conditions.

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

EAL Momentarily Met But The Condition Is Corrected Prior To An Emergency Declaration - If an operator takes prompt manual action to address a condition, and the action is successful in correcting the condition prior to the emergency declaration, then the applicable EAL is not considered met and the associated emergency declaration is not required. For illustrative purposes, consider the following example:

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

It is important to stress that the 15-minute emergency classification assessment period (process clock) is not a grace period during which a classification may be delayed to allow the performance of a corrective action that would obviate the need to classify the event. Emergency classification assessments must be deliberate and timely, with no undue delays. The provision discussed above addresses only those rapidly evolving situations when an operator is able to take a successful corrective action prior to the Emergency Coordinator completing the review and steps necessary to make the emergency declaration. This provision is included to ensure that any public protective actions resulting from the emergency classification are truly warranted by the plant conditions.

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EIP-ZZ-00101 ADDENDUM 2 Rev. 021 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 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.

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

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EIP-ZZ-00101 ADDENDUM 2 Rev. 021 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT

4.0 REFERENCES

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

4.1.3. NUREG-1022 Event Reporting Guidelines: 10CFR50.72 and 50.73 4.1.4. 10 CFR 50.72 Immediate Notification Requirements for Operating Nuclear Power Reactors 4.1.5. 10 CFR 50.73 License Event Report System 4.1.6. Drawing 8600-X-88100 Property-Site Layout Owner Controlled Area and Surrounding Area 4.1.7. Callaway FSAR Figure 1.2-44 Plant Area Layout 4.1.8. Technical Specifications Table 1.1-1 MODES 4.1.9. OSP-GT-00003 Containment Closure 4.1.10. Procedure Writers Manual Callaway Plant Procedure Writers Manual 4.1.11. NSIR/DPR-ISG-01 Interim Staff Guidance, Emergency Planning for Nuclear Power Plants 4.1.12. Callaway Plant Radiological Emergency Response Plan Emergency Plan (RERP) 4.1.13. OTG-ZZ-00007 Refueling Preparation, Performance and Recovery 4.1.14. APA-ZZ-00520, Reporting Requirements and Responsibilities 4.2. Implementing 4.2.1. EIP-ZZ-00101 Classification of Emergencies 4.2.2. NEI 99-01 Rev. 6 to Callaway EAL Comparison Matrix 4.2.3. Callaway EAL Matrix 4.2.4. CR 201702763, NOS Insight - EP Risk Significant Planning Standard Performance Upper Tier Cause Evaluation Needed Page 18 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 5.0 DEFINITIONS, ACRONYMS, & ABBREVIATIONS 5.1. Definitions (ref. 4.1.1 except as noted)

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

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

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

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 Callaway, Containment Closure is established when the requirements of OSP-GT-00003 Containment Closure are met. (ref. 4.1.9)

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).

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

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.

Page 19 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Step 5.1 Cont'd Fire - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.

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

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

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 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.

High Winds - Winds in excess of 40 mph (18 m/s) sustained, or 58 mph (26 m/s) gusting.

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

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

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

Impede(d) - Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g.,

requiring use of protective equipment, such as SCBAs, that is not routinely employed).

Independent Spent Fuel Storage Installation (ISFSI) - A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage.

Page 20 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Step 5.1 Cont'd Initiating Condition - An event or condition that aligns with the definition of one of the four emergency classification levels by virtue of the potential or actual effects or consequences.

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

Owner Controlled Area (OCA) - The fenced area contiguous to the Protected Area, designated by AmerenUE (Callaway Plant) to be controlled for security purposes (ref 4.1.6).

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 Drawing 8600-X-88100 Property-Site Layout, Owner Controlled Area and Surrounding Area. The lower level of the Main Access Facility (MAF), including the Central Alarm Station (CAS) is considered inside the Protected Area. (ref. 4.1.7).

RCS Intact - The RCS should be considered intact when the RCS pressure boundary is in its normal condition for the Cold Shutdown MODE of operation (e.g., no freeze seals or nozzle dams). The RCS is capable of being placed in an intact condition by Operator Action, i.e.,

pressurized to support natural circulation cooling.

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 (< 64.0 in.) (ref. 4.1.13).

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

Restore - Take the appropriate action required to return the value of an identified parameter to the applicable limits.

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

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

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

1) The integrity of the reactor coolant pressure boundary.

2) The capability to shut down the reactor and maintain it in a safe shutdown condition.

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

Page 21 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Step 5.1 Cont'd 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 Building and the canceled Unit 2 Reactor Building. Control of access to this is by virtue of ownership and in accordance with 10CFR100 (ref. 4.1.12).

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 indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.

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

Page 22 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 5.2. Abbreviations/Acronyms

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

° .................................................................................................................................................................................................................. Degrees AC ................................................................................................................................................................................Alternating Current ATWS .......................................................................................................................... Anticipated Transient Without Scram Callaway .................................................................................................................................................... Callaway Energy Center CDE ......................................................................................................................................................Committed Dose Equivalent CFR ...................................................................................................................................................... Code of Federal Regulations CMT ........................................................................................................................................................................................... Containment CSFST ............................................................................................................................. Critical Safety Function Status Tree DBA.................................................................................................................................................................... Design Basis Accident DBT ......................................................................................................................................................................... Design Bases Threat DC ............................................................................................................................................................................................ Direct Current EAL............................................................................................................................................................... Emergency Action Level ECCS ...................................................................................................................................... Emergency Core Cooling System ECL ..............................................................................................................................................Emergency Classification Level EOF ................................................................................................................................................Emergency Operations Facility EOP ............................................................................................................................................ Emergency Operating Procedure EPA ........................................................................................................................................ Environmental Protection Agency EPIP .................................................................................................................... Emergency Plan Implementing Procedure ERG .............................................................................................................................................Emergency Response Guideline ESF ............................................................................................................................................................ Engineered Safety Feature ESW................................................................................................................................................................. Essential Service Water FAA ........................................................................................................................................... Federal Aviation Administration FBI ................................................................................................................................................ Federal Bureau of Investigation FEMA ................................................................................................................ Federal Emergency Management Agency FSAR ................................................................................................................................................. Final Safety Analysis Report GE ............................................................................................................................................................................... General Emergency IC.................................................................................................................................................................................. Initiating Condition IPEEE..................................... Individual Plant Examination of External Events (Generic Letter 88-20)

Keff............................................................................................................................. Effective Neutron Multiplication Factor LCO ........................................................................................................................................... Limiting Condition of 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...................................................................................... milli-Roentgen Equivalent Man MSL ................................................................................................................................................................................. Main Steam Line MW .....................................................................................................................................................................................................Megawatt NEI ................................................................................................................................................................. Nuclear Energy Institute Page 23 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Step 5.2 Cont'd NESP....................................................................................................................... National Environmental Studies Project NPP .......................................................................................................................................................................... Nuclear Power Plant NRC .......................................................................................................................................... Nuclear Regulatory Commission NSSS .............................................................................................................................................. Nuclear Steam Supply System NORAD............................................................................................. North American Aerospace Defense Command (NO)UE ........................................................................................................................................ Notification of Unusual Event OBE .......................................................................................................................................................Operating Basis Earthquake OCA..................................................................................................................................................................Owner Controlled Area ODCM .................................................................................................................................... Offsite Dose Calculation Manual ORO................................................................................................................................................Offsite Response Organization OTO ......................................................................................................................................... Off-Normal Operating Procedure PA ............................................................................................................................................................................................ Protected Area PAG .......................................................................................................................................................Protective Action Guideline PRA/PSA ............................................... Probabilistic Risk Assessment / Probabilistic Safety Assessment PWR ......................................................................................................................................................... Pressurized Water Reactor PSIG .............................................................................................................................................. Pounds per Square Inch Gauge R..............................................................................................................................................................................................................Roentgen RCC ...............................................................................................................................................................Reactor Control Console RCS ................................................................................................................................................................. Reactor Coolant System Rem, rem, REM ............................................................................................................................. Roentgen Equivalent Man RETS......................................................................................................Radiological Effluent Technical Specifications RPS............................................................................................................................................................ Reactor Protection System R(P)V ........................................................................................................................................................ Reactor (Pressure) Vessel RVLIS ................................................................................................................... Reactor Vessel Level Indicating System SAR....................................................................................................................................................................Safety Analysis Report SBO.................................................................................................................................................................................... Station Blackout SCBA ............................................................................................................................ Self-Contained Breathing Apparatus SG ........................................................................................................................................................................................Steam Generator SI ............................................................................................................................................................................................ Safety Injection SPDS ....................................................................................................................................... Safety Parameter Display System SRO.................................................................................................................................................................Senior Reactor Operator SSF .................................................................................................................................................................... Safe Shutdown Facility TEDE......................................................................................................................................... Total Effective Dose Equivalent TOAF ......................................................................................................................................................................... Top of Active Fuel TSC ............................................................................................................................................................. Technical Support Center WOG ................................................................................................................................................. Westinghouse Owners Group Page 24 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 6.0 CALLAWAY-TO-NEI 99-01 REV. 6 EAL CROSS-REFERENCE This cross-reference is provided to facilitate association and location of a Callaway EAL within the NEI 99-01 IC/EAL identification scheme. Further information regarding the development of the Callaway EALs based on the NEI guidance can be found in the EAL Comparison Matrix.

Callaway NEI 99-01 Rev. 6 Callaway NEI 99-01 Rev. 6 Callaway NEI 99-01 Rev. 6 Example Example Example EAL IC EAL IC EAL IC I I I EAL I I I I EAL I I I I EAL I RU1.1 AU1 1, 2 CU5.1 CU5 1, 2, 3 HA5.1 HA5 1 RU1.2 AU1 3 CA1.1 CA1 1 HA6.1 HA6 1 RU2.1 AU2 1 CA1.2 CA1 2 HA7.1 HA7 1 RA1.1 AA1 1 CA2.1 CA2 1 HS1.1 HS1 1 RA1.2 AA1 2 CA3.1 CA3 1, 2 HS6.1 HS6 1 RA1.3 AA1 3 CA6.1 CA6 1 HS7.1 HS7 1 RA1.4 AA1 4 CS1.1 CS1 1 HG7.1 HG7 1 RA2.1 AA2 1 CS1.2 CS1 2 SU1.1 SU1 1 RA2.2 AA2 2 CS1.3 CS1 3 SU3.1 SU2 1 RA2.3 AA2 3 CG1.1 CG1 1 SU4.1 SU3 2 RA3.1 AA3 1 CG1.2 CG1 2 SU5.1 SU4 1, 2, 3 RA3.2 AA3 2 FA1.1 FA1 1 SU6.1 SU5 1 RS1.1 AS1 1 FS1.1 FS1 1 SU6.2 SU5 2 RS1.2 AS1 2 FG1.1 FG1 1 SU7.1 SU6 1, 2, 3 RS1.3 AS1 3 HU1.1 HU1 1, 2 3 SU8.1 SU7 1, 2 RS2.1 AS2 1 HU2.1 HU2 1 SA1.1 SA1 1 RG1.1 AG1 1 HU3.1 HU3 1 SA3.1 SA2 1 RG1.2 AG1 2 HU3.2 HU3 2 SA6.1 SA5 1 RG1.3 AG1 3 HU3.3 HU3 3 SA9.1 SA9 1 RG2.1 AG2 1 HU3.4 HU3 4 SS1.1 SS1 1 CU1.1 CU1 1 HU4.1 HU4 1 SS2.1 SS8 1 CU1.2 CU1 2 HU4.2 HU4 2 SS6.1 SS5 1 CU2.1 CU2 1 HU4.3 HU4 3 SG1.1 SG1 1 CU3.1 CU3 1 HU4.4 HU4 4 SG1.2 SG8 1 CU3.2 CU3 2 HU7.1 HU7 1 EU1.1 E-HU1 1 CU4.1 CU4 1 HA1.1 HA1 1, 2 Page 25 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 7.0 ATTACHMENTS 7.1. Attachment 1, Emergency Action Level Technical Bases 7.2. Attachment 2, Fission Product Barrier Loss / Potential Loss Matrix and Bases 8.0

SUMMARY

OF CHANGES Section or Page(s) Description Step Number 90, 93, 143, Attachment 1 Corrected Unit Aux transformer name to XMA02. CR 202305763 145, 147, 149 Page 26 of 225 INFORMATION USE

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EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 Emergency Action Level Technical Bases Sheet 1 of 152 Category R - Abnormal Rad Release / Rad Effluent EAL Group: ANY (EALs in this category are applicable in All, Hot or Cold plant conditions.)

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 31 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 2 of 152 Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer EAL:

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

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

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

Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE Unit Vent GT-RE-21B 1.35E+8 µCi/sec 1.35E+7 µCi/sec 9.65E+5 µCi/sec 2 X Hi-Hi alarm AB-RE-111/112/

ASD Monitors (A/B/C/D) 12 mR/hr 1.2 mR/hr ---- ----

Gaseous 113/114 TD AFW Steam FC-RE-385 163 mR/hr 16.3 mR/hr 1.6 mR/hr ----

Discharge Radwaste Bldg Vent GH-RE-10B ---- ---- ---- 2 X Hi-Hi alarm Liquid Liquid Radwaste HB-RE-18 ---- ---- ---- 2 X Hi-Hi alarm Discharge MODE Applicability:

All Definition(s):

None Basis:

Page 32 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 3 of 152 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).

The effluent monitor Hi-Hi alarm setpoints correspond to the Hi-Hi alarm (red) setpoint as displayed on RM-11.

The Hi-Hi alarm value setpoints are available by examining channel 9 on the RM-23.

The RM-11 Channel Number 213 is utilized for the Unit Vent (GT-RE-21B) reading for Table R-1. This channel is read out in µCi/sec while all others are read out in µCi/ml.

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

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

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

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

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

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

Escalation of the emergency classification level would be via IC RA1.

Callaway Basis Reference(s):

1. APA-ZZ-01003, Callaway Plant Offsite Dose Calculation Manual Section 2.2.3

2. FSAR Section 16.11.1.3, Radioactive Effluent Monitoring Instrumentation LCO

3. EPCI 1402, EAL Table R-1 Calculations

4. NEI 99-01, AU1 Page 33 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 4 of 152 Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer.

EAL:

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

> 2 x ODCM limits for 60 min.

(Notes 1, 2)

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

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

MODE Applicability:

All Definition(s):

None Basis:

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

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.).

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.

-Continued Below-Page 34 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 5 of 152 Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL.

All water runoff from the plant eventually flows into Logan Creek and then to the Missouri River. If radioactive liquid flows offsite, begin hourly grab samples at the Portland River Sample Location and analyze for tritium and gamma spectrum. Send results to the Dose Assessment Technician or Dose Assessment Coordinator for evaluation.

Escalation of the emergency classification level would be via IC RA1.

Callaway Basis Reference(s):

1. APA-ZZ-01003, Callaway Plant Offsite Dose Calculation Manual Section 2.2.3

2. NEI 99-01, AU1 Page 35 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 6 of 152 Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL:

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

(Notes 1, 2, 3, 4)

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

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

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

Note 4: The pre-calculated effluent monitor values presented in EALs 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 GT-RE-21B 1.35E+8 µCi/sec 1.35E+7 µCi/sec 9.65E+5 µCi/sec 2 X Hi-Hi alarm AB-RE-111/112/

ASD Monitors (A/B/C/D) 12 mR/hr 1.2 mR/hr ---- ----

Gaseous 113/114 TD AFW Steam FC-RE-385 163 mR/hr 16.3 mR/hr 1.6 mR/hr ----

Discharge Radwaste Bldg Vent GH-RE-10B ---- ---- ---- 2 X Hi-Hi alarm Liquid Liquid Radwaste HB-RE-18 ---- ---- ---- 2 X Hi-Hi alarm Discharge MODE Applicability:

All Definition(s):

None Page 36 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 7 of 152 Basis:

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

• 10 mRem TEDE

• 50 mRem CDE Thyroid 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.

To clarify Note 4, if a threshold value is met in Table R-1 for a classification, there is a 15-minute time limit to make the classification (RA1.1). If Dose Assessment (URI/RASCAL) is available, it should be used instead (RA1.2) since it is more accurate than the values in Table R-1. However, the Dose Assessment personnel must be able to calculate results within 15 minutes of the Table R-1 value being exceeded OR the classification should be made using Table R-1 (RA1.1).

The RM-11 Channel Number 213 is utilized for the Unit Vent (GT-RE-21B) reading for Table R-1. This channel is read out in µCi/sec while all others are read out in µCi/ml.

The column ALERT gaseous effluent release values in Table R-1 correspond to calculated doses of 1%

(10% of the SAE thresholds) of the EPA Protective Action Guidelines (TEDE or CDE Thyroid) (ref. 1).

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

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

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

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

Callaway Basis Reference(s):

1. EPCI 1402, EAL Table R-1 Calculations

2. NEI 99-01, AA1 Page 37 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 8 of 152 Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL:

RA1.2 Alert Dose assessment using actual meteorology indicates doses > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY.

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 Unit 1 Reactor Building and the canceled Unit 2 Reactor Building. Control of access to this is by virtue of ownership and in accordance with 10CFR100.

Basis:

Dose assessments are performed by computer-based method (ref. 1, 2).

This IC is used based on results from the Unified RASCAL Interface software (URI) regardless of the input source. This value is in mrem TEDE or thyroid CDE.

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

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

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

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

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

Callaway Basis Reference(s):

1. EIP-ZZ-01211, Accident Dose Assessment

2. EPCI 1402, EAL Table R-1 Calculations

3. NEI 99-01, AA1 Page 38 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 9 of 152 Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE.

EAL:

RA1.3 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 CDE at or beyond the SITE BOUNDARY for 60 min.

of exposure.

(Notes 1, 2)

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

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

MODE Applicability:

All Definition(s):

SITE BOUNDARY - 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 Building and the canceled Unit 2 Reactor Building.

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 is based on liquid sample analysis by the Count Room.

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.

-Continued Below-Page 39 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 10 of 152 The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

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

All water runoff from the plant eventually flows into Logan Creek and then to the Missouri River. If radioactive liquid flows offsite, begin hourly grab samples at the Portland River Sample Location and analyze for tritium and gamma spectrum. Send results to the Dose Assessment Technician or Dose Assessment Coordinator for evaluation.

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

Callaway Basis Reference(s):

1. APA-ZZ-01003, Callaway Plant Offsite Dose Calculation Manual Section 2.2.3

2. NEI 99-01, AA1 Page 40 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 11 of 152 Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE.

EAL:

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

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

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

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

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

MODE Applicability:

All Definition(s):

SITE BOUNDARY - 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 Building and the canceled Unit 2 Reactor Building.

Control of access to this is by virtue of ownership and in accordance with 10CFR100.

Basis:

EIP-ZZ-00211, Field Monitoring provides guidance for emergency or post-accident radiological environmental monitoring (ref. 1).

This IC is based solely on field monitoring team results without performing calculations using the Unified RASCAL Interface software (URI).

The closed window value is in mR/hr. The analysis of field survey samples is in mrem thyroid CDE for 60 minutes.

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).

-Continued Below-Page 41 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 12 of 152 Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

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

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

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

Callaway Basis Reference(s):

1. EIP-ZZ-00211, Field Monitoring

2. NEI 99-01, AA1 Page 42 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 13 of 152 Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL:

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

(Notes 1, 2, 3, 4)

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

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

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

Note 4: The pre-calculated effluent monitor values presented in EALs 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 GT-RE-21B 1.35E+8 µCi/sec 1.35E+7 µCi/sec 9.65E+5 µCi/sec 2 X Hi-Hi alarm AB-RE-111/112/

ASD Monitors (A/B/C/D) 12 mR/hr 1.2 mR/hr ---- ----

Gaseous 113/114 TD AFW Steam FC-RE-385 163 mR/hr 16.3 mR/hr 1.6 mR/hr ----

Discharge Radwaste Bldg Vent GH-RE-10B ---- ---- ---- 2 X Hi-Hi alarm Liquid Liquid Radwaste HB-RE-18 ---- ---- ---- 2 X Hi-Hi alarm Discharge MODE Applicability:

All Definition(s):

None Page 43 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 14 of 152 Basis:

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

• 100 mRem TEDE

• 500 mRem CDE Thyroid 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.

To clarify Note 4, if a threshold value is met in Table R-1 for a classification, there is a 15-minute time limit to make the classification (RS1.1). If Dose Assessment (URI/RASCAL) is available, it should be used instead (RS1.2) since it is more accurate than the values in Table R-1. However, the Dose Assessment personnel must be able to calculate results within 15 minutes of the Table R-1 value being exceeded OR the classification should be made using Table R-1 (RS1.1).

The RM-11 Channel Number 213 is utilized for the Unit Vent (GT-RE-21B) reading for Table R-1. This channel is read out in µCi/sec while all others are read out in µCi/ml.

The column SAE gaseous effluent release value in Table R-1 corresponds to calculated doses of 10% of the EPA Protective Action Guidelines (TEDE or CDE Thyroid) (ref. 1).

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

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

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

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

Callaway Basis Reference(s):

1. EPCI 1402, EAL Table R-1 Calculations

2. NEI 99-01, AS1 Page 44 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 15 of 152 Category:R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL:

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

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 Unit 1 Reactor Building and the canceled Unit 2 Reactor Building. Control of access to this is by virtue of ownership and in accordance with 10CFR100.

Basis:

Dose assessments are performed by computer-based method (ref. 1, 2)

This IC is used based on results from the Unified RASCAL Interface software (URI) regardless of the input source.

This value is in mrem TEDE or thyroid CDE.

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

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

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

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

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

Callaway Basis Reference(s):

1. EIP-ZZ-01211, Accident Dose Assessment

2. EPCI 1402, EAL Table R-1 Calculations

3. NEI 99-01, AS1 Page 45 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 16 of 152 Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL:

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

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

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

(Notes 1, 2)

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

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

MODE Applicability:

All Definition(s):

SITE BOUNDARY - 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 Building and the canceled Unit 2 Reactor Building.

Control of access to this is by virtue of ownership and in accordance with 10CFR100.

Basis:

EIP-ZZ-00211, Field Monitoring provides guidance for emergency or post-accident radiological environmental monitoring (ref. 1).

This IC is based solely on field monitoring team results without performing calculations using the Unified RASCAL Interface software (URI).

The closed window value is in mR/hr. The analysis of field survey samples is in mrem thyroid CDE for 60 minutes.

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.

-Continued Below-Page 46 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 17 of 152 Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

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

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

Callaway Basis Reference(s):

1. EIP-ZZ-00211, Field Monitoring

2. NEI 99-01, AS1 Page 47 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 18 of 152 Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE EAL:

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

(Notes 1, 2, 3, 4)

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

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

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

Note 4: The pre-calculated effluent monitor values presented in EALs 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 GT-RE-21B 1.35E+8 µCi/sec 1.35E+7 µCi/sec 9.65E+5 µCi/sec 2 X Hi-Hi alarm AB-RE-111/112/

ASD Monitors (A/B/C/D) 12 mR/hr 1.2 mR/hr ---- ----

Gaseous 113/114 TD AFW Steam FC-RE-385 163 mR/hr 16.3 mR/hr 1.6 mR/hr ----

Discharge Radwaste Bldg Vent GH-RE-10B ---- ---- ---- 2 X Hi-Hi alarm Liquid Liquid Radwaste HB-RE-18 ---- ---- ---- 2 X Hi-Hi alarm Discharge MODE Applicability:

All Definition(s):

None Page 48 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 19 of 152 Basis:

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

• 1000 mRem TEDE

• 5000 mRem CDE Thyroid 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.

To clarify Note 4, if a threshold value is met in Table R-1 for a classification, there is a 15-minute time limit to make the classification (RG1.1). If Dose Assessment (URI/RASCAL) is available, it should be used instead (RG1.2) since it is more accurate than the values in Table R-1. However, the Dose Assessment personnel must be able to calculate results within 15 minutes of the Table R-1 value being exceeded OR the classification should be made using Table R-1 (RG1.1).

The RM-11 Channel Number 213 is utilized for the Unit Vent (GT-RE-21B) reading for Table R-1. This channel is read out in µCi/sec while all others are read out in µCi/ml.

The column GE gaseous effluent release values in Table R-1 correspond to calculated doses of 100% of the EPA Protective Action Guidelines (TEDE or CDE Thyroid) (ref. 1).

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

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

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

Callaway Basis Reference(s):

1. EPCI 1402, EAL Table R-1 Calculations

2. NEI 99-01, AG1 Page 49 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 20 of 152 Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE EAL:

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

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 Unit 1 Reactor Building and the canceled Unit 2 Reactor Building. Control of access to this is by virtue of ownership and in accordance with 10CFR100.

Basis:

Dose assessments are performed by computer-based method (ref. 1, 2)

This IC is used based on results from the Unified RASCAL Interface software (URI) regardless of the input source.

This value is in mrem TEDE or thyroid CDE.

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

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

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

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.

Callaway Basis Reference(s):

1. EIP-ZZ-01211, Accident Dose Assessment

2. EPCI 1402, EAL Table R-1 Calculations

3. NEI 99-01, AG1 Page 50 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 21 of 152 Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE EAL:

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

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

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

(Notes 1, 2)

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

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

MODE Applicability:

All Definition(s):

SITE BOUNDARY - 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 Building and the canceled Unit 2 Reactor Building.

Control of access to this is by virtue of ownership and in accordance with 10CFR100.

Basis:

EIP-ZZ-00211, Field Monitoring provides guidance for emergency or post-accident radiological environmental monitoring (ref. 1).

This IC is based solely on field monitoring team results without performing calculations using the Unified RASCAL Interface software (URI).

The closed window value is in mR/hr. The analysis of field survey samples is in mrem thyroid CDE for 60 minutes.

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.

-Continued below-Page 51 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 22 of 152 Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

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

Callaway Basis Reference(s):

1. EIP-ZZ-00211, Field Monitoring

2. NEI 99-01, AG1 Page 52 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 23 of 152 Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Unplanned loss of water level above irradiated fuel EAL:

RU2.1 Unusual Event UNPLANNED water level drop in the REFUELING PATHWAY as indicated by low water level alarm or indication (EC LI-0039A, EC LI-0039B, local observation of SFP level).

AND UNPLANNED rise in corresponding area radiation levels as indicated by any Table R-2 radiation monitors.

Table R-2 Fuel Building & Containment Area Radiation Monitors Fuel Building: Containment:

SD-RE-34, Cask Handle Area Radiation SD RE 40, Personnel Access Hatch Area SD-RE-35, New Fuel Storage Area Radiation SD RE 41, Manipulator Crane Radiation Monitor SD-RE-36, New Fuel Storage Area Radiation SD RE 42, Containment Building Radiation SD-RE-37, Fuel Pool Bridge Crane Radiation GT RE 59 Containment High Area Radiation Monitor SD-RE-38, Spent Fuel Pool Area Radiation 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.

-Continued Below-Page 53 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 24 of 152 Basis:

The low water level alarm in this EAL refers to the Spent Fuel Pool (SFP) low level alarm (Annunciator 76D, 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 remotely monitored by level indicator EC LI-0039A. The level switch initiates high- and low-level annunciators.

Technical Specification 3.7.15 (ref. 2) requires at least 23 ft. of water above the Spent Fuel Pool storage racks.

Technical Specification 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). Increasing radiation indications on these monitors in the absence of indications of decreasing REFUELING CAVITY level are not classifiable under this EAL.

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

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

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

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

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

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

Callaway Basis Reference(s):

1. OTO-EC-00001, Loss of Spent Fuel Pool/Refuel Pool Level

2. Technical Specification 3.7.15, Fuel Storage Pool Water Level

3. Technical Specification 3.9.7, Refueling Pool Water Level

4. NEI 99-01, AU2 Page 54 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 25 of 152 Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL:

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

MODE Applicability:

All Definition(s):

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

Basis:

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

This EAL escalates from RU2.1 in that the loss of level, in the affected portion of the REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation (e.g., reports from personnel or camera images), as well as significant changes in water and radiation levels, or other plant parameters.

Computational aids may also be used (e.g., a boil-off curve). Classification of an event using this EAL should be based on the totality of available indications, reports, and observations.

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

A drop in water level above irradiated fuel within the reactor vessel may require classification in accordance Recognition Category C during the Cold Shutdown and Refueling MODES.

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

Callaway Basis Reference(s):

1. OTO-EC-00001, Loss of Spent Fuel Pool/Refuel Pool Level

2. NEI 99-01, AA2 Page 55 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 26 of 152 Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL:

RA2.2 Alert Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by any of the following:

• Hi-Hi Alarm on Fuel Building exhaust monitors (GG-RE-27 or 28).

• Manipulator crane radiation monitor (SD-RE-41) >100 mR/hr.

• Fuel Pool Bridge Crane OR Spent Fuel Pool Area radiation monitor (SD-RE-37 or 38) > 30 mR/hr.

MODE Applicability:

All Definition(s):

None Basis:

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

The bases for the SFP ventilation radiation Hi-Hi alarm and the SFP and containment area radiation readings are a spent fuel handling accident (ref. 2, 3). 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, 3, 4).

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

This EAL applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed.

Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with EU1.1. Cask is sealed when welding is complete.

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

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

Callaway Basis Reference(s):

1. OTO-EC-00001, Loss of Spent Fuel Pool/Refuel Pool Level

2. OTO-KE-00001, Fuel Handling Accident

3. Calc. EPCI 98-01, Emergency Action Level Bases

4. Calc. HPCI 05-02, Gaseous and Liquid Radiation Monitor Setpoints

5. NEI 99-01, AA2 Page 56 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 27 of 152 Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL:

RA2.3 Alert Lowering of spent fuel pool level to 120 as indicated on EC-LI-0059A or EC-LI-0060A.

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 Callaway Plant SFP Level 2 is plant elevation 2031 ft. 1.25 in. (~9 ft. 11 in. above the top of the spent fuel racks) as indicated by 120 on EC-LI-0059A in the Auxiliary Building Hallway 2026. Backup indication is also available on EC-LI-0060A in the Auxiliary Building hallway 2026.

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

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

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

Escalation of the emergency classification level would be via IC RS2.

Callaway Basis Reference(s):

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

2. SFPIS Mod Overview for EP 3 NEI 99-01, AA2 Page 57 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 28 of 152 Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Spent fuel pool level at the top of the fuel racks.

EAL:

RS2.1 Site Area Emergency Lowering of spent fuel pool level to 12 as indicated on EC-LI-0059A or EC-LI-0060A.

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 Callaway Plant SFP Level 3 has been set at a plant elevation 2022 ft. 1.25 in. (~11 in. above the top of the spent fuel racks) as indicated by 12 on EC-LI-0059A in the Auxiliary Building Hallway 2026. Backup indication is also available on EC-LI-0060A in the Auxiliary Building hallway 2026.

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

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

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

Callaway Basis Reference(s):

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

2. SFPIS Mod Overview for EP

3. NEI 99-01, AS2 Page 58 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 29 of 152 Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Spent fuel pool level cannot be restored to at least the top of the fuel racks for 60 minutes or longer.

EAL:

RG2.1 General Emergency Spent fuel pool level cannot be restored to at least 12 as indicated on EC-LI-0059A or EC-LI-0060A for 60 min.

(Note 1)

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

MODE Applicability:

All Definition(s):

None Basis:

Post-Fukushima order EA-12-051 (ref. 1) required the installation of reliable SFP level indication capable of identifying normal level (Level 1), 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 Callaway Plant SFP Level 3 has been set at a plant elevation 2022 ft. 1.25 in. (~11 in. above the top of the spent fuel racks) as indicated by 12 on EC-LI-0059A in the Auxiliary Building Hallway 2026. Backup indication is also available on EC-LI-0060A in the Auxiliary Building hallway 2026.

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.

Callaway Basis Reference(s):

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

2. SFPIS Mod Overview for EP

3. NEI 99-01, AG2 Page 59 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 30 of 152 Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 3 - Area Radiation Levels Initiating Condition: Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown.

EAL:

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

• Control Room (SD-RE-33).

• Central Alarm Station (by survey).

MODE Applicability:

All Definition(s):

None 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 Coordinator should consider the cause of the increased radiation levels and determine if another IC may be applicable.

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

Callaway Basis Reference(s):

1. FSAR Section 12.3, Table 12.3-2, Area Radiation Monitors

2. NEI 99-01, AA3 Page 60 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 31 of 152 Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 3 - Area Radiation Levels Initiating Condition: Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown.

EAL:

RA3.2 Alert An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to EITHER of the following: (Note 5)

• North Electrical Penetration Room. (Room 1410)

• South Electrical Penetration Room. (Room 1409)

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.

MODE Applicability:

4 - Hot Shutdown Definition(s):

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

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

Basis:

The only rooms/areas external to the Control Room that require access to perform field actions consistent with the above criteria for Callaway are the North and South Electrical Penetration Rooms when in MODE 4 to support isolating SI accumulators and placing RHR in service for RCS cooldown to Cold Shutdown (ref.

1, 2, 3). The equipment required is:

For SI Accumulators:

• NG01BGF3, FDR BKR TO EPHV8808A SI ACC A OUT ISO, (Room 1410)

• NG02BGF3, FDR BKR TO EPHV8808B SI ACC B OUT ISO, (Room 1409)

• NG01BGF2, FDR BKR TO EPHV8808C SI ACC C OUT ISO, (Room 1410)

• NG02BHF2, FDR BKR TO EPHV8808D SI ACC D OUT ISO, (Room 1409)

For "A" RHR:

• NG02BCF2, FDR BKR TO BBPV8702A RCS LOOP 1 HOT LEG TO RHR PMPS ISO, (Room 1409)

• NG01BEF2, FDR BKR TO EJHV8701A A RHR PMP SUCT FROM RCS HOT LEG 1 ISO, (Room 1410)

-Continue Below-Page 61 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 32 of 152 For "B" RHR:

• NG02BBF3, FDR BKR TO BBPV8702B RCS LOOP 4 HOT LEG TO RHR PMPS ISO, (Room 1409)

• NG01BDF3, FDR BKR TO EJHV8701B B RHR PMP SUCT FROM RCS HOT LEG 4 ISO, (Room 1410)

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

Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating MODE in effect at the time of the elevated radiation levels. The emergency classification is not contingent upon whether entry is actually necessary at the time of the increased radiation levels.

Access should be considered as IMPEDED if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., installing temporary shielding, requiring use of non-routine protective equipment, requesting an extension in dose limits beyond normal administrative limits).

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

• The plant is NOT in MODE 4.

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

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

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

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

Callaway Basis Reference(s):

1. OTG-ZZ-00006 Addendum 06, Securing Safety Injection Accumulators 2 OTN-EJ-00001 Addendum 3, Placing A RHR Train In Service for RCS Cooldown

3. OTN-EJ-00001 Addendum 4, Placing B RHR Train In Service for RCS Cooldown

4. NEI 99-01, AA3 Page 62 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 33 of 152 Category E - Independent Spent Fuel Storage Installation (ISFSI)

EAL Group: Any (EALs in this category are applicable to All, Hot, or Cold plant conditions.)

An independent spent fuel storage installation (ISFSI) is a facility that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. A significant amount of the radioactive material contained within a canister must escape its packaging and enter the biosphere for there to be a significant environmental effect resulting from an accident involving the dry storage of spent nuclear fuel.

An Unusual Event is declared on the basis of the occurrence of an event of sufficient magnitude that a loaded cask confinement boundary is damaged or violated.

Page 63 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 34 of 152 Category: E - Independent Spent Fuel Storage Installation (ISFSI)

Subcategory: 1 - Confinement Boundary Initiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARY EAL:

EU1.1 Unusual Event Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading > EITHER of the following:

• 60 mrem/hr (gamma + neutron) on the top of the closure lid of the Overpack/VVM.

• 7,000 mrem/hr (gamma + neutron) on the side of the Transfer Cask.

MODE Applicability:

All Definition(s):

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

OVERPACK - For the HI-Storm UMAX, the term OVERPACK is synonyms with the term VVM.

TRANSFER CASK - Containers designed to contain the MPC during and after loading of spent fuel assemblies, and prior to and during unloading and to transfer the MPC to or from the overpack/VVM VERTICAL VENTILATED MODULE (VVM) - A subterranean type overpack which receives and contains the sealed MPC for interim storage at the ISFSI. The VVM supports the MPC in a vertical orientation and provide gamma and neutron shielding and also provides air flow through cooling passages to promote heat transfer from the MPC to the environs.

Basis:

Confinement boundary is established at Callaway when the Multi-Purpose Canister welding is complete.

Overpacks/VVM casks receive and contain the sealed MPCs for interim storage in the ISFSI. They provide gamma and neutron shielding and provide for ventilated air flow to promote heat transfer from the MPC to the environs. The term overpack/VVM does not include the transfer cask (ref. 1).

The values shown represents 2 times the limits specified in the ISFSI Certificate of Compliance Technical Specification 5.3.4 for radiation external to either a loaded MPC overpack/VVM or transfer cask (ref. 1).

-Continued Below-Page 64 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 35 of 152 This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of a storage cask containing spent fuel. It applies to irradiated fuel that is licensed for dry storage beginning at the point that the loaded storage cask is sealed. The issues of concern are the creation of a potential or actual release path to the environment, degradation of one or more fuel assemblies due to environmental factors, and configuration changes which could cause challenges in removing the cask or fuel from storage. The cask is sealed when the welding is complete.

The existence of damage is determined by radiological survey. The technical specification multiple of 2 times, which is also used in Recognition Category R IC RU1, is used here to distinguish between non-emergency and emergency conditions. The emphasis for this classification is the degradation in the level of safety of the spent fuel cask and not the magnitude of the associated dose or dose rate. It is recognized that in the case of extreme damage to a loaded cask, the fact that the on-contact dose rate limit is exceeded may be determined based on measurement of a dose rate at some distance from the cask.

Security-related events for ISFSIs are covered under IC HS1.

Callaway Basis Reference(s):

1. Certificate of Compliance No. 1040 Appendix A Technical Specifications for the HI-STORM UMAX Canister Storage System

2. NEI 99-01, E-HU1 Page 65 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 36 of 152 Category C - Cold Shutdown / Refueling System Malfunction EAL Group: Cold Conditions (RCS temperature 200ºF); 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.

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

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

Page 66 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 37 of 152 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: UNPLANNED loss of RCS inventory for 15 minutes or longer 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 Coordinator should declare the event promptly upon determining that time limit has been exceeded or will likely be exceeded.

MODE Applicability:

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

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

Basis:

With the plant in Cold Shutdown, RCS water level is normally maintained above the pressurizer low level setpoint of 17% (ref. 1). However, if RCS level is being controlled below the pressurizer low level setpoint, or if level is being maintained in a designated band in the reactor vessel it is the inability to maintain level above the low end of the designated control band due to a loss of inventory resulting from a leak in the RCS that is the concern.

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

The Plant Computer System Display called Refuel Level Indications (turn on code RLI) is available to assist in monitoring important parameters crucial to RCS draining operations (ref. 3).

This IC addresses the inability to restore and maintain water level to a required minimum level. This condition 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.

-Continued Below-Page 67 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 38 of 152 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 RCS level, specified for the current plant conditions, cannot be maintained for 15 minutes or longer.

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.

Callaway Basis Reference(s):

1. FR-I.2, Response to Low Pressurizer Level

2. OTN-BB-00002, Reactor Coolant System Draining

3. Technical Specification 3.9.7, Refueling Pool Water Level

4. NEI 99-01, CU1 Page 68 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 39 of 152 Category: C - Cold Shutdown / 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 / Tanks

• Containment Sumps

• Containment Normal Sumps

• Containment Instrument Sump

• PRT

• RCDT

• Auxiliary Building Sump MODE Applicability:

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

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). The RCS is capable of being placed in an intact condition by Operator Action, i.e., pressurized to support natural circulation cooling.

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

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

Basis:

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

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

-Continued Below-Page 69 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 40 of 152 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 connected to the RCS that cannot be isolated could also be indicative of a loss of RCS inventory (ref. 1, 2).

The Plant Computer System Display called Refuel Level Indications (turn on code RLI) is available to assist in monitoring important parameters crucial to RCS draining operations (ref. 3).

This IC addresses a loss of the ability to monitor RCS level concurrent with indications of coolant leakage.

This condition 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 level have been lost. In this condition, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels (Table C-1). Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RCS.

Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA1 or CA3.

Callaway Basis Reference(s):

1. OTO-BB-00003, R014, Excess RCS Leakage

2. OSP-BB-00009, RCS Inventory Balance

3. OTN-BB-00002, Reactor Coolant System Draining

4. NEI 99-01, CU1

5. OOA-BB-00003, Refuel Level Indications

6. TDB-001, Callaway Energy Center Tank Data Book Page 70 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 41 of 152 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: Loss of RCS inventory EAL:

CA1.1 Alert Loss of RCS inventory as indicated by Reactor Vessel level < bottom of RCS hot leg ID (RVLIS Pumps Off < 73%).

MODE Applicability:

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

None Basis:

When Reactor Vessel water level lowers to 2013.29 ft. (ref. 1), the inside diameter (ID) of the bottom of the RCS hot leg penetration is uncovered. The elevation of the bottom of the RCS hot leg penetration can be monitored only by RVLIS. (Note that this threshold is the loop penetration at the Reactor Vessel not the low point of the loop.) (ref. 3) When RVLIS is out of service, classification should be based on CA1.2 if RCS inventory cannot be monitored.

The RVLIS Pumps Off threshold has been determined as follows (ref. 1, 2):

Elevation of bottom of Reactor Vessel (ft) A 1987.150 Elevation of bottom ID of RCS hot leg penetration (ft) B 2013.290 Hot leg penetration (above vessel bottom) C = B - A (ft) 26.140 Height of vessel D (ft) 41.245 RVLIS indication corresponding to the top of the core: H = 100 x C / D (%) 63.377 RVLIS overall channel accuracy: OCA = 7.48% + (0.0104 x H) + 0.81% ---

OCA at H (%) 8.949 Bottom ID of RCS loop, including channel uncertainties: H + OCA (%) 72.327 Rounded upward to nearest 1% (RVLIS range is 0 - 120% in 2% increments) 73 The threshold was chosen because level indication may be lost (RVLIS is normally inoperable in Refueling MODE (ref. 2)) and loss of suction to decay heat removal systems has occurred. The inability to restore and maintain level after reaching this setpoint infers a failure of the RCS barrier.

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.

-Continued Below-Page 71 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 42 of 152 For this EAL, a lowering of RCS water level below the specified level 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 Decay Heat Removal suction point). An increase in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3.

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

Callaway Basis Reference(s):

1. OOA-BB-00003, Refuel Level Indications

2. Calculation No. BB-177 (387.1 - CAL RVLIS Setpoints)

3. OTN-BB-00002, Reactor Coolant System Draining

4. NEI 99-01, CA1 Page 72 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 43 of 152 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: Loss of RCS inventory EAL:

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

AND EITHER

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

• Visual observation of UNISOLABLE RCS leakage.

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

Table C-1 Sumps / Tanks

• Containment Sumps

• Containment Normal Sumps

• Containment Instrument Sump

• PRT

• RCDT

• Auxiliary Building Sump MODE Applicability:

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

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). The RCS is capable of being placed in an intact condition by Operator Action, i.e., pressurized to support natural circulation cooling.

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.

-Continued Below-Page 73 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 44 of 152 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). 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).

The Plant Computer System Display called Refuel Level Indications (turn on code RLI) is available to assist in monitoring important parameters crucial to RCS draining operations (ref. 3).

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.

Callaway Basis Reference(s):

1. OTO-BB-00003-R014, Excess RCS Leakage

2. OSP-BB-00009, RCS Inventory Balance

3. OTN-BB-00002, Reactor Coolant System Draining

4. NEI 99-01, CA1

5. OOA-BB-00003, Refuel Level Indications

6. TDB-001, Callaway Energy Center Tank Data Book Page 74 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 45 of 152 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: Loss of RCS inventory affecting core decay heat removal capability EAL:

CS1.1 Site Area Emergency With CONTAINMENT CLOSURE not established, RVLIS Pumps Off < 72%.

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 Callaway, Containment Closure is established when the requirements of OSP-GT-00003, Containment Closure are met.

Basis:

When Reactor Vessel water level lowers to 2012.79 ft. (ref. 1), water level is six inches below the bottom of the RCS hot leg penetration. When Reactor Vessel water level drops significantly below the bottom of the RCS hot leg penetration, all sources of RCS injection have failed or are incapable of making up for the inventory loss. Six inches below the bottom of the RCS hot leg penetration can be monitored only by RVLIS. Level monitoring instruments BB LI-53A/B and Computer Point BBL0053BB cannot sense level changes in the Reactor Vessel below the RCS loop hot leg penetration. The Plant Computer System Display called Refuel Level Indications (turn on code RLI) is available to assist in monitoring important parameters crucial to RCS draining operations (ref. 3). When RVLIS is out of service, classification should be based on CS1.3 if RCS inventory cannot be monitored.

The RVLIS Pumps Off threshold has been determined as follows (ref. 1, 2):

Elevation of bottom of Reactor Vessel (ft) A 1987.150 Elevation of bottom ID of RCS hot leg penetration (ft) B 2013.290 Six inches below hot leg penetration (above vessel bottom) C = B - A - 0.5 (ft) 25.640 Height of vessel D (ft) 41.245 RVLIS indication corresponding to the top of the core: H = 100 x C / D (%) 62.165 RVLIS overall channel accuracy: OCA = 7.48% + (0.0104 x H) + 0.81% ---

OCA at H (%) 8.937 Six inches below Bottom ID of RCS loop, including channel uncertainties: H + OCA (%) 71.102 Rounded upward to nearest 1% (RVLIS range is 0 - 120% in 2% increments) 72

-Continued Below-Page 75 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 46 of 152 Under the conditions specified by this EAL, continued lowering of Reactor Vessel water level is indicative of a loss of inventory control. Inventory loss may be due to a vessel breach, RCS pressure boundary leakage or continued boiling in the Reactor Vessel. The magnitude of this loss of water indicates that makeup systems have not been effective and may not be capable of preventing further RCS or Reactor Vessel water level drop and potential core uncovery. The inability to restore and maintain level after reaching this setpoint infers a failure of the RCS barrier and Potential Loss of the Fuel Clad barrier.

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

This IC addresses a significant and prolonged loss of reactor vessel/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 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.

Callaway Basis Reference(s):

1. OOA-BB-00003, Refuel Level Indications

2. Calculation No. BB-177, (387.1 - CAL RVLIS Setpoints)

3. OTN-BB-00002, Reactor Coolant System Draining

4. OSP-GT-00003, Containment Closure

5. NEI 99-01, CS1 Page 76 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 47 of 152 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: Loss of RCS inventory affecting core decay heat removal capability EAL:

CS1.2 Site Area Emergency With CONTAINMENT CLOSURE established, RVLIS Pumps Off < 65% (Top of Fuel).

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 Callaway, Containment Closure is established when the requirements of OSP-GT-00003, Containment Closure are met.

Basis:

When Reactor Vessel water level drops below RVLIS Pumps Off indication of 65% (2010.29 ft.), core uncovery is about to occur. The Plant Computer System Display called Refuel Level Indications (turn on code RLI) is available to assist in monitoring important parameters crucial to RCS draining operations (ref. 3). When RVLIS is out of service, classification should be based on CS1.3 if RCS inventory cannot be monitored.

-Continued Below-Page 77 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 48 of 152 Under the conditions specified by this EAL, continued lowering of Reactor Vessel water level is indicative of a loss of inventory control. Inventory loss may be due to a vessel breach, RCS pressure boundary leakage or continued boiling in the Reactor Vessel. The magnitude of this loss of water indicates that makeup systems have not been effective and may not be capable of preventing further RCS or Reactor Vessel water level drop and potential core uncovery. The inability to restore and maintain level after reaching this setpoint infers a failure of the RCS barrier and Potential Loss of the Fuel Clad barrier.

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

This IC addresses a significant and prolonged loss of reactor vessel/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 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.

Callaway Basis Reference(s):

1. OOA-BB-00003, Refuel Level Indications

2. Calculation No. BB-177, (387.1 - CAL RVLIS Setpoints)

3. OTN-BB-00002, Reactor Coolant System Draining

4. OSP-GT-00003, Containment Closure

5. NEI 99-01, CS1 Page 78 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 49 of 152 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: Loss of RCS inventory affecting core decay heat removal capability EAL:

CS1.3 Site Area Emergency RCS water level cannot be monitored for 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 crane radiation monitor SD-RE-41 > 10,000 mR/hr.

• Erratic Source Range Monitor indication.

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

Table C-1 Sumps / Tanks

• Containment Sumps

• Containment Normal Sumps

• Containment Instrument Sump

• PRT

• RCDT

• Auxiliary Building Sump MODE Applicability:

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

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). The RCS is capable of being placed in an intact condition by Operator Action, i.e., pressurized to support natural circulation cooling.

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.

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

-Continued Below-Page 79 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 50 of 152 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).

The Plant Computer System Display called Refuel Level Indications (turn on code RLI) is available to assist in monitoring important parameters crucial to RCS draining operations (ref. 3).

The Reactor Vessel inventory loss may be detected by the manipulator 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 manipulator crane radiation monitor (SD-RE-41) indication (ref.

4,5,6).

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. 7, 8).

This IC 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 level cannot be restored, fuel damage is probable.

The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level),

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

The inability to monitor 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 Callaway Basis Reference(s):

1. OTO-BB-00003-R014, Excess RCS Leakage

2. OSP-BB-00009, RCS Inventory Balance

3. OTN-BB-00002, Reactor Coolant System Draining

4. FSAR, Section 12.3.3.4

5. FSAR, Table 12.3-2

6. Calc. No. HPCI -0701, SD-RE-41 Response to Core Uncovery in Refueling MODE

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

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

9. NEI 99-01, CS1

10. OOA-BB-00003, Refuel Level Indications

11. TDB-001, Callaway Energy Center Tank Data Book Page 80 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 51 of 152 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: Loss of RCS inventory affecting fuel clad integrity with containment challenged EAL:

CG1.1 General Emergency RVLIS Pumps Off < 65% (Top of Fuel) for 30 min. (Note 1)

AND Any Containment Challenge indication, Table C-2.

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

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

Table C-2 Containment Challenge Indications

• CONTAINMENT CLOSURE not established (Note 6)

• Containment hydrogen concentration 4%

• Unplanned rise in Containment pressure 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 Callaway, Containment Closure is established when the requirements of OSP-GT-00003 Containment Closure are met.

Basis:

When Reactor Vessel water level drops below RVLIS Pumps Off indication of 65% (2010.29 ft.), core uncovery is about to occur. The Plant Computer System Display called Refuel Level Indications (turn on code RLI) is available to assist in monitoring important parameters crucial to RCS draining operations (ref. 3). When RVLIS is out of service, classification should be based on CG1.2 if RCS inventory cannot be monitored.

-Continued Below-Page 81 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 52 of 152

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

2. Containment hydrogen 4% - The 4% hydrogen concentration threshold is generally considered the lower limit for hydrogen deflagrations. Callaway 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 hydrogen monitors (GS AI-10 and GS AI-19) with a range of 0% to 10% provide indication on Control Room Panel RL020 and Emergency Response Facilities Information System (ERFIS) (ref. 6, 7). The hydrogen monitors require a 2-hour warmup period when starting from the OFF position and 15 minutes when starting from STANDBY (ref. 8, 9).

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 (ref. 4).

-Continued Below-Page 82 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 53 of 152 Under the conditions specified by this EAL, continued lowering of Reactor Vessel water level is indicative of a loss of inventory control with a challenge to the Containment. Inventory loss may be due to a vessel breach, RCS pressure boundary leakage or continued boiling in the Reactor Vessel. The magnitude of this loss of water indicates that makeup systems have not been effective and may not be capable of preventing further RCS or Reactor Vessel water level drop and potential core uncovery. The inability to restore and maintain level inventory within 30 minutes after reaching this condition in combination with a Containment challenge infers a failure of the RCS barrier, Loss of the Fuel Clad barrier and a Potential Loss of Containment.

This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.

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

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

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

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

During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged (Table C-2, Containment Challenge Indications).

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.

-Continued Below-Page 83 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 54 of 152 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.

Callaway Basis Reference(s):

1. OOA-BB-00003, Refuel Level Indications

2. Calculation No. BB-177, (387.1 - CAL RVLIS Setpoints)

3. OTN-BB-00002, Reactor Coolant System Draining

4. OSP-GT-00003, Containment Closure

5. FSAR, Section 6.2.5

6. FSAR, Table 7A-3 (Sheet 31)

7. Technical Specifications 3.3.3

8. OTN-GS-00001, Containment Hydrogen Control System

9. Calc No. 392.2 XX-95 Callaway Containment Parameters EOP Action Values, Setpoint ID T101 & T102

10. NEI 99-01, CS1 Page 84 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 55 of 152 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: Loss of RCS inventory affecting fuel clad integrity with containment challenged EAL:

CG1.2 General Emergency RCS level cannot be monitored for 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 crane radiation monitor SD-RE-41 > 10,000 mR/hr.

• Erratic Source Range Monitor indication.

AND Any Containment Challenge indication, Table C-2.

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

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

Table C-1 Sumps / Tanks

• Containment Sumps

• Containment Normal Sumps

• Containment Instrument Sump

• PRT

• RCDT

• Auxiliary Building Sump Table C-2 Containment Challenge Indications

• CONTAINMENT CLOSURE not established (Note 6)

• Containment hydrogen concentration 4%

• Unplanned rise in Containment pressure MODE Applicability:

5 - Cold Shutdown, 6 - Refueling Page 85 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 56 of 152 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 Callaway, Containment Closure is established when the requirements of OSP-GT-00003 Containment Closure are met.

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). The RCS is capable of being placed in an intact condition by Operator Action, i.e., pressurized to support natural circulation cooling.

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.

In this EAL, all RCS water level indication would be unavailable for greater than 30 minutes, and the RCS inventory loss must be detected by indirect leakage indications (Table C-1). Surveillance procedures provide instructions for calculating primary system leak rate by manual or computer-based water inventory balances.

Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage. If the make-up rate to the RCS unexplainably rises above the pre-established rate, a loss of RCS inventory may be occurring even if the source of the leakage cannot be immediately identified. 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).

The Plant Computer System Display called Refuel Level Indications (turn on code RLI) is available to assist in monitoring important parameters crucial to RCS draining operations (ref. 3).

The Reactor Vessel inventory loss may be detected by the manipulator 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 manipulator crane radiation monitor (SD-RE-41) indication (ref. 4, 5, 6).

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. 7, 8).

Page 86 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 57 of 152 Three conditions are associated with a challenge to Cotainment integrity:

1. CONTAINMENT CLOSURE not established - The status of Containment closure is tracked if plant conditions change that could raise the risk of a fission product release as a result of a loss of decay heat removal (ref. 15). If containment closure is re-established prior to exceeding the 30-minute core uncovery time limit, then escalation to GE would not occur.

2. Containment hydrogen 4% - The 4% hydrogen concentration threshold is generally considered the lower limit for hydrogen deflagrations. Callaway 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. 9). Two Containment hydrogen monitors (GS AI-10 and GS AI-19) with a range of 0% to 10% provide indication on Control Room Panel RL020 and ERFIS (ref. 10, 11). The hydrogen monitors require a 2-hour warmup period when starting from the OFF position and 15 minutes when starting from STANDBY (ref. 12, 13).

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 (ref. 15).

This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.

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

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

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

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

During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged. (Table C-2, Containment Challenge Indications).

-Continued Below-Page 87 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 58 of 152 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.

Callaway Basis Reference(s):

1. OTO-BB-00003-R014, Excess RCS Leakage

2. OSP-BB-00009, RCS Inventory Balance

3. OTN-BB-00002, Reactor Coolant System Draining

4. FSAR, Section 12.3.3.4

5. FSAR, Table 12.3-2

6. Calc. No. HPCI -0701, SD-RE-41 Response to Core Uncovery in Refueling MODE

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

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

9. FSAR, Section 6.2.5

10. FSAR, Table 7A-3 (Sheet 31)

11. Technical Specifications 3.3.3

12. OTN-GS-00001, Containment Hydrogen Control System

13. Calc No. 392.2 XX-95 Callaway Containment Parameters EOP Action Values, Setpoint ID T101 & T102

14. OSP-GT-00003, Containment Closure

15. NEI 99-01, CG1

16. OOA-BB-00003, Refuel Level Indications

17. TDB-001, Callaway Energy Center Tank Data Book Page 88 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 59 of 152 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 2 - Loss of Emergency AC Power Initiating Condition: Loss of all but one AC power source to 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 Coordinator should declare the event promptly upon determining that time limit has been exceeded or will likely be exceeded.

Table C-3 AC Power Sources Offsite:

• Safeguards XFMR A or B via ESF LTC XFMR XNB01

• Startup XFMR XMR01 via ESF LTC XFMR XNB02

• Main XFMR XMA01 backfed via UAT XFMR XMA02 (in-service)

• Alternate Emergency Power Supply (in-service or stand-by alignment)

Onsite:

• EDG NE01

• EDG NE02 MODE Applicability:

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

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

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

1. The integrity of the reactor coolant pressure boundary.

2. The capability to shut down the reactor and maintain it in a safe shutdown condition.

3. The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

-Continued Below-Page 89 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 60 of 152 Basis:

For emergency classification purposes, capability means that an offsite AC power source(s) is available to the emergency buses and can be aligned within 15 minutes, whether or not the buses are powered from it.

The criteria for Standby Alignment is that the source can be supplying the station with power within 15 minutes. Obviously, the Main Transformer could not be aligned for backfeed in 15-minutes during normal power operations. But, in an outage, and if already aligned for backfeed, the Main Transformer could be supplying power to the station within 15 minutes, and credit could be taken for it. The same applies for the Alternate Energy Power System (AEPS). Timed control room actions have shown that Callaway can supply power from AEPS to the station in approximately 9 minutes, if AEPS is aligned in standby. If AEPS cannot be aligned to a bus within 15 minutes, then it is not considered a capable AC power source.

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 13.8 KV safeguards transformer A or B via ESF Load Tap Changing (LTC) transformer XNB01 and the other source is from the startup transformer XMR01 via ESF LTC transformer XNB02. Transformer XNB01 is the normal supply to bus NB01; XNB02 is the normal supply to bus NB02 (ref. 1, 2, 3).

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. 1).

Another method to obtain offsite power is by backfeeding the emergency buses through the main transformer XMA01 and unit auxiliary transformer XMA02. This is only done during Cold Shutdown unless nuclear safety considerations require it to be done during hot shutdown when no other power sources are available (ref. 4).

An additional source of offsite power is the Alternate Emergency Power Supply (AEPS). AEPS consists of Co-op Power or AEPS Diesel Generators. Credit can be taken for this source only if it can be aligned within 15 minutes.

This cold condition EAL is equivalent to the hot condition EAL SA1.1.

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

When in the Cold Shutdown, refueling, or defueled MODE, this condition is not classified as an Alert because of the increased time available to restore another power source to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these MODES, this condition is considered to be a potential degradation of the level of safety of the plant.

-Continued Below-Page 90 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 61 of 152 An AC power source is a source recognized in AOPs and EOPs, and capable of supplying required power to an essential bus. Some examples of this condition are presented below.

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

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

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

If the offsite circuits for NB01 and/or NB02 are INOPERABLE due to low voltage, that does NOT mean the offsite sources are incapable of supplying AC power per EAL CU2.1. This EAL is concerned with if the required equipment would work when needed. Since there is clearly a difference between operability and functionality, the following applies:

Both NB buses have degraded voltage protection which will strip the bus, open the offsite source supply breaker, and start their respective diesel generator to power the bus. The degraded voltage protection on NB01 and NB02 is designed to protect the bus and motors powered by each bus from damage due to undervoltage. If the bus was stripped and reenergized by its diesel, at that point it would be considered a loss of AC power capability from the offsite source. Note, if AEPS were still available as an AC power supply, then the threshold for CU2.1 is NOT met.

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.

Callaway Basis Reference(s):

1. E-21001(Q) Main Single Line Diagram (Electrical Distribution Diagram)

2. FSAR Site Addenda, Section 8.2.1.2

3. FSAR, Section 8.3.1

4. OTS-MA-00001-R011, Main Step-Up Transformer Backfeed - IPTE

5. NEI 99-01, CU2 Page 91 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 62 of 152 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 2 - Loss of Emergency AC Power Initiating Condition: Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer EAL:

CA2.1 Alert Loss of all offsite and all onsite AC power capability, Table C-3, to emergency 4.16KV buses NB01 and NB02 for 15 min. (Note 1)

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

Table C-3 AC Power Sources Offsite:

• Safeguards XFMR A or B via ESF LTC XFMR XNB01

• Startup XFMR XMR01 via ESF LTC XFMR XNB02

• Main XFMR XMA01 backfed via UAT XFMR XMA02 (in-service)

• Alternate Emergency Power Supply (in-service or stand-by alignment)

Onsite:

• EDG NE01

• EDG NE02 MODE Applicability:

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

For emergency classification purposes, capability means that an offsite AC power source(s) is available to the emergency buses and is aligned within 15 minutes.

If AEPS is not aligned to a bus within 15 minutes, then it is not considered a capable AC power source.

The criteria for Standby Alignment is that the source can be supplying the station with power within 15 minutes.

Obviously, the Main Transformer could not be aligned for backfeed in 15 minutes during normal power operations.

But, in an outage, and if already aligned for backfeed, the Main Transformer could be supplying power to the station within 15 minutes, and credit could be taken for it. The same applies for AEPS. Timed control room actions have shown that Callaway can supply power from AEPS to the station in approximately 9 minutes, if AEPS is aligned in standby.

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).

-Continued Below-Page 92 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 63 of 152 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 13.8 KV safeguards transformer A or B via ESF Load Tap Changing (LTC) transformer XNB01 and the other source is from the startup transformer XMR01 via ESF LTC transformer XNB02. Transformer XNB01 is the normal supply to bus NB01; XNB02 is the normal supply to bus NB02 (ref. 1, 2, 3).

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. 1).

Another method to obtain offsite power is by backfeeding the emergency buses through the main transformer XMA01 and unit auxiliary transformer XMA02. This is only done during Cold Shutdown unless nuclear safety considerations require it to be done during hot shutdown when no other power sources are available (ref. 4).

An additional source of offsite power is the Alternate Emergency Power Supply (AEPS). AEPS consists of Co-op Power or AEPS diesel generators. Credit can be taken for this source only if it can be aligned within 15 minutes.

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

If the offsite circuits for NB01 and/or NB02 are INOPERABLE due to low voltage, that does NOT mean the offsite sources are incapable of supplying AC power per EAL CA2.1. This EAL is concerned with if the required equipment would work when needed. Since there is clearly a difference between operability and functionality, the following applies:

Both NB buses have degraded voltage protection which will strip the bus, open the offsite source supply breaker, and start their respective diesel generator to power the bus. The degraded voltage protection on NB01 and NB02 is designed to protect the bus and motors powered by each bus from damage due to undervoltage. If the bus was stripped and reenergized by its diesel, at that point it would be considered a loss of AC power capability from the offsite source. Note, if AEPS were still available as an AC power supply, then the threshold for CA2.1 is NOT met.

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

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

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

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

Callaway Basis Reference(s):

1. E-21001(Q) Main Single Line Diagram (Electrical Distribution Diagram)

2. FSAR Site Addenda, Section 8.2.1.2

3. FSAR, Section 8.3.1

4. OTS-MA-00001-R011, Main Step-Up Transformer Backfeed - IPTE

5. NEI 99-01, CA2 Page 93 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 64 of 152 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: UNPLANNED increase in RCS temperature EAL:

CU3.1 Unusual Event UNPLANNED increase in RCS temperature to > 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):

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). The RCS is capable of being placed in an intact condition by Operator Action, i.e., pressurized to support natural circulation cooling.

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 (e.g., MODE3, HEATU, COOLD, MODE4, ACCUM, and RHR). The most limiting temperature indication should be used. For example, the highest valid reading temperature indication should be used (ref. 2, 3, 4).

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 IC addresses an UNPLANNED increase in RCS temperature above the Technical Specification Cold Shutdown temperature limit and represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Coordinator should also refer to IC CA3.

-Continued Below-Page 94 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 65 of 152 A momentary UNPLANNED excursion above the Technical Specification Cold Shutdown temperature limit when the heat removal function is available DOES NOT warrant a classification.

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

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

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

Callaway Basis Reference(s):

1. Callaway Technical Specifications, Table 1.1-1

2. OTG-ZZ-00001, Plant Heatup Cold Shutdown to Hot Standby

3. OSP-BB-00007, RCS Heatup and Cooldown Limitations, Note before Section 6.1 and Attachment 2

4. FSAR, Section 7.2.2.3.2

5. NEI 99-01, CU3 Page 95 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 66 of 152 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: UNPLANNED increase in RCS temperature EAL:

CU3.2 Unusual Event Loss of all RCS temperature and RCS level indication for 15 min.

(Note 1)

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

MODE Applicability:

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

None Basis:

Reactor Vessel water level is normally monitored using the following instruments (ref. 2):

• RCS Loop level indications):

• Indicators on RL018:

BB LI-53A, RCS (LOOP 1) HOT LEG LEV BB LI-53B, RCS (LOOP 4) HOT LEG LEV

• Computer points:

BBL0053A, RCS LOOP 1 HOT LEG LEVEL BBL0053B, RCS LOOP 4 HOT LEG LEVEL BBL053BB, RCS LOOP LEVEL - CTMT VENTED

• RVLIS BB LI-1311, BB LI-1312, BB LI-1321, and BB LI-1322 (if in service) (ref. 3, 4)

• Visual observation (if vessel head is removed) (ref. 5)

The Plant Computer System Display called Refuel Level Indications (turn on code RLI) is available to assist in monitoring important parameters crucial to RCS draining operations (ref. 3).

-Continued Below-Page 96 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 67 of 152 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 (e.g., MODE3, HEATU, COOLD, MODE4, ACCUM, and RHR). The most limiting temperature indication should be used. For example, the highest valid reading temperature indication should be used (ref. 6, 7, 8).

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 Coordinator should also refer to IC CA3.

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

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

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

Callaway Basis Reference(s):

1. Callaway Technical Specifications Table 1.1-1

2. OOA-BB-00003, Refuel Level Indications

3. OTN-BB-00002, Reactor Coolant System Draining

4. FSAR, Section 18.2.13.2

5. OTS-KE-00018, Draining the Refueling Pool

6. OTG-ZZ-00001, Plant Heatup Cold Shutdown to Hot Standby

7. OSP-BB-00007, RCS Heatup and Cooldown Limitations, Note before Section 6.1 and Attachment 2

8. FSAR, Section 7.2.2.3.2

9. NEI 99-01, CU3 Page 97 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 68 of 152 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: Inability to maintain plant in Cold Shutdown EAL:

CA3.1 Alert UNPLANNED increase in RCS temperature to > 200°F for > Table C-4 duration. (Notes 1, 10)

OR UNPLANNED RCS pressure increase > 10 psig. (This EAL does not apply during water-solid plant conditions)

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

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

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

REDUCED INVENTORY)

RCS 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 Callaway, Containment Closure is established when the requirements of OSP-GT-00003 Containment Closure are met.

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). The RCS is capable of being placed in an intact condition by Operator Action, i.e., pressurized to support natural circulation cooling.

-Continued Below-Page 98 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 69 of 152 UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change, or event may be known or unknown.

REDUCED INVENTORY - Plant condition when fuel is in the reactor vessel and Reactor Coolant System level is lower than 3 feet below the Reactor Vessel flange (< 64.0 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. Plant computer screens are available for monitoring heatup and cooldown (e.g., MODE3, HEATU, COOLD, MODE4, ACCUM, and RHR). The most limiting temperature indication should be used. For example, the highest valid reading temperature indication should be used (ref. 2, 3, 4).

RCS pressure instrument BB PI-403A is capable of measuring pressure to less than 10 psig (ref. 5).

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

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). 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 RCS pressure increase threshold provides a pressure-based indication of RCS heat-up in the absence of RCS temperature monitoring capability.

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

Callaway Basis Reference(s):

1. Callaway Technical Specifications, Table 1.1-1

2. OTG-ZZ-00001, Plant Heatup Cold Shutdown to Hot Standby

3. OSP-BB-00007, RCS Heatup and Cooldown Limitations, Note before Section 6.1 and Attachment 2

4. FSAR, Section 7.2.2.3.2

5. OTG-ZZ-00006, Plant Cooldown Hot Standby To Cold Shutdown

6. NEI 99-01, CA3 Page 99 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 70 of 152 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 4 - Loss of Vital DC Power Initiating Condition: Loss of Vital DC power for 15 minutes or longer EAL:

CU4.1 Unusual Event

< 107 VDC bus voltage indications on Technical Specification required 125 VDC buses for 15 min.

(Note 1)

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

MODE Applicability:

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

None Basis:

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

As used in this EAL, required means the vital DC buses necessary to support operation of the in-service, or operable, train or trains of SAFETY SYSTEM equipment. For example, if Train A is out-of-service (inoperable) for scheduled outage maintenance work and Train B is in-service (operable), then a loss of Vital DC power affecting Train B (NK02 or NK04) would require the declaration of an Unusual Event. A loss of Vital DC power to Train A would not warrant an emergency classification.

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

Division 1: Division 2:

NK01 NK02 NK03 NK04 There are four battery banks (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.

-Continued Below-Page 100 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 71 of 152 Due to the load distribution on each of the 125VDC buses, the four batteries for each bus do not have the same rating. All four of the 125VDC buses supply inverters for 120VAC NN bus power as well as control power for various safety related systems. NK01 and NK04 supply additional DC loads such as diesel field flashing, breaker control power, main control board power and emergency lighting. These loads are not supplied by the other two buses, NK02 and NK03. For this reason, batteries NK11 and NK14 require additional capacity. 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 107.0 VDC (ref. 4, 5). Bus voltage may be obtained from the following instruments (ref. 6):

• NK EI-1 (NK01)

• NK EI-2 (NK02)

• NK EI-3 (NK03)

• NK EI-4 (NK04)

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.

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

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

Callaway Basis Reference(s):

1. E-21010(Q), DC Single Line Diagram

2. FSAR, Tables 8.3-1, -2, -3

3. FSAR, Section 8.3.2

4. Calculation NK-10, NK System DC Voltage Drop

5. FSAR, Table 8.3A-1 III.B

6. ECA-0.0, Loss of All AC Power

7. NEI 99-01, CU4 Page 101 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 72 of 152 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 5 - Loss of Communications Initiating Condition: Loss of all onsite or offsite communications capabilities EAL:

CU5.1 Unusual Event Loss of all Table C-5 onsite communication methods.

OR Loss of all Table C-5 ORO communication methods.

OR Loss of all Table C-5 NRC communication methods.

Table C-5 Communication Methods System Onsite ORO NRC Gaitronics X Plant Radios X Plant Emergency Dedicated Phones X Plant Telephone System X X X ENS (Red Phone) Line X X Back-Up Radio System X Event Notification System X MODE Applicability:

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

OFFSITE RESPONSE ORGANIZATIONS (ORO) - The State of Missouri (SEMA/MIAC), Callaway County 911/EOC, Gasconade County 911/EOC, Montgomery County 911/EOC and Osage County 911/EOC.

-Continued Below-Page 102 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 73 of 152 Basis:

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

1. Gaitronics system The Gaitronics system provides six separate independent communication channels--one general page, one Control Room page and four party lines. Communication between parties within the plant can be easily and quickly established by using the 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. The party line channel is normally used after the page call is completed. As many as four party lines may communicate simultaneously. The portion of the PA system connecting the fuel transfer area in the Containment, the spent fuel area and new fuel handling area in the fuel building, and the control room can be isolated from the remainder of the PA system from the control room.

This permits extended use of the fuel handling communications system without disruption to the remainder of the system.

2. Plant Radios A six channel 800 MHZ trunked radio system for overall plant site area coverage reaches out as far as the intake structure. This two-way radio system provides communications for operating purposes with plant radio-equipped vehicles and plant hand-held portable radios. These systems are for use during normal operation or during a plant emergency. This radio system is available on the Control Room radio consoles, on the security radio consoles, on the EOF radio console, and the TSC radio console. This system is also in the field monitoring team vehicles and is used to communicate during emergencies.

3. Plant Emergency Dedicated Phones Three independent telephone systems are available for communications between the Emergency Response Facilities: the Technical Assessment Bridge Line, the Dose Assessment Bridge Line and the Emergency Management Bridge Line. Each system operates independently from the other systems and allows for conference calls between the members of that bridge line group.

4. Plant telephone system The telephone system consists of digital automatic switchboard (DPBX) equipment and telephone stations. The DPBX is provided with redundant processors for reliability. The telephone stations are located throughout the power block, in the main control room, in the various buildings around the site, in the security building, and in the service building where the administrative offices are located.

For emergency use, unlisted telephone numbers are provided for direct access to the outside local public telephone system. Company provided cell phones ARE considered part of the Plant Telephone System. The FLEX response satellite phones are in place for beyond design basis accidents and ARE NOT considered part of the Plant Telephone System.

-Continued Below-Page 103 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 74 of 152

5. ENS (Red Phone) line The NRC Emergency Notification System (ENS) is an 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.

6. Back-Up Radio System (BURS)

The Back-up Radio System is a communication link between the Callaway Plant and offsite emergency response agencies. The primary use of this system is the back-up notification of offsite agencies and the coordination of offsite activities during a radiologica1 emergency. The system uses 800 MHz radios. There are radio control base: units in the Plant Control Room, TSC and EOF, as well as each county EOC and the State EOC. The backup to this system is the commercial touchtone telephone system Notifications may also be initiated through the Callaway County/City of Fulton EOC via the Security radio.

7. Event Notification System A computerized notification system linked between the Callaway Plant, the State Emergency Management Agency and the four (4) EPZ risk counties. It allows the Communicator to fill out a notification form on screen and transmit the data simultaneously. Notifications on the Event Notification System can be initiated from the Control Room, the Emergency Operations Facility (EOF), or the Technical Support Center (TSC).

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

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

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

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

The second EAL condition addresses a total loss of the communications methods used to notify all OROs of an emergency declaration.

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

Callaway Basis Reference(s):

1. Callaway Plant Radiological Emergency Response Plan (RERP), Section 7.2

2. FSAR, Section 9.5.2

3. NEI 99-01, CU5 Page 104 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 75 of 152 Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 6 - Hazardous Event Affecting Safety Systems Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating MODE.

EAL:

CA6.1 Alert The occurrence of any Table C-6 hazardous event AND Event damage has caused indication of degraded performance on one train of a SAFETY SYSTEM needed for the current operating MODE.

AND EITHER:

• Event damage has caused indications of degraded performance in a second train of a SAFETY SYSTEM needed for the current operating MODE.

• Event damage has resulted in VISIBLE DAMAGE to a second train of a SAFETY SYSTEM needed for the current operating MODE.

(Notes 11, 12)

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

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

Table C-6 Hazardous Events

• EXPLOSION

• FIRE

• HIGH WINDS or tornado strike

• Internal or external FLOODING event

• Seismic event (earthquake)

• Other events with similar hazard characteristics as determined by the Emergency Coordinator MODE Applicability:

5 - Cold Shutdown, 6 - Refueling

-Continued Below-Page 105 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 76 of 152 Definition(s):

EXPLOSION - A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or over pressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events require a post-event inspection to determine if the attributes of an explosion are present.

FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.

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

HIGH WINDS - Winds in excess of 40 mph (18 m/s) sustained, or 58 mph (26 m/s) gusting.

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

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

1. The integrity of the reactor coolant pressure boundary;

2. The capability to shut down the reactor and maintain it in a safe shutdown condition;

3. The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

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

Basis:

• Annunciator 98D, OBE will illuminate if the seismic instrument detects ground motion in excess of the OBE threshold. OTO-SG-00001, Seismic Event provides the guidance for determining if an OBE earthquake threshold is exceeded and any required response actions (ref. 1).

• Internal FLOODING may be caused by events such as component failures, equipment misalignment, or outage activity mishaps (ref. 2).

• External flooding may be due to high rainfall. Callaway plant grade elevation is 840.0 ft. MSL.

(ref. 3).

• Seismic Category I structures are analyzed to withstand a sustained, design wind velocity of at least 100 mph. (ref. 4).

• Areas containing functions and systems required for safe shutdown of the plant are identified by fire area (ref. 5).

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

-Continued Below Page 106 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 77 of 152 This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the current operating mode. In order to provide the appropriate context for consideration of an ALERT classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train, and there must be either indications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance, and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE such that the potential exists for performance issues. Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has indications of degraded performance; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events.

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

Operators will make a determination of VISIBLE DAMAGE based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.

An event affecting equipment common to two or more safety systems or safety system trains (i.e., there are indications of degraded performance and/or VISIBLE DAMAGE affecting the common equipment) should be classified as an Alert. By affecting the operability or reliability of multiple system trains, the loss of the common equipment effectively meets the two-train impact criteria that underlines the EALs and Bases. For example, a unit with a tank that is the water source for multiple safety injection systems or trains, such as a Refueling Water Storage Tank.

An event that affects two trains of safety system (e.g., one train has indications of degraded performance and the other VISIBLE DAMAGE) that also has one or more additional trains should be classified as an Alert.

This approach maintains consistency with the two-train impact criteria that underlines the EALs and Bases and is warranted because the event was severe enough to affect the operability or reliability of two trains of a safety system despite plant design criteria associated with system and system train separation and protection.

Such an event may have caused other plant impacts that are not immediately apparent. For example, a unit that has an Auxiliary/Emergency Feedwater system with three trains.

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

Callaway Basis Reference(s):

1. OTO-SG-00001, Seismic Event

2. IPE Section 3.4.2.3 Results of the Vulnerability Screening

3. FSAR, Section 3.4 Water Level (Flood) Design Table 3.4-1 PMF, Groundwater, Reference, and Actual Plant Elevations

4. FSAR, Section 3.3.1.1 Design Wind Loadings

5. FSAR, Section 9.5.1 Fire Protection System

6. NEI 99-01, CA6 Page 107 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 78 of 152 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.)

NOTE Hazards are non-plant, system related events that can directly or indirectly affect plant operation, reactor plant safety or personnel safety.

1. Security Unauthorized entry attempts into the Protected Area, bomb threats, sabotage attempts, and actual security compromises threatening loss of physical control of the plant.

2. Seismic Event Natural events such as earthquakes have potential to cause plant structure or equipment damage of sufficient magnitude to threaten personnel or plant safety.

3. Natural or 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 Non-naturally occurring events that can cause damage to plant facilities and include toxic, corrosive, asphyxiant or flammable gas leaks.

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.

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

Page 108 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 79 of 152 Category: H - Hazards Subcategory: 1 - Security Initiating Condition: Confirmed SECURITY CONDITION or threat EAL:

HU1.1 Unusual Event A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the Security Shift Supervisor.

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 Callaway 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 Callaway.

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 Supervisor.

This EAL is based on the Callaway Plant Security Plan and DBT (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.

-Continued Below-Page 109 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 80 of 152 Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event (ref. 2, 3, 4). Classification of these events will initiate appropriate threat-related notifications to plant personnel and Offsite Response Organizations.

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

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

The second threshold addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with the Callaway Plant Security Plan and DBT.

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 Callaway Plant Security Plan and DBT (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 Callaway Plant Security Plan and DBT (ref. 1).

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

Callaway Basis Reference(s):

1. Callaway Plant Security Plan and DBT (Safeguards)

2. EIP-ZZ-SK001, Response to Security Threat

3. SDP-CP-00003, Security Contingency Events

4. OTO-SK-00002, Plant Security Event - Aircraft Threat

5. NEI 99-01, HU1 Page 110 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 81 of 152 Category: H - Hazards Subcategory: 1 - Security Initiating Condition: HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes EAL:

HA1.1 Alert A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the Security Shift Supervisor.

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 Callaway 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 Callaway.

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 Supervisor.

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

Timely and accurate communications between the Security Shift Supervisor and the Control Room is essential for proper classification of a security-related event (ref. 2, 3, 4).

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

-Continued Below-Page 111 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 82 of 152 As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Alert declaration will also heighten the awareness of Offsite Response Organizations (OROs), allowing them to be better prepared should it be necessary to consider further actions.

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

The first threshold is applicable for any HOSTILE ACTION occurring, or that has occurred, in the OWNER CONTROLLED AREA.

The second threshold addresses the threat from the impact of an aircraft on the plant, and the anticipated arrival time is within 30 minutes. The intent of this EAL is to ensure that threat-related notifications are made in a timely manner so that plant personnel and OROs are in a heightened state of readiness. This EAL is met when the threat-related information has been validated in accordance with site-specific security procedures.

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

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

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

Callaway Basis Reference(s):

1. Callaway Plant Security Plan and DBT (Safeguards)

2. EIP-ZZ-SK001, Response to Security Threat

3. SDP-CP-00003, Security Contingency Events

4. OTO-SK-00002, Plant Security Event - Aircraft Threat

5. NEI 99-01, HA1 Page 112 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 83 of 152 Category: H - Hazards Subcategory: 1 - Security Initiating Condition: HOSTILE ACTION within the PROTECTED AREA EAL:

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

MODE Applicability:

All Definition(s):

HOSTILE ACTION - An act toward Callaway 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 Callaway.

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 Drawing 8600-X-88100 Property-Site Layout, Owner Controlled Area and Surrounding Area. The lower level of the Main Access Facility (MAF), including the Central Alarm Station (CAS) is considered inside the Protected Area.

Basis:

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

These individuals are the designated on-site personnel qualified and trained to confirm that a security event is occurring or has occurred. Training on security event classification confirmation is closely controlled due to the strict secrecy controls placed on the Callaway Plant Security Plan and DBT (Safeguards) information.

(ref. 1)

This IC addresses the occurrence of a HOSTILE ACTION within the PROTECTED AREA. This event will require rapid response and assistance due to the possibility for damage to plant equipment.

Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event (ref. 2, 3, 4 5).

-Continued Below Page 113 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 84 of 152 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 Site Area Emergency declaration will mobilize Offsite Response Organization (ORO) resources and have them available to develop and implement public protective actions in the unlikely event that the attack is successful in impairing multiple safety functions.

This IC does not apply to 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 Callaway Plant Security Plan and DBT (ref. 1).

Escalation of the emergency classification level would be via ICs RG1, RG2, SG1, SG2, FG1, and CG1.

Callaway Basis Reference(s):

1. Callaway Plant Security Plan and DBT (Safeguards)

2. EIP-ZZ-SK001, Response to Security Threat

3. SDP-CP-00003, Security Contingency Events

4. OTO-SK-00001, Plant Security Event - Hostile Intrusion

5. OTO-SK-00002, Plant Security Event - Aircraft Threat

6. NEI 99-01, HS1 Page 114 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 85 of 152 Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 2 - Seismic Event Initiating Condition: Seismic event greater than OBE level EAL:

HU2.1 Unusual Event Seismic event > OBE as indicated by Seismic Activity, Annunciator 98D.

MODE Applicability:

All Definition(s):

None Basis:

Annunciator 98D, OBE will illuminate if the seismic instrument detects ground motion in excess of the OBE threshold (ref. 4).

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

Earthquakes of this magnitude should be readily felt by on-site personnel and recognized as a seismic event.

Shift Manager or Emergency Coordinator may seek external verification if deemed appropriate (e.g., a call to the USGS, check internet news sources, etc.); however, the verification action must not preclude a timely emergency declaration.

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

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

Additional actions after EAL declaration:

When the seismic recorder indicates that the OBE has been exceeded, as verified by ETP-SG-00001, the reactor must be shut down and remain shut down until inspection of the facility shows that no damage has been incurred which would jeopardize safe operation of the facility or until such damage is repaired.

Callaway was designed such that, for ground motion less than the OBE, those features of the plant necessary for continued operation without undue risk to the health and safety of the public will remain functional. Any ground motion in excess of this results in an uncertainty as to the extent of the damage which must be resolved before continued operation can be considered safe (ref. 1). Ground motion of this magnitude is unmistakably a felt earthquake.

-Continued Below-Page 115 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 86 of 152 OTO-SG-00001, Seismic Event provides the guidance for determining if the OBE earthquake threshold is exceeded and any required response actions. (ref. 2)

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

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

Callaway Basis Reference(s):

1. FSAR Section 3.7(B).1.1 Design Response Spectra

2. OTO-SG-00001, Seismic Event

3. NEI 99-01, HU2

4. FSAR Table 16.3-3, Seismic Monitoring Instrumentation Page 116 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 87 of 152 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 Drawing 8600-X-88100 Property-Site Layout, Owner Controlled Area and Surrounding Area. The lower level of the Main Access Facility (MAF), including the Central Alarm Station (CAS) is considered inside the Protected Area.

Basis:

Response actions associated with a tornado onsite is provided in OTO-ZZ-00012 Severe Weather (ref. 1).

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

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

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

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

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

Callaway Basis Reference(s):

1. OTO-ZZ-00012, Severe Weather

2. NEI 99-01, HU3 Page 117 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 88 of 152 Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technology Hazard Initiating Condition: Hazardous event EAL:

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

MODE Applicability:

All Definition(s):

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

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

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

1. The integrity of the reactor coolant pressure boundary;

2. The capability to shut down the reactor and maintain it in a safe shutdown condition;

3. The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

Basis:

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

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

This EAL addresses FLOODING of a building room or area that results in operators isolating power to a SAFETY SYSTEM component due to water level or other wetting concerns. Classification is also required if the water level or related wetting causes an automatic isolation of a SAFETY SYSTEM component from its power source (e.g., a breaker or relay trip). To warrant classification, operability of the affected component must be required by Technical Specifications for the current operating MODE.

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

Callaway Basis Reference(s):

1. IPE Section 3.4.2.3 Results of the Vulnerability Screening

2. NEI 99-01, HU3 Page 118 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 89 of 152 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 Drawing 8600-X-88100 Property-Site Layout, Owner Controlled Area and Surrounding Area. The lower level of the Main Access Facility (MAF), including the Central Alarm Station (CAS) is considered inside the Protected Area.

Basis:

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

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

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

The process of flushing chemicals to the sump at the Circ and Service Water building is in a normal plant operations area, but since it is a planned maintenance activity it is excluded provided the process was controlled. That means looking at extra ventilation and barrier tape to control access. Then as long as the process by which the gasses are being generated is controlled (not expanding beyond the boundary set), is short in duration, and we are able to monitor the atmosphere at the boundary we should be outside the EAL threshold. If we lose control of the process and as a result, we had to evacuate part of the Protected Area, then that could meet the IC for HU3.3.

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

Callaway Basis Reference(s):

1. NEI 99-01, HU3 Page 119 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 90 of 152 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.

If ALL access roads to the plant are impassable, and local authorities are no longer clearing roads, this EAL applies.

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

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

Calhoun Station in 2011.

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

Callaway Basis Reference(s):

1. NEI 99-01, HU3 Page 120 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 91 of 152 Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:

HU4.1 Unusual Event A FIRE is not extinguished within 15 min. of any of the following FIRE detection indications:

(Note 1)

• Report from the field (i.e., visual observation).

• Receipt of multiple (more than 1) fire alarms or indications.

• Field verification of a single fire alarm.

AND The FIRE is located within any Table H-1 area.

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

Table H-1 Fire Areas

• Area 5

• Auxiliary Building

• Aux Feedwater Pump Rooms

• Containment

• Control Building/Communications Corridor

• Diesel Generator Building

• ESW Pumphouse

• Fuel Building

• RWST

• UHS Cooling Tower MODE Applicability:

All Definition(s):

FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.

-Continued Below-Page 121 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 92 of 152 Basis:

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

Actual field reports must be made within the 15-minute time limit, or a classification must be made. If a fire is verified to be occurring by field report, the 15-minute time limit is from the original receipt of the fire detection alarm.

Table H-1 Fire Areas are based on FSAR Section 5.4A.2 System Required to Go From Hot Standby to Cold Shutdown. Table H-1 Fire Areas include those structures containing functions and systems required for safe shutdown of the plant (SAFETY SYSTEMS) (ref. 1). The Laundry Decon Facility is NOT part of the Aux Building.

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

For EAL HU4.1 the intent of the 15-minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., smoldering wastepaper 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.

The Shift Manager needs to ask some specific questions to ensure they have the information needed to evaluate the situation.

• Is there visible flame?

• Are there copious quantities of smoke still being generated?

A smoked component (subject to overheating) should show blackened areas or signs that the component itself had been very hot (e.g., paint peeling). It can be expected to generate some lower level of smoke. If there is so much smoke present that entry to inspect the component is not possible without an SCBA, that would probably be an indication that a fire existed and determine if EAL HA5.1 is applicable (MODE 4 only). If a breaker truly suffered a fault local to the breaker, the damage and fire ball would be such that consideration of the Hazardous Event EAL SA9.1 would be recommended, if a required Safety System was affected.

In the case of a fire alarm in Containment, OTA-KC-01008 states that at the discretion of the Shift Manager/Operating Supervisor, either:

• INSPECT detectors for operation AND INSPECT the Reactor Building for the presence of smoke/fire, OR

• INSPECT other containments parameters available in the Control Room, such as other detection zones, containment temperature or equipment failure, for evidence of a fire.

-Continued Below-Page 122 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 93 of 152 Other items to monitor would be Containment Radiation Monitors such as GTRE0031 and GTRE0032 for loss of flow due to filters plugging. The important thing is to make the initial declaration timely with respect to the time of the initial indication. In all cases, document the indications considered for the decision made.

If indications of failing safety related equipment are attributable to the fire, consider Hazardous Event EAL SA9.1 Upon receipt, operators will take prompt actions to confirm the validity of an initial fire alarm, indication, or report. For EAL assessment purposes, the emergency declaration clock starts at the time that the initial alarm, indication, or report was received, and not the time that a subsequent verification action was performed. Similarly, the fire duration clock also starts at the time of receipt of the initial alarm, indication or report.

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

Callaway Basis Reference(s):

1. FSAR, Section 5.4A.2 System Required to Go From Hot Standby to Cold Shutdown

2. NEI 99-01, HU4 Page 123 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 94 of 152 Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:

HU4.2 Unusual Event Receipt of a single fire alarm (i.e., no other indications of a FIRE).

AND The fire alarm is indicating a FIRE within any Table H-1 area.

AND The existence of a FIRE is not verified within 30 min. of alarm receipt.

(Note 1)

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

Table H-1 Fire Areas

• Area 5

• Auxiliary Building

• Aux Feedwater Pump Rooms

• Containment

• Control Building/Communications Corridor

• Diesel Generator Building

• ESW Pumphouse

• Fuel Building

• RWST

• UHS Cooling Tower MODE Applicability:

All Definition(s):

FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.

-Continued Below-Page 124 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 95 of 152 Basis:

The 30-minute requirement begins upon receipt of a single valid fire detection system alarm. The alarm is to be validated using available Control Room indications or alarms to prove that it is not spurious, or by reports from the field. Actual field reports must be made within the 30-minute time limit or a classification must be made. If a fire is verified to be occurring by field report, classification shall be made based on EAL HU4.1.

Table H-1 Fire Areas are based on FSAR Section 5.4A.2 System Required to Go From Hot Standby to Cold Shutdown. Table H-1 Fire Areas include those structures containing functions and systems required for safe shutdown of the plant (SAFETY SYSTEMS) (ref. 1). The Laundry Decon Facility is NOT part of the Aux Building.

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

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

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

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

The Shift Manager needs to ask some specific questions to ensure they have the information needed to evaluate the situation.

• Is there visible flame?

• Are there copious quantities of smoke still being generated?

A smoked component (subject to overheating) should show blackened areas or signs that the component itself had been very hot (e.g., paint peeling). It can be expected to generate some lower level of smoke. If there is so much smoke present that entry to inspect the component is not possible without an SCBA, that would probably be an indication that a fire existed and determine if EAL HA5.1 is applicable (MODE 4 only). If a breaker truly suffered a fault local to the breaker, the damage and fire ball would be such that consideration of the Hazardous Event EAL SA9.1 would be recommended, if a required Safety System was affected.

-Continued Below-Page 125 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 96 of 152 In the case of a fire alarm in Containment, OTA-KC-01008 states that at the discretion of the Shift Manager/Operating Supervisor, either:

• INSPECT detectors for operation AND INSPECT the Reactor Building for the presence of smoke/fire, OR

• INSPECT other containments parameters available in the Control Room, such as other detection zones, containment temperature or equipment failure, for evidence of a fire.

Other items to monitor would be Containment Radiation Monitors such as GTRE0031 and GTRE0032 for loss of flow due to filters plugging. The important thing is to make the initial declaration timely with respect to the time of the initial indication. In all cases, document the indications considered for the decision made.

If indications of failing safety related equipment are attributable to the fire, consider Hazardous Event EAL SA9.1 Basis-Related Requirements from 10 CFR 50 10 CFR 50 Appendix A, Criterion 3 states in part:

"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."

10 CFR 50.48 Fire Protection states under (2) (iii) "The means to limit fire damage to structures, systems, or components important to safety so that the capability to shut down the pant safely is ensured.

NFPA 805 Section 1.3.1 states "The Nuclear Safety Goal is to provide reasonable assurance that a fire during any plant operational mode and plant configuration will not prevent the plant from achieving and maintaining the fuel in a safe and stable condition."

In addition, NFPA 805 Section 4.2.3.3, 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. 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.

Callaway Basis Reference(s):

1. FSAR, Section 5.4A.2 System Required to Go From Hot Standby to Cold Shutdown

2. NEI 99-01, HU4 Page 126 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 97 of 152 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 Coordinator should declare the event promptly upon determining that time limit has been exceeded or will likely be exceeded.

MODE Applicability:

All Definition(s):

FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.

PROTECTED AREA - An area encompassed by physical barriers and to which access is controlled. The Protected Area refers to the designated security area around the process buildings and is depicted in Drawing 8600-X-88100 Property-Site Layout, Owner Controlled Area and Surrounding Area. The lower level of the Main Access Facility (MAF), including the Central Alarm Station (CAS) is considered inside the Protected Area.

Basis:

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

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

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

Callaway Basis Reference(s):

1. NEI 99-01, HU4 Page 127 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 98 of 152 Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:

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

MODE Applicability:

All Definition(s):

FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.

PROTECTED AREA - An area encompassed by physical barriers and to which access is controlled. The Protected Area refers to the designated security area around the process buildings and is depicted in Drawing 8600-X-88100 Property-Site Layout, Owner Controlled Area and Surrounding Area. The lower level of the Main Access Facility (MAF), including the Central Alarm Station (CAS) is considered inside the Protected Area.

Basis:

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

If a FIRE within the plant PROTECTED AREA is of sufficient size to require a response by an offsite firefighting agency (e.g., a local town Fire Department), then the level of plant safety is potentially degraded.

The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish. Declaration is NOT necessary if the agency resources are placed on stand-by or supporting post-extinguishment recovery or investigation actions.

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

Callaway Basis Reference(s):

1. NEI 99-01, HU4 Page 128 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 99 of 152 Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 5 - Hazardous Gases Initiating Condition: Gaseous release IMPEDING access to equipment necessary for normal plant operations, cooldown or shutdown EAL:

HA5.1 Alert Release of a toxic, corrosive, asphyxiant or flammable gas that prohibits or IMPEDES access to EITHER of the following: (Note 5)

• North Electrical Penetration Room. (Room 1410)

• South Electrical Penetration Room. (Room 1409)

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.

MODE Applicability:

4 - Hot Shutdown Definition(s):

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

Basis:

The only rooms/areas external to the Control Room that require access to perform field actions consistent with the above criteria for Callaway are the North and South Electrical Penetration Rooms when in MODE 4 to support isolating SI accumulators and placing RHR in service for RCS cooldown to Cold Shutdown (ref.

1, 2, 3). The equipment required is:

For SI Accumulators:

• NG01BGF3, FDR BKR TO EPHV8808A SI ACC A OUT ISO, (Room 1410)

• NG02BGF3, FDR BKR TO EPHV8808B SI ACC B OUT ISO, (Room 1409)

• NG01BGF2, FDR BKR TO EPHV8808C SI ACC C OUT ISO, (Room 1410)

• NG02BHF2, FDR BKR TO EPHV8808D SI ACC D OUT ISO, (Room 1409)

For "A" RHR:

• NG02BCF2, FDR BKR TO BBPV8702A RCS LOOP 1 HOT LEG TO RHR PMPS ISO, (Room 1409)

• NG01BEF2, FDR BKR TO EJHV8701A A RHR PMP SUCT FROM RCS HOT LEG 1 ISO, (Room 1410)

For "B" RHR:

• NG02BBF3, FDR BKR TO BBPV8702B RCS LOOP 4 HOT LEG TO RHR PMPS ISO, (Room 1409)

• NG01BDF3, FDR BKR TO EJHV8701B B RHR PMP SUCT FROM RCS HOT LEG 4 ISO, (Room 1410)

-Continued Below-Page 129 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 100 of 152 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 Coordinators judgment that the gas concentration in the affected room/area is sufficient to preclude or significantly impede procedurally required access. This judgment may be based on a variety of factors including an existing job hazard analysis, report of ill effects on personnel, advice from a subject matter expert or operating experience with the same or similar hazards. Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g.,

requiring use of protective equipment, such as SCBAs, that is not routinely employed).

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

• The plant is NOT in MODE 4.

• The gas release is a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., fire suppression system testing).

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

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

An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels. Most commonly, asphyxiants work by merely displacing air in an enclosed environment. This reduces the concentration of oxygen below the normal level of around 19%, which can lead to breathing difficulties, unconsciousness or even death.

This EAL DOES NOT apply to firefighting activities that automatically or manually activate a fire suppression system in an area.

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

Callaway Basis Reference(s):

1. OTG-ZZ-00006 Addendum 06, Securing Safety Injection Accumulators

2. OTN-EJ-00001 Addendum 3, Placing A RHR Train In Service For RCS Cooldown

3. OTN-EJ-00001 Addendum 4, Placing B RHR Train In Service For RCS Cooldown

4. NEI 99-01, AA3 Page 130 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 101 of 152 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 uninhabitable and requires evacuation. Control Room inhabitability may be caused by fire, dense smoke, noxious fumes, bomb threat in or adjacent to the Control Room, or other life-threatening conditions. OTO-ZZ-00001 Control Room Inaccessibility provides the instructions for tripping the unit and maintaining RCS inventory and Hot Shutdown conditions from outside the Control Room (Ref. 1).

For the purpose of this EAL, the 15-minute clock starts after determination that Control Room evacuation is necessary, not when OTO-ZZ-00001 Control Room Inaccessibility, is entered.

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.

Callaway Basis Reference(s):

1. OTO-ZZ-00001, Control Room Inaccessibility

2. NEI 99-01, HA6 Page 131 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 102 of 152 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 (MODE 1, 2, and 3 only).

• Core Cooling.

• RCS heat removal.

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

MODE Applicability:

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, to re-establish control of key safety functions, starts when the last licensed operator leaves the Control Room.

Key Safety Functions, Core Cooling and Heat Sink, are evaluated under this EAL when plant control has been transferred to the ASP and NOT under the Fission Product Barrier Matrix.

The Shift Manager (SM) determines if the Control Room is uninhabitable and requires evacuation. Control Room inhabitability may be caused by fire, dense smoke, noxious fumes, bomb threat in or adjacent to the Control Room, or other life-threatening conditions. OTO-ZZ-00001, Control Room Inaccessibility, provides the instructions for tripping the unit, and maintaining RCS inventory and Hot Shutdown conditions from outside the Control Room (Ref. 1, 2).

-Continued Below-Page 132 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 103 of 152 The intent of this EAL is to capture events in which control of the plant cannot be reestablished in a timely manner. The time interval is based on how quickly control must be reestablished without core uncovery and/or core damage. The determination of whether or not control is established from outside the Control Room is based on Emergency Coordinator judgment. The Emergency Coordinator is expected to make a reasonable, informed judgment that control of the plant from outside the Control Room cannot be established within the fifteen-minute interval.

Once the Control Room is evacuated, the objective is to establish control of important plant equipment and maintain knowledge of important plant parameters in a timely manner. Primary emphasis should be placed on components and instruments that supply protection for and information about safety functions. Typically, these safety functions are reactivity control (ability to shut down 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 Coordinator judgment. The Emergency Coordinator is expected to make a reasonable, informed judgment within 15 minutes whether or not the operating staff has control of key safety functions from the remote safe shutdown location(s).

Escalation of the emergency classification level would be via IC FG1 or CG1.

Callaway Basis Reference(s):

1. OTO-ZZ-00001, Control Room Inaccessibility

2. OTS-ZZ-00001, Cooldown from Outside the Control Room

3. NEI 99-01, HS6 Page 133 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 104 of 152 Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Coordinator Judgment Initiating Condition: Other conditions existing that in the judgment of the Emergency Coordinator warrant declaration of a UE.

EAL:

HU7.1 Unusual Event Other conditions exist which in the judgment of the Emergency Coordinator indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of SAFETY SYSTEMS occurs.

MODE Applicability:

All Definition(s):

None Basis:

The Emergency Coordinator is the designated onsite individual having the responsibility and authority for implementing the Callaway Radiological Emergency Response Plan (ref. 1). The Shift Manager (SM) initially acts in the capacity of the Emergency Coordinator 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 Coordinator, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency.

This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Coordinator to fall under the emergency classification level description for an Unusual Event.

Callaway Basis Reference(s):

1. Callaway Radiological Emergency Response Plan, Section 5.2.1, Emergency Coordinator

2. Callaway Radiological Emergency Response Plan, Section 5.1.1, Shift Manager

3. NEI 99-01, HU7 Page 134 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 105 of 152 Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Coordinator Judgment Initiating Condition: Other conditions exist that in the judgment of the Emergency Coordinator warrant declaration of an Alert.

EAL:

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

MODE Applicability:

All Definition(s):

HOSTILE ACTION - An act toward Callaway 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 Callaway.

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

Basis:

The Emergency Coordinator is the designated onsite individual having the responsibility and authority for implementing the Callaway Radiological Emergency Response Plan (ref. 1). The Shift Manager (SM) initially acts in the capacity of the Emergency Coordinator 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 Coordinator, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency.

This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Coordinator to fall under the emergency classification level description for an Alert.

Callaway Basis Reference(s):

1. Callaway Radiological Emergency Response Plan, section 5.2.1 Emergency Coordinator

2. Callaway Radiological Emergency Response Plan, section 5.1.1 Shift Manager

3. NEI 99-01, HA7 Page 135 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 106 of 152 Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Coordinator Judgment Initiating Condition: Other conditions existing that in the judgment of the Emergency Coordinator warrant declaration of a Site Area Emergency.

EAL:

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

MODE Applicability:

All Definition(s):

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

Basis:

The Emergency Coordinator is the designated onsite individual having the responsibility and authority for implementing the Callaway Radiological Emergency Response Plan (ref. 1). The Shift Manager (SM) initially acts in the capacity of the Emergency Coordinator 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 Coordinator, emergency response personnel are notified and instructed to report to their emergency response locations.

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

This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Coordinator to fall under the emergency classification level description for a Site Area Emergency.

Callaway Basis Reference(s):

1. Callaway Radiological Emergency Response Plan, Section 5.2.1 Emergency Coordinator

2. Callaway Radiological Emergency Response Plan, Section 5.1.1 Shift Manager

3. NEI 99-01, HS7 Page 136 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 107 of 152 Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Coordinator Judgment Initiating Condition: Other conditions exist which in the judgment of the Emergency Coordinator warrant declaration of a General Emergency.

EAL:

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

MODE Applicability:

All Definition(s):

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

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

Basis:

The Emergency Coordinator is the designated onsite individual having the responsibility and authority for implementing the Callaway Radiological Emergency Response Plan (ref. 1). The Shift Manager (SM) initially acts in the capacity of the Emergency Coordinator 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 Coordinator, emergency response personnel are notified and instructed to report to their emergency response locations.

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

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

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

Callaway Basis Reference(s):

1. Callaway Radiological Emergency Response Plan, Section 5.2.1 Emergency Coordinator

2. Callaway Radiological Emergency Response Plan, Section 5.1.1 Shift Manager

3. NEI 99-01, HG7 Page 137 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 108 of 152 Category S - System Malfunction EAL Group: Hot Conditions (RCS temperature > 200ºF); EALs in this category are applicable only in one or more hot operating MODES.

NOTE 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 VDC power sources.

3. Loss of Control Room Indications Certain events that degrade plant operator ability to effectively assess plant conditions within the plant warrant emergency classification. Losses of indicators are in this subcategory.

4. RCS Activity During normal operation, reactor coolant fission product activity is very low. Small concentrations of fission products in the coolant are primarily from the fission of tramp uranium in the fuel clad or minor perforations in the clad itself. Any significant increase from these base-line levels (2% - 5%

clad failures) is indicative of fuel failures and is covered under the Fission Product Barrier Degradation category. However, lesser amounts of clad damage may result in coolant activity exceeding Technical Specification limits. These fission products will be circulated with the reactor coolant and can be detected by coolant sampling.

5. RCS Leakage The reactor vessel provides a volume for the coolant that covers the reactor core. The reactor pressure vessel and associated pressure piping (reactor coolant system) together provide a barrier to limit the release of radioactive material should the reactor fuel clad integrity fail. Excessive RCS leakage greater than Technical Specification limits indicates potential pipe cracks that may propagate to an extent threatening fuel clad, RCS and containment integrity.

-Continued Below-Page 138 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 109 of 152

6. RTS Failure This subcategory includes events related to failure of the Reactor Trip System (RTS) to initiate and complete reactor trips. In the plant licensing basis, postulated failures of the RTS to complete a reactor trip comprise a specific set of analyzed events referred to as Anticipated Transient Without Scram (ATWS) events. For EAL classification, however, ATWS is intended to mean any trip failure event that does not achieve reactor shutdown. If RTS actuation fails to assure reactor shutdown, positive control of reactivity is at risk and could cause a threat to fuel clad, RCS and containment integrity.

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

8. Containment Failure Failure of containment isolation capability (under conditions in which the containment is not currently challenged) warrants emergency classification. Failure of containment pressure control capability also warrants emergency classification.

9. Hazardous Event Affecting Safety Systems Various natural and technological events that result in degraded plant safety system performance or significant VISIBLE DAMAGE warrant emergency classification under this subcategory.

Page 139 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 110 of 152 Category: S - System Malfunction Subcategory: 1 - Loss of Emergency AC Power Initiating Condition: Loss of all offsite AC power capability to emergency buses for 15 minutes or longer EAL:

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

(Note 1)

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

Table S-1 AC Power Sources Offsite:

• Safeguards XFMR A or B via ESF LTC XFMR XNB01

• Startup XFMR XMR01 via ESF LTC XFMR XNB02

• Main XFMR XMA01 backfed via UAT XFMR XMA02 (in-service)

• Alternate Emergency Power Supply (in-service or stand-by alignment)

Onsite:

• EDG NE01

• EDG NE02 MODE Applicability:

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

None Basis:

For emergency classification purposes, capability means that one of the Table S-1 offsite sources remains available and can be aligned within 15 minutes.

The criteria here is that the source can be supplying the station with power within 15 minutes. Obviously, the Main Transformer could not be aligned for backfeed in 15 minutes during normal power operations. But, in an outage, and if already aligned for backfeed, the Main Transformer could be supplying power to the station within 15 minutes, and credit could be taken for it. The same applies for AEPS. Timed control room actions have shown that Callaway can supply power from AEPS to the station in approximately 9 minutes, if AEPS is aligned in standby. If AEPS cannot be aligned to a bus within 15 minutes, then it is not considered a capable AC power source.

-Continued Below-Page 140 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 111 of 152 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).

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 13.8 KV safeguards transformer A or B via ESF Load Tap Changing (LTC) transformer XNB01 and the other source is from the startup transformer XMR01 via ESF LTC transformer XNB02. Transformer XNB01 is the normal supply to bus NB01; XNB02 is the normal supply to bus NB02 (ref. 1, 2 3).

Another method to obtain offsite power is by backfeeding the emergency buses through the main transformer XMA01 and unit auxiliary transformer XMA02. However, this is only done during Cold Shutdown unless nuclear safety considerations require it to be done during hot shutdown when no other power sources are available (ref. 4).

An additional source of offsite power is the Alternate Emergency Power Supply (AEPS). AEPS consists of Co-op Power or AEPS diesel generators. Credit can be taken for this source only if it can be aligned within 15 minutes.

If the offsite circuits for NB01 and/or NB02 are INOPERABLE due to low voltage, that does NOT mean the offsite sources are incapable of supplying AC power per EAL SU1.1. This EAL is concerned with if the required equipment would work when needed. Since there is clearly a difference between operability and functionality, the following applies:

Both NB buses have degraded voltage protection which will strip the bus, open the offsite source supply breaker, and start their respective diesel generator to power the bus. The degraded voltage protection on NB01 and NB02 is designed to protect the bus and motors powered by each bus from damage due to undervoltage. If the bus was stripped and reenergized by its diesel, at that point it would be considered a loss of AC power capability from the offsite source. Note, if AEPS were still available as an AC power supply, then the threshold for SU1.1 is NOT met.

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.

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 SA1.

Callaway Basis Reference(s):

1. E-21001(Q) Main Single Line Diagram (Electrical Distribution Diagram)

2. FSAR Site Addenda Section 8.2.1.2

3. FSAR, Section 8.3.1

4. OTS-MA-00001-R011, Main Step-Up Transformer Backfeed - IPTE

5. NEI 99-01, SU1 Page 141 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 112 of 152 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 Coordinator should declare the event promptly upon determining that time limit has been exceeded or will likely be exceeded.

Table S-1 AC Power Sources Offsite:

• Safeguards XFMR A or B via ESF LTC XFMR XNB01

• Startup XFMR XMR01 via ESF LTC XFMR XNB02

• Main XFMR XMA01 backfed via UAT XFMR XMA02 (in-service)

• Alternate Emergency Power Supply (in-service or stand-by alignment)

Onsite:

• EDG NE01

• EDG NE02 MODE Applicability:

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

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

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

1. The integrity of the reactor coolant pressure boundary;

2. The capability to shut down the reactor and maintain it in a safe shutdown condition;

3. The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

-Continued Below-Page 142 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 113 of 152 Basis:

For emergency classification purposes, capability means that one of the Table S-1 offsite sources remains available and can be aligned within 15 minutes.

If AEPS cannot be aligned to a bus within 15 minutes, then it is not considered a capable AC power source.

The criteria here is that the source can be supplying the station with power within 15 minutes. Obviously, the Main Transformer could not be aligned for backfeed in 15 minutes during normal power operations. But, in an outage, and if already aligned for backfeed, the Main Transformer could be supplying power to the station within 15 minutes, and credit could be taken for it. The same applies for AEPS. Timed control room actions have shown that Callaway can supply power from AEPS to the station in approximately 9 minutes, if AEPS is aligned in standby.

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).

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 13.8 KV safeguards transformer A or B via ESF Load Tap Changing (LTC) transformer XNB01 and the other source is from the startup transformer XMR01 via ESF LTC transformer XNB02. Transformer XNB01 is the normal supply to bus NB01; XNB02 is the normal supply to bus NB02 (ref. 1, 2 3).

Another method to obtain offsite power is by backfeeding the emergency buses through the main transformer XMA01 and unit auxiliary transformer XMA02. However, this is only done during Cold Shutdown unless nuclear safety considerations require it to be done during hot shutdown when no other power sources are available (ref. 4).

An additional source of offsite power is the Alternate Emergency Power Supply (AEPS). AEPS consists of Co-op Power or AEPS diesel generators. Credit can be taken for this source only if it can be aligned within 15 minutes.

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

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

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

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

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

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

-Continued Below-Page 143 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 114 of 152 If the offsite circuits for NB01 and/or NB02 are INOPERABLE due to low voltage, that does NOT mean the offsite sources are incapable of supplying AC power per EAL SA1.1. This EAL is concerned with if the required equipment would work when needed. Since there is clearly a difference between operability and functionality, the following applies:

Both NB buses have degraded voltage protection which will strip the bus, open the offsite source supply breaker, and start their respective diesel generator to power the bus. The degraded voltage protection on NB01 and NB02 is designed to protect the bus and motors powered by each bus from damage due to undervoltage. If the bus was stripped and reenergized by its diesel, at that point it would be considered a loss of AC power capability from the offsite source. Note, if AEPS were still available as an AC power supply, then the threshold for SA1.1 is NOT met.

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.

Callaway Basis Reference(s):

1. E-21001(Q) Main Single Line Diagram (Electrical Distribution Diagram)

2. FSAR, Site Addenda Section 8.2.1.2

3. FSAR, Section 8.3.1

4. OTS-MA-00001-R011, Main Step-Up Transformer Backfeed - IPTE

5. NEI 99-01, SA1 Page 144 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 115 of 152 Category: S - System Malfunction Subcategory: 1 - Loss of Emergency AC Power Initiating Condition: Loss of all offsite power and all onsite AC power to emergency buses for 15 minutes or longer.

EAL:

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

(Note 1)

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

MODE Applicability:

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

None Basis:

For this emergency classification this means NB01 and NB02 are deenergized for greater than or equal to 15 minutes.

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).

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 13.8 KV safeguards transformer A or B via ESF Load Tap Changing (LTC) transformer XNB01 and the other source is from the startup transformer XMR01 via ESF LTC transformer XNB02. Transformer XNB01 is the normal supply to bus NB01; XNB02 is the normal supply to bus NB02 (ref. 1, 2 3).

In addition, NB01 and NB02 each have an emergency diesel generator (onsite power supply) which supply electrical power to the bus automatically in the event that the preferred source becomes unavailable (ref. 1).

Another method to obtain offsite power is by backfeeding the emergency buses through the main transformer XMA01 and unit auxiliary transformer XMA02. However, this is only done during Cold Shutdown unless nuclear safety considerations require it to be done during hot shutdown when no other power sources are available (ref. 4).

-Continued Below-Page 145 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 116 of 152 Additional sources of offsite power are available from diesel generators such as the Alternate Emergency Power System (AEPS) or portable generation sources. AEPS consists of Co-op Power or AEPS diesel generators. Credit can be taken for these sources only if they are capable of carrying an NB bus and are aligned within 15 minutes. (ref. 5).

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

If the offsite circuits for NB01 and/or NB02 are INOPERABLE due to low voltage, that does NOT mean the offsite sources are incapable of supplying AC power per EAL SS1.1. This EAL is concerned with if the required equipment would work when needed. Since there is clearly a difference between operability and functionality, the following applies:

Both NB buses have degraded voltage protection which will strip the bus, open the offsite source supply breaker, and start their respective diesel generator to power the bus. The degraded voltage protection on NB01 and NB02 is designed to protect the bus and motors powered by each bus from damage due to undervoltage. If the bus was stripped and reenergized by its diesel, at that point it would be considered a loss of AC power capability from the offsite source. Note, if AEPS were still available as an AC power supply, then the threshold for SS1.1 is NOT met.

This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. In addition, fission product barrier monitoring capabilities may be degraded under these conditions. This IC represents a condition that involves actual or likely major failures of plant functions needed for the protection of the public.

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

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

Callaway Basis Reference(s):

1. E-21001(Q) Main Single Line Diagram (Electrical Distribution Diagram)

2. FSAR, Site Addenda Section 8.2.1.2

3. FSAR, Section 8.3.1

4. OTS-MA-00001-R011, Main Step-Up Transformer Backfeed - IPTE

5. ECA-0.0, Loss of All AC Power

6. NEI 99-01, SS1 Page 146 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 117 of 152 Category: S - System Malfunction Subcategory: 1 - Loss of Emergency AC Power Initiating Condition: Prolonged loss of all offsite and all onsite AC power to emergency buses EAL:

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

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

• CSFST Core Cooling RED Path conditions met.

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

MODE Applicability:

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

None Basis:

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 than the Callaway Station Blackout (SBO) coping analysis time (4 hrs.) (ref. 1, 2) 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. 3).

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).

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 13.8 KV safeguards transformer A or B via ESF Load Tap Changing (LTC) transformer XNB01 and the other source is from the startup transformer XMR01 via ESF LTC transformer XNB02. Transformer XNB01 is the normal supply to bus NB01; XNB02 is the normal supply to bus NB02 (ref. 4, 5 6).

In addition, NB01 and NB02 each have an emergency diesel generator (onsite power supply) which supply electrical power to the bus automatically in the event that the preferred source becomes unavailable (ref. 4).

Another method to obtain offsite power is by backfeeding the emergency buses through the main transformer XMA01 and unit auxiliary transformer XMA02. However, this is only done during Cold Shutdown unless nuclear safety considerations require it to be done during hot shutdown when no other power sources are available (ref. 7).

Additional sources of offsite power are available from diesel generators such as the Alternate Emergency Power System (AEPS) or portable generation sources. AEPS consists of Co-op Power or AEPS diesel generators. Credit can be taken for these sources only if they are capable of carrying an NB bus and it can be aligned within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. (ref. 8).

Four hours is the station blackout coping time (ref 1, 2).

-Continued Below Page 147 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 118 of 152 Indication of continuing core cooling degradation must be based on fission product barrier monitoring with particular emphasis on Emergency Coordinator judgment as it relates to imminent Loss of fission product barriers and degraded ability to monitor fission product barriers. Indication of continuing core cooling degradation is manifested by CSFST Core Cooling RED PATH conditions being met (ref. 3).

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.

If the offsite circuits for NB01 and/or NB02 are INOPERABLE due to low voltage, that does NOT mean the offsite sources are incapable of supplying AC power per EAL SG1.1. This EAL is concerned with if the required equipment would work when needed. Since there is clearly a difference between operability and functionality, the following applies:

Both NB buses have degraded voltage protection which will strip the bus, open the offsite source supply breaker, and start their respective diesel generator to power the bus. The degraded voltage protection on NB01 and NB02 is designed to protect the bus and motors powered by each bus from damage due to undervoltage. If the bus was stripped and reenergized by its diesel, at that point it would be considered a loss of AC power capability from the offsite source. Note, if AEPS were still available as an AC power supply, then the threshold for SG1.1 is NOT met.

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.

Callaway Basis Reference(s):

1. FSAR, Section 8.3A.5

2. BO-01, Station Blackout (SBO) Coping Duration, sh 1

3. CSF-1, Critical Safety Function Status Trees (CSFST) Figure 2, Core Cooling

4. E-21001(Q) Main Single Line Diagram (Electrical Distribution Diagram)

5. FSAR, Site Addenda Section 8.2.1.2 6 FSAR, Section 8.3.1

7. OTS-MA-00001-R011, Main Step-Up Transformer Backfeed - IPTE

8. ECA-0.0, Loss of All AC Power

9. NEI 99-01, SG1 Page 148 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 119 of 152 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 to emergency 4.16KV buses NB01 and NB02 for 15 min.

AND Loss of all 125 VDC power based on battery bus voltage indications < 107 VDC on all vital DC buses NK01, NK03 (Division 1) and NK02, NK04 (Division 2) for 15 min (Note 1)

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

MODE Applicability:

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

None Basis:

For this emergency classification this means NB01 and NB02 are deenergized for greater than or equal to 15 minutes.

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.

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).

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 13.8 KV safeguards transformer A or B via ESF Load Tap Changing (LTC) transformer XNB01 and the other source is from the startup transformer XMR01 via ESF LTC transformer XNB02. Transformer XNB01 is the normal supply to bus NB01; XNB02 is the normal supply to bus NB02 (ref. 1, 2 3).

In addition, NB01 and NB02 each have an emergency diesel generator (onsite power supply) which supply electrical power to the bus automatically in the event that the preferred source becomes unavailable (ref. 1).

Another method to obtain offsite power is by backfeeding the emergency buses through the main transformer XMA01 and unit auxiliary transformer XMA02. However, this is only done during Cold Shutdown unless nuclear safety considerations require it to be done during hot shutdown when no other power sources are available (ref. 4).

Additional sources of offsite power are available from diesel generators such as the Alternate Emergency Power System (AEPS) or portable generation sources. AEPS consists of Co-op Power or AEPS diesel generators. Credit can be taken for these sources only if they are capable of carrying an NB bus and it can be aligned within 15 minutes. (ref.

5).

-Continued Below Page 149 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 120 of 152 The vital DC buses are the following 125 VDC Class 1E buses (ref. 6):

Division 1: Division 2:

NK01 NK02 NK03 NK04 There are four battery banks, (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.

Due to the load distribution on each of the 125VDC buses, the four battery banks for each bus do not have the same rating. All four of the 125VDC buses supply inverters for 120VAC NN bus power as well as control power for various safety related systems. NK01 and NK04 supply additional DC loads such as diesel field flashing, breaker control power, main control board power and emergency lighting. These loads are not supplied by the other two buses, NK02 and NK03. For this reason, batteries NK11 and NK14 require additional capacity. 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. 8, 9, 10).

Minimum DC bus voltage is 107.0 VDC (ref. 9, 10). Bus voltage may be obtained from the following instruments (ref. 6):

• NK EI-1 (NK01)

• NK EI-2 (NK02)

• NK EI-3 (NK03)

• NK EI-4 (NK04)

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.

Callaway Basis Reference(s):

1. E-21001(Q) Main Single Line Diagram (Electrical Distribution Diagram)

2. FSAR, Site Addenda Section 8.2.1.2

3. FSAR, Section 8.3.1

4. OTS-MA-00001-R011, Main Step-Up Transformer Backfeed - IPTE

5. ECA-0.0, Loss of All AC Power

6. E-21010(Q) DC Single Line Diagram

7. FSAR, Tables 8.3-1, -2, -3

8. FSAR, Section 8.3.2

9. Calculation NK-10, NK System DC Voltage Drop

10. FSAR, Table 8.3A-1 III.B

11. NEI 99-01, SG8 Page 150 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 121 of 152 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 < 107 VDC on all vital DC buses NK01, NK03 (Division 1) and NK02, NK04 (Division 2) for 15 min.

(Note 1)

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

MODE Applicability:

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

None Basis:

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

Division 1: Division 2:

NK01 NK02 NK03 NK04 There are four battery banks, (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.

Due to the load distribution on each of the 125VDC buses, the four battery banks for each bus do not have the same rating. All four of the 125VDC buses supply inverters for 120VAC NN bus power as well as control power for various safety related systems. NK01 and NK04 supply additional DC loads such as diesel field flashing, breaker control power, main control board power and emergency lighting. These loads are not supplied by the other two buses, NK02 and NK03. For this reason, batteries NK11 and NK14 require additional capacity. 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).

-Continued Below Page 151 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 122 of 152 Minimum DC bus voltage is 107.0 VDC (ref. 4, 5). Bus voltage may be obtained from the following instruments (ref. 6):

• NK EI-1 (NK01)

• NK EI-2 (NK02)

• NK EI-3 (NK03)

• NK EI-4 (NK04)

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.

Callaway Basis Reference(s):

1. E-21010(Q) DC Single Line Diagram

2. FSAR, Tables 8.3-1, -2, -3

3. FSAR, Section 8.3.2

4. Calculation NK-10, NK System DC Voltage Drop

5. FSAR, Table 8.3A-1 III.B

6. ECA-0.0, Loss of All AC Power

7. NEI 99-01, SS8 Page 152 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 123 of 152 Category: S - System Malfunction Subcategory: 3 - Loss of Control Room Indications Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer.

EAL:

SU3.1 Unusual Event An UNPLANNED event results in the inability to monitor one or more Table S-2 parameters from within the Control Room for 15 min.

(Note 1)

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

Table S-2 Safety System Parameters

• Reactor power

• RCS level

• RCS pressure

• Core Exit T/C temperature

• Level in at least one S/G

• Auxiliary or emergency feedwater flow in at least one S/G MODE Applicability:

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

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

Basis:

SAFETY SYSTEM parameters listed in Table S-2 are monitored in the Control Room through a combination of hard control panel indicators as well as computer-based information systems. The Plant Computer, which displays SPDS required information, serves as a redundant compensatory indicator which may be utilized in lieu of normal Control Room indicators (ref. 1, 2).

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.

-Continued Below Page 153 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 124 of 152 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, or any loss of monitoring capabilities of feedwater flow to ALL Steam Generators.

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 the plant computer, the availability of other parameter values may be compromised as well.

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

Escalation of the emergency classification level would be via IC SA3.

Callaway Basis Reference(s):

1. FSAR, Section 7.5 Safety-Related Display Instrumentation

2. OTO-RJ-00001, Loss of Plant Computer

3. NEI 99-01, SU2 Page 154 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 125 of 152 Category: S - System Malfunction Subcategory: 3 - Loss of Control Room Indications Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress.

EAL:

SA3.1 Alert An UNPLANNED event results in the inability to monitor one or more Table S-2 parameters from within the Control Room for 15 min. (Note 1)

AND Any significant transient is in progress, Table S-3.

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

Table S-2 Safety System Parameters

• Reactor power

• RCS level

• RCS pressure

• Core Exit T/C temperature

• Level in at least one S/G

• Auxiliary or emergency feedwater flow in at least one S/G Table S-3 Significant Transients (Automatically or manually initiated)

• Reactor trip

• Runback 25% thermal power

• Electrical load rejection > 25% electrical load

• ECCS actuation MODE Applicability:

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

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

-Continued Below-Page 155 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 126 of 152 Basis:

SAFETY SYSTEM parameters listed in Table S-2 are monitored in the Control Room through a combination of hard control panel indicators as well as computer-based information systems. The Plant Computer, which displays SPDS required information, serves as a redundant compensatory indicator which may be utilized in lieu of normal Control Room indicators (ref. 1, 2).

Significant transients are listed in Table S-3 and include response to automatic or manually initiated functions such as reactor trips, runbacks involving greater than or equal to 25% thermal power change, electrical load rejections of greater than 25% full electrical load or ECCS (SI) injection actuations.

This IC addresses the difficulty associated with monitoring rapidly changing plant conditions during a transient without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. During this condition, the margin to a potential fission product barrier challenge is reduced. It thus represents a potential substantial degradation in the level of safety of the plant.

As used in this EAL, an inability to monitor means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room.

An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.

This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for reactor vessel level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well.

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

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

Callaway Basis Reference(s):

1. FSAR, Section 7.5 Safety-Related Display Instrumentation

2. OTO-RJ-00001, Loss of Plant Computer

3. NEI 99-01 SA2 Page 156 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 127 of 152 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 3.4.16 limits (listed below):

• > 60 µCi/gm Dose Equivalent I-131.

OR

• > 1.0 µCi/gm Dose Equivalent I-131 for a > 48 hr. continuous period.

OR

• > 225 µCi/gm Dose Equivalent Xe-133 for a > 48 hr. continuous period.

MODE Applicability:

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

None Basis:

This EAL should be entered when the Shutdown Action Statement for Tech Spec 3.4.16 is applied. These values are:

• > 60 µCi/gm Dose Equivalent I-131.

OR

• > 1.0 µCi/gm Dose Equivalent I-131 for a > 48 hr. continuous period.

OR

• > 225 µCi/gm Dose Equivalent Xe-133 for a > 48 hr. continuous period.

(ref 1, 2)

This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant.

Escalation of the emergency classification level would be via ICs FA1 or the Recognition Category R ICs.

Callaway Basis Reference(s):

1. Callaway Technical Specifications 3.4.16 RCS Specific Activity

2. OTO-BB-00005, High Coolant Activity

3. NEI 99-01, SU3 Page 157 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 128 of 152 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 for 15 min.

(Note 1)

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

MODE Applicability:

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

None Basis:

Manual or computer-based methods of performing an RCS inventory balance are normally used to determine RCS leakage. The Personal Computer (PC) is preferred method of calculating RCS leak rate. When the PC is used, plant status information and all calculations are generated by the Plant Process Computer. When the PC software is not available, procedural guidance is available to perform the manual RCS inventory balance (ref. 1).

Identified leakage includes:

• Leakage such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water 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. 2).

Unidentified leakage is all leakage (except RCP seal water leakoff) that is not identified leakage (ref. 2).

-Continued Below-Page 158 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 129 of 152 Pressure Boundary leakage is leakage (except SG tube leakage) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall (ref. 2)

RCS leakage outside of the containment that is not considered identified or unidentified leakage per Technical Specifications includes leakage via interfacing systems such as RCS to the Component Cooling Water, or systems that directly see RCS pressure outside containment such as Chemical & Volume Control System, Nuclear Sampling system and Residual Heat Removal system (when in the shutdown cooling MODE) (ref. 3, 4)

Escalation of this EAL to the Alert level is via Category F, Fission Product Barrier Degradation, EAL FA1.1.

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.

The first and second EAL conditions are focused on a loss of mass from the RCS due to unidentified leakage", "pressure boundary leakage" or "identified leakage (as these leakage types are defined in the plant Technical Specifications). The third condition addresses an RCS mass loss caused by an UNISOLABLE leak through an interfacing system. These conditions thus apply to leakage into the containment, a secondary-side system (e.g., steam generator tube leakage) or a location outside of containment.

The leak rate values for each condition were selected because they are usually observable with normal Control Room indications. Lesser values typically require time-consuming calculations to determine (e.g., a mass balance calculation). The first condition uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage.

The release of mass from the 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. If the leak is isolated, the RCS barrier was never lost.

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

Callaway Basis Reference(s):

1. OSP-BB-00009, RCS Inventory Balance

2. Callaway Technical Specifications, Definitions Section 1.1

3. FSAR, Section 5.2.5.2.1 Intersystem Leakage

4. OTO-BB-00003-R014, Excess RCS Leakage

5. NEI 99-01, SU4 Page 159 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 130 of 152 Category: S - System Malfunction Subcategory: 6 - RTS Failure Initiating Condition: Automatic or manual trip fails to shut down the reactor.

EAL:

SU6.1 Unusual Event An automatic trip did not shut down the reactor as indicated by reactor power 5% after any RTS setpoint is exceeded.

AND A subsequent automatic trip or manual trip action taken at the reactor control consoles (SB-HS-1 or SB-HS-42) is successful in shutting down the reactor as indicated by reactor power < 5%.

(Note 8)

Note 8: A manual trip action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods 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).

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, such as opening PG19 and PG20 supply breakers, emergency boration or manually driving control rods, do not constitute a successful manual trip (ref. 4). A successful manual turbine trip that subsequently automatically trips the reactor does constitute a successful trip.

-Continued Below Page 160 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 131 of 152 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).

Following any automatic RTS trip signal, E-0 (ref. 2) and FR-S.1 (ref. 4) prescribe insertion of redundant manual trip signals to back up the automatic RTS trip function and ensure reactor shutdown is achieved.

Even if the first subsequent manual trip signal inserts all control rods to the full-in position immediately after the initial failure of the automatic trip, the lowest level of classification that must be declared is an Unusual Event (ref. 4).

The ATWS Mitigation System Actuation Circuitry (AMSAC) logic will automatically initiate 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 or manual reactor trip that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic trip is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.

Following the failure on an automatic reactor trip, operators will promptly initiate manual actions at the reactor control consoles to shut down 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 plants decay heat removal systems.

-Continued Below-Page 161 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 132 of 152 If an initial manual reactor trip is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shut down 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 plants decay heat removal systems.

A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip). This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be at the reactor control consoles.

The plant response to the failure of an automatic or manual reactor trip will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC SA6. Depending upon the plant response, escalation is also possible via IC FA1. Absent the plant conditions needed to meet either IC SA6 or FA1, an Unusual Event declaration is appropriate for this event.

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 shut down 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.

Note 8 is a generic note applicable to the EAL's as approved by the NRC.

Callaway Basis Reference(s):

1. Callaway Technical Specifications, Section 3.3.1 Reactor Trip System (RTS) Instrumentation

2. E-0, Reactor Trip or Safety Injection

3. F-0, Critical Safety Function Status Trees - Subcriticality

4. FR-S.1, Response to Nuclear Power Generation/ATWS

5. FSAR, Section 7.7.1 6 NEI 99-01, SU5 Page 162 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 133 of 152 Category: S - System Malfunction Subcategory: 6 - RTS Failure Initiating Condition: Automatic or manual trip fails to shut down the reactor.

EAL:

SU6.2 Unusual Event A manual trip did not shut down the reactor as indicated by reactor power 5% after any manual trip action was initiated.

AND A subsequent automatic trip or manual trip action taken at the reactor control 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 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).

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, such as opening PG19 and PG20 supply breakers, emergency boration or manually driving control rods, do not constitute a successful manual trip (ref. 4). A successful manual turbine trip that subsequently automatically trips the reactor does constitute a successful trip.

-Continued Below-Page 163 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 134 of 152 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).

Following the failure of any manual trip signal, E-0 (ref. 2) and FR-S.1 (ref. 4) prescribe insertion of redundant manual trip signals to back up the RTS trip function and ensure reactor shutdown is achieved.

Even if a subsequent automatic trip signal or the first subsequent manual trip signal inserts all control rods to the full-in position immediately after the initial failure of the manual trip, the lowest level of classification that must be declared is an Unusual Event (ref. 4).

The ATWS Mitigation System Actuation Circuitry (AMSAC) logic will automatically initiate 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 an automatic or manual reactor trip that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic trip is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.

Following the failure on an automatic reactor trip, operators will promptly initiate manual actions at the reactor control consoles to shut down 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 plants decay heat removal systems.

If an initial manual reactor trip is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shut down 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 plants 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.

-Continued Below-Page 164 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 135 of 152 The plant response to the failure of an automatic or manual reactor trip will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC SA6. Depending upon the plant response, escalation is also possible via IC 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 RPS fails to automatically shut down 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.

Note 8 is a generic note applicable to the EAL's as approved by the NRC.

Callaway Basis Reference(s):

1. Callaway Technical Specifications, Section 3.3.1 Reactor Trip System (RTS) Instrumentation

2. E-0, Reactor Trip or Safety Injection

3. F-0, Critical Safety Function Status Trees - Subcriticality

4. FR-S.1, Response to Nuclear Power Generation/ATWS

5. FSAR, Section 7.7.1

6. NEI 99-01, SU5 Page 165 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 136 of 152 Category: S - System Malfunction Subcategory: 2 - RTS Failure Initiating Condition: Automatic or manual trip fails to shut down the reactor and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor.

EAL:

SA6.1 Alert An automatic or manual trip fails to shut down the reactor as indicated by reactor power 5%.

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 or implementation of boron injection strategies.

MODE Applicability:

1 - Power Operation Definition(s):

None Basis:

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 has escalated to this EAL.

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, such as opening PG19 and PG20 supply breakers, emergency boration or manually driving control rods, do not constitute a successful manual trip (ref. 4). A successful manual turbine trip that subsequently automatically trips the reactor does constitute a successful trip.

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).

The ATWS Mitigation System Actuation Circuitry (AMSAC) logic will automatically initiate auxiliary feedwater and a turbine trip under conditions indicative of an Anticipated Transient Without Scram (ATWS) event (ref. 5).

-Continued Below-Page 166 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 137 of 152 5% rated power is a minimum reading on the power range scale that indicates continued power production.

It also approximates the decay heat which the shutdown systems were designed to remove and is indicative of a condition requiring immediate response to prevent subsequent core damage. Below 5%, plant response will be similar to that observed during a normal shutdown. Nuclear instrumentation can be used to determine if reactor power is greater than 5 % power (ref. 3, 4).

Escalation of this event to a Site Area Emergency would be under EAL SS6.1 or Emergency Coordinator judgment.

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and subsequent operator manual actions taken at the reactor control consoles to shut down the reactor are also unsuccessful. This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if the reactor is subsequently shutdown by an action taken away from the reactor control consoles since this event entails a significant failure of the RTS.

A manual action at the reactor control console is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip). This action does not include manually driving in control rods or implementation of boron injection strategies. If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control console (e.g., locally opening breakers). Actions taken at 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 console.

The plant response to the failure of an automatic or manual reactor trip will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If the failure to shut down the reactor is prolonged enough to cause a challenge to the core cooling or 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.

Note 8 is a generic note applicable to the EAL's as approved by the NRC.

Callaway Basis Reference(s):

1. Callaway Technical Specifications, Section 3.3.1 Reactor Trip System (RTS) Instrumentation

2. E-0, Reactor Trip or Safety Injection

3. F-0, Critical Safety Function Status Trees - Subcriticality

4. FR-S.1, Response to Nuclear Power Generation/ATWS

5. FSAR, Section 7.7.1

6. NEI 99-01, SA5 Page 167 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 138 of 152 Category: S - System Malfunction Subcategory: 2 - RTS Failure Initiating Condition: Inability to shut down the reactor causing a challenge to core cooling or RCS heat removal.

EAL:

SS6.1 Site Area Emergency An automatic or manual trip fails to shut down the reactor as indicated by reactor power 5%.

AND All actions to shut down the reactor are not successful as indicated by reactor power 5%.

AND EITHER:

• CSFST Core Cooling RED Path conditions met.

• CSFST Heat Sink RED Path conditions met AND Heat Sink is required.

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 opening PG19 and PG20 supply breakers, 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, 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 normal shutdown. Nuclear instrumentation can be used to determine if reactor power is greater than 5 % power (ref. 1, 5).

Indication of continuing core cooling degradation is manifested by CSFST Core Cooling RED PATH conditions being met (ref. 2).

-Continued Below-Page 168 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 139 of 152 Indication of inability to adequately remove heat from the RCS is manifested by CSFST Heat Sink RED PATH conditions being met (ref. 3).

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, FR-H.1 is entered from CSFST Heat Sink Red. Step 1 tells the operator to determine if heat sink is required by checking that RCS pressure is greater than any non-faulted SG pressure and either RCS temperature is greater than 350ºF or RCS pressure is greater than 360 psig. 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. 4).

This IC addresses a failure of the RPS 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 may be higher than that resulting from an assessment of the plant responses and symptoms against the Recognition Category F ICs. This is appropriate in that the Recognition Category F ICs do not address the additional threat posed by a failure to shut down the reactor. The inclusion of this IC ensures the timely declaration of a Site Area Emergency in response to prolonged failure to shut down 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.

Callaway Basis Reference(s):

1. CSF-1, Critical Safety Function Status Trees - Figure 1 Subcriticality

2. CSF-1, Critical Safety Function Status Trees - Figure 2 Core Cooling

3. CSF-1, Critical Safety Function Status Trees - Figure 3 Heat Sink

4. FR-H.1, Response to Loss of Secondary Heat Sink

5. FR-S.1, Response to Nuclear Power Generation/ATWS

6. NEI 99-01, SS5 Page 169 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 140 of 152 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 ORO communication methods.

OR Loss of all Table S-4 NRC communication methods.

Table S-4 Communication Methods System Onsite ORO NRC Gaitronics X Plant Radios X Plant Emergency Dedicated Phones X Plant Telephone System X X X ENS (Red Phone) Line X X Back-Up Radio System X Event Notification System X MODE Applicability:

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

OFFSITE RESPONSE ORGANIZATIONS (ORO) - The State of Missouri (SEMA/MIAC), Callaway County 911/EOC, Gasconade County 911/EOC, Montgomery County 911/EOC and Osage County 911/EOC.

-Continued Below-Page 170 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 141 of 152 Basis:

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

1. Gaitronics system The Gaitronics system provides six separate independent communication channels--one general page, one Control Room page and four party lines. Communication between parties within the plant can be easily and quickly established by using the 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. The party line channel is normally used after the page call is completed. As many as four party lines may communicate simultaneously. The portion of the PA system connecting the fuel transfer area in the Containment, the spent fuel area and new fuel handling area in the fuel building, and the control room can be isolated from the remainder of the PA system from the control room.

This permits extended use of the fuel handling communications system without disruption to the remainder of the system.

2. Plant Radios A six channel 800 MHZ trunked radio system for overall plant site area coverage reaches out as far as the intake structure. This two-way radio system provides communications for operating purposes with plant radio-equipped vehicles and plant hand-held portable radios. These systems are for use during normal operation or during a plant emergency. This radio system is available on the Control Room radio consoles, on the security radio consoles, on the EOF radio console, and the TSC radio console. This system is also in the field monitoring team vehicles and is used to communicate during emergencies.

3. Plant Emergency Dedicated Phones Three independent telephone systems are available for communications between the Emergency Response Facilities: the Technical Assessment Bridge Line, the Dose Assessment Bridge Line and the Emergency Management Bridge Line. Each system operates independently from the other systems and allows for conference calls between the members of that bridge line group

4. Plant telephone system The telephone system consists of digital automatic switchboard (DPBX) equipment and telephone stations. The DPBX is provided with redundant processors for reliability. The telephone stations are located throughout the power block, in the main control room, in the various buildings around the site, in the security building, and in the service building where the administrative offices are located.

For emergency use, unlisted telephone numbers are provided for direct access to the outside local public telephone system. Company provided cell phones ARE considered part of the Plant Telephone System. The FLEX response satellite phones are in place for beyond design basis accidents and ARE NOT considered part of the Plant Telephone System.

-Continued Below-Page 171 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 142 of 152

5. ENS (Red Phone) line The NRC Emergency Notification System (ENS) is an 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.

6. Back-Up Radio System (BURS)

The Back-up Radio System is a communication link between the Callaway Plant and offsite emergency response agencies. The primary use of this system is the backup notification of offsite agencies and the coordination of offsite activities during a radiologica1 emergency. The system uses 800 MHz radios. There are radio control base: units in the Plant Control Room, TSC and EOF, as well as each county EOC and the State EOC. The backup to this system is the commercial touchtone telephone system Notifications may also be initiated through the Callaway County/City of Fulton EOC via the Security radio.

7. Event Notification System A computerized notification system linked between the Callaway Plant, the State Emergency Management Agency and the four (4) EPZ risk counties. It allows the Communicator to fill out a notification form on screen and transmit the data simultaneously. Notifications on the Event Notification System can be initiated from the Control Room, the Emergency Operations Facility (EOF), or the Technical Support Center (TSC).

This EAL is the hot condition equivalent of the cold condition EAL CU5.1.

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

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

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

The second EAL condition addresses a total loss of the communications methods used to notify all OROs of an emergency declaration.

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

Callaway Basis Reference(s):

1. Callaway Plant Radiological Emergency Response Plan (RERP), Section 7.2

2. FSAR, Section 9.5.2

3. NEI 99-01, SU6 Page 172 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 143 of 152 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.

(Notes 1, 9)

Note 1: The Emergency Coordinator 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 indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment.

Basis:

This EAL addresses a failure of one or more containment penetrations to automatically isolate (close) when required by an actuation signal. It also addresses an event that results in high containment pressure with a concurrent failure of containment pressure control systems. Absent challenges to another fission product barrier, either condition represents potential degradation of the level of safety of the plant.

For the first condition, the containment isolation signal must be generated as the result of 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. This includes inadvertent actuation or operation. The determination of containment and penetration status - isolated or not isolated - should be made in accordance with the appropriate criteria contained in the plant AOPs and EOPs. The 15-minute criterion is included to allow operators time to manually isolate the required penetrations, if possible.

-Continue Below-Page 173 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 144 of 152 The second condition addresses a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design. The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible. The inability to start the required equipment indicates that containment heat removal/depressurization systems (e.g., containment spray system or containment cooling system) are either lost or performing in a degraded manner.

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 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 in the lower areas of containment. During normal operation, all four fan units may be operating. In post-accident operation following an actuation signal, the Containment Cooling System fans are designed to start automatically in slow speed if not already running (ref. 3).

The Containment pressure Hi-3 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 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.

Callaway 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. CSF-1, Critical Safety Function Status Trees (CSFST) Figure 5, Containment

5. FR-Z.1, Response to High Containment Pressure

6. Technical Specifications, Table 3.3.2-1

7. Technical Specifications, B3.6.6

8. NEI 99-01, SU7 Page 174 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 145 of 152 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 Event damage has caused indication of degraded performance on one train of a SAFETY SYSTEM needed for the current operating MODE.

AND EITHER:

• Event damage has caused indications of degraded performance in a second train of a SAFETY SYSTEM needed for the current operating MODE.

• Event damage has resulted in VISIBLE DAMAGE to a second train of a SAFETY SYSTEM needed for the current operating MODE.

(Notes 11, 12)

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

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

Table S-5 Hazardous Events

• EXPLOSION

• FIRE

• HIGH WINDS or tornado strike

• Internal or external FLOODING event

• Seismic event (earthquake)

• Other events with similar hazard characteristics as determined by the Emergency Coordinator MODE Applicability:

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

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

-Continue Below-Page 175 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 146 of 152 FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.

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

HIGH WINDS - Winds in excess of 40 mph (18 m/s) sustained, or 58 mph (26 m/s) gusting.

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

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

1. The integrity of the reactor coolant pressure boundary;

2. The capability to shut down the reactor and maintain it in a safe shutdown condition;

3. The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

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

Basis:

• Annunciator 98D, OBE will illuminate if the seismic instrument detects ground motion in excess of the OBE threshold. OTO-SG-00001, Seismic Event provides the guidance for determining if an OBE earthquake threshold is exceeded and any required response actions (ref. 1).

• Internal FLOODING may be caused by events such as component failures, equipment misalignment, or outage activity mishaps (ref. 2).

• External flooding may be due to high lake level. Callaway plant grade elevation is 840.0 ft. MSL.

(ref. 3).

• Seismic Category I structures are analyzed to withstand a sustained, design wind velocity of at least 100 mph. (ref. 4).

• Areas containing functions and systems required for safe shutdown of the plant are identified by fire area (ref. 5).

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

A single FAULTED steam generator would NOT require declaration per this EAL. Technical Specification Bases 3.7.4 explains that two intact Steam Generators are required for cooldown of the RCS and a third Steam Generator is assumed to be RUPTURED. If more than one Steam Generator is FAULTED, then this EAL is applicable.

-Continue Below-Page 176 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 147 of 152 This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the current operating mode. In order to provide the appropriate context for consideration of an ALERT classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train, and there must be either indications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance, and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE such that the potential exists for performance issues. Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has indications of degraded performance; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events.

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

Operators will make a determination of VISIBLE DAMAGE based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.

An event affecting equipment common to two or more safety systems or safety system trains (i.e., there are indications of degraded performance and/or VISIBLE DAMAGE affecting the common equipment) should be classified as an Alert. By affecting the operability or reliability of multiple system trains, the loss of the common equipment effectively meets the two-train impact criteria that underlines the EALs and Bases. For example, a unit with a tank that is the water source for multiple safety injection systems or trains, such as a Refueling Water Storage Tank.

An event that affects two trains of safety system (e.g., one train has indications of degraded performance and the other VISIBLE DAMAGE) that also has one or more additional trains should be classified as an Alert.

This approach maintains consistency with the two-train impact criteria that underlines the EALs and Bases, and is warranted because the event was severe enough to affect the operability or reliability of two trains of a safety system despite plant design criteria associated with system and system train separation and protection.

Such an event may have caused other plant impacts that are not immediately apparent. For example, a unit that has an Auxiliary/Emergency Feedwater system with three trains.

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

Callaway Basis Reference(s):

1. OTO-SG-00001, Seismic Event

2. IPE Section 3.4.2.3 Results of the Vulnerability Screening

3. FSAR, Section 3.4 Water Level (Flood) Design Table 3.4-1 PMF, Groundwater, Reference, and Actual Plant Elevations

4. FSAR, Section 3.3.1.1 Design Wind Loadings

5. FSAR, Section 9.5.1 Fire Protection System

6. NEI 99-01, SA9 Page 177 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 148 of 152 Category F - Fission Product Barrier Degradation EAL Group: Hot Conditions (RCS temperature > 200ºF); 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

-Continue Below-Page 178 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 149 of 152 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.

• 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 Callaway design and operating characteristics.

• As used in this category, the term RCS leakage encompasses not just those types defined in Technical Specifications but also includes the loss of RCS mass to any location- inside the primary containment, an interfacing system, or outside of the primary containment. The release of liquid or steam mass from the RCS due to the as-designed/expected operation of a relief valve is not considered to be RCS leakage.

• At the Site Area Emergency level, EAL users should maintain cognizance of how far present conditions are from meeting a threshold that would require a General Emergency declaration. For example, if the Fuel Clad and RCS fission product barriers were both lost, then there should be frequent assessments of containment radioactive inventory and integrity. Alternatively, if both the Fuel Clad and RCS fission product barriers were potentially lost, the Emergency Coordinator would have more assurance that there was no immediate need to escalate to a General Emergency.

Page 179 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 150 of 152 Category: Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Any loss or any potential loss of either Fuel Clad or RCS.

EAL:

FA1.1 Alert Any loss or any potential loss of either Fuel Clad or RCS (Table F-1).

MODE Applicability:

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

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.

Callaway Basis Reference(s):

1. NEI 99-01, FA1 Page 180 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 151 of 152 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 Coordinator would have greater assurance that escalation to a General Emergency is less imminent.

Callaway Basis Reference(s):

1. NEI 99-01, FS1 Page 181 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 1 (Cont'd.)

Emergency Action Level Technical Bases Sheet 152 of 152 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.

Callaway Basis Reference(s):

1. NEI 99-01, FG1 Page 182 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Sheet 1 of 43 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 / RCS Activity D. CMT Integrity or Bypass E. Emergency Coordinator Judgment Each category occupies a row in Table F-1 thus forming a matrix defined by the categories. The intersection of each row with each Loss/Potential Loss column forms a cell in which one or more fission product barrier thresholds appear.

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 Losses and Potential Losses are then applied to the algorithms given in EALs FG1.1, FS1.1, and FA1.1 to determine the appropriate emergency classification.

In the remainder of this Attachment, the Fuel Clad barrier threshold bases appear first, followed by the RCS barrier and finally the Containment barrier threshold bases. In each barrier, the bases are given according category Loss followed by category Potential Loss beginning with Category A, then B,,E Page 183 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 INFORMATION USE

1. CSFST Core 1. Containment 1. CSFST 1. Any condition in the Tble F-1 Containmen Potential Cooling-RED Path radiation Containment-RED opinion of the conditions met > 14,000 R/hr on Path conditions met Emergency Fission t (CMT) Loss AND GT-RE-59 (591) or 2. Containment Coordinator that GT-RE-60 (601) hydrogen indicates potential Product Barrier None Restoration loss of the procedures not concentration 4%

Barrier effective within 15 3. Containment pressure Containment barrier Rev. 021 Threshold min. > 27 psig with < one Fission Product Barrier Loss/Potential Loss Matrix and Bases (Note 1) full train of Matrix depressurization equipment operating per design for

1. A leaking or 1. Containment isolation 1. Any condition in the Loss RUPTURED SG is is required opinion of the FAULTED outside AND EITHER: Emergency of containment

• Containment Coordinator that None None integrity has been indicates loss of the lost based on Containment barrier Emergency Coordinator judgment

1. Operation of a 1. CSFST Heat 1. Any condition in the Reactor Potential standby charging Sink-RED Path opinion of the Attachment 2 (Cont'd) pump is required by conditions met Emergency Coolant Loss EITHER: AND None None Coordinator that indicates potential System

• UNISOLABLE Heat sink is loss of the RCS RCS leakage required (RCS)

• SG tube barrier Page 184 of 225 Sheet 2 of 43 Barrier leakage

2. CSFST

1. An automatic or 1. Containment 1. Any condition in the Loss manual ECCS (SI) radiation opinion of the actuation required > 59 R/hr on Emergency by EITHER: None GT-RE-59 (591) or None Coordinator that

• UNISOLABLE GT-RE-60 (601) indicates loss of the RCS leakage RCS barrier

• SG tube RUPTURE

1. CSFST Core 1. Any condition in the Fuel Clad Potential Cooling-ORANGE opinion of the Path conditions met Emergency (FC) Barrier Loss None 2. CSFST Heat None None Coordinator that Sink-RED Path indicates potential conditions met loss of the fuel clad barrier AND Heat sink is

1. CSFST Core 1. Containment 1. Any condition in the Loss Cooling-RED Path radiation opinion of the conditions met > 840 R/hr on Emergency GT-RE-59 (591) or Coordinator that None GT-RE-60 (601) None indicates loss of the

2. Dose equivalent fuel clad barrier I-131 coolant activity > 300 µCi/cc

3. CVCS letdown radiation Category A B C D E RCS or SG Tube Inadequate Heat CMT Radiation / CMT Integrity or Judgment Leakage Removal RCS Activity Bypass

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 2 (Cont'd)

Fission Product Barrier Loss/Potential Loss Matrix and Bases Sheet 3 of 43 Barrier: Fuel Clad Category: A. RCS or SG Tube Leakage Degradation Threat: Loss Threshold:

None.

Page 185 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 2 (Cont'd)

Fission Product Barrier Loss/Potential Loss Matrix and Bases Sheet 4 of 43 Barrier: Fuel Clad Category: A. RCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold:

None.

Page 186 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 2 (Cont'd)

Fission Product Barrier Loss/Potential Loss Matrix and Bases Sheet 5 of 43 Barrier: Fuel Clad Category: B. Inadequate Heat Removal Degradation Threat: Loss Threshold:

1. CSFST Core Cooling-RED Path conditions met.

Definition(s):

None Basis:

Critical Safety Function Status Tree (CSFST) Core Cooling-RED path indicates significant core exit superheating and core uncovery. The CSFSTs are normally monitored using the SPDS display on the Plant Computer (ref. 1).

This reading indicates temperatures within the core are sufficient to cause significant superheating of reactor coolant.

Callaway Basis Reference(s):

1. CSF-1 Critical Safety Function Status Trees

2. FR-C.1 Response to Inadequate Core Cooling

3. FR-C.2 Response to Degraded Core Cooling

4. NEI 99-01 Inadequate Heat Removal Fuel Clad Loss 2.A Page 187 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 2 (Cont'd)

Fission Product Barrier Loss/Potential Loss Matrix and Bases Sheet 6 of 43 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 are normally monitored using the SPDS display on the Plant Computer (ref. 1).

This reading indicates a reduction in reactor vessel water level sufficient to allow the onset of heat-induced cladding damage.

Callaway Basis Reference(s):

1. CSF-1 Critical Safety Function Status Trees

2. FR-C.1 Response to Inadequate Core Cooling

3. FR-C.2 Response to Degraded Core Cooling

4. NEI 99-01 Inadequate Heat Removal Fuel Clad Loss 2.A Page 188 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 2 (Cont'd)

Fission Product Barrier Loss/Potential Loss Matrix and Bases Sheet 7 of 43 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 heat sink function is under extreme challenge and that some fuel clad damage may potentially occur (ref. 1).

The CSFSTs are normally monitored using the SPDS display on the Plant Computer (ref. 1).

The phrase and heat sink required precludes the need for classification for conditions in which RCS pressure is less than SG pressure or Heat Sink-RED path entry was created through operator action directed by an EOP. For example, FR-H.1 is entered from CSFST Heat Sink-Red. Step 1 tells the operator to determine if heat sink is required by checking that RCS pressure is greater than any non-faulted SG pressure and either RCS temperature is greater than 350ºF or RCS pressure is greater than 360 psig. 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.

Callaway Basis Reference(s):

1. CSF-1 Critical Safety Function Status Trees Figure 3 Heat Sink

2. FR-H.1 Response to Loss of Secondary Heat Sink

3. NEI 99-01 Inadequate Heat Removal Fuel Clad Loss 2.B Page 189 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 2 (Cont'd)

Fission Product Barrier Loss/Potential Loss Matrix and Bases Sheet 8 of 43 Barrier: Fuel Clad Category: C. CMT Radiation / RCS Activity Degradation Threat: Loss Threshold:

1. Containment radiation > 840 R/hr on GT-RE-59 (591) or GT-RE-60 (601).

Definition(s):

None Basis:

Containment radiation monitor readings greater than 840 R/hr (ref. 1) indicate the release of reactor coolant, with elevated activity indicative of fuel damage, into the Containment. The reading is derived assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with a concentration of 300 µCi/cc dose equivalent I-131 into the Containment atmosphere. Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations (including iodine spiking) allowed within technical specifications and are therefore indicative of fuel damage (approximately 5% clad failure depending on core inventory and RCS volume).

Monitors used for this fission product barrier loss threshold are the Containment High Range Radiation Monitors GT-RE-59 (Panel RM-11 channel 591) and GT-RE-60 (Panel RM-11 channel 601). The threshold value of 840 R/hr is the HI-HI (RED) alarm setpoint (ref. 2).

The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals 300 µCi/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since 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 ECL to a Site Area Emergency.

Callaway Basis Reference(s):

1. EPCI-1701

2. OTA-SP-RM011 Radiation Monitor Control Panel RM-11

2. NEI 99-01 CMT Radiation / RCS Activity Fuel Clad Loss 3.A Page 190 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 2 (Cont'd)

Fission Product Barrier Loss/Potential Loss Matrix and Bases Sheet 9 of 43 Barrier: Fuel Clad Category: C. CMT Radiation / RCS Activity Degradation Threat: Loss Threshold:

2. Dose equivalent I-131 coolant activity > 300 µCi/cc.

Definition(s):

None Basis:

Dose Equivalent Iodine (DEI) is determined by Chemistry procedure CDP-ZZ-08100, Post Accident Sampling Guidelines (ref. 1).

Elevated reactor coolant activity represents a potential degradation in the level of safety of the plant and a potential precursor of more serious problems. The threshold dose equivalent I-131 concentration is well above that expected for iodine spikes and corresponds to about 2% to 5% fuel clad damage. When reactor coolant activity reaches this level the Fuel Clad barrier is considered lost. (ref. 2).

This threshold indicates that RCS radioactivity concentration is greater than 300 µCi/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.

There is no Potential Loss threshold associated with RCS Activity / Containment Radiation.

Callaway Basis Reference(s):

1. CDP-ZZ-08100 Post Accident Sampling Guidelines

2. NEI 99-01 CMT Radiation / RCS Activity Fuel Clad Loss 3.B Page 191 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 2 (Cont'd)

Fission Product Barrier Loss/Potential Loss Matrix and Bases Sheet 10 of 43 Barrier: Fuel Clad Category: C. CMT Radiation / RCS Activity Degradation Threat: Loss Threshold:

3. CVCS letdown radiation > 2.50E+01 µCi/ml on SJ-RE-01 (016).

Definition(s):

None Basis:

The normal Chemical and Volume Control System (CVCS) charging and letdown flow path allows purification of the reactor coolant and control of the RCS volume while maintaining a continuous feed and bleed flow between the RCS and the CVCS. Reactor coolant is first "letdown" from the RCS through a regenerative heat exchanger, which minimizes heat losses from the RCS. Additional cooling takes place in a letdown heat exchanger that acts as the heat sink for the system. Downstream of the letdown heat exchanger pressure control valve and upstream of the mixed bed demineralizers, the letdown stream passes by radiation monitor SJ-RE-01, which will warn of fission products in the letdown coolant if a fuel element failure occurs. The monitor is located in the Primary Sample Sink Room.

The CVCS letdown monitor SJ-RE-01 provides indication in the Control Room on Panel RM-11 channel 016 with a range of 1.7E-03 to 1.7E+03 µCi/ml (ref. 2, 3). The HI-HI (RED) alarm is 5E0 + background +

(background x 0.05) (ref. 4) and represents a total fuel clad failure in excess of 1% in 30 minutes (ref. 2, 3).

Five times this alarm setpoint corresponds to approximately 5% fuel clad failure. 5% clad failure is also the basis for the coolant activity and Containment radiation Fuel Clad loss thresholds.

Callaway Basis Reference(s):

1. FSAR Section 9.3.4.2

2. FSAR Table 11.5-1

3. OTA-SP-RM011 Radiation Monitor Control Panel RM-11

4. HPCI-05-02 Gaseous and Liquid Radiation Monitor Setpoints Rev. 0, Note 11

5. NEI 99-01 Other Indications Fuel Clad Loss 5.A Page 192 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 2 (Cont'd)

Fission Product Barrier Loss/Potential Loss Matrix and Bases Sheet 11 of 43 Barrier: Fuel Clad Category: C. CMT Radiation / RCS Activity Degradation Threat: Potential Loss Threshold:

None.

Page 193 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 2 (Cont'd)

Fission Product Barrier Loss/Potential Loss Matrix and Bases Sheet 12 of 43 Barrier: Fuel Clad Category: D. CMT Integrity or Bypass Degradation Threat: Loss Threshold:

None.

Page 194 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 2 (Cont'd)

Fission Product Barrier Loss/Potential Loss Matrix and Bases Sheet 13 of 43 Barrier: Fuel Clad Category: D. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:

None.

Page 195 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 2 (Cont'd)

Fission Product Barrier Loss/Potential Loss Matrix and Bases Sheet 14 of 43 Barrier: Fuel Clad Category: E. Emergency Coordinator Judgment Degradation Threat: Loss Threshold:

1. Any condition in the opinion of the Emergency Coordinator that indicates loss of the Fuel Clad barrier.

Definition(s):

None Basis:

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

• Imminent barrier degradation exists if the degradation will likely occur within relatively short period of time based on a projection of current safety system performance. The term imminent refers to recognition of the inability to reach safety acceptance criteria before completion of all checks.

• Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators. This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.

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

This threshold addresses any other factors that are to be used by the Emergency Coordinator in determining whether the Fuel Clad barrier is lost.

Callaway Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment Fuel Clad Loss 6.A Page 196 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 2 (Cont'd)

Fission Product Barrier Loss/Potential Loss Matrix and Bases Sheet 15 of 43 Barrier: Fuel Clad Category: E. Emergency Coordinator Judgment Degradation Threat: Potential Loss Threshold:

1. Any condition in the opinion of the Emergency Coordinator that indicates potential loss of the Fuel Clad barrier.

Basis:

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

• Imminent barrier degradation exists if the degradation will likely occur within relatively short period of time based on a projection of current safety system performance. The term imminent refers to recognition of the inability to reach safety acceptance criteria before completion of all checks.

• Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators. This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.

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

This threshold addresses any other factors that are to be used by the Emergency Coordinator in determining whether the Fuel Clad barrier is potentially lost. The Emergency Coordinator should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

Callaway Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment Potential Fuel Clad Loss 6.A Page 197 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 2 (Cont'd)

Fission Product Barrier Loss/Potential Loss Matrix and Bases Sheet 16 of 43 Barrier: Reactor Coolant System Category: A. RCS or SG Tube Leakage Degradation Threat: Loss Threshold:

1. An automatic or manual ECCS (SI) actuation required by EITHER:

• UNISOLABLE RCS leakage.

• SG tube RUPTURE.

Definition(s):

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

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

Basis:

ECCS (SI) actuation is caused by (ref. 1):

• Pressurizer low pressure < 1849 psig

• Steamline low pressure < 615 psig

• Containment high pressure > 3.5 psig

• Manual 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.

If EOPs directs Operators to open the Pressurizer Pressure Relief Valves to implement a core cooling strategy (i.e., a "feed and bleed" cooldown), then there will exist a Reactor Coolant flow path from the RCS, past the "Pressurizer Safety and Relief Valves" and into the Containment that Operators cannot isolate without compromising the effectiveness of the strategy (i.e., for the strategy to be effective, the valves must be kept in the open position);

therefore, the flow through the pressure relief line is UNISOLABLE. In this case, the ability of the RCS pressure boundary to serve as an effective barrier to a release of fission products has been eliminated and thus this condition constitutes a loss of RCS barrier.

A steam generator with primary-to-secondary leakage of sufficient magnitude to require a safety injection is considered to be RUPTURED. If a RUPTURED steam generator is also FAULTED outside of containment, the declaration escalates to a Site Area Emergency since the Containment Barrier Loss threshold 1.A will also be met.

Callaway Basis Reference(s):

1. E-0 Reactor Trip or Safety Injection

2. E-3 Steam Generator Tube Rupture

3. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Loss 1.A Page 198 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 2 (Cont'd)

Fission Product Barrier Loss/Potential Loss Matrix and Bases Sheet 17 of 43 Barrier: Reactor Coolant System Category: A. RCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold:

1. Operation of a standby charging pump is required by EITHER:

• UNISOLABLE RCS leakage.

• SG tube leakage.

Definition(s):

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

Basis:

This threshold is based on the inability to maintain liquid inventory within the RCS by normal operation of the Chemical and Volume Control System (CVCS). The CVCS includes three charging pumps: one Normal Charging Pump with a design flow capacity of 130 gpm, and two centrifugal charging pumps each with a design flow capacity of 150 gpm (ref. 1). Approximately 12 gpm of charging flow bypasses the RCS due to leakage through the RCP seals; thus, the Normal Charging Pump can deliver 130 gpm - 12 gpm = 118 gpm (rounded to 120 gpm for readability) (ref. 2). A second charging pump being required is indicative of a substantial RCS leak exceeding the capacity of one charging pump (120 gpm) in the normal charging MODE with letdown isolated.

This threshold is based on an UNISOLABLE RCS leak that results in the inability to maintain pressurizer level within specified limits by operation of a normally used charging (makeup) pump, but an ECCS (SI) actuation has not occurred. The threshold is met when an operating procedure, or operating crew supervision, directs that a standby charging (makeup) pump be placed in service, indicating a substantial RCS leak exceeding the capacity of one charging pump (120 gpm) in the normal charging MODE with letdown isolated, to restore and maintain pressurizer level.

This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacing system. The mass loss may be into any location - inside containment, to the secondary-side (i.e., steam generator tube leakage) or outside of containment.

If a leaking steam generator is also FAULTED outside of containment, the declaration escalates to a Site Area Emergency since the Containment Barrier Loss threshold 1.A will also be met.

Callaway Basis Reference(s):

1. FSAR, Table 9.3-9

2. FSAR, Section 9.3.4 Chemical and Volume Control System

3. E-3 Steam Generator Tube Rupture

4. NEI 99-01, RCS or SG Tube Leakage Reactor Coolant System Potential Loss 1.A Page 199 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 2 (Cont'd)

Fission Product Barrier Loss/Potential Loss Matrix and Bases Sheet 18 of 43 Barrier: Reactor Coolant System Category: A. RCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold:

2. CSFST Integrity-RED Path conditions met.

Definition(s):

None Basis:

The "Potential Loss" threshold is defined by the CSFST Reactor Coolant Integrity - RED path. CSFST RCS Integrity - Red Path plant conditions and associated Pressurized Thermal Shock (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).

Callaway Basis Reference(s):

1. CSF-1, Critical Safety Function Status Trees Figure 4 Integrity and 4a Limit A Curve

2. FR-P.1, Response to Imminent Pressurized Thermal Shock Condition

3. NEI 99-01, RCS or SG Tube Leakage Reactor Coolant System Potential Loss 1.B Page 200 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 2 (Cont'd)

Fission Product Barrier Loss/Potential Loss Matrix and Bases Sheet 19 of 43 Barrier: Reactor Coolant System Category: B. Inadequate Heat Removal Degradation Threat: Loss Threshold:

None.

Page 201 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 2 (Cont'd)

Fission Product Barrier Loss/Potential Loss Matrix and Bases Sheet 20 of 43 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:

In combination with FC Potential Loss B.2, meeting this threshold results in a Site Area Emergency.

Critical Safety Function Status Tree (CSFST) Heat Sink-RED path indicates the heat sink function is under extreme challenge and that some fuel clad damage may potentially occur (ref. 1).

The CSFSTs are normally monitored using the SPDS display on the Plant Computer (ref. 1).

The phrase and heat sink required precludes the need for classification for conditions in which RCS pressure is less than SG pressure or Heat Sink-RED path entry was created through operator action directed by an EOP. For example, FR-H.1 is entered from CSFST Heat Sink-Red. Step 1 tells the operator to determine if heat sink is required by checking that RCS pressure is greater than any non-faulted SG pressure and either RCS temperature is greater than 350ºF or RCS pressure is greater than 360 psig. 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.

Callaway Basis Reference(s):

1. CSF-1, Critical Safety Function Status Trees Figure 3 Heat Sink

2. FR-H.1, Response to Loss of Secondary Heat Sink

3. NEI 99-01, Inadequate Heat Removal RCS Loss 2.B Page 202 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 2 (Cont'd)

Fission Product Barrier Loss/Potential Loss Matrix and Bases Sheet 21 of 43 Barrier: Reactor Coolant System Category: C. CMT Radiation/ RCS Activity Degradation Threat: Loss Threshold:

1. Containment radiation > 59 R/hr on GT-RE-59 (591) or GT-RE-60 (601).

Definition(s)

None Basis:

Containment radiation monitor readings greater than 59 R/hr (ref. 1) indicate the release of reactor coolant to the Containment. The readings assume the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with normal operating concentrations (i.e., within Technical Specifications) into the Containment atmosphere. Because of the very high fuel clad integrity, only small amounts of noble gases would be dissolved in the primary coolant.

Monitors used for this fission product barrier loss threshold are the Containment High Range Radiation Monitors GT-RE-59 (Panel RM-11 channel 591) and GT-RE-60 (Panel RM-11 channel 601). The threshold value of 59 R/hr is the HI (YELLOW) alarm setpoint (ref. 2).

The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for Fuel Clad Barrier Loss threshold C.1 since it indicates a loss of the RCS Barrier only.

There is no Potential Loss threshold associated with RCS Activity / Containment Radiation.

Callaway Basis Reference(s):

1. EPCI-1701

2. OTA-SP-RM011, Radiation Monitor Control Panel RM-11

3. NEI 99-01, CMT Radiation / RCS Activity RCS Loss 3.A Page 203 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 2 (Cont'd)

Fission Product Barrier Loss/Potential Loss Matrix and Bases Sheet 22 of 43 Barrier: Reactor Coolant System Category: C. CMT Radiation/ RCS Activity Degradation Threat: Potential Loss Threshold:

None.

Page 204 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 2 (Cont'd)

Fission Product Barrier Loss/Potential Loss Matrix and Bases Sheet 23 of 43 Barrier: Reactor Coolant System Category: D. CMT Integrity or Bypass Degradation Threat: Loss Threshold:

None.

Page 205 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 2 (Cont'd)

Fission Product Barrier Loss/Potential Loss Matrix and Bases Sheet 24 of 43 Barrier: Reactor Coolant System Category: D. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:

None.

Page 206 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 2 (Cont'd)

Fission Product Barrier Loss/Potential Loss Matrix and Bases Sheet 25 of 43 Barrier: Reactor Coolant System Category: E. Emergency Coordinator Judgment Degradation Threat: Loss Threshold:

1. Any condition in the opinion of the Emergency Coordinator that indicates loss of the RCS barrier.

Definition(s):

None Basis:

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

• Imminent barrier degradation exists if the degradation will likely occur within relatively short period of time based on a projection of current safety system performance. The term imminent refers to recognition of the inability to reach safety acceptance criteria before completion of all checks.

• Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators. This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.

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

This threshold addresses any other factors that may be used by the Emergency Coordinator in determining whether the RCS Barrier is lost.

Callaway Basis Reference(s):

1. NEI 99-01, Emergency Director Judgment RCS Loss 6.A Page 207 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 2 (Cont'd)

Fission Product Barrier Loss/Potential Loss Matrix and Bases Sheet 26 of 43 Barrier: Reactor Coolant System Category: E. Emergency Coordinator Judgment Degradation Threat: Potential Loss Threshold:

1. Any condition in the opinion of the Emergency Coordinator that indicates potential loss of the RCS barrier.

Definition(s):

None Basis:

The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the RCS barrier is potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

• Imminent barrier degradation exists if the degradation will likely occur within relatively short period of time based on a projection of current safety system performance. The term imminent refers to recognition of the inability to reach safety acceptance criteria before completion of all checks.

• Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators. This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.

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

This threshold addresses any other factors that may be used by the Emergency Coordinator in determining whether the RCS Barrier is potentially lost. The Emergency Coordinator should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

Callaway Basis Reference(s):

1. NEI 99-01, Emergency Director Judgment RCS Potential Loss 6.A Page 208 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 2 (Cont'd)

Fission Product Barrier Loss/Potential Loss Matrix and Bases Sheet 27 of 43 Barrier: Containment Category: A. RCS or SG Tube Leakage Degradation Threat: Loss Threshold:

1. A leaking or RUPTURED SG is FAULTED outside of containment.

Definition(s):

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

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

Basis:

This threshold addresses a leaking or RUPTURED Steam Generator (SG) that is also FAULTED outside of containment. The condition of the SG, whether leaking or RUPTURED, is determined in accordance with the thresholds for RCS Barrier Potential Loss A.1 and Loss A.1, respectively. This condition represents a bypass of the containment barrier.

FAULTED is a defined term within the NEI 99-01 methodology; this determination is not necessarily dependent upon entry into, or diagnostic steps within, an EOP. For example, if the pressure in a steam generator is decreasing uncontrollably (part of the FAULTED definition) and the FAULTED steam generator isolation procedure is not entered because EOP user rules are dictating implementation of another procedure to address a higher priority condition, the steam generator is still considered FAULTED for emergency classification purposes.

The FAULTED criterion establishes an appropriate lower bound on the size of a steam release that may require an emergency classification. Steam releases of this size are readily observable with normal Control Room indications. The lower bound for this aspect of the containment barrier is analogous to the lower bound criteria specified in IC SU4 for the fuel clad barrier (i.e., RCS activity values) and IC SU5 for the RCS barrier (i.e., RCS leak rate values).

This threshold also applies to prolonged steam releases necessitated by operational considerations such as the forced steaming of a leaking or RUPTURED steam generator directly to atmosphere to cooldown the plant, or to drive an auxiliary (emergency) feedwater pump. These types of conditions will result in a significant and sustained release of radioactive steam to the environment (and are thus similar to a FAULTED condition). If the TDAFP is running and being supplied by a ruptured steam generator that has not been isolated, this threshold is met. Manual Operator action can NOT be credited.

-Continued Below-Page 209 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 2 (Cont'd)

Fission Product Barrier Loss/Potential Loss Matrix and Bases Sheet 28 of 43 Steam releases associated with the expected operation of a Steam Generator Atmospheric Steam Dump or Main Steam Safety Valve do not meet the intent of this threshold. Such releases may occur intermittently for a short period of time following a reactor trip as operators process through emergency operating procedures to bring the plant to a stable condition and prepare to initiate a plant cooldown. Steam releases associated with the unexpected operation of a valve (e.g., a stuck-open safety valve) do meet this threshold.

Following an SG tube leak or rupture, there may be minor radiological releases through a secondary-side system component (e.g., air ejectors, gland 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 ECLs resulting from primary-to-secondary leakage, with or without a steam release from the FAULTED SG, are summarized below.

Affected SG is FAULTED Outside of Containment?

P-to-S Leak Rate Yes No Less than or equal to 25 gpm No classification No classification Unusual Event per Unusual Event per Greater than 25 gpm SU5.1 SU5.1 Requires operation of a standby charging (makeup) Site Area Emergency Alert per FA1.1 pump per FS1.1 (RCS Barrier Potential Loss)

Requires an automatic or Site Area Emergency manual ECCS (SI) actuation Alert per FA1.1 per FS1.1 (RCS Barrier Loss)

There is no Potential Loss threshold associated with RCS or SG Tube Leakage.

Callaway Basis Reference(s):

1. E-2, Faulted Steam Generator Isolation

2. E-3, Steam Generator Tube Rupture

3. NEI 99-01, RCS or SG Tube Leakage Containment Loss 1.A Page 210 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 2 (Cont'd)

Fission Product Barrier Loss/Potential Loss Matrix and Bases Sheet 29 of 43 Barrier: Containment Category: A. RCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold:

None.

Page 211 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 2 (Cont'd)

Fission Product Barrier Loss/Potential Loss Matrix and Bases Sheet 30 of 43 Barrier: Containment Category: B. Inadequate Heat Removal Degradation Threat: Loss Threshold:

None.

Page 212 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 2 (Cont'd)

Fission Product Barrier Loss/Potential Loss Matrix and Bases Sheet 31 of 43 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 Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Definition(s):

None Basis:

Critical Safety Function Status Tree (CSFST) Core Cooling-RED path indicates significant core exit superheating and core uncovery. The CSFSTs are normally monitored using the SPDS display on the Plant Computer (ref. 1).

The function restoration procedures are those emergency operating procedures that address the recovery of the core cooling critical safety functions. The procedure is considered effective if the temperature is decreasing or if the vessel water level is increasing (ref. 1, 2, 3).

A direct correlation to status trees can be made if the effectiveness of the restoration procedures is also evaluated. If core exit thermocouple (TC) readings are greater than 1,200°F (ref. 1), Fuel Clad barrier is also lost.

This threshold addresses any other factors that may be used by the Emergency Coordinator in determining whether the RCS Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

Callaway Basis Reference(s):

1. CSF-1, Critical Safety Function Status Trees Figure 2 Core Cooling

2. FR-C.1, Response to Inadequate Core Cooling

3. FR-C.2, Response to Degraded Core Cooling

4. NEI 99-01, Inadequate Heat Removal Containment Potential Loss 2.A Page 213 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 2 (Cont'd)

Fission Product Barrier Loss/Potential Loss Matrix and Bases Sheet 32 of 43 Barrier: Containment Category: C. CMT Radiation/RCS Activity Degradation Threat: Loss Threshold:

None.

Page 214 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 2 (Cont'd)

Fission Product Barrier Loss/Potential Loss Matrix and Bases Sheet 33 of 43 Barrier: Containment Category: C. CMT Radiation/RCS Activity Degradation Threat: Potential Loss Threshold:

1. Containment radiation > 14,000 R/hr on GT-RE-59 (591) or GT-RE-60 (601).

Definition(s):

None Basis:

Containment radiation monitor readings greater than 14,000 R/hr (ref. 1) indicate significant fuel damage well in excess of that required for loss of the RCS barrier and the Fuel Clad barrier.

The readings are higher than that specified for Fuel Clad barrier Loss C.1 and RCS barrier Loss C.1.

Containment radiation readings at or above the Containment barrier Potential Loss threshold, therefore, signify a loss of two fission product barriers and Potential Loss of a third, indicating the need to upgrade the emergency classification to a General Emergency.

Monitors used for this fission product barrier loss threshold are the Containment High Range Radiation Monitors GT-RE-59 (Panel RM-11 channel 591) and GT-RE-60 (Panel RM-11 channel 601) (ref. 2).

The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that 20% of the fuel cladding has failed. This level of fuel clad failure is well above that used to determine the analogous Fuel Clad Barrier Loss and RCS Barrier Loss thresholds.

NUREG-1228, Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents, indicates the fuel clad failure must be greater than approximately 20% in order for there to be a major release of radioactivity requiring offsite protective actions. For this condition to exist, there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential loss of containment which would then escalate the ECL to a General Emergency.

Callaway Basis Reference(s):

1. EPCI-1701

2. OTA-SP-RM011, Radiation Monitor Control Panel RM-11

3. NEI 99-01, CMT Radiation / RCS Activity Containment Potential Loss 3.A Page 215 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 2 (Cont'd)

Fission Product Barrier Loss/Potential Loss Matrix and Bases Sheet 34 of 43 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 Coordinator judgment.

• UNISOLABLE pathway from containment to the environment exists.

Definition(s):

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

Basis:

These thresholds address a situation where containment isolation is required and one of two conditions exists as discussed below. Users are reminded that there may be accident and release conditions that simultaneously meet both bulleted thresholds.

First Threshold - Containment integrity has been lost, i.e., the actual containment atmospheric leak rate likely exceeds that associated with allowable leakage (or sometimes referred to as design leakage).

Following the release of 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 Coordinator will assess this threshold using judgment, and with due consideration given to current plant conditions, and available operational and radiological data (e.g., containment pressure, readings on radiation monitors outside containment, operating status of containment pressure control equipment, etc.).

Refer to the middle piping run of Figure 1. Two simplified examples are provided. One is leakage from a penetration and the other is leakage from an in-service system valve. Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure.

Another example would be a loss or potential loss of the RCS barrier, and the simultaneous occurrence of two FAULTED locations on a steam generator where one fault is located inside containment (e.g., on a steam or feedwater line) and the other outside of containment. In this case, the associated steam line provides a pathway for the containment atmosphere to escape to an area outside the containment.

Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable (design) containment leakage through various penetrations or system components. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R ICs.

-Continued Below-Page 216 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 2 (Cont'd)

Fission Product Barrier Loss/Potential Loss Matrix and Bases Sheet 35 of 43 Second Threshold - Conditions are such that there is an UNISOLABLE pathway for the migration of radioactive material from the containment atmosphere to the environment. As used here, the term environment includes the atmosphere of a room or area, outside the containment, that may, in turn, communicate with the outside-the-plant atmosphere (e.g., through discharge of a ventilation system or atmospheric leakage). Depending upon a variety of factors, this condition may or may not be accompanied by a noticeable drop in containment pressure.

Refer to the top piping run of Figure 1. In this simplified example, the inboard and outboard isolation valves remained open after a containment isolation was required (i.e., containment isolation was not successful).

There is now an UNISOLABLE pathway from the containment to the environment.

The existence of a filter is not considered in the threshold assessment. Filters do not remove fission product noble gases. In addition, a filter could become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream.

Leakage between two interfacing liquid systems, by itself, does not meet this threshold.

Refer to the bottom piping run of Figure 1. In this simplified example, leakage in an RCP seal cooler is allowing radioactive material to enter the Auxiliary Building. The radioactivity would be detected by the Process Monitor. If there is no leakage from the closed water cooling system to the Auxiliary Building, then no threshold has been met. If the pump developed a leak that allowed steam/water to enter the Auxiliary Building, then second threshold would be met. Depending upon radiation monitor locations and sensitivities, this leakage could be detected by any of the four monitors depicted in the figure and cause the first threshold to be met as well.

Following the leakage of 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.

The status of the containment barrier during an event involving steam generator tube leakage is assessed using Loss Threshold A.1.

Callaway Basis Reference(s):

1. NEI 99-01, CMT Integrity or Bypass Containment Loss 4.A Page 217 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 2 (Cont'd)

Fission Product Barrier Loss/Potential Loss Matrix and Bases Sheet 36 of 43 Barrier: Containment Category: D. CMT Integrity or Bypass Degradation Threat: Loss Threshold:

2. Indications of RCS leakage outside of containment.

Definition(s):

None Basis:

ECA-1.2 LOCA Outside Containment (ref. 1) provides instructions to identify and isolate a LOCA outside of the containment. Potential RCS leak pathways outside containment include (ref. 1, 2):

• Residual Heat Removal

• Safety Injection

• Chemical & Volume Control

• RCP seals

• PZR/RCS Loop 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.

The sum of the leakage rates of less than or equal to 1 gpm are acceptable outside of containment per Technical Specification. These systems include the recirculation portion of the Containment Spray, Safety Injection, Chemical and Volume Control, and Residual Heat Removal.

Refer to the middle piping run of Figure 1. In this simplified example, a leak has occurred at a reducer on a pipe carrying reactor coolant in the Auxiliary Building. Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure and cause threshold D.1 to be met as well.

To ensure proper escalation of the emergency classification, the RCS leakage outside of containment must be related to the mass loss that is causing the RCS Loss and/or Potential Loss threshold A.1 to be met.

Callaway Basis Reference(s):

1. ECA-1.2, LOCA Outside Containment

2. E-1, Loss of Reactor or Secondary Coolant

3. NEI 99-01, CMT Integrity or Bypass Containment Loss

4. ESP-ZZ-00356, Technical Specification 5.5.2.B Verification Integrated Leak Rate Requirements for Primary Coolant Sources Outside Containment.

5. Technical Specification 5.5.2.B Page 218 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 INFORMATION USE Rev. 021 2 nd * *..

Fission Product Barrier Loss/Potential Loss Matrix and Bases

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Sheet 37 of 43 Page 219 of 225 Ooen *va lve~

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EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 2 (Cont'd)

Fission Product Barrier Loss/Potential Loss Matrix and Bases Sheet 38 of 43 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 48 psig and represents an extreme challenge to safety function. The CSFSTs are normally monitored using the SPDS display on the Plant Computer (ref. 1).

48 psig is the containment pressure that is expected to occur following a design basis Loss of Coolant Accident (LOCA) (ref. 2) and is the pressure used to define CSFST Containment Red Path conditions.

If containment pressure exceeds the pressure that is expected to occur following a design basis Loss of Coolant Accident (LOCA), 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.

Callaway Basis Reference(s):

1. CSF-1, Critical Safety Function Status Trees Containment Figure 5

2. Calc No. 392.2 XX-95, Callaway Containment Parameters EOP Action Values, Setpoint ID T.03

3. NEI 99-01, CMT Integrity or Bypass Containment Potential Loss 4.A Page 220 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 2 (Cont'd)

Fission Product Barrier Loss/Potential Loss Matrix and Bases Sheet 39 of 43 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).

Callaway 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 EOPs if Containment hydrogen concentration should reach 0.5% by volume (ref. 4). 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 AI-10 and GS AI-19) with a range of 0% to 10% provide indication on Control Room Panel RL020 and ERFIS (ref. 3). The hydrogen monitors require a 2-hour warmup period when starting from the OFF position and 15 minutes when starting from STANDBY (ref. 4, 5). If an actual hydrogen concentration measurement is unavailable, CA-3 (ref. 6) may be used to estimate the Containment atmosphere hydrogen concentration.

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.

Callaway Basis Reference(s):

1. FSAR, Section 6.2 Containment Systems

2. FR-Z.4, Response to High Containment Hydrogen Concentration

3. FSAR, Table 7A-3 (Sheet 32 Data Sheet 6.4)

4. OTN-GS-00001, Containment Hydrogen Control System

5. Calc No. 392.2 XX-95, Callaway Containment Parameters EOP Action Values, Setpoint ID T101 & T102

6. CA-3, Hydrogen Flammability in Containment

7. NEI 99-01, CMT Integrity or Bypass Containment Potential Loss 4.B Page 221 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 2 (Cont'd)

Fission Product Barrier Loss/Potential Loss Matrix and Bases Sheet 40 of 43 Barrier: Containment Category: D. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:

3. Containment pressure > 27 psig with < one full train of containment depressurization equipment operating per design for 15 min.

(Notes 1, 9)

Note 1: The Emergency Coordinator 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 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 in the lower areas of containment. During normal operation, all four fan units may be operating. In post-accident operation following an actuation signal, the Containment Cooling System fans are designed to start automatically in slow speed if not already running (ref. 3).

The Containment pressure setpoint (27 psig, ref. 4, 5, 6) is the pressure at which the equipment should actuate and begin performing its function. The design basis accident analyses and evaluations assume the loss of one Containment Spray System train and one Containment Cooling System train (ref. 7). Consistent with the design requirement, one full train of depressurization equipment is therefore defined to be the availability of one train of each system. If less than this equipment is operating and Containment pressure is above the actuation setpoint, the threshold is met.

-Continued Below-Page 222 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 2 (Cont'd)

Fission Product Barrier Loss/Potential Loss Matrix and Bases Sheet 41 of 43 This threshold describes a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design. The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible. This threshold represents a potential loss of containment in that containment heat removal/depressurization systems (e.g., containment sprays, ice condenser fans, etc., but not including containment venting strategies) are either lost or performing in a degraded manner.

Callaway 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. CSF-1, Critical Safety Function Status Trees (CSFST) Figure 5, Containment

5. FR-Z.1, 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 223 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 2 (Cont'd)

Fission Product Barrier Loss/Potential Loss Matrix and Bases Sheet 42 of 43 Barrier: Containment Category: E. Emergency Coordinator Judgment Degradation Threat: Loss Threshold:

1. Any condition in the opinion of the Emergency Coordinator that indicates loss of the Containment barrier.

Definition(s):

None Basis:

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

• Imminent barrier degradation exists if the degradation will likely occur within relatively short period of time based on a projection of current safety system performance. The term imminent refers to recognition of the inability to reach safety acceptance criteria before completion of all checks.

• Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators. This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.

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

This threshold addresses any other factors that may be used by the Emergency Coordinator in determining whether the Containment Barrier is lost.

Callaway Basis Reference(s):

1. NEI 99-01, Emergency Director Judgment PC Loss 6.A Page 224 of 225 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 021 Attachment 2 (Cont'd)

Fission Product Barrier Loss/Potential Loss Matrix and Bases Sheet 43 of 43 Barrier: Containment Category: E. Emergency Coordinator Judgment Degradation Threat: Potential Loss Threshold:

1. Any condition in the opinion of the Emergency Coordinator that indicates potential loss of the Containment barrier.

Definition(s):

None Basis:

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

• Imminent barrier degradation exists if the degradation will likely occur within relatively short period of time based on a projection of current safety system performance. The term imminent refers to recognition of the inability to reach safety acceptance criteria before completion of all checks.

• Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators. This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.

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

This threshold addresses any other factors that may be used by the Emergency Coordinator in determining whether the Containment Barrier is lost.

Callaway Basis Reference(s):

1. NEI 99-01, Emergency Director Judgment PC Potential Loss 6.A Page 225 of 225 INFORMATION USE