Text

Revised Emergency Action Level Technical Basis Document
	ML18190A473
Person / Time
Site:	Callaway
Issue date:	07/09/2018
From:	Wink R; Ameren Missouri, Union Electric Co
To:	; Document Control Desk, Office of New Reactors, Office of Nuclear Reactor Regulation
References
	ULNRC-06445
	Download: ML18190A473 (246)
	v • d • e

Am MISSOURI eren Callaav Plant July 9,2018 ULNRC-06445 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 10 CFR 50.47(b) 10 CFR 50.54(q)

Ladies and Gentlemen:

DOCKET NUMBERS 50-483 AND 72-1045 CALLA WAY PLANT UNIT 1 UNION ELECTRIC CO.

RENEWED FACILITY OPERATING LICENSE NPF-30 REVISED EMERGENCY ACTION LEVEL TECHNICAL BASIS DOCUMENT Enclosed (as Attachment 2) is one copy of EIP-ZZ-0010l Addendum 2, Emergency Action Level Technical Basis Document, Revision 015.

This revision to a procedure that is included by reference in the RERP does not represent a reduction in the effectiveness of emetgency preparedness for the Callawav Plant. The RERP continues to meet the standards of 10 CFR 50.47(b) arid the requirements of 10 CFR 50.54(q).

This letter does not contain new commitments. If there are any questions concerning this letter, please contact Gene Juricic at 573-676-4489 or Getiy Rauch at 573-676-8504.

Sinceiy,

/

/Roger C. Wink t Manager, Regulatory Affairs JP l/

Attachments:

1. EIP-ZZ-OOlOl Addendum I, Emergency Action Level Technical Basis Document, Revision 015 Summary of Changes

2. EIP-ZZ-00l01 Addendum 2, Emergency Action Level Technical Basis Document. Revision 015 P.O. Box 620 fukon. MO 65251 ArnerenMissouri.com

ULNRC-06445 July 9,201$

Page 2 of 3 cc: Mr. Kriss M. Kennedy Regional Administrator U. S. Nuclear Regulatory Commission Region IV 1600 East Lamar Boulevard Arlington. TX 76011-45 11 Director Division of Spent Fuel Management Office of Nuclear Material Safety and Safeguards U. S. Nuclear Regulatory Commission Washington, DC 20555-000 1 Senior Resident Inspector Callaway Resident Office U.S. Nuclear Regulatory Commission

$201 NRC Road Steedman, MO 65077 Mr. L. John Kios Project Manager, Callaway Plant Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop O9E3 Washington, DC 20555-0001 Senior Emergency Preparedness Analyst U.S. Nuclear Regulatory Commission Region IV 1600 East Lamar Boulevard Arlington, TX 76011-4511

ULNRC-06445 July 9, 2018 Page 3 of 3 Index and send hardcopy to QA File A160.0761 Hardcopy:

Certrec Corporation 6100 Western Place. Suite 1050 Fort Worth, TX 76107 (Certrec receives ALL attachments as long as they are non-safeguards and may be publicly disclosed.)

Electronic distribution for the following can be made via RERP ULNR C Distribution:

F. M. Diya T. E. Herrmann B. L. Cox S. P. Banker

0. P. Rauch R. C. Wink T. B. Elwood Corporate Communications NSRB Secretary STARS Regulatory Affairs Mr. Jay Silberg (Pillsbuiy Winthrop Shaw Pittman LLP)

Missouri Public Service Commission Mrs. Collette Linder, REP Manager (SEMA) to EIP-ZZ-00101 Addendum 1, Emergency Action Level ULNRC-06445 Technical Basis Document, Revision 015 Sumniary of Changes Page 1 of I Section or Pige(s) Descriptioii Step Number These pages were used for Summary Of Changes in the previous revision. In 8 this revision, the wording This page intentional/v Ic/i blank was added to the pages N A as administrative place-keepers to keep the EAL Basis page numbering the same

- I and remain aligned with the page references on the EAL wall chart. This is strictly administrative and does not have any effect on the document.

Moved the last sentence of what was the paragraph (now the 3 paragraph) to a new stand-alone 2 paragraph, which reads, J/AEPS cannot he aligned to i/ic bits 97 CA2. I Basis within 15 minutes, i/ten it is not considered a cttpcthlc ACpouer source.

155 SAI.I Basis In CA2.1 cannot be was replaced with is not for clarity so that the sentence reads if;IEPS is not cilignccl to the bits tithin 15 minute.c, then it 15 not considered a capable .4 C potter source.

Moved the old 3td paragraph to become the new 2 paragraph and removed the reference to FRS.i in the old sentence Reactor shutdoitn achieved by usc of other (ii]) ctctions specified itt FRS. / Respoii.sc to Nuclear Generatiou:4 T115 (site/i as opening 173 SU6. I Basis PGI 9 and PG2O supply breake,c, cnlergcncl hoicition 01 ,iittiiiialli c/in mg control rods) 1 76 SU6.2 Basis c/u not constitute a successful manual trip (re/ 4), to create a new sentence that reads 179 SA6. I Basis Reactor shutdown cichieuecl hi list of otlict trip actions, such as opcning PG/V and PG2O tipph hreakc,c, emclgcnci horietion or mantictth driving control rods, c/u not constitute ci sitecc.cs/itl mcrnucd trip (ref 4), This addressed an Operations procedure change that removed directions to open the PG19 and PG2O supply breakers from FR-Si.

In the 3rd paragraph, removed the reference to FR-S.1 in the old sentence Reactor shutdown achieved h usc of fRS. 1 Responsc to Nttckar Pont, GencrcttionATWS (such as opening PG/V and PG2U supp/u hreczkcrs, emnergcnci horatio,,

or manna/h c/ruing contivl tic/sI cue ct/so credited cis a .cuccesvful ntamtctl trip proiiclccl reactor power cciii be reduced be/on 5,j he/arc i ucliccitions ofcm cxtrcnle challenge to either curt cooling or heat ,emoicil exist (ref. 1, 4). to create a new 181 SS6.l Basis sentence that reads Reactor shittdouun c,chievcc/ hr opcning PGJ9 and PG2t) sitpph breakers.

emergency horatio,, or nicmucilli driving control rods, cire ct/co ciedited as a success/td incinucil trip provided reactor poluem cciii be methiceci he/mi 5o belome incliccitions of cut extremc c/iallcngc to cit/icr core cooling or heat ,enuovcd ceist (ief 1,

41. This addressed an Operations procedure change that removed directions to open the PG1 9 and PG2O supply breakers from FR-Si.

Attachment 2 to ULNRC-06445 EIP-ZZ-OO1O1 Addendum 2, Emergency Action Level Technical Basis Document, Revision 015 241 pages

Ameren MISSOURI CaIIawa Energy Center EIP-ZZ-OO1O1 ADDENDUM 2 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT MINOR Revision 015 Page 1 of 24t INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 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 Or%anization 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 Dtiring 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. I. Developmental 18 4.2. Implementing 18 5.0 DEFINITIONS, ACRONYMS, & ABBREVIATIONS 19 5.1. Definitions 19 5.2. Abbreviations/Acronyms 23 6.0 CALLA WAY-TO-NET 99-01 REV. 6 EAL CROSS-REFERENCE 25 7.0 ATTACHMENTS 26 Page 2 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT TABLE OF CONTENTS Section Page Number 8.0

SUMMARY

Of CHANGES 26 ATTACHMENT 1 Emergency Action Level Technical Bases 31 Category R Abnormal Rad Release / Rad Effluent 31 RU!. 1 Unusual Event 32 RU 1.2 Unusual Event 34 RAI.1 Alert 36 RAI.2 Alert 38 RAI.3 Alert 40 RAI.4 Alert 42 RS 1.1 Site Area Emergency 44 RS 1 .2 Site Area Emergency 46 RSI.3 Site Area Emergency 48 RG 1 1 General Emergency 50 RG1.2 General Emergency 52 RGI .3 General Emergency 54 RU2. I Unusual Event 56 RA2. I Alert 5$

RA2.2 Alert 59 RA2.3 Alert 61 RS2. I Site Area Emergency 62 RG2.l General Emergency 63 RA3.I Alert 64 RA3.2 Alert 65 Category E Independent Spent Fuel Storage Installation (ISFSI) 67 EU I .1 Unusual Event 6$

Category C Cold Shutdown / Refueling System Malfunction 70 CUI.1 Unusual Event 71 CUI .2 Untisual Event 73 CAI.l Alert 75 CAI.2 Alert 77 CS 1. 1 Site Area Emergency 79 CS I .2 Site Area Emergency 81 CS] .3 Site Area Emergency 83 CGJ.t General Emergency 86 CGI.2 General Emergency 90 CU2.1 Unusual Event 94 CA2. 1 Alert 97 CU3.l Unusual Event 99 Page 3 of 241 INFORMATION USE

EIP-ZZ-0001 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT TABLE OF CONTENTS Section Page Number CU3.2 Unusual Event 101 CA3.l Alert 103 CU4.l Unusual Event 106 CU5.1 Unustial Event 108 CA6.1 Alert 111 Category H Hazards and Other Conditions Affecting Plant Safety 114 HU1.1 Unusual Event 115 HA1.1 Alert 117 HS 1.1 Site Area Emergency 119 HG 1 I General Emergency 121 HU2.l Unusual Event 123 HU3.l Unusual Event 125 HU3.2 Unusual Event 126 HU3.3 Unusual Event 128 KU3.4 Unusual Event 129 HU4.1 Unusual Event 130 HU4.2 Unusual Event 133 HU4.3 Untisual Event 136 l-1U4.4 Unusual Event 137 HA5.l Alert 138 HA6.l Alert 140 1-156.1 Site Area Emergency 141 HU7.l Unusual Event 143 HA7.l Alert 144 HS7.1 Site Area Emergency 146 HG7.1 General Emergency 148 Category S System Malfunction 150 SUI.I Unusual Event 152 SAI.1 Alert 154 551.1 Site Area Emergency 157 SGI.l General Emergency 159 SG1.2 General Emergency 161 SS2.l Site Area Emergency 164 SU3.l Unusual Event 166 SA3.l Alert 168 SU4.l Unusual Event 170 SU5.1 Unusual Event 171 SU6.l Unustial Event 173 SU6.2 Unusual Event 176 Page 4 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT TABLE OF CONTENTS Section Page Number SA6. 1 Alert 179 SS6.1 Site Aiea Emergency 181 SU7.1 Unusual Event t$3 SLJ8.1 Unusual Event 186 SA9.1 Alert 18$

Category F fission Product Barrier Degradation 191 FAI.1 Alert 193 fS1.1 Site Area Emergency 194 fG 1. 1 General Emergency 195 ATTACHMENT 2 - Fission Product Barrier Loss/Potential Loss Matrix and Bases 196 Page 5 of 241 INFORMATION USE

EIP-ZZ-00l01 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 1.0 PURPOSE This doctirnent provides an explanation and rationale for each Emergency Action Level (EAL).

Decision-makers responsible for implementation of EIP-ZZ-00l0 1, Classification of Emergencies, shotild 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).

Page 6 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 2.2. Fission Product Barriers Fission product harrier 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 mtdtipte physical barriers, any one of which. if maintained intact, precludes the release of significant amounts of radioactive fission prodttcts 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 battier 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 huildiiw and connections up to and including the outermost containment isolation valves. This barrier also includes the main steam. feedwater, and blowdewn 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 harrier 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 241 INFORMATION USE

EIP-ZZ-0010l ADDENDUM 2 Rev. 015 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 pj operating MODES This group wottid 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 grotip would only he reviewed by the EAL-user when the plant is in Cold Shutdown, Refueling or Deluded MODE.

The ptirpose 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 grotip. assignment of EALs to categories and subcategories:

Category and subcategory titles are selected tc) represent conditions that are operationally significant to the EAL-user. The Callaway hAL categories are aligned to and represent the NET 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.

Page 8 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT EAL Groups, Categories and Subcategories EAL Group I Category EAL Subcategory ALl Conditions:

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

S System Malfunction I 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 failtire 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 Temperattire 4 Loss of Vital DC Power 5 Loss of Communications 6 Hazardous Event Affecting Safety Sy stems The primary tool for determining the emergency classification level is the EAL Classification Matrix. The user of the EAL Classification Matrix may (but is not required to) consult the EAL Technical Bases Document in order to obtain additional information concerning the EALs under classification consideration. The user should consult Section 3.0 and Attachments I & 2 of this document for such information.

Page 9 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 2.5. Technical Bases Information EAL technical bases are provided in Attachment I for each EAL according to EAL group (All, Hot, Cold), EAL category (R, C, H. 5, E and F) and EAL subcategory. A summary explanation of each category and stLbcategory is given at the beginning of the technical bases disccissions 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 rectangjcl Each EAL is assigned a unique identifier to support accurate communication of the

-r emergency classification to onsite and offsite personnel. Four characters define each EAL identifier:

I. First character (letter): Corresponds to the EAL category as described above (R, C, H, 5, 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): Sitbcategory number within the given category. Stibcategories are sequentially numbered beginning with the number one (I). If a category does not have a stibcategory, this character is assigned the nuimber one (I).

4. Fourth character (number): The nctmerical sequtence of the EAL within the EAL subcategory. If the subcategory has only one EAL, it is given the number one (I).

Classification (enclosed in rectangle):

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

EAL (encLosed in rectangle)

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

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

5 Cold Shutdown, 6 Refueling, D Defueled, or 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.

Page 10 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 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.

Catlaway 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 Kii? 0.99 and reactor thermal power> 5%.

2 Startup Ken E 0.99 and reactor thermal potver 5%.

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

4 Hot Shutdown Kii < 0.99 and average coolant temperature 350°F > Lv> 200°F and at least 53 of 54 reactor vessel head closure bolts ftilly tensioned 5 Cold Shutdown Kn <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.

Page 11 of 241 INFORMATION USE

EIP-ZZ-00l01 ADDENDUM 2 Rev. 015 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 NSIRIDPR-ISG-Ol. 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 reqciired. 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 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.

Page 12 of 241 INFORMATION USE

EIP-ZZ-00lt)l ADDENDUM 2 Rev. 015 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. maniptilate.

repair. maintain or tiiodify 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 he subject to the reporting requirements of 10 CFR 50.72 (ref. 4.1.1).

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 Jitdment 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 tising EALs that are consistent with the Emergency Classification Level (ECL) definitions (refer to Category H). The Emergency Coordinator wilt 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.

Page 13 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 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 evalitation 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 I 5 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 rims concurrently with the emergency classification process clock. For a full discussion of this timing requirement, refer to NSIRIDPR-ISG-0l (ref. 4.1.11).

3.2. 1. Classification of Multiple Events and Conditions When multiple emergency events or conditions are present, the tiser 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 he declared and the other Alert should he 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. En particular, the fission product barrier EALs are applicable only to events that initiate in the Hot Shutdown MODE or higher.

Page 14 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 3.2.3. Classification of Imminent Conditions Aithotigh EALs provide specific thresholds, the Emergency Coordinator must remain alert to events or conditions that could lead to meeting or exceeding ati 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 he made as if the EAL has been met. While applicable to all emergency classification levels, this approach is partictilarly important at the higher emergency classification levels since it provides additional time for implementation of protective measures.

3.2.4. Emergency Classification Level Up%radin and Downradin An ECL may he 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 he based on a lower applicable ICts) and EAL(s). The ECL may also simply he 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 contintied 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.

Page 15 of 241 INFORMATION USE

EIP-ZZ-00l01 ADDENDUM 2 Rev. 015 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 1Iet 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 occtirs 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 l)rior 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 he 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 ptiblic protective actions rescilting from the emergency classification are truly warranted by the plant conditions.

Page 16 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 0t5 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 he met bctt the emergency classification was not made at the time of the event or condition. This sittiation 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 lonter 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).

-END OF SECTION Page 17 of 241 INFORMATION USE

EIP-ZZ-00l0l ADDENDUM 2 Rev. 015 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 ML I 2326A805

4. 1.2. RIS 2007-02 Clarification of NRC Gttidance for Emergency Notifications During Quickly Changing Events, February 2, 2007.

4.1.3. NUREG-1022 Event Reporting Guidelines. IOCFR5O.72 and 50.73 4.1.4. 10 CER 50.72 Immediate Notification Reciuirements for Operating Nuclear Power Reactors 4.1 .5. 10 CFR 50.73 License Event Report System

4. I .6. Drawing $600-X-88 100 Property-Site Layout Owner Controlled Area and Surrounding Area

4. 1.7. Caltaway FSAR Figure 1 .2-44 Plant Area Layout 4.1.8. Technical Specifications Table 1.1-1 MODES

4. 1 .9. OSP-GT-0t)003 Containment Closure

4. 1.10. Procedure Writers Manual Callaway Plant Procedure Writers Manual 4.1.11. NSIRIDPR-ISG-0 I Interim Staff Guidance, Emergency Planning for Nticlear Power Plants 4.1.12. Callaway Plant Radiological Emergency Response Plan Emergency Plan (RERP) 4.1.13. OTG-ZZ-000t)7 Refueling Preparation. Performance and Recovery 4.1.14. APA-ZZ-00520. Reporting Requirements and Responsibilities 4.2. tmplementiiw 4.2.1. EIP-ZZ-00 101 Classification of Emergencies 4.2.2. NEt 99-01 Rev. 6 to Callaway EAL Comparison Matrix 4.2.3. Calf away EAL Matrix 4.2.4. CR 201702763. NOS Insight EP Risk Significant Planning Standard Performance Upper Tier Cause Evaluation Needed Page 1$ of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 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 [nitiating 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 acttial 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 Gtiideline exposui-e 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 [SF51. the CONFINEMENT BOUNDARY is defined to be the Multi-Purpose Canister (MPC).

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

Asappliecl 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 Nticlear 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. I 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 241 INFORMATION USE

El P-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Fire Combustion characterized by heat and light. Souices 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 actions that result in an actual loss of physical control of the facility.

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

High Winds Winds in excess of 40 mph (1$ m/s) sustained, or 58 mph (26 mIs) 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 incltided. 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 caiis ing 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 roomlarea (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 nctclear fuel and other radioactive materials associated with spent fuel storage.

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 conseqtlences.

Page 20 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 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 ArnerenUE (Callaway Plant) to he controlled for security purposes (reI4.1.6).

Projectile An object directed toward a Nuclear Power Plant that cotild 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 antI is depicted in Drawing 8600-X-88 100 Property-Site Layout. Owner Controlled Area and Surrounding Area (ref. 4. 1 .7).

RCS Intact The RCS should he 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 darns). 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 conditioti 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 IOCFR5O.2):

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

I) The integrity of the reactor coolant pressure boundary:

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

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

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

Page 21 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Site Area Emergency Events are in process or have occurred which involve acittal 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 raditis of 1.200 meters (3,937 feet) from the midpoint of the Unit I Reactor Building and the canceled Unit 2 Reactor Building. Control of access to this is by virtue of ownership and in accordance with IOCFRIOO (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 I) the restilt 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 occttrred which indicate a potential degradation in the level of safety of the plant or indicate a security threat to facility protection has been initiated. NC) releases of radioactive material requiring offsite response or monitoring are expected tinless further degradation of safety systems occurs.

Valid An indication, report. or condition, is considered to be valid when it is verified by (I) 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 acctiracy is removed. Implicit in this definition is the need for timely assessment.

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

Page 22 of 241 INFORMATION USE

EIP-ZZ-0010I ADDENDUM 2 Rev. 015 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 Procedtire 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)

K0t1 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 rnilli-Roentgen Equivalent Man MSL Main Steam Line MW Megawatt NEI Nuclear Energy Institute Page 23 of 241 INFORMATION USE

EIP-ZZ-00l0l ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 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 Untisual Event OBE Operating Basis Earthcluake 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 Press tin zed 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 Eftluent 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 lOAF Top of Active Fuel TSC Technical Support Center WOG Westinghouse Owners Group Page 24 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT 6.0 CALLAWAY-TO-NEI 99-0 1 REV. 6 EAL CROSS-REFERENCE This cross-reference is provided to facilitate association and location of a Callaway EAL within the NEE 99-01 IC/EAL identification scheme. Further information regarding the development of the Callaway EALs based on the NEI guidance can he found in the EAL Comparison Matrix.

Callaway Nfl 99-01 Rev. 6 Callaway fNEI 99-01 Rev. 6 Callaway Nfl 99-01 Rev. 6 Example Example Example EAL IC EAL EAL IC RU1.1 AU1 1,2 CU5.1 CU5 1,2,3 HA5.1 HA5 1 RU1.2 AU1 3 CA1.1 CAl 1 HA6.1 HA6 1 RU2.1 AU2 1 CA1.2 CAl 2 HA7.l 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 HG1.1 HG1 1 RA21 AA2 1 CS1.2 CS1 2 HG7.1 HG7 1 RA2.2 AA2 2 CS1.3 CS1 3 SU1.1 SU1 1 RA2.3 AA2 3 CG1.1 CG1 1 SU3.1 SU2 1 RA3.1 AA3 1 CG1.2 CG1 2 SU4.1 SU3 2 RA3.2 AA3 2 FA1.1 FA1 1 SU5.1 SU4 1,2,3 RS1.1 AS1 1 FS1.1 FS1 1 SU6.1 SU5 1 RS1.2 AS1 2 FG1.i FG1 1 SU6.2 SU5 2 RS1.3 AS1 3 HU1.1 HU1 1,23 SU7.1 SU6 1,2,3 RS2.1 AS2 1 HU2.1 HU2 1 SU8.1 SU7 1,2 RG1.1 AG1 1 HU3.1 HU3 1 SA1.1 SAl 1 RGJ.2 AG1 2 HU3.2 HU3 2 SA3.l SA2 1 RG1.3 AG1 3 HU3.3 HU3 3 SA6.1 SA5 1 RG2.1 AG2 1 HU3.4 HU3 4 SA9.1 SA9 1 CU1.1 GUi 1 HU4.1 HU4 1 SS1.1 551 1 CU1.2 CU1 2 HU4.2 HU4 2 SS2.l 558 1 CU2.1 CU2 1 HU4.3 HU4 3 556.1 SS5 1 CU3.1 CU3 1 HU4.4 HU4 4 SG1.1 SG1 1 CU3.2 CU3 2 HU7.1 HU7 1 SG1.2 SG8 1 CU4.1 GU4 1 HA1.l HAl 1,2 EUJ.1 E-HU1 1 Page 25 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 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 These pages were used for Summary Of Changes in the previous revision. In this revision we placed the wording Thi.c page in!entionfil!V lefi blank as

27. 28, 29, & 30 N/A administrative place-keepers to keep the EAL Basis page numbering the same and remained aligned with the page references on the EAL wall chart. This is strictly administrative and does not have any effect on the document.

Moved the last sentence of what was the 2,d paragraph (now the 31d paragraph) to a new stand-alone 2 paragraph. Which reads !f 4 EPS cannot be aligned to the bus 97 CA2. I Basis within 15 minutes. then it is not considered ci capahic ACpoiiei source.

155 SAIl Basis In CA2.1 cannot be was replaced with is not for clarity to end with the sentence H

that reads i[AEPS is not ciligiiccl to the bu,r iiithiii I) ilifluites. then it is lit)! considered ci capablc AC poller ctnirce Moved the old 3td paragraph to become the new 2nd paragraph and removed the reference to FR-S. 1 in the old sentence of Reactor shutdown achieved by use of other trip actions cpcci/wd in FRS. I Response to Nticlear Generation/A Ti VS. (.vcicli as opcuing 173 SU6.1 Basis PGI 9 and PG2O supply breakers. emergency boratiomi or manually driving control rods) 176 SU6.2 Basis do not constitute a success/ui manitdil trip (ref 4) to create the new sentence of 179 SA6.I Basis Reactor shutdown achieved by use of other trip actions. stuch as opening PGI9 and PG2O supply breakers, enier,çency bo ration or manna/tv drimg control myds, fit? hot comisfltttte a

,cuiccesstiul manual trip (ref 4). This was done to address an Operations procedure change that removed opening the PG19 and PG2O from FR-Si.

In the 3td paragraph, removed the reference to FR-S.1 in the old sentence of Reactor .chutciouvn achieved bs ruse of FR-S.] Response to Nuclear Poiucr Generation/A TI VS (such as opening PGI 9 and PG2O supph breakers. emergency ho ration or nuamally thimg control rods) are aLso credited as a success/ui manual trip iiulccl reactor poller can bc iethrcecl bclouu )f be/ore i;ucflccttions of an evtremc challenge to either core cooling or hecit reniovctl e.vist (tef 1, 4 to create the new sentence of 181 SS6. I Basis Reactor shutdown achieved by opening PGI9 and PG2O supply breakers, emerccncv horatio,, or niantually driving control rods, arc also credited as ci successful manual trip proiidecl recictor poller cciii be rethwed belmi 5% before mdiccttio,i.c cf an e.vtteme challenge to either come cotulimug or hear removal evict (ref 1. 4). This was done to address an Operations procedure change that removed opening the PG19 and PG2O from FR-Si.

Page 26 of 241 INFORMATION USE

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EIP-ZZ-OOtOl ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category R Abnormal Rad Release / Rad Effluent EAL Group: ANY tEALs 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 otfsite protective actions. Elevated area radiation levels in plant may also he 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 eftluent radiation monitoring systems provides a rapid assessment mechanism to determine releases in excess of classifiable limits. Projected offsite doses, actual oftsite field measurements or measured release rates via sampling indicate doses or dose rates above classifiable Ii m its.

2. Irradiated Ftiel Event Conditions indicative of a loss of adequate shielding or damage to irradiated fuel may preclude access to vital plant areas or restilt 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 241 INFORMATION USE

EIP-ZZ-OO 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Category: R Abnormal Rad Levels I Rad Effluent Subcategory: 1 Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer EAL:

RU 1.1 Unusual Event Reading on any Table R- I effluent radiation monitor> column UE for ? 60 mm.

(Nous 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 tiow past an effluent monitor is known to have stopped. indicating that the release path is isolated, the effluent monitor readinu 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-218 659E÷7 iiCi/sec 6.59E÷6 .ICi/sec 6.59E+5 pCi/sec 2 X Hi-Hi alarm ASD Monitors fNB/C/D) AB-RE-1 71/112!

113/114 72 mR/hr 1.2 mR/hr 0

G)

TD AFW Steam

° Discharge FC-RE-385 163 mR/hr 16.3 mR/hr 1.6 mR/hr Radwaste Bldg Vent GH-RE-1OB ----

2 X Hi-Hi alarm Liquid Radwaste Discharge HB-RE-18 ----

2 X Hi-Hi alarm MODE Applicability:

All Definition(s):

None Basis:

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

Page 32 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases The effluent monitor Hi-Hi alarm setpoints correspond to the Hi-Hi alarm (red) setpoint as displayed on RM-l 1.

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

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

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 cenditioti and radiological effluent EALS more ftilly 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 ptlrposes.

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):

I. APA-ZZ-01003. Catlaway Plant Offsite Dose Calculation Manual Section 2.2.3

2. FSAR Section 16.11.1.3, Radioactive Effluent Monitorine Instrumentation LCO

3. EPCI 1402. EAL Table R-l Calculations

4. NEI 99-01. AUI Page 33 of 241 INFORMATION USE

EIP-ZZ-00l0l ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Category: R Abnormal Rad Levels / Rad Effittent Subcategory: I Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or Ion cer.

EAL:

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

> 2 x ODCM limits for 60 mm.

(Notes I, 2)

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

Note 2: It an on2oiIle 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):

N one 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 regtilatory 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 feattires and/or controls.

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

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

Page 34 of 241 INFORMATION USE

EIP-ZZ-0010l ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases 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 tritiurn 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 RA I.

Callaway Basis Reference(s):

1. APA-ZZ-0 1003. Callaway Plant Offsite Dose Calculation Manual Section 2.2.3

2. NEI 99-01, AUI Page 35 of 241 INFORMATION USE

EIP-ZZ-00J01 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Category: R Abnormal Rad Levels / Rad Effluent Subcategory: I Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrern TEDE or 50 mrem thyroid CDE EAL:

RAI.1 Alert Reading on any Table R-l effluent radiatioti monitor > column ALERT for ? 15 mm.

(Notes 1, 2, 3, 4)

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

Note 2: It 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 eftluent monitor is known to have stopped. indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.

Note 4: The pre-calculated effluent monitor values presented in EALs RAI .1. RS 1.1 and RG 1.1 shoLild be used for emergency classification assessments until the restilts from a dose assessment using acttial meteorology are available.

Table R-1 Effluent Monitor Classification Thresholds Release Point

[ Monitor GE SAE Alert

[ UE Unit Vent GT-RE-21 B 6.59E+7 pCi/sec 6.59E+6 pCi/sec 6.59E+5 pCi/sec 2 X Hi-Hi alarm AB-RE-111/112/

ASD Monitors (A/B/C/D) 12 mR/hr 1.2 mR/hr 0

113/114 w

TD AFW Steam Discharge FC-RE-385 163 mR/hr 16.3 mR/hr 1.6 mR/hr Radwaste Bldg Vent GH-RE-1 OB ---- ---- ---- 2 X Hi-Hi alarm Liquid Radwaste Discharge HB-RE-18 ---- ---- -- 2 X Hi-Hi alarm MODE Applicability:

All Definition(s):

None Page 36 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Basis:

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

JO mRem TEDE

50 rnRem 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] for a classification, there is a 15 minute time limit to make the classification (RAIl). If Dose Assessment (URIJRASCAL) is available it should be used instead (RAI.2) since it is more accurate than the values in Table R-l. However the Dose Assessment personnel must he able to calculate results within 15 minutes of the Table R- I value being exceeded OR the classification should be macic using Table R- 1 (RA 1.1).

The RM- II Channel Number 213 is utilized for the Unit Vent tGT-RE-2 I B) reading for Table R- I. This channel is read out in iiCi/sec while all others are read out in pCi/mI.

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

Oç 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 I c 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 inrem 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 he via IC RSI.

Callaway Basis Reference(s):

I. EPCI 1402. EAL Table R-l Calculations

2. NE199-0l.AAI Page 37 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: R Abnormal Rad Levels / Rad Effluent Subcategory: I Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 rnrem TEDE or 50 mrem thymici CDE EAL:

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

MODE Applicability:

All Definition(s):

SITE BOUNDARY Exclusion Area Boundary is a synonymous term for Site Botindary. 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 I Reactor Building and the canceled Unit 2 Reactor Building.

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

Basis:

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

This IC is used based on resttlts 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 restilts 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 regtilatory 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 RS I.

Page 38 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EIERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Callawav Basis Reference(s):

1. EIP-ZZ-0121 1. Accident Dose Assessment

2. EPCI 1402. EAL Table R-l Calculations

3. NEI 99-01. AAI Page 39 of 241 INFORMATION USE

EIP-ZZ-0010l ADDENDUM 2 Rev. 0t5 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Category: R Abnormal Rad Levels / Rad Effluent Subcategory: 1 Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE.

EAL:

RA1.3 Alert Analysis of a liqttid effittent sample indicates a concentration or release rate that would result in doses> 10 mrem TEDE or 50 mrern thyroid CDE at or beyond the SITE BOUNDARY for 60 mm.

of exposure.

(Notes 1. 2)

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

Note 2: Ii an onooino 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 scirrounding 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 IOCFR 100.

Basis:

Dose assessments based on liquid releases are performed per Offsite Dose Calctilation Manual (ref. I).

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 ttreater 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 he 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 I % of the EPA PAG of I ,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

Page 40 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path. then the effluent monitor reading is no longer valid for classification purposes.

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 restilts to the Dose Assessment Technician or Dose Assessment Coordinator for evaluation.

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

Callaway Basis Reference(s):

1. APA-ZZ-0l003. Callaway Plant Offsite Dose Calculation Manual Section 2.2.3

2. NE199-Ol.AAI Page 41 of 241 INFORMATION USE

EIP-ZZ-0010l ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Category: R Abnormal Rad Levels / Rad Effluent Subcategory: 1 Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE.

EAL:

RA1.4 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 mm.

Analyses of field survey samples indicate thyroict CDE> 50 mrem for 60 mm. of inhalation.

(Notes 1. 2)

Now I: The Ernerency Coordinator should declare the event promptly cipon 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 Botindary. The Exclusion Area is defined as the area that encompasses the laud surrounding the Plant to a radius of 1.200 meters (3.937 feet) from the midpoint of the Unit I Reactor Building and the canceled Unit 2 Reactor Building.

Control of access to this is by virtue of ownership and in accordance with IOCFR 100.

Basis:

EIP-ZZ-002 11, Field Monitoring provides guidance for emergency or post-accident radiological environmental monitoring (ref. t).

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

The ctosecl window value is in mR/hr. The analysis of field stirvey samples is in mrern 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 I % of the EPA Protective Action Guides (PAGs). It incltides 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).

Page 42 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot he 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 TIDE dose is set at I of the EPA PAG of 1 .000 rnrem while the 50 rnrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TIDE 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 tt) 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 RS I Callaway Basis Reference(s):

I. EIP-ZZ-0021 I. Field Monitoring

2. NEI 99-01. AAI Page 43 of 241 INFORMATION USE

EIP-ZZ00l0l ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Category: R Abnormal Rad Levels / Rad Effluent Subcategory: I Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 inrem TEDE or 500 mrem thyroid CDE E AL:

RS1.1 Site Area Emergency Reading on any Table R- 1 effluent radiation nionitor> column SAE for? 15 mm.

(Notes I. 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 eft]uent monitor is known to have stopped. indicating that the release path is isolated, the effluent monitor reading is no longer VALID br classttication purposes.

Note 4: The pre-calculated effluent monitor values presented in EALs RAIl. RS I. I and RG I I 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 6.59E+7 j.tCi/sec 6.59E+6 pCi/sec 6.59E+5 pCi/sec 2 X Hi-Hi alarm ASD Monitors (A/B/C/D) AB-RE-11J/112/

12 mR/hr 1.2 mR/hr 0 113/114 w

ID AFW Steam Discharge FC-RE-385 163 mR/hr 16.3 mR/hr 1.6 mR/hr Radwaste Bldg Vent GH-RE-1OB ---- ----

2 X Hi-Hi alarm Liquid Radwaste

, Discharge HB-RE-18 ---- ----

2 X Hi-Hi alarm MODE Applicability:

All Definition(s):

None Page 44 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Basis:

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

IOOrnRernTEDE

500 rnRem CDE Thyroid Classification based on efficient 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 ptlrposes.

To clarify Note 4. if a threshold value is met in Table R- I for a classification, there is a 15 minute time limit to make the classification tRS 1.1). If Dose Assessment (URI/RASCAL) is available it should he used instead (RS 1.2) since it is more accurate than the values in Table R- I. However the Dose Assessment personnel must he able to calculate results within 15 minutes of the Table R- I value being exceeded OR the classification should be made using Table R I RS 1. U.

The RM- 11 Channel Number 213 is titilized for the Unit Vent (GT-RE-2 I B) reading for Table R- I. This channel is read out in ).iCi/sec while alt others are read out in iCi/rnl.

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

This IC addresses a release of gaseous radioactivity that results in projected or actcial offsite doses greater than or equal to 10% of the EPA Protective Action Gciicles (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 incicided 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 efficient 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 RG 1.

Callaway Basis Reference(s):

1. EPCI 1402. EAL Table R-l Calculations

2. NEI 99-01, ASL Page 45 of 241 INFORMATION USE

EIP-ZZ-0010l ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Category:R Abnormal Rad Levels / Rad Effluent Subcategory: I Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 rnrern TEDE or 500 mrem thyroid CDE EAL:

RSI.2 Site Area Emergency Dose assessment using actual meteorology indicates doses> 100 rnrem 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 I 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 acttial 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 effltient EALs more fully addresses the spectrum of possible accident events atid 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 effitient monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isoLate the release path. then the effluent monitor reading is no longer valid for classification purposes.

Escalation of the emergency cLassification level would be via IC RGI.

Page 46 of 241 INFORMATION USE

EIP-ZZ-0Ol01 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Callatvay Basis Reference(s):

1. EIP-ZZ-0121 1. Accident Dose Assessment

2. EPCI 1402. EAL Table R-1 Calculations

3. NE199-0LASI Page 47 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Category: R Abnormal Rad Levels / Rad Effluent Subcategory: 1 Radiological Effitient Initiating Condition: Release of gaseous radioactivity resulting in oftstte 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 mm.

Analyses of field stirvey samples indicate thyroid CDE > 500 rnrem for 60 mm. of inhalation.

(Note.c 1, 2)

Note I: 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 onoiniz release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.

MODE Applicability:

Alt 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 raditis of 1,200 meters (3.937 feet) from the midpoint of the Unit I Reactor Building and the canceled Unit 2 Reactor Building.

Control of access to this is by virtue of ownership and in accordance with IOCFR 100.

Basis:

EIP-ZZ-002 1 1, 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 calctilations tising 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 acttial offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and tin-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.

Page 48 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Radiological emiient EALs are also incitided to provide a basis for classifying events and conditions that cannot he 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 l0 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 RG 1.

Callaway Basis Reference(s):

1. EIP-ZZ-002 11. Field Monitoring

2. NEI 99-01. ASI Page 49 of 241 INFORMATION USE

EIP-ZZ-00l0l ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: R Abnormal Rad Levels / Rad Effluent Subcategory: 1 Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in ofisite dose greater than I ,000 rnrem TEDE or 5,000 mrem thyroid CDE EAL:

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

(Notes 1, 2, 3, 4)

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

Note 2. Il 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 classihcation purposes.

Note 4: The pre-calculated effluent monitor values presented in EALs RA I I, RS 1.1 and RG I I should he used for emergency classification assessments until the results trom a dose assessment using actual meteorology are available.

t Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor

[ GE

] SAE Alert

[ UE Unit Vent GT-RE-21 B 6.59E+7 l.ICi/sec 6.59E+6 MCi/sec 6.59E+5 pCi/sec 2 X Hi-Hi alarm AB-RE-li 1/112/

ASD Monitors (A/B/C/D) 12 mR/hr 1.2 mR/hr 0 113/114 w

ID AFW Steam FC-RL-385 163 mR/hr 16.3 mR/hr 1.6 mR/hr Discharge Radwaste Bldg Vent GH-RE-JOB ---- ---- - 2 X Hi-Hi alarm Liquid Radwaste HB-RE-18 2 X Hi-Hi alarm Discharge ---- ----

MODE Applicability:

All Definition(s):

None Page 50 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Basis:

This EAL address gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to site botindary 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-t for a classification, there is a 15 minute time limit to make the classification (RG I I). [f Dose Assessment fURl/RASCAL) is available it should be used instead (RGI .2) since it is more accurate than the values in Table R-1. However the Dose Assessment personnel must he able to calculate results within 15 minutes of the Table R- I value being exceeded OR the classification should be made using Table R- I (RG 1. 1).

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

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

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

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

The TEDE dose is set at the EPA PAG of 1.000 rnrern while the 5.000 nuem 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. AG!

Page 51 of 241 INFORMATION USE

EIP-ZZ-OOlOl ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Category: R Abnormal Rad Levels / Rad Effluent Subcategory: I Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1.000 mi-em 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 BOUiVDARY 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 I Reactor Building and the canceled Unit 2 Reactor Building.

Control of access to this is by virttie of ownership and in accordance with IOCFR 100.

Basis:

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

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

This IC addresses a release of gaseous radioactivity that restilts 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 requtire implementation of protective actions for the ptiblic.

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 asscimes 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.

Page 52 of 241 INFORMATION USE

E[P-ZZ-O0 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Callaway Basis Reference(s):

1. EIP-ZZ-0121 1. Accident Dose Assessment

2. EPCI 1402. EAL Table R-1 Calculations

3. NE199-OLAG1 Page 53 of 241 INFORMATION USE

EIP-ZZ-00I01 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Category: R Abnormal Rad Levels / Rad Effitient Subcategory: 1 Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite close greater than 1.000 rnrern TEDE or 5,000 mrern thyroid CDE EAL:

RGI.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 mm.

Analyses of field survey samples indicate thyroid CDE> 5000 mrem for 60 mm. of inhalation.

(Notes 1, 2) iVote 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 I Reactor Building and the canceled Unit 2 Reactor Building.

Control of access to this is by virtue of ownership and in accordance with IOCFR 100.

Basis:

EIP-ZZ-002 11, Field Monitoring provides gtiidance for emergency or post-accident radiological environmental monitoring (ref. 1).

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

The closed window value is in mR/hr. The analysis of field survey samples is in mrern 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 incltides both monitored and un-monitored releases. Reteases of this magnitude will require implementation of protective actions for the public.

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

Page 54 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCU1VIENT Attachment 1 - Emergency Action Level Technical Bases The TEDE dose is set at the EPA PAG of 1.000 mrem while the 5.000 rnrem 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-0021 I. Field Monitoring

2. NEI 99-01, AGI Page 55 of 241 INFORMATION USE

EIP-ZZ-0010l ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Category: R Abnormal Rad Levels I Rad Effluent Subcategory: 2 Irradiated Fuel Event Initiating Condition: Unplanned loss of water level above irradiated fuel EAL:

RU2.1 Unusual Event UNPLANNED water level drop in the REFUELING PATHWAY as indicated by low water level alarm or indication (EC 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

CT RE 60 Containment High Area Radiation Monitor MODE Applicability:

All Detlnition(s):

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

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

Basis:

The low water level alarm in this EAL refers to the Spent Fitel Pool (SEP) low level alarm (Annunciator 76D, SFP LEV HI LO) (ref. 1). Dtiring 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 Page 56 of 241 INFORMATION USE

EIP-ZZ-00l01 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases 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 oii these monitors in the absence of indications of decreasiiw REFUELING CAVITY level are not classifiable tinder 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 fttel 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 he 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 stich 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-0000 I, Loss of Spent Ftiel Pool/Refuel Pool Level

2. Technical Specification 3.7.15. Ftiel Storage Pool Water Level

3. Technical Specification 3.9.7, Refueling Pool Water Level

4. NEI 99-0 I. AU2 Page 57 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Category: R Abnormal Rad Levels / Rad Effluent Subcategory: 2 Irradiated Fuel Event Initiating Condition: Significant towering 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 ftiel pool and fuel transfer canal comprise the rettieling 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 pooi. 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. I 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 tincovery 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 dtie 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 he 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 RS I.

Callaway Basis Reference(s):

1. OTO-EC-0000 I. Loss of Spent Ftiel Pool/Refuel Pool Level

2. NEI 99-01, AA2 Page 58 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: R Abnormal Rad Levels / Rad Effluent Subcategory: 2 Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to. irradiated fuel EAL:

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

Hi-Hi Alarm on Fuel Building exhaust monitors (GG-RE-27 or 2$).

Manipulator ci-ane radiation monitor (SDRE4 1) >100 mRlhr.

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 fLiel 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 EU 1 I. 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 tC RSl.

Page 59 of 24] INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachnient 1 Emergency Action Level Technical Bases Catlaway Basis Reference(s):

I. OTO-EC-0000l, Loss of Spent Fuel Pool/Refuel Pool Level

2. OTO-KE-0000l, Fuel Handling Accident

3. CaIc. EPCC 98-01. Emergency Action Level Bases

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

5. NEI 99-01. AA2 Page 60 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: R Abnormal Rad Levels / Rad Effluent Subcategory: 2 Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to. irradiated fuel EAL:

RA2.3 Alert Lowering of spent fuel pool level to I 20 as indicated on EC-LI-0059A or EC-LI-0060A.

MODE Applicability:

All Definition(s):

None Basis:

Post-Fukushirna order LA- 12-051 (ref. I) required the installation of reliable SFP level indication capable of identifying normal level (Level 1). SEP level lOft. above the top of the fuel racks (Level 2) and SFP level at the top of the fuel racks (Level 3).

For Callaway Plant SEP Level 2 is plant elevation 2031 ft. I .25 in. (-9 ft. Ii in. above the top of the spent fuel racks) as indicated by 120 on EC-L1-0059A in the Auxiliary Building Hallway 2026. Backup indication is also available on EC-Ll-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 acttial 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 ftiel 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 ftiel pool. This condition reflects a significant loss of spent fuel pool water inventory and thtis 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- 12-5 1, 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 61 of 241 INFORMATION USE

EIP-ZZ-0010l ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: R Abnormal Rad Levels / Rad Effluent Subcategory: 2 Irradiated Fciel 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 I 2 as indicated on EC-LI-0059A or EC-L1-0060A.

MODE Applicability:

All Definition(s):

None Basis:

Post-fukushima order EA- t 2-05 1 t ceI. I) 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 ftiel racks (Level 3).

For Callaway Plant SFP Level 3 has been set at plant elevation 2022 ft. 1 .25 in. (l I in. above the top of the spent fuel racks) as indicated by ] 2 on EC-Ll-0059A in the Auxiliary Building Hallway 2026. Backup indication is also available on EC-LI-0060A in the Atixiliary 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 wotild 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 he via IC RGI or RG2.

CaIlawy 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 62 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emetgency Action Level rfechnicaI Bases Category: R Abnormal Rad Levels I Rad Effluent Subcategory: 2 Irradiated Fuel Event Initiating Condition: Spent fttel pool level cannot he restored to at least the top of the fuel racks for 60 minutes or lonier EAL:

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

(Vote I)

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

MODE Applicability:

All Dell ni tionC s):

None Basis:

PostFukushirna order EA- 12-051 (ref. I ) required the installation of reliable SFP level indication capable of identifying normal level (Level I). SFP level 10 ft. above the top of the fuel racks (Level 2) and SFP level at the top of the fuel racks (Level 3).

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 1-lallway 2026. Backup indication is also available on EC-LI-0060A in the Atixiliary Building hallway 2026.

This EAL addresses a significant loss of spent fuel pooi inventory control and makeup capability leading to a prolonged uncovery of spent fuel. This condition will lead to fuel damage and a radiological telease 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):

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

2. SFP[S Mod Overview for EP

3. NE199-01. AG2 Page 63 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level efechnical Bases Category: R Abnormal Rad Levels / Rad Effluent Subcategory: 3 Area Radiation Levels Initiating Condition: Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown EAL:

RA3.1 Alert Dose rates > 15 mRlhr 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 fCAS). 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 he assessed via local radiation survey for the CAS (ref. I).

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 he applicable.

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

Callaway Basis Reference(s):

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

2. NEt 99-01, AA3 Page 64 of 241 INFORMATION USE

EIP-ZZ-00I0I ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Category: R Abnormal Rad Levels I Rad Effluent Subeategory: 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 followin: (Notc 5)

North Electrical Penetration Room. (Room 1410)

South Electrical Penetration Room. (Room 14C)9)

Note 5: If the equipment in the listed room or area was already inoperable ot oututservice before the event occurred.

then no emcrgenc 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).

UNPL4NNED - 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 tin known.

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 RI-fR in service for RCS cooldown to Cold Shutdown (ief.

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

For St Accumulators:

NGO I BGF3. FDR BKR TO EPfIVS8O8A SI ACC A OUT ISO. (Room 1110)

NGO2BGF3. FDR BKR TO EPHVSSOXB SI ACC B OUT ISO. (Room 1409)

NGOI BGF2. FDR BKR TO EPHV8SOC SI ACC C OUT ISO. (Room 1410)

NGO2BHF2, FDR BKR TO EPHVSSOSD SI ACC D OUT ISO. (Room 1409)

For A RHR:

NGO2BCF2. FDR BKR TO BBPV87O2A RCS LOOP I HOT LEG TO RHR PMPS ISO. (Room 1409)

NGOIBEF2. FDR BKR TO EJHVS7OIA A RHR PMP SUCT FROM RCS HOT LEG I ISO. (Room 1410) for B RHR:

NGO2BBF3. FDR BKR TO BBPV87t)2B RCS LOOP 4 HOT LEG TO RHR PMPS ISO. (Room 1409)

NGO] BDF3. EDR BKR TO EJFIV87t)IB B RHR PMP SUCT FROM RCS HOT LEG 4 ISO. (Room 1410)

Page 65 of 241 INFORMATION USE

EIP-ZZ-0010J ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technica] Bases This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to l)reclclde 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 radiatioti levels and determine if another IC may be applicable.

Alert declaration is warranted if entry into the affected room/area is, or may be, procedtiratly reqttired 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-rotitine 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 roomlarea entry is required is of an administrative or record keeping nattire (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 requited action.

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

Callaway Basis Reference(s):

I. OTG-ZZ-00006 Addendum 06, Securing Safety Injection Accumulators 2 OTN-EJ-t)000l Addendum 3. Placing A RHR Train In Service for RCS Cooldown

3. OTN-EJ-0000l Addendum 4. Placing B RHR Train In Service for RCS Cooldown

4. NET 99-01, AA3 Page 66 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases 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 ftiel and other radioactive materials associated with spent fttel storage. A significant amount of the radioactive material contained within a canister must escape its packaging and enter the biosphere for there to he 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 67 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases 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 rnremlhr (gamma + neutron) on the side of the Transfer Cask.

MODE Applicability:

All Definition(s):

COiVEINEMENT 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 overpacklVVM VERTIQAL 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 Caltaway when the Mtilti-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 overpacklVVM 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 overpacklVVM or transfer cask (ref. I).

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.

Page 68 of 241 INFORMATION USE

EIP-ZZ-00l01 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases 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 RU I, 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 he determined based on measurement of a dose rate at some distance from the cask, Security-related events for ISESIs are covered tinder ICs HG 1 and HS 1.

Callaway Basis Reference(s):

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

2. NIl 99-01. E-HUI Page 69 of 241 INFORMATION USE

EIP-ZZ-0010I ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Category C Cold Shutdown / Refueling System Malfunction EAL Group: Cold Conditions (RCS temperature 200°F); EALs in this category are applicable only in one or more cold operating MODES.

Category C EALs are directly associated with Cold Shutdown or refueling system safety functions. Given the variability of plant configurations (e.g., systems out-of-service for maintenance, containment open, reduced AC power redundancy, time since shutdown) during these periods, the consequences of any given initiating event can vary greatly. For example. a loss of decay heat removal capability that occurs at the end of an extended outage has less significance than a similar loss occurring during the first week after shutdown. Compounding these events is the likelihood that instrumentation necessary for assessment may also he inoperable. The Cotd 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 Deluded).

The events of this category pertain to the following subcategories:

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

2. Loss of Emereency AC Power Loss of emergency plant electrical power can compi-omise plant safety system operability including decay heat removal and emergency core coc)liflg 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 fitnctions.

4. Loss of Vital DC Power Loss of emergency plant electrical power can compromise pliit 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 70 of 241 INFORMATION USE

EIP-ZZ-0010l ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: C Cold Shutdown I Refueling System Malfunction Subcategory: 1 RCS Level Initiating Condition: UNPLANNED loss of RCS inventory for 15 minutes or longer EAL:

CULl Unusual Event UNPLANNED loss of reactor coolant results in RCS water level less than a required lower limit for

? 15 mm.

(Now 1)

Note I: The Emergency Coordinator should declare the event promptly tipon determining that time limit has been exceeded, or will bk-ely be exceeded MODE Applicability:

5 - Cold Shtitdown. 6 - Reftteling Definition(s):

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

Basis:

With the plant in Cold Shutdown. RCS water level is normally maintained above the pressurizer low level setpointof 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 RU) 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.

Page 71 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases This EAL recognizes that the minimum required RCS level can change several times during the course of a refueling outage as different plant configttrations and system lineiips 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 15minute threshold dtiration allows sufficient time for prompt operator actions to restore and maintain the expected water level. This criterion excludes transient conditions causing a brief lowering of water level.

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

Callaway Basis Reference(s):

I. FR-1.2. Response to Low Pressurizer Level

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

3. Technical Specification 3.9.7. Refueling Pool Water Level

4. NET 99-01, CUI Page 72 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Category: C Cold Shtitdown / Refueling System Malfunction Subcategory: I RCS Level Initiating Condition: UNPLANNED Loss of RCS inventory for 15 minutes or longer EAL:

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

UNPLANNED iticrease in any Table C-i sump/tank level Utie to loss of RCS inventory.

Visual observation of UNISOLABLE RCS leakage.

L

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

LJNISOL4BLE - An open or breached system line that cannot be isolated. remotely or locally.

UNPL4NNED A parameter change or an event that is not I) the result of an intended evolution or 2) an expected plant response to a transient. The catise of the parameter change or event may be known or u nkno w ii.

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.

Page 73 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases In this EAL, all water level indication is tinavailable and the RCS inventory loss must be detected by indirect leakage indications. Level increases must be evaluated against other potential socirces of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage. If the make-tip rate to the RCS unexplainably rises above the pre-established rate, a loss of RCS inventory may be occtirring 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 crticial 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 he 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 procedtirally required limit warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the core c overecl.

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-i). Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RCS.

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

Callawav Basis Reference(s):

I. 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, CUI Page 74 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: C Cold Shutdown / Refueling System Malfunction Subcategory: I 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 (RVLJS 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. I ), 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 he based on CA 1.2 if RCS inventory cannot he monitored.

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

Elevation of bottom of Reactor Vessel (H) A 1987.150 Elevation of bottom ID of RCS hot leg penetration (tt) 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 (3/4) 8.949 Bottom ID of RCS ioop, including channel uncertainties: H ÷ OCA (3/4) 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 occttrred. 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 ptant safety.

Page 75 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases For this EAL a lowering of RCS water level below the specified level indicates that operator actions have not beeii successful in restoring and maintaining RCS water level. The heat-tip rate of the coolant will increase as the available water inventory is reduced. A continuing decrease in water level wilt lead to core ii nc o very.

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 CS I.

Callaway Basis Reference(s):

I. 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. CAl Page 76 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: C Cold Shtttdown I Refueling System Malfunction Subcategory: I RCS Level Initiating Condition: Loss of RCS invernory EAL:

CA1.2 Alert RCS water level cannot he monitored for 15 mm. iVote 1)

AND EITHER

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

Visual observation of UNISOLABLE RCS leakage.

.Vni 1: The Emergenc) Coordinator should declare the event promptlr upon determining that time limit has been exceeded. or vill likely he eweeded.

Table C-i Sumps / Tanks

Containment Sumps

Containment Normal Sumps

Containment Instrument Sump

PRT

RCDT

Auxiliary Building Sump MODE Applicability:

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

RCS INTACT The RCS should be considered intact when the RCS pre sure 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 evolittion or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or in known.

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.

Page 77 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases 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-I). Stirveillance procedtires provide instrLictions for calculating primary system leak rate by manual or comptiter-based water inventory balances.

Level increases must he 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 he isolated coctld also he indicative of a toss 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 harrier). This condition represents a potential sutbstantial 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 poteitial sources of water flow to ensure they are indicative of leakage from the RCS.

The 15-minute dtiration 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 he via IC CS I.

Callaway Basis Reference(s):

I. OTO-BB-00003-R0l4. Excess RCS Leakage

2. OSP-BB-00f)09, RCS Inventory Balance

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

4. NE199-0J,CAI Page 78 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: C Cold Shutdown / Refueling System Malfunction Subcategory: I RCS Level Initiating Condition: Loss of RCS inventory affecting core decay heat removal capability EAL:

CS1.1 Site Area Eniergency With CONTAINMENT CLOSURE not established, RVLIS Ptimps Off< 72çc.

MODE Applicability:

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

COVT4INMENT 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-GT00003, 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 monitoritig instruments RB LI-53A1B and Computer Point BBLOO53BB cannot sense tevel 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 dratning operations (ref. 3). When RVLIS is out of service. classification should be based on CS 1.3 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 (tt) B 2013.290 Six inches below hot leg penetration (above vessel bottom) C B A 0.5 (ft)

- - 25.640 Height of vessel D (It) 41 .245 RVLIS indication corresponding to the top of the core: H = 100 x C! D (3/4) 62.165 RVLIS overall channel accuracy: OCA = 7.48% + (0.0104 x H) ÷ 0.81%

OCA at H (3/4) 8.937 Six inches below Bottom ID of RCS loop, including channel uncertainties: H + OCA (3/4) 71 .102 Rounded upward to nearest 1% (RVLIS range is 0 120% in 2% increments)

- 72 Page 79 of 241 INFORMATION USE

EIP-ZZ-00l0l ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Cinder 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 dtie to a vessel breach, RCS pressure botitidary 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 tincovery. The inability to restore and maintain level after reaching this setpoint infers a failure of the RCS harrier and Potential Loss of the Fuel Clad harrier.

The status of Containment clostire is tracked if plant conditions change that could raise the risk of a fission prodtict release as a result of a loss of decay heat removal (i-ef. 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 clue to a RCS component failure.

a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failtires of plant functions needed for protection of the public and thus warrant a Site Area Etnergency 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/shtitdown 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 CS I I and CS 1 .2 reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission prodtict release to the environment.

This EAL addresses concerns raised by Generic Letter 88-17. Loss of Decay Heat Removal: SECY 9 1-283, Evaluation of Shutdown and Low Power Risk Issues: NUREG-l449. Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States: and NUMARC 9t-06, Guidelines for Industry Actions to Assess Shtitdown Management.

Escalation of the emergency classification level would be via IC CGI or RGI.

Callaway Basis Reference(s):

I. OOA-BB-00003, Refuel Level Indications

2. Calculation No. BB-l77, (387.1 - CAL RVLIS Setpoints)

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

4. OSP-GT-00003, Containment Closure

5. NE199-Ol,CSI Page 80 of 241 INFORMATION USE

EIP-ZZ-00I01 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: C Cold Shutdown / Refueling System Malfunction Suhcategory: I RCS Level Initiating Condition: Loss of RCS inventory atfecting core decay heat removal capability EAL:

CS 1.2 Site Area Emergency With CONTAINMENT CLOSURE established. RVLIS Pumps Off< 65 (Top of Fuel).

MODE Applicability:

5 - Cold Shutdown. 6 - Reftieling Definition(s):

CONTAINMENT CLOSURE The procedurally defined conditions or actions taken to sectire 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 Clostire is established when the requirements of OSP-GT00003, Containment Closure are met.

Basis:

When Reactor Vessel water level drops below RVLIS Pumps Off indication of 65cc (201t).29 ft.). core uncovety is about to occur. The Plant Computer System Display called Refuel Level Indications (turn on code RLD is available [C) assist in inorntoring important parameters crucial to RCS draining operations (ref. 3). When RVLIS is out of service, classification should he based on CS 1.3 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 1987.750 Elevation of top of liel (if) B 2010.290 Height of top of core (above vessel bottom) C B -A (if) 23.740 Height of vessel 0 (if) 41.245 RVLIS indication corresponding to the top of the core: H = 100 x Cl D (3/4) 56.104 RVLIS overall channel accuracy: OCA = 7.48% + (0.0104 x H) + 0.87%

OCA at H (%) 6.873 Top of core, including channel uncertainties: H + OCA (%) 64.977 Rounded upward to nearest 1% (RVLIS range is 0- 120% in 2% increments) 65 Page 81 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases 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 dtie to a vessel hi-each, 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 he capable of preventing further RCS or Reactor Vessel water level drop and potential core lincovery. The inability to restore and maintain level after reaching this setpomt 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 ftiel 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 CS 1.1 and CS 1 .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 9 1-06, Guidelines for Indtistry Actions to Assess Shutdown Management.

Escalation of the emergency classification level would he via IC CG I or RG 1.

Callaway Basis Reference(s):

I. OOA-BB-00003, Refuel Level Indications

2. Calculation No. BB-l77. (387.1 - CAL RVLIS Setpoints)

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

4. OSP-GT-000t)3, Containment Closure

5. NEI 99-01. CSI Page 82 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 ETvIERGENCY ACTION LEVEL TECHNICAL BASES DOCUPvIENT Attachment 1 Emergency Action Level Technical Bases Category: C Cold Shutdown / Refueling System Malfunction Subcategory: I RCS Level Initiating Condition: Loss of RCS inventory affecting core decay heat removal capabilit EAL:

CS 1.3 Site Area Emergency RCS water level cannot be monitored for> 30 mm. (Nt)te 1)

AND Core uncovery is indicated by any of the following:

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

Manipulator crane radiation monitor SD-RE4 I > 10,000 mR/hr.

Erratic Source Range Monitor indication.

Vow 1: The Eniergencv Coordinator should declare the event promptly upon determining that time limit has been exceeded. or vill lilelv he exceeded.

Table C-i Sumps / Tanks

Containment Sumps

Containment Normal Sumps

Containment Instrument Sump

PRT

RCDT

Auxiliary Building Sump NIODE Applicability:

5 - Cold Shutclown. 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 circtilation cooling.

UNPL4NA1ED 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 catise 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.

Page 83 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases 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-l). Surveillance procedures provide instructions for calculating primary system leak rate by manual or computer-based water inventory balances.

Level increases must he evaluated against other potential sources of leakge 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 he 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 close rate above the core will rise. The dose rate due to this core shine should result in upscaled manipulator crane radiation monitor (SD-RE-4l) imlication (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 asa 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 he restored, ftiel 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 acttially occurred (i.e., to account for various* ccideit 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 instrttmentation 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. Stimp and/or tank level changes must be evaltiated against ether 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 9 1-06. Guidelines for Industry Actions to Assess Shutclown Management.

Escalation of the emergency classification level would be via IC CGI or RGI Page 84 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Catlatvay Basis Reference(s):

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

2. OSP-BB-00009. RCS [nventorv Balance

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

4. FSAR. Section 12.3.3.4

5. FSAR. Table 12.3-2

6. CaIc. 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. Nticlear Safety Analysis Center (NSAC). 1980. Analysis of Three Mile tsland - Unit 2 Accident, NSAC-1

9. NE! 99-01. CSI Page 85 of 241 INFORMATION USE

EIP-ZZ-OOtOl ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachnient 1 Emergency Action Level Technical Bases Category: C Cold Shutclown / Reftteling System Malfunction Subcategory: I RCS Level Initiating Condition: Loss of RCS inventory affecting fuel clad integrity with containment chatlenged EAL:

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

AND Any Containment Challenge indication. Table C-2.

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

Note 6: It CONTAINMENT CLOSURE is re-established prior to exceedine 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 procedtirally defined conditions or actions taken to secure Primary or Secondary Containment and its associated structures, systems, and components as a fctnctional barrier to fission product release under shutdown conditions.

As applied to Caltaway, Containment Closure is established when the requirements of OSP-GT-00003 Containment Closttre ate met.

Basis:

When Reactor Vessel water level drops below RVLIS Pttmps 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 crticial to RCS draining operations (ref. 3). When RVLIS is octt of service, classification should he based on CG 1.2 if RCS inventory cannot be moni toted.

Page 86 of 241 INFORMATION USE

E[P-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases The RVLIS Pumps Off threshold has been determined as follows (ref. 1.2):

Elevation of bottom of Reactor Vessel (fl)A 1987.150 Elevation of top ofiel (ft) B 207 0.290 Height oftop of core (above vessel bottom) C = B - A (if) 23.740 Height of vessel D (if) 41.245 RVLIS indication corresponding to the top of the core: H = 100 x Cl D (3/4) 56.104 RVLIS overall channel accuracy: OCA = 7.48% + (0.0704 x H) ÷ 0.81%

OCA at H (%) 8.873 Top of core, including channel uncertainties: H + OCA (%) 64.977 Rounded upward to neatest 1% (RVLIS range isO- 120% in 2% increments) 65 Three conditions are associated with a challenge to Containment integrity:

CONTAINMENT CLOSURE not established The status of Containment closure is tracked if plant conditions change that cotild raise the risk of a fission product release as a result of a loss of decay heat removal (ref. 4). 11 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 1-ICS which serves to limit or reduce combtistihle gas concetitrations in the Containment. The HCS is an engineered safety feature with redtmdant 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 AT-JO and GS A1-19) with a range of 0% to 10% provide indication on Control Room Panel RLO2O and Emergency Response Facilities Information System (ERFIS) (ref. 6.

7). The hydrogen monitors require a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 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).

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 dtie to a vessel breach. RCS pressure boundary leakage or contintied 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.

Page 87 of 241 INFORMATION USE

EIP-ZZ-00l01 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases 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 sttbstantial 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 tmrnonitored 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 stipport a hydrogen hum (i.e.. at the tower 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 (Tcthle C-2, Contctinmeni Challenge Ijic/kations).

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 acttially 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 he catised 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 9 1-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG- 1449, Shtitdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 9 1-06, Guidelines for Indtistry Actions to Assess Shutdown Management.

Callaway Basis Reference(s):

I. OOA-BB-00003. Refuel Level Indications

2. Calculation No.33-177. (387. 1 - CAL RVLIS Setpoints)

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

4. OSP-GT-00003, Containment Closure Page 88 of 241 INFORIIATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases

5. FSAR. Section 6.2.5

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

7. Technical Specifications 3.3.3

8. OTN-GS-0000t. Containment Hydrogen Control System

9. CaIc No. 392.2 XX-95 Callaway Containment Parameters EOP Action Values. Setpoint ID T1t)l & T102

10. NEI 99-01. CSI Page 89 of 24] INFORMATION USE

EIP-ZZ-00l0I ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: C Cold ShLttdown / Refueling System Malfunction Subcategory: I RCS Level Initiating Condition: Loss of RCS inventoty affecting fuel clad integrity with containment challenged EAL:

CG1.2 General Emergency RCS level cannot be monitored for> 30 miii. (Note I)

AND Core uncovery is indicated by any of the following:

UNPLANNED increase in any Table C-I sump/tank level of sufficiern magnitude to indicate core uncovery.

Manipulator crane radiation monitor SDRE4 I > I 0.000 mRlhr.

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

Table C-i 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 90 of 241 INFORMATION USE

EIP-ZZ-001OI ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Definition(s):

CONTAJyVMEiVT CLOSURE The procedurally defined conditions or actions taken to secttre 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 I) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

Basis:

In Cold Shutdown MODE, the 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-I ). Surveillance procedures provide instructions for calculating primary system leak rate by manual or computer-based water inventory balances.

Level increases must he evaluated against other potential sources of leakage suich 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 he 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 tref. 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).

Three conditions are associated with a challenge to Containment integrity:

1. CONTAINMENT CLOSURE not established - The status of Containment clostire 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 cincovery time limit then escalation to GE would not occur.

Page 91 of 241 INFORMATION USE

EIP-ZZ-00l01 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases

2. Containment hydrogen 4% The 4% hydrogen concentration threshold is generally considered the lower limit for hydrogen deflagrations. CaB away is equipped with a Hydrogen Control System tHCS) which serves to limit or reduce combustible gas concentrations in the Containment. The KCS is an engineered safety feature with redundant hydrogen recombiners. hydrogen mixing system.

hydrogen monitoring subsystem. and a hackLip hydrogen purge subsystem. The HCS is designed to maintain the Containment hydrogen concentration below 4% by volume (ref. 9). Two Containment hydrogen monitors (GS Al-lO and GS A1-19) with a range of 0% to 10% provide indication on Control Room Panel RLO2O and ERf IS (ref. 10, 11). The hydrogen monitors require a 2 hotir 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 Shcitdown or Refueling MODES can threaten Containment Closure capability and fluis Containment potentially cannot he relied upon as a barrier to fission product release (ref. 15).

This [C 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 he 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 t-elease 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 detlagration 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 tmcovery event, it is unlikely that hydrogen buildup dtte 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 he 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.

Page 92 of 241 INFORMATION USE

EIP-ZZ-00l0l ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases The inability to monitor RCS level may he caused by instrumentation and/or power failures, or water level dropping below the range ol 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. ESAR. Table 12.3-2

6. CaIc. No. HPCI -t)70l, 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), 1 980. 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-0000I, Containment Hydrogen Control System

13. CaIc No. 392.2 XX-95 Callaway Containment Parameters EOP Action Values. Setpoint ID TIOl & T102

14. OSP-GT-00003, Containment Closure

15. NEI 99-01. CGI Page 93 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: C Cold Shutdown I Refueling System Maiftinction Subcategory: 2 Loss of Emergency AC Power Initiating Condition: Loss of all hut one AC power soctrce to emergency buses for 15 minutes or longer EAL:

CU2.1 Unusual Event AC power capability, Table C-3, to emergency 4. 16KV buses NB0 I and NBO2 reduced to a single power source for? 15 mm. (Note 1)

AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS.

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

Table C-3 AC Power Sources Offsite:

Safeguards XFMR A or B via ESF LTC XFMR XNBO1

Startup XFMR XMRO1 via ESF LTC XFMR XNBO2

Main XFMR XMAO1 backfed via UAT XFMR XMAO2 (in-service)

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

Onsite:

EDGNEO1

EDGNEO2 MODE Applicability:

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

SAFETY SYSTEiVJ 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-i-elated (as defined in IOCFR5O.2):

Those structures, systems and components that are relied tipon 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 94 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Basis:

For emergency classification purposes. capability means that an offsite AC power source(s) is available to the emergency buses and can he 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 haekfeed 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 fot- 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 he 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 ci 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 NBOI and NBO2 are the emergency (essential) buses. NBOI supplies power to Load Group I (Red Train) safety related toads and NBO2 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 XNBO I and the other source is from the startup transformer XMROI via ESF LTC transformer XNBO2. Transformer XNBOI is the normal supply to bus NB0l XNBO2 is the normal supply to bus NBO2 (ref. 1,2.3).

In addition. NBO I and NBO2 each have an emergency diesel generator which supply electrical power to the btis automatically in the event that the preferred source becomes unavailable (ref. I).

Another method to obtain offsite potver is by backfeeding the emergency btises through the main transformer XMAOI and unit auxiliary transformer XMRO2. This is only done during Cold Shutdown unless nuclear safety considerations require it to he 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 he aligned within 15 minutes.

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

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 safetyrelated 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. Thuts. when in these MODES, this condition is considered to he a potential degradation of the levet of safety of the plant.

Page 95 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases An AC power source is a source recognized in AORs 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 hack-fed from the tinit main generator.

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

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

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

Callaway Basis Reference(s):

I . E-2 1001(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-0000 I -Roll. Main Step-Up Transformer Backfeed - IPTE

5. NEI 99-01. CU2 Page 96 of 241 INFORMATION USE

EIP-ZZ-00l0l ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: C Cold Shutdown / Refueling System Malfunction Subcategory: 2 Loss of Emergency AC Power Initiating Condition: Loss of all offsite and all onsite AC power to emergeticy buses for 15 minutes or Ion ter EAL:

CA2.1 Alert Loss of all offsite and all onsite AC power capability. Table C-3. to emergency 4.16KV buses NBO1 and NBO2 for> 15 mm. (Note I) 1VOtL 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has heen exceeded. or will lite!v he exceeded.

Table C-3 AC Power Sources Offsite:

Safeguards XFMR A or B via ESF LTC XFMR XNBOJ

Startup XFMR XMRO1 via ESF LTC XFMR XNBO2

Main XFMR XMAO1 backfed via UAT XFMR XMAO2 (in-service)

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

Onsite:

EDO NEO1

EDG NEO2 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 poer source.

The criteria for Standby Alignment is that the source can he supplying the station with power within 15 minutes. Obviously the Main Transformer could not be aligned for hackfeed 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 he 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 NBO1 and NBO2 (ref. 1).

Page 97 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases 4.16KV buses NBOI and NBO2 are the emergency (essential) buses. NBOI supplies power to Load GroLip 1 (Red Train) safety related loads and NBO2 supplies power to Load Group 2 (YelLow Train) safety related loads. Each bus has two sources ot ofisite power. One source is from 13.8 KV safeguards transformer A or B via ESF Load Tap Changing tLTC transformer XNBO 1 and the other source is from the startup transformer XMROI via ESF LTC transformer XNBO2. Transformer XNBOI is the normal supply to btms NBOI: XNBO2 is the normal supply to btis NBO2 (ref. 1,2,3).

In addition, NBOI and NBO2 each have an emergency diesel generator which stipply 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 XMAO I and unit auxiliary transformer XMRO2. This is only done during Cold Shutdown unless nuclear safety considerations reqtiire it to be done during hot shutdown when no other power sotirces 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 SS I 1.

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

When in the Cold Shutdown, refueling, or defueled MODE, this condition is not classified as a Site Area Emergency because of the increased time available to restore an 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 stibstantial 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 he via IC CSI or RS1.

Callaway Basis Reference(s):

E2 1001(Q) Main Single Line Diagram (Electrical Distribution Diagram)

2. FSAR Site Addenda, Section S.2.1.2

3. FSAR, Section 8.3.1

4. OTS-MA-00001-R01 t, Main Step-Up Transformer Backfeed - IPTE

5. NEI 99-01, CA2 Page 98 of 241 INfORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Category: C Cold Shutdown / Refueling System Malfunction Subcategory: 3 RCS Temperature Initiating Condition: UNPLANNED increase in RCS temperature EAL:

CU3.1 tJnusual Event UNPLANNED increase in RCS temperature to > 200°F.

(Note 10)

Now 10: Beinn monitorinC hot condition EALs concurrently for an new event or condition not related to the Itos e[

decay heat removal.

MODE Applicability:

5 - Cold Shutdown. 6 - Refuelinti Definition(s):

RCS INT4CT The RCS should be considered intact when the RCS pressure boundary is in its normal condition for the Cold Shutdown MODE of operation C 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 nattiral circulation cooling.

t]NPL4NNED A parameter change or an event that is not I ) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may he known or unknown.

Basis:

Several instruments are capable of providing indication of RCS temperature with respect to the Technical Specification Cold Shutdown temperature limit t200°F. ref. I ). 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 he used. For example. the highest valid reading temperature indication should he used (ref. 2. 3. 4).

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

This IC addresses an UNPLANNED increase in RCS temperattire above the Technical Specification Cold Sluitdown temperattire 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.

A momentary UNPLANNED excursion above the Technical Specification Cold Shutdown temperature limit when the heat removal function is available DOES NOT warrant a classification.

Page 99 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases 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 he 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 toss of forced decay heat removal at reduced inventory may result in a rapid increase in reactor coolant temperature depending on the time after shutdown.

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

Catlaway Basis Reference(s):

I. Callaway Technical Specifications. Table 1 .1 -

2. OTG-ZZ-0000 I, 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 72.2.3.2

5. NEI99-01,CU3 Page 100 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Category: C Cold Shutdown / Refueling System Malfunction Subcategorv: 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 mm.

(iVote 1)

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

MODE Applicability:

S - 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 RLOI8:

BB Ll-53A. RCS (LOOP I) HOT LEG LEV BB L1-53B. RCS (LOOP 4) HOT LEG LEV

Computer points:

BBLOO53A, RCS LOOP I HOT LEG LEVEL BBLOO53B, RCS LOOP 4 HOT LEG LEVEL BBLO53BB. RCS LOOP LEVEL - CTMT VENTED

RVLIS BB Lt-l31 1. BB LI-1312. BB Ll-132l. 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 Reftiel Level Indications (turn on code RLI) is available to assist in monitoring important parameters crucial to RCS draining operations (ref. 3).

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 he used. for example. the highest valid reading temperature indication should be used (ref. 6, 7, 8).

Page 101 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases 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 durin 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 wottid be unable to monitor key parameters necessary to assure core decay heat removal. Dtiring this condition, there is no immediate threat of fuel damage because the core decay heat toad has been reduced since the cessation of power operation.

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

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

Callaway Basis Reference(s):

1. Callaway Technical Specifications Table 1 .1 -

2. OOA-BB-t)0003. Refuel Level Indications

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

4. FSAR, Section 18.2.13.2

5. OTS-KE-000 18, Draining the Refueling Pool

6. OTGZZ-0000l, Plant Heatup Cold Shutdown to Hot Standby

7. OSP-BB-00007, RCS Heattip and Cooldown Limitations, Note before Section 6. I and Attachment 2

8. FSAR, Section 7.2.2.3.2

9. NE! 99-01, CLT3 Page 102 of 241 INFORMATION USE

EIP-ZZ-0010l ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Category: C Cold Shtitdown / Refueling System Malfunction Subcategorv: 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), 10)

OR UNPLANNED RCS pressure increase > 10 psig. (This EAL does 110! ctpp/v diiriiiç water-solid plant cc),Iclttions)

Note 1: The Eniereencv Coordinator should declare the event prt.mptly upon determining that the applicable time has been exceeded. or will likely he exceeded.

Note 10: Begin monitoring hot condition EALs concurrently for any new event ot condition not related to the loss ut 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 mm. .

REDUCED INVENTORY)

RCS Not INTACT established 20 min.*

OR I REDUCED INVENTORY J not established 0 mm.

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 Deli nition(s):

CONTAINtiENT CLOSURE The procedurally defined conditions or actions taken to secure Primary or Secondary Containment and its associated structures. systems. and components as a functional harrier 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 ShiLtdown 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 nattiral circulation cooling.

Page 103 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases 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 he known or unknown.

RED LICED 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 t< 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 incltide core exit thermocouples (TICs) 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 he used (ref. 2. 3. 4).

RCS pressure instrument BB P1-403A is capable of measuring pressure to less than It) psig (ref. 5).

In the absence of reliable RCS temperattire 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 degradatioti of the level of safety of the plant.

A momentary UNPLANNED excursion above the Technical Specification Cold Shutdown temperattire limit when the heat removal function is available DOES NOT warrant a classification.

The RCS Heat-tip 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 presstire harrier 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 I ) the evaporated reactor coolant may he 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-tip in the absence of RCS temperature monitoring capability.

Escalation of the emergency classification level would be via IC CS 1 or RS I.

Page 104 of 241 INFORMATION USE

ELP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Callaway Basis Reference(s):

Callaway Technical Specifications. Table I. I-I

2. OTG-ZZ-000t) 1. 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-0t)006. Plant Cooldown Hot Standby To Cold Shutdown

6. NEI 99-01. CA3 Page 105 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: C Cold Shutdown / Refueling System Malfunction Subcategory: 4 Loss of Vital DC Power Initiating Condition: Loss of Vital DC power for 15 minutes or longer EAL:

CU4.l Untisual Event

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

> 15 mm.

(Note 1)

No 1: The Emerency 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 ol 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 toss. The fifteen minute interval is intended to exclude transient or momentary power losses.

As used in this EAL, required means the vital DC btises 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 inservice (operable), then a loss of Vital DC power affecting Train B (NKO2 or NKO4) wotild 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 I E buses (ref. I):

Division 1: Division 2:

NKO1 NKO2 NKO3 NKO4 There are fotir battery banks (NKI I. NK 12. NK 13 and NKI4) 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.

Page 106 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCtJMENT Attachment 1 Emergency Action Level Technical Bases Due to the load distribution on each of the I 25VDC buses, the four batteries for each bus do not have the same rating. All four of the I 25VDC buses supply inverters for I 2OVAC NN bus poer as well as control power for various safety related systems. NKO1 and NKO4 stipply additional DC loads such as diesel field flashing. breaker control power. main control hoard power and emergency lighting. These loads are not supplied by the other two buses, NKO2 and NKO3. For this reason. batteries NK1 I and NKI4 require additional capacity. Each battery is designed to have stifficient stored energy to supply the required ernergeticy loads for 240 minutes following a loss of AC power (station blackout) (ref. 2. 3, 4).

Minirntim DC bus voltage is 107.0 VDC (ref. 4, 5). Bus voltage may be obtained from the following instruments (ref. 6):

NKEI-l (NKOI)

NK ET-2 (NKO2)

NK EI-3 NKO3)

NKEI-4NKO1)

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. anti 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 CAl 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-I. -2. -3

3. FSAR, Section 8.3.2

4. Calculation NK-lO. NK System DC Voltage Drop

5. FSAR. Table 8.3A- I [t[.B

6. ECA-0.0. Loss of All AC Power

7. NET 99-01, CU-I Page 107 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: C Cold Shutdown / Refueling System Malfcinction Subcategory: 5Loss of Communications Initiating Condition: Loss of all onsite or offsite communications capabilities EAL:

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

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

OR Loss of all Table C-S 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________

Sentry Notification System X MODE Applicability:

S - Cold Shutdown, 6 - Refueling, D Defueled Detmnition(s);

OFESITE RESPONSE ORGAiVIZATIONS (ORO) The State of Missouri (SEMA/MIAC). Callaway County 91 l/EOC. Gasconade County 91 1/EOC, Montgomery County 91 l/EOC and Osage County 91 IIEOC.

Page 108 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Basis:

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

Gaitronics system The Gaitronics system provides six separate independent communication channels--one general page.

one Control Room page and four party tines. Cot;mtinication between parties within the ptant can he 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 tine channel is normally used after the page catt is completed. As many as focir party lines may communicate sinuiltaneously. The portion of the PA system connecting the fuel transfer area in the Containment, the spent fuel area and new ftiet 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 tise of the fuel handling communications system without disruption to the remainder of the system.

2. Plant Radios A six channel Xt)0 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 radioequipped vehicles and plant handheld portable radios. These systems are for use during normal operation or dtiring 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 indeJ)endently 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 pait of the Plant Telephone System.

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.

Page 109 of 241 INFORMATION USE

EIP-ZZ-0010I ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases

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 hack-up notification of offsite agencies and the coordination of offsite activities during a radio logic al emergency. The system uses $00 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 he initiated through the Callaway County/City of Fulton EOC via the Security radio.

7. Sentry Notification System A computerized notification system linked between the Catlaway 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 simttltaneoiisly. Notifications on Sentry 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 5U7. I.

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. individtials 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. NE199-t)l,CU5 Page LI 0 of 241 INFORMATION USE

EIP-ZZ-0010l ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: C Cold Shutdown / Refueling System Malfunction Subcategory: 6 Hazardous Event Affecting Safety Systems Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating MODE EAL:

CA6.1 Alert The occurrence of any Table C-6 hazardous event AND EITHER:

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

The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating MODE.

Table C-6 Hazardous Events

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 Appticability:

5 - Cold Shutdown. 6 Reitteling 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 circciits. grounding. arcing. etc.) should not automatically he 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 hut 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 mIs) sustained, or58 mph (26 m/s) gusting.

Page Ill of 241 INFORMATION USE

EIP-ZZ-DO1O1 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases SAFETY SYSTEM A system tequired for safe plant operation, cooling down the plant and/or placing it in the Cold Shutdown condition. including the ECCS. These are typically systems classified as safety-related (as defined in IOCFR5O.2):

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

The integrity of the reactor coolant pressure boundary;

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

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

VISIBLE DAMAGE Damage to a component or structure that is readily observable without measurements.

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

Basis:

Annunciator 98D. OBE will illuminate if the seismic instrument detects ground motion in excess of the OBE threshold. OTO-SG-0000l, Seismic Event provides the guidance for determining if an OBE earthquake threshold is exceeded and any required response actions (ref. I).

Internal FLOODING may he catised by events suich as component failures, equipment misalignment.

or outage activity mishaps (ref. 2).

External flooding may be due to high rainfall. Callaway plant grade elevation is 810.0 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 tinder this EAL.

This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure containing SAFETY SYSTEM components, needed for the current operating MODE. This condition significantly reduces the margin to a loss or potential loss of a fission produict barrier, and therefore represents an actual or potentiat stthstantial degradation of the level of safety of the plant.

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

The second conditional addresses damage to a SAFETY SYSTEM component that is not in service/operation or readily apparent through indications alone, or to a structure containing SAFETY SYSTEM components.

Operators will make this determination based on the totality of available event and damage report information. This is intended to he a brief assessment not requiring lengthy analysis or quantification of the damage.

Escalation of the emergency ctassification level would be via IC CS 1 or RS 1.

Page 112 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases 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 113 of 241 INFORMATION USE

EIP-ZZ-0010l ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Category H Hazards and Other Conditions Affecting Plant Safety EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)

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

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

2. Seismic Event Natural events sttch as earthquakes have potential to cause plant strticture 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 Nonnaturally 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 uipon Emergency Coordinator judgment.

Page 114 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: H Hazards Subcategory: I Security Initiating Condition: Confirmed SECURITY CONDITION or threat EAL:

HULl 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 providine information of an aircraft threat.

MODE Applicability:

All Definition(s):

SECURIT} CONDITIOiV 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. I).

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. Sectirity events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR 73.71 or 10 CFR 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HAl, HS 1 and HG 1.

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.

Page 115 of 241 INFORMATION USE

EIP-ZZ-00I01 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Security plans and terminology are based on the guidance provided by NEI 03-12. Template for the Sectirity Plan. Training and Qualification Plan, Safeguards Contingency Plan.

The first threshold references the Shift Security Stipervisor because these are the individuals trained to confirm that a security event is occurring or has occurred. Training on sectirity event confirmation and classification is controlled due to the nature of Safeguards and JO 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 commtinicate to the licensee if the threat involves an aircraft. The status and size of the plane may also he 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 Securitysensitive information. This includes information that may be advantageous to a potential adversary. such as the particulars concerning a specific threat or threat location. Securitysensitive 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 he via IC HAl.

Callaway Basis Reference(s):

Callaway Plant Security Plan and DBT (Safeguards)

2. EIP-ZZ-SKOOI, Response tO Security Threat

3. S DP-CP-00003, Security Contingency Events

4. OTO-SK-00002. Plant Security Event Aircraft Threat

5. NE199-0l. HUI Page 116 of 241 INFORMATION USE

EIP-ZZ-00I0I ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: H Hazards Stibcategory: I Security Initiating Condition: HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes EAL:

HAl.! Atert A HOSTILE ACTION is occurrin% 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 mm. 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. oi- other devices used to deliver destrttctive force.

Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts ot 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 olTicial 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 NEJ 03-12, Template for the Security Plan, Training and Qualification Plan. Safegtiards Contingency Plan.

Page 117 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases 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 EAL, 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 threatrelated information has been validated in accordance with sitespecific 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 Securitysensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Securitysensitive information should be contained in nonpublic documents such as the Callaway Plant Security Plan and DBT (ref. I).

Callaway Basis Reference(s):

I. Callaway Plant Security Plan and DBT (Safeguards)

2. EIP-ZZ-SKOOI, Response to Sectirity Threat

3. S DP-CP-00003, Security Contingency Events

4. OTO-SK-00002, Plant Security Event - Aircraft Threat

5. NEI 99-01, HAl Page 118 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: H Hazards Subcategory: I Secw-ity Initiating Condition: HOSTILE ACTION within the PROTECTED AREA EAL:

HS1.1 Site Area Emergency A HOSTILE ACTION is occurring or has occtirred 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 Lising guns. explosives, projectiles, vehicles, or other devices used to deliver destructive force.

Other acts that satisfy the overall intent may he included. 1lostile action should not be construed to include acts of civil disobedience or felonious acts that aie not part of a concerted attack on Callaway.

Nonterrorism-based EALs should he 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 harriers anti to which access is controlled. The Protected Area refers to the designated securityarea around the process buildings and is depicted in Drawing 8600-X-8$ 100 Property-Site Layout. Owner Controlled Area and Surrounding Area.

Basis:

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

These individuals are the designated onsite 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. I 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 commuinications between Security Shift Supervision anci the Control Room is essential for proper classification of a security-related event (ref. 2, 3, 4 5).

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.

Page 119 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases 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 he advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Securitysensitive information should be contained in non-public documents such as the Cal laway Plant Security Plan and DBT (ref. I).

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

Callaway Basis Reference(s):

I. Callaway Plant Security Plan and DBT (Safeguards)

2. EIP-ZZ-SKOO1. Response to Security Threat

3. SDP-CP-00003. Security Contingency Events

4. OTO-SK-f)000l, Plant Security Event Hostile Intrusion

5. OTO-SK-00002, Plant Security Event - Aircraft Threat

6. NEt 99-01. HSI Page 120 of 241 INFORMATION USE

EIP-ZZ-0010I ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Category: H Hazards Subcategory: I Security Initiating Condition: HOSTILE ACTION resulting in loss of physical control of the facility EAL:

HG1.1 General Emergency A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the Security Shift Supervisor.

AND EITHER of the following has occurred:

Any of the iollowint safety functions cannot he controlled or maintained Reactivity control.

Core cooling.

RCS heat removal.

OR

Damage to spent fuel has occurred or is IMMINENT.

MODE Applicability:

All Definition(s):

HOSTILEACTION 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 itsed 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 he 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 EAI. will be met within a relatively short period of time regardless of mitigation or corrective actions.

PROTECTED AREA An area encompassed by physical barriers and to which access is controlled. The Protected Area refers to the designated security area around the process huiildings and is depicted in Drawing 8600-X-88 100 Property-Site Layout. Owner Controlled Area and Surrounding Area.

Page 121 of 241 INFORMATION USE

EIP-ZZ-0010I ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Basis:

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

This IC addresses an event in which a HOSTILE FORCE has taken physical control of the facility to the extent that the plant staff can no longer operate equipment necessary to maintain key safety functions. It also addresses a HOSTILE ACTION leading to a loss of physical control that restilts in actual or IMMINENT damage to spent fuel due to 1) damage to a spent fuel pool cooling system (e.g., pumps, heat exchangers, controls, etc.) or, 2) loss of spent fuel 1)001 integrity such that sufficient water level cannot be maintained.

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

Sectirity plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan. Training and Qualification Plan, Safeguards Contingency Plan and Independent Spent Fuel Storage Installation Security Program.

Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Sectirity-sensitive information. This includes information that may be advantageous to a potential adversary, such as the partictilars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Callaway Plant Security Plan & DBT (ref.!).

Callaway Basis Reference(s):

I. Caltaway Plant Security Plan and DBT (Safeguards)

2. EIP-ZZ-SKOOI, Response to Sectirity Threat

3. SDP-CP-00003. Security ConU ngency Events

4. OTO-SK-0t)00l, Plant Security Event Hostile Intrusion

5. OTO-SK-00002, Plant Security Event - Aircraft Threat

6. NEI 99-01, HG!

Page t 22 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: H Hazards and Other Conditions Affecting Plant Safety Subcategory: 2 Seismic Event Initiating Condition: Seismic event greater than OBE level F AL:

HU2.l Unusual Event Seismic event> OBE as indicated by Seismic Activity, Annunciator 98D.

MODE Applicability:

All Definition(s):

None Basis:

Annunciator 98D. OBE will illuminate it the seismic instrument detects ground motion in excess of the OBE threshold (ref. 4).

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

Earthquakes of this magnitude should he 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, preclttde a timely emergency declaration based on receipt of the OBE alarm. The NEIC can he 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://earthguake.usgs. ov/

Additional actions after EAL declaration:

When the seismic recorder indicates that the OBE has been exceeded, as verified by ETP-SG-0000l, 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 grotind 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.

Page 123 of 241 INFORMATION USE

EIP-ZZ-0010l ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases OTO-SG-000t)1, Seismic Event provides the guidance ft)r determining if the OBE eatthquake 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 acttial 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 degi-adation 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 he via IC CA6 or SA9.

Callaway Basis Reference(s):

I. FSAR Section 3.7(B). 1.1 Design Response Spectra

2. OTO-SG-0000l, Seismic Event

3. NEI 99-01, HU2

4. FSAR Table 16.3-3. Seismic Monitoring Instrumentation Page 124 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: H Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 Nattiral 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-8$ 100 Property-Site Layout. Owner Controlled Area and Surrounding Area.

Basis:

Response actions associated with a tornado onsite is provided in OTO-ZZ-000l2 Severe Weather ref. 1).

If damage is confirmed visually or by other inplant indications, the event may be escalated to an Alert under EAL CA6. 1 oi SAQ. 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 he based on ICs in Recognition Categories R. F. S or C.

Callaway Basis Reference(s):

I. OTO-ZZ-000 I 2. Severe Weather

2. NE199-01. HU3 Page 125 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Category: H Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 Natural or Technology Hazard Initiating Condition: Kazardous event EAL:

HU3.2 Unusual Event Internal room or area FLOODING of a magnitude sufficient to require manual or atitornatic 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 instatlecl equipment is capable of removal, restitting 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 IOCFR5O.2):

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

The integrity of the reactor coolant presstire 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 olisite exposures.

Basis:

Refer to EAL CA6.l or SA9.l 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 le\el 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 classitication level would be based on tCs in Recognition Categories R, F, S or C.

Page 126 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Callaway Basis Reference(s):

1. IPE Section 3.4.2.3 Results of the Vulnerability Screening

2. NEt 99-01. HU3 Page 127 of 241 INFORMATION USE

EIP-ZZ-00l01 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: H Hazards and Other Conditions Affecting Plant Safety Subeategory: 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 ofisite 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-88 100 Property-Site Layout, Owner Controlled Area and Surrounding Area.

Basis:

As used here, the term offsite is meant to he 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 looLing 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 evactiate part of the Protected Area.

then that could meet the IC for KU3.3.

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

Callaway Basis Reference(s):

t. NEt 99-01, HU3 Page 128 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technica] Bases Category: H Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 Natural or Technology Hazard Initiating Condition: Hazardous event EAL:

HU3.4 Unusual Event A hazardous event that results in onsite 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 trattic impediments such as log. snow. ice, or vehicle breakdowns or accidents.

MODE Applicability:

All Definition(s):

None Basis:

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

This EAL addresses a hazardous event that causes an onsite impediment tc) 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 catised 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, btit 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.

If ALL access roads to the plant are impassable. and local authorities are no longer clearing roads, this EAL applies.

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

Callaway Basis Reference(s):

1. NEI 99-t) I. HU3 Page 129 of 24] INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Category: H Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:

11U4.l Unusual Event A FIRE is not extinguished within 15 mm. of any of the following FIRE detection indications:

(Note I

Report from the field (i.e.. visual observation).

Receipt of multiple (more than 1) fire alarms or indications.

Field verification of a single fire alarm.

AND The FIRE is located within any Table H-I area.

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

Table H-i Fire Areas

Area5

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 stich as slipping drive belts or overheated electrical equipment do not constitute fit-es. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.

Page 130 of 241 INFORMATION USE

EIP-ZZ-0010l ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases 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 tising available Control Room indications or alarms to prove that it is not spurious, or by reports from the field.

Actual field reports must he made within the 15 minute time limit or a classification must be made. If a fire is verified to be occurrine by field report, the t 5 minute time limit is from the original receipt of the fire detection alarm.

Table H-i Fire Areas aie based on FSAR Section 5.4A.2 System Required to Go from Hot Standby to Cold Shutdown. Table H- I Fire Areas include those structures containing functions and systems required for safe shutdown of the plant (SAFETY SYSTEMS) (ref. I). The Laundry Decon Facility is NOT part of the Aux Building.

This [C addresses the magnitude and extent of FIRES that may he indicative of a potential degradation of the level of safety of the plant.

For EAL HU4. I the intent of the I 5minitte duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished e.g.. smoldering waste paper basket). [11 addition to alarms. other indications of a FIRE could he a drop in fire main pressure, automatic activation of a suppression system.

etc.

The Shift Manager needs to ask some specific qiLestions 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 HAS. 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. I would be recommended, if a required Safety System was affected.

In the case of a fire alarm in Containment. OTA-KC-t)1008 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 he Containment Radiation Monitors such as GTREOO3 I and GTREOO32 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.l Page 131 of 241 INFORMATION USE

EIP-ZZ-00l0l ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Upon receipt, operators will take protipt 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 he via IC CA or SA9.

Callaway Basis Reference(s):

1. FSAR. Section 5.4A.2 System Required to Go From Hot Standby to Cold Shutdown

2. NE199-0l,HU4 Page 132 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUrvIENT Attachment 1 - Emergency Action Level Technical Bases Category: H Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:

HU4.2 Unusual Event Receipt of a single fire alarm (i.e.. no other indications of a FIRE).

AND The fire alarm is indicating a FIRE within any Table H-I area.

AND The existence of a FIRE is not verified within 30 mm. of alarm receipt.

(iVote 1) iVotc 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will liLely he exceeded.

Table H-i Fire Areas

Area5

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 Ores. Observation of flame is preferred btit is NOT required if large quantities of smoke and heat are observed.

Basis:

The 30 minute requirement begins upon receipt of a single valid fire detection system alarm. The alarm is to he 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. It a fire is verified to be occurring by field report, classification shall be made based on EAL HU4. 1.

Page 133 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Table H-I Fire Areas are based on FSAR Section 5.4A.2 System Required to Go From Hot Standby to Cold Shcitclown. Table H-I Fire Areas incitide those structures containing functions and systems requited for safe shutdown of the plant (SAFETY SYSTEMS) (ref. I). The Latmdry Decon Facility is NOT part of the Aux B u ldi ng.

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

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

A single fire alarm, absent other indication(s) of a FIRE, may he 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 contiary, it is assumed that an actual FIRE is in progress.

If an actutal FIRE is verified by a report from the field, then HU4. I 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 he due to an ecjuipment failure or a spurious activation, and this verification occurs within 30minutes 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 Illiestions to ensure they have the information needed to evaluate the situation.

Is there visible flame?

Are there copious qutantities 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 he 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 he an indication that a fire existed and determine if EAL HAS. I is applicable (MODE 4 only). If a breaker truly suffered a fault locaL to the breaker, the damage and fire ball woutld be such that consideration of the Hazardous Event EAL SA9. I woukl be recommended, if a required Safety System was affected.

In the case of a fire alarm in Containment, OTAKC0 100$ 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 GTREOO3 I and GTREOO32 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 Page 134 of 241 INFORMATION USE

EIP-ZZ-0010] ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Basis-Related Requirements from 10 Cf R 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 50.4$ 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 shtit down the pant safely is ens ureci.

NFPA 805 Section 1.3.1 states The Nuclear Safety Goal is to provide reasonable assurance that a Ore 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. NEPA 805 Section 4.2.3.3. requires. among other considerations, the use of 1-hour fire barriers for the enclosure of cable and ecjuipment and associated nonsafety circuits of one redundant train. As used in HU4.2. the 30minutes to verify a single alarm is well within this worstcase Ihour 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):

I. ESAR. Section 5.1A.2 System Required to Go From Hot Standby to Cold Shutdown

2. NEI 99-01. HU4 Page 135 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: H Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:

11U4.3 Unusual Event A FIRE within the plant PROTECTED AREA not extinguished within 60 mm. 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 hut is NOT required if large quantities of smoke and heal are observed, PROTECTED AREA An area encompassed by physical barriers and to which access is controlled. The Pi-otected Area refers to the designated security area arotLnd the process buildings and is depicted in Drawing X600-X-8$ 100 Property-Site Layout. Owner Control led Area and Surrounding Area.

Basis:

This IC acldi-esses 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 HU1. I or HU4.2. a FIRE within the plant PROTECTED AREA not extinguished within 60-t;inutes 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. NEI99-0l.HU4 Page 136 of 241 INFORMATION USE

EIP-ZZ-00I01 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: H Hazards and Other Conditions Affecting Plant Safety Subcategorv: 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 stipport by au 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 sectirity area around the process buildings and is depicted in Drawing

$6t)0-X-88 100 Property-Site Layout. Owner Controlled Area and SurroLinding Area.

Basis:

This IC addresses the magnitude and extent of FIRES that flay he 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 effsite 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.

Callawat Basis Reference(s):

I. NE199-0l, HU4 Page 137 of 241 INFORMATION USE

EIP-ZZ-00I01 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: H Hazards and Other Conditions Affecting Plant Safety Subcategory: 5 Hazardous Gases Initiating Condition: Gaseous release IMPEDING access to equipment necessary for normal plant operations. cooldown or shutdown EAL:

11A5.l Alert Release of a toxic, corrosive, asphyxiant or flammable gas that prohibits or IMPEDES access to EITHER of the following: (iVole 5)

North Electrical Penetration Room. (Rooni 1110)

South Electrical Penetration Room. (Room 1409) iVote 5: If the equipment in the listed mom or area was already inoperable or out-ofservice betore the event occurred.

then no emergency classihcation is warranted.

MODE Applicability:

4 Hot Shtitdown 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:

NGOIBGF3, FDR BKR TO EPHVSSOSA SI ACC A OUT ISO. (Room 1410)

NGO2BGF3. FDR BKR TO EPIIVSXOSB SI ACC B OUT ISO. (Room /409)

NGO1BGF2. FDR BKR TO EPHVS$OSC SI ACC C OUT ISO. (Room 14/0)

NGO2BHF2, FDR BKR TO EPHVS8O8D SI ACC D OUT ISO. (Room /409)

For A RHR:

NGO2BCF2, FDR BKR TO BBPV87O2A RCS LOOP I HOT LEG TO RHR PMPS ISO. (Room /409)

NGO I BEF2, FDR BKR TO EJHV$70 IA A RHR PMP SUCT FROM RCS HOT LEG IISO, (Room 1410)

For B RHR:

NGO2BBF3. FDR BKR TO BBPV87O2B RCS LOOP 4 HOT LEG TO RHR PMPS ISO. (Room 1409)

NGO I BDF3, FDR BKR TO EJHV$70 I B B RHR PMP SUCT FROM RCS HOT LEG 1 tSO, (Room 1410)

Page t 38 of 241 INFORMATION USE

EIP-ZZ-00l01 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases This IC addresses an event invoIvin 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 ii entry into the affected room/area is. or may he, procedurally reqtiired 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 roornlarea is sufficient to preclude or significantly impede procedtirally required access. This judgment may he 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 roonilarea (e.g.,

requiring tise of protective ecluiprnent. 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 roomlarea entr 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 l9c, which can lead to breathing difficulties, tinconsciousness or even death.

This EAL DOES NOT apply to firefighting activities that automatically or manually activate a fii-e stippression system in an area.

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

Callaway Basis Reference(s):

I. OTG-ZZ-00006 Addendum 06. Securing Safety Injection Accumulators

2. OTN-EJ-0000l Addendum 3. Placing A RHR Train In Service For RCS Cooldown

3. OTN-EJ-0000l Addendum 4. Placing B RHR Train In Service For RCS Cooldown

4. NEI 99-01. AA3 Page 139 of 24 t INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Category: H Hazards and Other Conditions Affecting Plant Safety Subcategory: 6 Control Room Evacuation Initiating Condition: Control Room evacuation resulting in transfer of plant control to alternate locations EAL:

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

MODE Applicability:

All Delinition(s):

None Basis:

The Shift Manager (SM) determines if the Control Room is uninhabitable and requires evacuation. Control Room inhabitability may he caused by lire, dense smoke, noxious fumes, bomb threat in or adjacent to the Control Room, or other life threatening conditions. OTO-ZZ-0000 I Control Room Inaccessibility, provides the instructions for tripping the unit, and maintaining RCS inventory and Hot Shutdown conditions from outside the Control Room (Ref. I ).

For the purpose of this EAL, the 15 minute clock starts after determination that Control Room evacuation is necessary. not when OTO-ZZ-0000l 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.l.

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 plait 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 recluired 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.

Caltaway Basis Reference(s):

I. OTO-ZZ-0000 I, Control Room Inaccessibility

2. NEt 99-01, HA6 Page 140 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Category: H Hazards and Other Conditions Affecting Plant Safety Subcategory: 6 Control Room Evacuation Initiating Condition: Inability to control a key safety ftinction from outside the Control Room EAL:

HS6.1 Site Area Emergency An event has resttlted 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 reestablished within IS mm.:

(Note 1)

Reactivity (MODE 1,2. and 3 only.

Core Cooling.

RCS heat removal.

Now I: The Emerencv Coordinator should declare the event promptly upon deternitnitig that Ume limit has been exceeded, or will likely he exceeded.

MODE Applicability:

- 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.

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-0000 1, 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).

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.

Page 141 of 241 INFORMATION USE

EEP-ZZ-0010I ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Once the Control Room is evacuated, the objective is to establish control of important plant eqclipmetlt and maintain knowledge of important plant parameters in a timely manner. Primary emphasis should he placed on components and insttuments that suppiy protection for and information abottt safety functions. Typically.

these safety functions are reactivity control (ability to shut down the teactor 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 he 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 harriers 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 he via IC FGI or CG 1.

Callaway Basis Reference(s):

I. OTO-ZZ-0000l, Control Room Inaccessibility

2. OTS-ZZ-00001. Cooldown from Outside the Control Room

3. NE199-0l, HS6 Page 142 of 241 INFORMATION USE

EIP-ZZ-00I01 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: H Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 Emergency Coordinator Judgment Initiating Condition: Other conditions existing that in the judgment of the Emergency Coordinator warrant declaration of a liE 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 otitlined 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 instrticted to report to their emergency response locations. In this manner, the individual ustially 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 tmanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency becatise conditions exist which are believed by the Emergency Coordinator to fall tinder the emergency classification level description for an Unusual Event.

Callaway Basis Reference(s):

I. 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 143 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Category: H Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 Emergency Coordinator Judgment Initiating Condition: Other conditions exist that in the judgment of the Emergency Coordinator warrant declaration of an Alert EAL:

HA7.1 Alert Other conditions exist which. in the judgment of the Emergency Coordinator. indicate that events are in progress or have occurred which involve an actual or potential sctbstantial 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 disohedietice or feloniotis acts that are not part of a concerted attack on Callaway.

Nonterrorism-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 hut 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.

Page 144 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Cattaway Basis Reference(s):

1 Callaway Radiological Emergency Response Plan, section 5.2.1 Emergency Coordinator

2. Caltawa Radiological Emergency Response Plan, section 5. 1 .1 Shift Managei

3. NEI 99-0 1. HA7 Page 145 of 241 INFORMATION USE

EIP-ZZ-D0l01 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Category: H Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 Emergency Coordinator Judgment Initiating Condition: Other conditions existing that in the judgment of the Emergency Coordinator warrant declaration of a Site Area Emergency EAL:

HS7.i Site Area Eniergencv 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, (I) toward site personnel or equipment that cotild 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 expostire 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. rojectiles, vehicles, or other devices tised to deliver destructive force.

Other acts that satisfy the overall intent may be included. Hostile action should not he 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. I). 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 reqtiired 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.

Page 146 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Catlaway Basis Reference(s):

1. Callaway Radiological Emergency Response Plan, Section 5.2.1 Emergency Coordinator

2. Callaway Radiological Emergency Response Plan. Section 5.1. I Shift Manager

3. NET 99-01. HS7 Page 147 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Category: H Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 Emergency Coordinator Judgment Initiating Condition: Other conditions exist which in the judgment of the Emergency Coordinator warrant declaration of a General Emergency EAL:

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

MODE Applicability:

All Definition(s):

HOSTILEACTION An act toward Catlaway or its personnel that includes the tise 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 tising guns, explosives, projectiles, vehicles, or other devices used to deliver desttuctive force.

Other acts that satisfy the overall intent may he included. Hostile action should not he 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 stLch activities (i.e.. this may include violent acts between individLials in the owner controlled area).

IMMINENT The trajectory of events or conditions is such that an EAL will he 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. I). 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 individtial 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 he expected to exceed EPA PAG plume exposure levels otitside 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.

Page 148 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Callaway Basis Reference(s):

I. Callaway Radiological Emergency Response Plan, Section 5.2.1 Emereency Coordinator

2. Caltawav Radiological Emergency Response Plan. Section 5.1.t Shift Manager

3. NEI 99-01. HG7 Page 149 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Category S System Malfunction EAL Group: Hot Conditions (RCS temperature > 200°F) EALs in this category are applicable only in one or more hot operating MODES.

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

The events of this category pertain to the following subcategories:

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 he necessary to ensure fission product harrier 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 coie cooling systems which may be necessary to ensure fission product barrier integrity. This category incltides loss of vital plant 125 VDC power sources.

3. Loss of Control Room Indications Certain events that degrade plant operator ability to effectively assess pliit conditions within the plant warrant emergency classification. Losses of indicators are in this suhcategory.

4. RCS Activity During normal operation, reactor coolant fission product activity is very tow. Small concentrations of fission products in the coolant are primarily from the fission of tramp uranium in the ftiel clad or minor perforations in the clad itself. Any significant increase from these base-line levels (2% 5% -

clad failures) is indicative of ftiel 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 circitlated with the reactor coolant and can be detected by coolant sampling.

5. RCS Leakae 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.

Page 150 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases

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 tATWS) 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 Affectine Safety Systems Variotis natural and technological events that result in degraded plant safety system performance or significant visible damage warrant emergency classification under this subcategory.

Page 151 of 241 INFORMATION USE

EIP-ZZ-0010I ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Category: S System Malfunction Subcategory: I Loss of Emergency AC Power Initiating Condition: Loss of all offsite AC power capability to emergency buses for t5 minutes or longer EAL:

StIl.1 Unusual Event Loss of all offsite AC power capabiLity, Table S-I, to emergency 4. 16KV htises NBO I and NBO2 for 15 mm.

(Ntiw 1)

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

Table S-i AC Power Sources Offsite:

Safeguards XFMR A or B via ESF LTC XFMR XNBO1

Startup XFMR XMRO1 via ESE LIC XFMR XNBO2

Main XFMR XMAO1 backfed via UAT XFMR XMAO2 (in-service)

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

Onsite:

EDGNEO1

EDGNEO2 MODE Applicability:

- 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-I offsite sources remains available and can be aligned within 15 mintites.

The criteria here is that the source can he supplying the station with power within 15 minutes. Obviously the Main Transformer could not he 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.

Page 152 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases The 4. 16KV AC System provides the power requirements for operation and safe shutdown of the plant. The essential switchear are buses NB01 and N302 (ref. 1).

N301 supplies power to Load Group 1 (Red Train) safety related loads and NBO2 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 XNBO1 and the other source is from the startup transformer XMROI via ESF LTC transformer XNBO2. Transformer XNBOI is the normal supply to bits NBO1: XNBO2 is the normal supply to bus NBO2 ref. 1, 2 3).

Another method to obtain offsite power is by backfeeding the emergency buses through the main transformer XMAOI and unit auxiliary transformer XMRO2. However, this is only done during Cold Shtttdown 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 Coop Power or AEPS diesel geilerators. Credit can be taken for this source only if it can be aligned within 15 mintttes.

This IC addresses a prolonged loss of offsite 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 asa threshold to exclude transient or momentary losses of offsite power.

Escalation of the emergency classification level would he via IC SA I.

Callaway Basis Reference(s):

I. 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-0000 1-ROll. Main Step-Up Transformer Backfeed IPTE

5. NE199-Ol.SUI Page 153 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: S System Malfunction Subcategory: I Loss of Emergency AC Power Initiating Condition: Loss of all but one AC potver sotirce to emergency buses for 1 5 minutes or longer EAL:

SAL! Alert AC power capability, Table S-I. to emergency 4.16KV buses NBOI and NBO2 reduced to a single power souice for ? 15 mm. (Now 1)

AND Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS.

Now I: The Emergency Coordinator should declare the event promptly cipon determining that time limit has been exceeded, or will likely be exceeded.

Table S-i AC Power Sources Oftsite:

Safeguards XFMR A or B via ESF LIC XFMR XNBO1

Startup XFMR XMRO1 via ESF LTC XFMR XNBO2

Main XFMR XMAOJ backfed via UAT XFMR XMAO2 (in-service)

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

Onsite:

EDG NEO1

EDG NEO2 MODE Applicability:

- 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 anti/or placing it in the Cold Shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in IOCFR5O.2):

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

I. The integrity of the reactor coolant pressure boundary:

2. The capabi[ity 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 oftsite exposures.

Page 154 of 241 INFORMATION USE

EIP-ZZ-00l0l ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Basis:

For emergency classification purposes. capability means that one of the Table S-I offsite sources remains available and can be aligned within 15 minutes.

If AEPS cannot he 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 he 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 btises NBOI and NBO2 (ref. I).

NBOI supplies power to Load Group I (Red Train) safety related loads and N802 supplies power to Load Group 2 (Yellow Train) safety related loads. Each bus has two sources of olfsite power. One source is from 13.8 KV safeguards transformer A or B via ESF Load Tap Changing tLTC) transformer XNBOI and the other soui-ce is from the startup transformer XMR() 1 via ESE LTC transformer XNBO2. Transformer XNBOI is the normal supply to bus NBOl: XNBO2 is the normal supply to btis NBO2 (ref. 1.2 3).

Another method to obtain offsite power is by backfeeding the emergency buses through the main transformer XMAOI and unit auxiliary transformer XMRO2. However, this is only done during Cold Shutdown unless nticlear safety considerations require it to be done during hot shtttdotvn 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 he taken for this source only if it can he 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 ofisite 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 patti from IC SUI.

An AC power source is a source recognized in AOPs and EOPs. and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.

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

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

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

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 SS 1.

Page 155 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Callaway Basis Reference(s):

E21 001(Q) Main Single Line Diagram (Electrical Distribution Diagram)

2. FSAR. Site Addenda Section 8.2. I .2

3. FSAR, Section 8.3.1

4. OTS-MA-00001-R01 1, Main Step-Up Transformer Backfeed IPTE

5. NE199-01,SAI Page 156 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: S System Malfunction Subcategory: I Loss of Emergency AC Power Initiating Condition: Loss of all offsite power and all onsite AC power to emergency buses for 15 minutes or longer EAL:

SSI.1 Site Area Enietgency Loss of all offsite and all onsite AC power to emergency 4.16KV buses NBt)l and NBO2 for? 15 mm.

(Vole 1) tote 1. The Emergenc\ Coordinator should declare the event pwmptlv upon determining that time limit has been exceeded, or will Likely he exceeded.

MODE Applicability;

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

None Basis:

For this emergency classification this means NBOI and NBO2 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 NBOI and NBO2 (ref. I).

NBO1 supplies power to Load Group I (Red Train) safety related loads and NBO2 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 tLTC) transformer XNBOI and the other source is from the startup transformer XMROI via ESF LTC transformer XNBO2. Transformer XNBOI is the normal stipply to bus NBOI; XNBO2 is the normal supply to bus N302 (ref. 1.23).

In addition. NBOI and NBO2 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. I).

Another method to obtain offsite power is by backfeeding the emergency buses through the main transformer XMAOI and unit auxiliary transformer XMRO2. 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 AlPS 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).

Page 157 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Enrergency Action Level Technical Bases The interval begins when both offsite anti onsite AC power capability are lost.

This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. In addition, fission prodttct harrier monitoring capabilities may he 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 he via ICs RG I. FGI or SG 1.

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-0000 1-ROll, Main Step-Up Transformer Backfled - IPTE

5. ECA-0.0. Loss of All AC Power

6. NEI 99-01. SSI Page 158 of 241 INFORMATION USE

EIP-ZZ-t)0101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: S System Malfunction Subcategory: I Loss of Emergency AC Power Initiating Condition: Prolonged loss of all offsite and all onsite AC power to emergency buses EA L:

SGI.1 General Emergency Loss of all ofTsite and all onsite AC power to emergency 4. 16KV buses NBOI and N302 AND EITHER:

Restoration of at least one emergency bus in < 4 hotirs is not likely. (iVotc I)

CSFST Core Cooling RED Path conditions met.

Now 1: The Emeriency Coordinator should declare the event promptly upon determininit that time limit has been exceeded. or vill litcly he exeeeded MODE Applicability:

- Power Operation. 2 - Starttip. 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 NBOI and NBO2 either for greater then 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 NBOI and NBO2 (ref. 1).

NBOI supplies power to Load Group 1 (Red Train) safety related loads and NBO2 stipplies 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 XNBOI and the other source is from the startup transformer XMROI via ESF LTC transformer XNBO2. Transformer XNBO1 is the normal supply to bus NBOI: XNBO2 is the normal supply to bus NBO2 (ref. 4.56).

In addition, NBOI and N302 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 XMAO1 and unit auxiliary transformer XMRO2. However, this is only done during Cold Shutdown unless nuclear safety considerations require it to be done during lint shutdown when no other power sources are availabLe (ref. 7).

Page 159 of 241 INFORMATION USE

EIP-ZZ-0010l ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action LeveL Technical Bases 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 he 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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . (ref. 8).

Four hours is the station blackout coping time (ref I, 2).

Indication of continuing core cooling degradation must be based on fission product barrier monitoring with particular emphasis on Emergency Coordinator judgment as it relates to imminent Loss of fission product harriers and degraded ability to monitor fission product barriers. Indication of continuing core cooling degradation is manifested by CSFST Con Cooling RED PATH contlitions 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 harriers.

In addition, fission product harrier monitoring capabilities may be ctegraded under these conditions.

The EAL should reqtlire declaration of a General Emergency prior to meeting the thresholds for IC FG 1.

This wilt allow additional time for implementation of ofisite 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 btis by the end of the analyzed station blackout coping period. Beyond this time, plant responses antI event trajectory are subject to greater uncertainty, and there is an increased likelihood of challenges to multiple fission product harriers.

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 tipgrade. 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 ilthe loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core.

Callaway Basis Reference(s):

I. FSAR, Section $.3A.5

2. BO-Ol, Station Blackout (580) Coping Duration, sh I

3. CSF- 1. Critical Safety Function Status Trees (CSFST) Figure 2. Core Cooling

4. E-2100l(Q) Main Single Line Diagram (Electrical Distribution Diagram)

5. FSAR, Site Addenda Section 8.2.1.2 6 ESAR. Section 8.3.1

7. OTS-MA-000t) I-ROll, Main Step-Up Transformer Backfeed IPTE

8. ECA-0.0, Loss of All AC Power

9. NE199-0l,SGI Page 160 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev, 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUfIENT Attachment 1 - Emergency Action Leve] Technical Bases Category: S System Malfunction Subcategory: I Loss of Emergency AC Power Initiating Condition: Loss of all AC and vital DC power socirces for 15 minutes or longer EAL:

SG1.2 General Emergency Loss ol all off.site and all onsite AC power to emergency 4.16KV buses NBOI and NBO2 for 15 mm.

AND Loss of all 125 VDC power based on battery bus voltage indications < 107 VDC on all vital DC buses NKOI, NKO3 (Division I) and NKO2. NKO4 (Division 2) for> 15 mm.

(Now 1) tVow I: The Emergency Coordinator should declare the event pmmptlv 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 NBOI and N802 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 NBOI and NBO2 for greater than 15 minutes in combination with degraded vital DC power voltage. This EAL addresses operating experience from the March 2011 accident at Fukushirna Daiichi.

The 4. 16KV AC System provides the power requirements for operation and safe shutdown of the plant. The essential switchgeur are buses NBOI and NBO2 (tef. I).

NBOI supplies power to Load Group I (Red Train) safety related loads and NBO2 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 XNBO1 and the other sotirce is from the startup transformer XMROI via ESF LTC transformer XNBO2. Transformer XNBOI is the normal supply to bus NBOI: XNBO2 is the normal supply to bus N302 (ref. 1.23).

In addition. NBO1 and NBO2 each have an emergency diesel generator (onsite power supply which supply electrical power to the bcis automatically in the event that the preferred source becomes unavailable (ref. 1).

Page 161 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Another method to obtain offsite power is by backfeeding the emergency buses through the main transformer XMAOI and tinit auxiliary transformer XMRO2. 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 he taken for these sources only if they are capable of carrying an NB bus and it can be aligned within 15 minutes. (ref. 5).

The vital DC buses are the following 125 VDC Class IE buses (ref. 6):

Division 1: Division 2:

NKO1 NKO2 NKO3 NKO4 There aie four battery banks. (NKI I. NKI2. NK13 and NKI4) 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 I 25VDC buses, the fotir battery banks for each bus do not have the same rating. All four of the I 25VDC buses supply inverters for I 2OVAC NN bus power as well as control power for various safety related systems. NKOI and NKO4 supply additional DC loads such as diesel field flashing. breaker control i wer. main control hoard power and emergency lighting. These loads are not supplied by the other two buses, NKO2 and NKO3. For this reason, batteries NKI 1 and NKI4 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 (siation blackout) (ref. 8, 9, 10).

Minimum DC bits voltage is 107.0 VDC (ref. 9. 10). Bus voltage may he obtained from the following instruments (ref. 6):

NKE-1 (NKOI)

NKEI-2(NKO2)

NKEI-3(NKO3)

NK EI-4 (NKO4)

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 I 5-minute emergency declaration clock begins at the point when BOTH EAL thresholds are met.

Page 162 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Callawav 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-0000 I-ROll, Main Step-Up Transformer Backfeed IPTE

5. ECA-0.0. Loss of All AC Power

6. E-2lOIOtQ) 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-l NiB

11. NEI 99-01, SG$

Page 163 of 24] INFORMATION USE

EIP-ZZ-OOlOt ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Category: S System Malfunction Subeategory: 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 NKOI, NKO3 (Division I) and NKO2, NKO4 (Division 2) for> 15 mm.

(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:

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

None Basis:

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

Division 1: Division 2:

NKO1 NKO2 NKO3 NKO4 There are four battery banks, (NKII, NKI2. NKI3 and NKI4) 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 I25VDC buses, the four battery banks for each bus do not have the same rating. All four of the 125 VDC buses supply inverters for I2OVAC NN bus power as well as control power for variotts safety related systems. NKOI and NKO4 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, NKO2 and NKO3. For this reason. batteries NKI I and NKI4 require additional capacity. Each battery is designed to have sufficient stored energy to stipply the required emergency loads for 240 minutes following a loss of AC power (station blackotit) (ref. 2. 3. 4).

Page 164 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Minimum DC bus voltage is 107.0 VDC (ref. 4. 5). Bus voltage may be obtained from the following instruments (ref. 6):

NKEI-l(NKOI)

NK EI-2 NKO2)

NK EI-3 (NKO3)

NKEI-4(NKO4)

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 tvas selected as a threshold to exclude transient or momentary power losses.

Escalation of the emergency classification level would be via ICs RGI, FG I or SC I.

Callaway Basis Reference(s):

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

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

3. FSAR. Section S.3.2

4. Calculation NK-l0, NK System DC Voltage Drop

5. FSAR. Table S.3A-l Ilt.B

6. ECA-0.0. Loss of All AC Power

7. NEI 99-01. SS8 Page 165 of 241 INFORMATION USE

EIP-ZZ-00I01 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Category: S System Malfunction Subcategory: 3 Loss of Control Room Indications Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer EAL:

SU3.1 Unusual Event An UNPLANNED event results in the inability to monitor one or more Table S-2 parameters from within the Control Room for 15 mm.

(iVotc I)

Note I: The Emerency 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 SIG MODE Applicability:

- Power Operation. 2 - Startup, 3 - Hot Standby. 4 - Hot Shutdown Dellnition(s):

UNPLANNED A parameter change or an event that is not 1) the restilt 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 tin known Basis:

SAFETY SYSTEM parameters listed in Table S2 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.

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.

Page 166 of 241 INFORMATION USE

EIP-ZZ-OOtOl ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular. emergency assessments necessary to implement abnormal operating procedures. emergency operating procedures. and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.

This EAL is focused on a selected subset of plant parameters associated with the key safety ftinctions 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 signitcant than simply a reportable condition, hi addition. if all indication sotirces 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 he 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 transietu or momentary losses of indication.

Escalation of the emergency classification level would be via IC SA3.

Caltaway Basis Reference(s):

FSAR, Section 7.5 Safety-Related Display Instrumentation

2. OTO-RJ-0000l, Loss of Plant Computer

3. NEI 99-01. 5U2 Page 167 of 241 INfOR1IATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Category: S System Malfunction Subcategory: 3 Loss of Control Room Indications Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress EAL:

SA3.1 Alert An UNPLANNED event results in the inability to monitor one or more Table S-2 parameters from within the Control Room for> 15 mm. (Note I)

AND Any significant transient is in progress. Table 5-3.

Note 1: The Emergency Coordinator should declare the event promptly upon detcrminin 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 SIC

Auxiliary or emergency feedwater flow in at least one S/C 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:

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

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

Page 168 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Basis:

SAFETY SYSTEM parameters listed in Table 5-2 are momtorecl 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 recluired 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.iitibacks involving greater than or eqtial to 25c 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 Iarameter( 5). 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 he reported if it significantly impaired the capability to perform emergency assessments. In partictilar, emergency assessments necessary to implement abnormal operating procedures. emergency operating procedures, and emergency plan implementing procedtires 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 he 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 he impacted as well. For example. if the valtte for reactor vessel level cannot he 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 FS 1 or IC RS 1.

Callaway Basis Reference(s):

I. FSAR, Section 7.5 Safety-Related Display Instrumentation

2. OTO-RJ-0000l, Loss of Plant Computer

3. NIl 99-01 SA2 Page 169 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: S System Malfunction Subcategory: 4 RCS Activity Initiating Condition: Reactor coolant activity greater than Technical Specification allowable limits EAL:

SU4.1 Unusual Event Sample analysis indicates RCS activity > Technical Specification 3.4. 16 limits (listed below):

> 60 pCi/gm Dose Equivalent I- 131.

OR

> 1 .0 pCi/gm Dose Equivalent 1-13 1 for a > 4% hr continuous period.

OR

> 225 pCi/gm Dose Eqttivalent Xe-i 33 for a > 48 hr continuous period.

MODE Applicability:

- Power Operation, 2 - Startup, 3 - Hot Standby, 4 Hot Shutdotvn 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 pCi/gm Dose Equivalent 1-131.

OR

> 1.0 pCi/gm Dose Equivalent 1-131 for a > 48 lit continuous period.

OR

> 225 pCi/gm Dose Equivalent Xe-133 for a > 48 hr continuotis 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 emergeticy classification level would be via ICs FA I or the Recognition Category R ICs.

CaRaway 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 170 of 241 INFORMATION USE

EIP-ZZ-00l01 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: S System Malfunctioti 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 gprn for? 15 mm.

OR RCS identified leakauie>25 pm for> 15 mm.

OR Leakage from the RCS to a location outside containment > 25 gpm for ? 15 mm.

(Vole 1)

\o,c 1: The Emergency CoordinLtor should declare the event promptly upon determining that time limit has been exceeded. Ot will likely he exceeded.

MODE Applicability:

- Power Operation. 2 - Starttip. 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. I).

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 stimp or collecting tank, or

Leakage into the containment atmosphere from sources that are both specifically located and known either net to interfere with the operation of leakage detection systems or not to he pressure bocLndary 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).

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)

Page 171 of 241 INFORiMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases 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 shtitdown cooling MODE) (ref. 3, 4)

Escalation of this EAL to the Alert level is via Category F. Fission Product Barrier Degradation, EAL FAI 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 otitside 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 hotindary 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 he reqtiired 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 isotated).

The 1 5minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible. If the leak is isolated, the RCS harrier was never lost.

Escalation of the emergency classification level would be via lCs of Recognition Category R or F.

Caltaway Basis Reference(s):

I. OSP-BB-00009. RCS Inventory Balance

2. Callaway Technical Specifications, DefinitionsSection I .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 172 of 241 INFORMATION USE

EIP-ZZ-00l01 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: S System Malfunction Subcategory: 6 RTS Failure Initiating Condition: Automatic or manual trip fails to shut down the reactor EAL:

SU6.1 Unusual Event An automatic trip did not shut down the reactor as indicated by reactor power? 5 after any RTS setpoint is exceeded.

AND A subsequent automatic trip or manual trip action taken at the reactor control consoles (SBHS- 1 or S3-HS-42) is successful in shutting down the reactor as indicated by reactor power < 5.

(\F0ft

)

iVote S: A TilanUal trip action is any operator action, Ut SCt of actions, which causes the control rods to he rapidR inserted into the core. and does not include manually driving in control rods or implementation ot boron injection strategies.

MODE Applicability:

- Power Operation Definition(s):

N one 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 RLOO3 or SB-HS-42 on Panel RLOO6. Reactor shutdown achieved by use of other trip actions, such as opening PGI9 and PG2O supply breakers, emergency horation or t;anttally driving control rods. do not constitute a successful manual trip (ref. 4). A successful manual turbine trip that suthseqtiently automatically trips the reactor does constitute a successful trip.

FoLlowing a successfut 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 severat 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 5c (ref. 2, 3, 4).

Page 173 of 241 INFORMATION USE

EIP-ZZ-00I0l ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Fo11owin any automatic RTS trip signal. E-0 (ref. 2) and FR-S.! (ref. 4) prescribe insertion of redundant manual trip signals to back tip 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 fullin 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 arttornatic 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 prodtict barrier toss. However, if subsequent manual reactor trip actions fail to redtice reactor power below 5%. the event escalates to the Alert under EAL SA6. I.

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 he 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 he 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 ate 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 suhseqtient effort to manually trip the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor trip signal. If a subseqtient 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 catises 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.

Page 174 of 241 INFORMATION USE

EIP-ZZ-00l01 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases The plant response to the failure of an automatic or manual reactor trip will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions. other concurrent plant conditions. etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via JC SA6. Depending upon the plant response. escalation is also possible via IC FAI. Absent the plant conditions needed to meet either IC SA6 or FAI. an Unusual Event declaration is appropriate for this event.

A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.

Should a reactor trip signal be generated as a result of plant work (e.g.. RTS setpoint testing). the following classification guidance should he 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 restilts). then this IC and the EALs are not applicable atid no classification is warran ted.

Note $ is a generic note applicable to the EALs as approved by the NRC.

Caltaway Basis Refei-ence(s):

1. Callaway Technical Specifications. Sectioii 3.3.1 Reactor Trip System (RTS) Instrumentation

2. E-0. Reactor Trip or Safety Injection

3. F-0. Critical Safety Function Status Trees - Suhcriticality

4. FR-S. I, Response to Nuclear Power GeneratioH/ATWS

5. FSAR. Section 7.7.1 6 NE199-0l.SU5 Page 175 of 241 INFORMATION USE

EIP-ZZ-00l01 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: S System Malfunction Subeategory: 6 RTS Failure Initiating Condition: Automatic or manual trip fails to shtit 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 (S3HS I or SB-HS-42) is successful in shutting down the reactor as indicated by reactor power <

(Note 8)

Note 8: A manual trip action is any operator action, or set of actions, which causes the control rods to he rapidly inserted into the core, and does not incitide 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 atitomatic RTS trip setpoint and a subsequent automatic or manual trip is successful in shutting down the reactor (reactor power < 5%). (ref. I).

For the purposes of emergency classification, successful manual trip actions are those which can be quickly performed from the reactor control console; SB-HS-l on Panel RLOO3 or SB-HS-42 on Panel RLOO6. Reactor shutdown achieved by ttse of other trip actions. stich as opening PG 19 and P020 supply breakers, emergency boration or inanttally driving control rods, do not constitute a sticcessful manual trip (ret. 4). A successful manual turbine trip that subsequently automatically trips the reactor does constitute a successftil trip.

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 sotirce netitrons 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 sotirce 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).

Page 176 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases 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 mantial trip, the lowest level of classification that must be declared is an Untistial 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 redtice 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 hAL SA. I This IC addresses a failure of the RTS to initiate or complete an automatic or manual reactor [rip 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 unsuccessftil, operators will promptly take mantial action at another location(s) on the reactor control consoles to shut down the reactor (e.g.. initiate a manuial reactor trip) using a different switch. Depending upon several factors. the initial or subseqttent effort to manually trip the reactor, or a concurrent plant condition. may lead to the generation of an automatic reactor trip signal. If a suibsequent mantial or automatic trip is successful in shutting down the reactor, core heat generation will clulickly 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., iiitiating 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 FAt. Absent the plant conditions needed to meet either IC SA6 or FAI. an Unttscial Event declaration is appropriate for this event.

A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.

Page 177 of 241 INFORMATION USE

EIP-ZZ-00l01 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases ShouLd a reactor trip signal be generated as a result of plant work (e.g., RTS setpoint testing), the following classification gtiidance 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 catise a plant transient and the trip failttre 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 EALs 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 - Subcritical ity

4. FR-S. 1. Response to Nuclear Power Generation/ATWS

5. FSAR. Section 7.7.1

6. NEI 99-01, SU5 Page 178 of 241 INFORMATION USE

EIP-ZZ-00l01 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases 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 F AL:

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-I-ISi or SB-HS-42) are not successful in shutting down the reactor as indicated by reactor power ? 5%.

(Note 8) iVore 8. A manual trip action is any operator action, or set ol actions, which causes the control rods to be rapidly inserted into the core, and does rn)t include manually driving in control rods or implementation of boron injection strategies.

MODE Applicability:

- 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 RLOO3 or SB-HS-42 on Panel RLOO6. Reactor shutdown achieved by use of other trip actions,s itch as opening PG 19 and PG2O supply breakers. emergency boration or manually driving control rods, do not constitute a successful manual trip (ref. 4). A successful manual ttirbine trip that subseqtiently 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 Acttiation 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).

Page 179 of 241 INFORMATION USE

EIP-ZZ-0010l ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases 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. 4).

Escalation of this event to a Site Area Emergency woitld be under EAL SS6. I 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 teactor control consoles to shut clown 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 shtmtdown 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 pttrstie 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 inclutding 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 ftinctions. 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 ES I. Absent the plant conditions needed to meet either IC 556 or ES 1, 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 al)plicable Emergency Operating Procedtmre criteria.

Note 8 is a generic note applicable to the EALs 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 tnjection

3. F-0. Critical Safety Function Status Trees - Subcriticality

4. FR-S. 1, Response to Niic tear Power GenerationlATWS

5. FSAR, Section 7.7.1

6. NEI 99-01, SA5 Page 180 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: S System Malfunction Subeategory: 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 57c.

AND EITHER:

CSFST Coie Cooling RED Path conditions met.

CSFST Heat Sink RED Path conditions met.

NIODE Applicability:

- 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 tEAL SA6.l), and

Indications that either core cooling is extremely challenged or heat removal is extremely challenged.

The combination of failure of both ftont 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 PGI9 and PG2O 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. 4).

5% rated power is a minimum reading on the power range scale that indicates continued power production.

It also approximates the decay heat tvhich 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 he used to determine if reactor power is greater than 5 % power (ref. I. 4).

Page 181 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Indication of continuing core cooling degradation is manifested by CSFST Core Cooling RED PATH conditions being met (ref. 2).

Indication of inability to adequately remove heat from the RCS is manifested by CSFST Heat Sink RED PATH conditions being niet (ref. 3).

This IC addresses a faitttre 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 restdting 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 failtire 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 RGI or FGI.

Callaway Basis Reference(s):

CSF-l, Critical Safety Function Status Trees Figure 1 Suberiticality

2. CSF-J, Critical Safety Function Status Tress Figure 2 Core Cooling

3. CSF-1. Critical Safety Ftinction Stattis Tress Figure 3 Heat Sink

4. FR-S. I, Response to Nuclear Power Generation/ATWS

5. NE199-01.SS5 Page 182 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: S System Maiftinction Suhcategory: 7 Loss of Communications Initiating Condition: Loss of all onsite or offsite communications capabilities EAL:

SU7.1 Unusual Event Loss of all Table 5-4 onsite communication methods.

OR Loss of all Table 5-4 ORO communication methods.

OR Loss of all Table 5-4 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 Sentry Notification System X MODE Applicability:

- Power Operation. 2 Startup.3

- - Hot Standby. 4 - Hot Shutdown Definition(s):

OFFSJTE RESPONSE ORGANIZ4 TIONS (ORO) The State of Missouri (SEMA/MIAC). Cal taway County 91 I/EOC. Gasconade Cotmnty 91 l/EOC, Montgomery County 91 l/EOC and Osage County 91 I/EOC.

Page 183 of 241 INFORMATION USE

EIP-ZZ-00l01 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Basis:

Onsite/otfsite communications include one or more of the systems listed in Table 5-4 (ref. 1. 2).

Gaitronics system The Gaitronics system provides six separate independent communication channels--one general page, one Control Room page and four party lines. Commttnication 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 he 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 disrttption 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 structctre. 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 tised 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 buildine 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.

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.

Page 184 of 241 INFORMATION USE

EIP-ZZ-0Ol0l ADDENDUM 2 Rev, 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases

6. Back-Up Radio System (BURS)

The Back-up Radio System is a communication link between the Caltaway 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 radio logic al emergency. The system uses $00 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. Sentry 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 otit a notification form on screen and transmit the data simtdtaneotisly. Notifications on Sentry 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. I.

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 he assessed only when extraordinary means are being utilized to make communications possible (e.g.. use of non-plant. privately owned eqtiipment, 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. NE199-0l.SU6 Page 185 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: S System Malfunction Subuategory: 8 Containment Failure Initiating Condition: Failure to isolate containment or loss of containment pressure control.

EAL:

SUS.1 Unusual Event Any penetration is not isolated within 15 mm. of a VALID containment isolation signal.

OR Containment pressure > 27 psig with < one full train of containment depressurization equipment operating per design for 15 mm.

(Notes 1, 9)

Note ]. 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 lull train of depresstirization equipment.

MODE Applicability:

- 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 (I) 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 mtist be generated as the restilt 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. 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 incltided to allow operators time to manually isolate the required penetrations, if possible.

Page 186 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases 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 eqtiiprnent 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 dtiring the injection phase of operation. In the recirculation MODE of operation. Containment Spray pump suction is transferred from the RWST to the Containment stimps (ref. 2).

The Containment Cooling System consists of two trains of Containment cooling, each of sufficient capacity to supply l00 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 presstlre at which the equipment should actuate and begin performing its ftinction. The design basis accident analyses and evaltiations assume the loss of one Containment Spray System train and one Containment Cooling System train treE 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 wotild escalate to a Site Area Emergency in accordance with IC FS I if there were a conctirrent 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-l. Critical Safety Function Stattis Trees (CSFST) FigureS, Containment

5. FR-Z.l, Response to High Containment Pressure

6. Technical Specifications, Table 3.3.2-I

7. Technical Specifications, B3.6.6

8. NEI 99-01, SU7 Page 187 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Category: S System Maiftinction Subcategory: 9 Hazardous Event Affecting Safety Systems Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating MODE EAL:

SA9.I Alert The occurrence of any Table 5-5 hazardous event AND EITHER:

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

The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the ctirrent operating MODE.

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:

- 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 prese lit.

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 reqttired 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 mIs) sustained, or 58 mph (26 mIs) gusting.

Page 188 of 241 INFORMATION USE

EIP-ZZ-00l0l ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attactiment 1 - Emergency Action Level Technical Bases SAFETY SYSTEM A system required for safe plant operation, cooling down the plant and/or placing it in the Cold Shutdown condition. including the ECCS. These are typically systems classified as safety-related (as defined in IOCFR5O.2):

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

The integrity of the reactor coolant pressure boundary:

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

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

VISIBLE DAMAGE Damage to a component or structure that is readily observable without measurements.

testing. or analysis. The vistial impact of the damage is sufficient to catise concern regarding the operability or reliability of the affected component or structure.

Basis:

Annunciator 9$D. OBE will illuminate if the seismic instrument detects ground motion in excess ot the OBE threshold. OTO-SG-0000l. Seismic Event provides the guidance for determining if an OBE earthqtiake threshold is exceeded and any required response actions (ref. 1).

Internal FLOODING may he caused by events such as component failures, equipment misalignment.

or outage activity mishaps (ref. 2).

External flooding may he due to high lake level. Callaway plaill 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 reqitired 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 wotild be classified tinder 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.

This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM. or a structure containing SAFETY SYSTEM components. needed for the current operating MODE. This condition significantly reduces the margin to a loss or potential loss of a fission product barrier, and therefore represents an actual or potential substantial degradation of the level of safety of the plant.

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

Page 189 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases The second condition addresses damage to a SAFETY SYSTEM component that is not in service/operation or readily apparent through indications alone, or to a structure containing SAFETY SYSTEM components.

Operators will make this determination based on the totality of available event and damage report information. This is intended to he a brief assessment not requiring lengthy analysis or quantification of the damage.

Escalation of the emergency classification level would be via IC FS I or RS 1.

Callaway Basis Reference(s):

1. OTO-SG-0000l, 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.].l Design Wind Loadings

5. FSAR, Section 9.5. I Fire Protection System

6. NE199-0l,SA9 Page 190 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category F fission Product Barrier Degradation EAL Group: Hot Conditions (RCS temperature> 200°F): EALs in this category are applicable only in one or more hot operating MODES.

EALs in this category represent threats to the defense in depth design concept that precludes the release of hih1y 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 fissioti prodtict harriers are:

A. fuel Clad tFC): 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 tip to and including the outermost secondary side isolation valve. Contaitiment 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 prodtict barrier matrix of Table F-I tAttachment 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 harrier 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:

Atert:

Aiiv loss or any potenttal loss of either Fuel Clad oi- RCS Site Area EmerRency:

Loss or potential loss 0/ any two barriers Genera! Emer,gency:

Loss of am tiio barriers and loss or potential loss oJthird barrier Page 191 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases 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 prodtict barrier thresholds may result in a Site Area Emergency classitication while a dose assessment may indicate that an EAL for General Emergency IC RGI has been exceeded.

The fission product harrier 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 proilict 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 wotild have more assurance that there was no immediate need to escalate to a General Emergency.

Page 192 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Any loss or any potential loss of either Ftiel Clad or RCS EAL:

FA1.l Alert Any loss or any potential loss of either Fuel Clad or RCS (Table f-i).

MODE Applicability:

- 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 FI (Attachment 2) lists the fission product barrier thresholds, bases and referetices.

At the Alert classilcation level. Fuel Clad and RCS barriers are weighted more heavily than the Containment barrier. Unlike the Containment barrier, loss or potential toss 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 FS I .

Callaway Basis Reference(s):

1. NE199-0i,FAI Page 193 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 Emergency Action Level Technical Bases Category: Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss or potential loss of any two barriers EAL:

fS.1 Site Area Emergency Loss or potential loss of any two barriers (Table F-i).

MODE Applicability:

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

None Basis:

Fuel Clad, RCS and Containment comprise the fission prodtict harriers. Table F-I (Attachment 2) lists the fission prodtict 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 toliowing conditions:

One barrier loss and a second barrier less (i.e., loss - toss)

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 levet, 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 recluire continual assessments of radioactive inventory and Containment integrity in anticipation of reaching a General Emergency classification. Alternatively, if both Ftiel Clad and RCS potential loss thresholds existed, the Emergency Coorclinator would have greater assurance that escalation to a General Emergency is less imminent.

Callaway Basis Reference(s):

I. NE199-0l.FSI Page 194 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 1 - Emergency Action Level Technical Bases Category: Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss of any two harriers and loss or potential loss of third barrier EAL:

FG1.1 General Emergency Loss of any two harriers.

AND Loss or potential loss of third harrier (Table FI).

MODE Applicability:

- Power Operation. 2 Startup. 3

- - Hot Standhy, 4 - Hot Shutdown Definition(s):

None Basis:

Ftiel Clad. RCS and Containment comprise the fission product barriers. Table FI (Attachment 2) lists the Fission product barrier thresholds, bases and references.

At the General Emergency classification level each harrier 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 Ftiel 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 Ftiel Clad and Containment harriers with potential loss of RCS barrier Catlaway Basis Reference(s):

I. NE199-01,FGI Page 195 of 241 INFORMATION USE

EIP-ZZ-00J0l ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 Fission Product Barrier Loss/Potentia] Loss Matrix and Bases Introduction Table F- I lists the threshold conditions that define the Loss and Potential Loss of the three fission product barriers (Fuel Clad, Reactor Coolant System, and Containment). The table is structured so that each of the three barriers occupies adjacent columns. Each fission product barrier column is further divided into two columns; one for Loss thresholds and one for Potential Loss thresholds.

The first column of the table (to the left of the Fuel Clad Loss column) lists the categories (types) of fission product harrier thresholds. The fission product barrier categories are:

A. RCS or SG Ttthe Leakage

3. 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-I thus forming a matrix defined by the categories. The intersection of each row with each Loss/Potential Loss column forms a cell in which one or more fission product barrier thresholds appear.

Thresholds are assigned sequential numbers within each Loss and Potential Loss coltirnn 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 he assigned FC Loss A. I the third Containment barrier Potential Loss in Category C would he assigned CMT P-Loss C.3, etc.

If a cell in Table F-I contains more than one numbered threshold, each of the numbered thresholds, if exceeded, signifies a Loss or Potential Loss of the harrier. It is not necessary to exceed all of the thresholds in a category before declaring a barrier Loss/Potential Loss.

Subdivision of Table F-I by category facilitates association of plant conditions to the applicable fission product barrier Loss and Potential Loss thresholds. This structure promotes a systematic approach to assessing the classification status of the fission product barriers.

When equipped with knowledge of plant conditions related to the fission product barriers, the EAL-user first scans down the category column of Table Fl, 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 scifficiently 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 FGI.l, FS 1 .1, and FA I .Ito 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 196 of 241 INFORMATION USE

EIP-ZZ-00 It)! ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 fission Product Barrier Loss/Potential Loss Matrix acid Bases Table F-i Fission Product Barrier Threshold Matrix Fuel Clad (FC) Barrier Reactor Coolant System (RCS) Barrier Containment (CMT) Barrier Category Loss Potential Loss Loss Potential Loss Loss Potential Loss I An automatic or manual 1 Operation of a standby 1 A leakIng or RUPTURED SC ECCS (Sit actuation required chargmg pump is required by is FAULTED outside of A by EITHER: EITHER containment UNISOCASLE ADS UNISOLABLE RCS RCS or SG None None None leakage leakage Tube SC lube RUPTURE SC tube leakage Leakage

2. CSFST Integrity-RED Path conditions met t CSFST Corn CoatIng-RED 1 CSFST Core 1 CSFST Heal Sink-RED Patlr I CSFST Core Cooling-RED Path conditions met Cooling- Path conditions met Path cooditons met t condltions met AND B AND

2. CSFST Heat Sink-RED Path None Heat 515k is required None na d equa e Restoration procedures not conditions met Heat Removal effective within IS miii AND ((dote It Heat sink is required

t. Corttainment radiation 1 Containment radiation I Contuinrncrit radiation

, 840 A/hr on >59 Rihr on c 4.000 A/hr on C DT-RE-59 (591) or GT-RE-59 15911 or GT-Rff-59 (591) or 01-RE-ho bOll CT-RE-hO (boll 61-RE-ho (bolt C T

- - Dose equivatenl -131 coolant a None None None a latlon activity, 300 pCi.cc RCS

3. CVCS letdown radlation Activity >2 50E*Qt pCi/mi on SJ-RE-Ot (0161 t Containment isolation is t CSFST Containment-RED req tilted Path conditions met AND EITHER 2. Containment hydrogen Containment integrity has concentratIon C 40 D been lost based on 3. Containment pressures 27 CMT None Emergency Coordinator None None None psig wilh -s one lull train 01 Inte g nt or depressorization equipment Bypass UNISOLABLE pathway operating per design tar tom Connto-nmenn to the t 5 mm vnvironmenl coats (Notes I. 9)

2. lrndicotlons 01 RCS leakage outside of Containment I Any candihon in lie opinion t Any coridiloni in Ihe opnron 1 Any c-and:lion in the opinion I Any condilion in lie opinion 1 Any condition In tIle opinion I Any condrlion in the opinion E 01 thu Emergency Coordinator 01 the Emergency Coordinator 01 the Emergency Coordinator ot the Emergency Coordinalot of lire Emergency Coordinator at the Emergency Coordinalon that indicates loss ot the fuel 11101 indicates potential loss ol Ihal indicates loss of the RCS that rndicales poteniial loss of that indicales loss at the that indicates potential loss 01 Judgment c/ad barrier Ihe fuel clad barrier barrier the RCS barrier Conlainmeni barrier the Contanment barrior Pate 197 of 241 INFORMATION USE

EIP-ZZ-0010l ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 - Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: A. RCS or SG Ttihe Leakage Degradation Threat: Loss Threshold:

N one.

Page 198 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: A. RCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold:

None.

Page 199 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: B. Inacteqtiate 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 stiperheating 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 stiperheating of reactor coolant.

Callaway Basis Reference(s):

I. CSF-l Critical Safety Ftinction Status Trees

2. FR-C. I Response to [nadequate Core Cooling

3. FR-C.2 Response to Degraded Core Cooling

4. NEI 99-0 1 Inadequate Heat Removal Ftiel Clad Loss 2.A Page 200 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: B. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold:

1. CSFST Core Cooling- tNC Path conditions met.

Detlnition(s):

No iie Basis:

Critical Safety Function Status Tree (CSFST) Core Cooling-ORANGE path indicates subcooling has been lost and that some fLiel clad damage may potentially occur. The CSFSTs are normally monitored using the SPDS display on the Plant Computer (ref. I).

This reaclin indicates a redtiction in reactor vessel water level sufficient to allow the onset of heatinduced 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 201 of 241 INFORMATION USE

EIP-ZZ-0010l ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: B. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold:

2. CSFST Heat Sink-RED Path conditions met.

AND Heat sink is reqtiired.

Definition(s):

None Basis:

In combination with RCS Potential Loss B.l, meetin this threshold results in a Site Area Emer2ency.

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 presstire 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 I tells the operator to determine if heat sink is recjuired by checking that RCS pressure is greater than any non-faulted SG pressure and either RCS temperature is greater than 350°F 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 classificalion. (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 beat sink). This condition represents a potential loss of the Fuel Clad Barrier. In accordance with lOPs, there may be unusual accident conditions during which operators intentionally redtice the heat removal capability of the steam geilerators: during these conditions, classification using threshold is not warranted.

Callaway Basis Reference(s):

t. CSF-l Critical Safety Function Status Trees Figure 3 Heat Sink

2. FR-H. I Response to Loss of Secondary Heat Sink

3. NEI 99-01 Inadequate Heat Removal Fuel Clad Loss 2.B Page 202 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attach;uent 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: C. CMT Radiation / RCS Activity Degradation Threat: Loss Threshold:

1. Containment radiation >840 R/hron GT-RE-59 (591) orGT-RE-60 (601).

Definition(s):

None Basis:

Containment radiation monitor readings greater than 840 RIhr (ref. I ) 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 iodine inventory associated with a concentration of 300 pCi/cc dose equivalent 1-131 into the Containment atmosphere. Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations (including iodine spiking) allowed within technical specilcations and are therefore indicative of ftiel 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-l I channel 591) and GT-RE-60 (Panel RM-1 I channel 601). The threshold value of $40 R/hr is the HIHI (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 pCi/gm dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of ftiel 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.l since it indicates a toss 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-RMO1 1 Radiation Monitor Control Panel RM-l1

2. NEI 99-01 CMT Radiation / RCS Activity Fuel Clad Loss 3.A Page 203 of 241 INFORMATION USE

EIP-ZZ-OOIOt ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: C. CMT Radiation / RCS Activity Degradation Threat: Loss Threshold:

2. Dose equivalent 1-13 t coolant activity > 300 pCi/cc.

Definition(s):

None Basis:

Dose Equivalent Iodine (DEl) is determined by Chemistry procedtire CDP-ZZ-08 100. Post Accident Sampling Guidelines (ref. I).

Elevated reactor coolant activity represents a potential degradation in the level of safety of the plant and a potential precursor of more serious problems. The threshold dose equivalent 1-13 I concentration is well above that expected for iodine spikes and corresponds to about 2 to 59c 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 coHceitration is greater than 300 pCi/gm dose equivalent 1-13 1. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2 to S%r 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.

Caliaway Basis Reference(s):

I. CDP-ZZ-08 100 Post Accident Sampling Guidelines

2. NEI 99-01 CMT Radiation / RCS Activity Fuel Clad Loss 3.3 Page 204 of 241 INFORMATION USE

EIP-ZZ-00I01 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 Fission Product Barrier Loss/Potentiat Loss Matrix and Bases Barrier: Ftiel Clad Category: C. CMT Radiation / RCS Activity Degradation Threat: Loss Threshold:

3. CVCS letdown radiation> 2.50E+t)I pCi/mi on SJ-RE-0I (016).

Detlnition(s):

N One 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 white maintaining a continilolts teed 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 presstire control valve and upstream of the mixed bed demineralizers. the letdown stream passes by radiation monitor SJRE-01 which will warn of fission products in the letdown coolant if a ftiet element failure occurs. The monitor is located in the Primary Sample Sink Room.

The CVCS letdown monitor SJ-RE-0l provides indication in the Control Room on Patiel RM-l I channel 016 with a range of l.7E-03 to l.7E+03 pCi/mI (ret. 2.3). The HI-HI (RED) alarm is 5E0 + background +

thackground x 0.05) (ref. 4) and represents a total fuel clad tailitre in excess of 1% in 30 minutes ret. 2. 3L 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-I

3. OTA-SP-RMO1 1 Radiation Monitor Control Panel RM-l 1

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 205 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 Fission Prodtict Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: C. CMT Radiation / RCS Activity Degradation Threat: Potential Loss Threshold:

None.

Page 206 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: D. CMT Integrity or Bypass Degradation Threat: Loss Threshold:

None.

Page 207 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clat Category: D. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:

None.

Page 208 of 241 INFORMATION USE

EIP-ZZ-00l0l ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: E. Emergency Coordinator Judgment Degradation Threat: Loss Threshold:

I. 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 harrier de2radation 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 he 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 he used by the Emergency Coordinator in determining whether the Fuel Clad harrier is lost.

Callatvay Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment Fuel Clad Loss 6.A Page 209 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: E. Emergency Coordinator Judgment Degradation Threat: Potential Loss Threshold:

I. Any condition in the opinion of the Emergency Coordinator that indicates potential loss of the Fuel Clad harrier.

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, harrier monitoring capability and dominant accident sequences.

Imminent barrier degradation exists if the degradation will likely occtir within relatively short period of time based on a projection of ctirrent 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 instrtimentation and consideration ol of f site monitoring results.

Dominant accident sequences lead to degradation of all fission prodttct harriers and likely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss of AC power (Station Blackottt) and ATWS EALs to asstire timely emergency classihcation decLarations.

This threshold addresses any other factors that are to he uised 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 harrier 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 210 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUTvIENT Attachment 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: A. RCS or SG Tube Leakage Degradation Threat: Loss Threshold:

1. An automatic oi manual ECCS (SI) actuation required by EITHER:

UNISOLABLE RCS leakae.

SG tube RUPTURE.

Delmnition(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 < 1 849 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 LINISOLABLE RCS leakage through an interfacing system. The mass loss may be into any location inside containment, to the secondary-side (i.e.. steam generator tttbe leakage) or outside of containment.

A steam generator with primary-to-secondary leakage of stifficient magnitude to require a safety injection is considered to he RUPTURED. If a RUPTURED steam generator is also FAULTED outside of containment.

the declaration escalates to a Site Area Emergency since the Containment Barrier Loss threshold I .A will also he met.

Callawav 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 I .A Page 211 of 241 INFORMATION USE

EIP-ZZ-00 701 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: A. RCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold:

I. Operation of a standby charging pump is required by EITHER:

UNISOLABLE RCS leakage.

SG ttihe leakage.

Definition(s):

UNISOMBLE - An open or breached system line that cannot he 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. I). Approximately 12 gpm of charging flow bypasses the RCS due to leakage through the RCP seals; thus, the Normal Charging Pump can deliver 130 gprn 12 gpm 118 gpm (rounded to 120 gprn 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 boutnclary 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 secondaryside (i.e.. steam generator tube leakage) or outside of containment.

If a leaking steam generator is also FAULTED outside of containment, the declaration escalates to a Site Area Emergency since the Containment Barrier Loss threshold I .A will also he met.

Caltaway Basis Reference(s):

I. 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 I .A Page 212 of 241 INFORMATION USE

EIP-ZZ-0010l ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 - Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: A. RCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold:

2. CSFST tnterity-RED Path conditions niet.

Deilnition(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 tPTS) Limit Curve A indicates an extreme challenge to the safety function tvhen plant parameters are to the left of the limit curve following excessive RCS coolciown under pressure (ref. 1, 2).

This condition indicates an extreme challenge to the integrity of the RCS pressure boundary due to presstirized 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):

CSF- 1, Critical Safety Function Statti.s Trees Figure 4 Integrity and 4a Limit A Ctirve

2. FR-P. 1. Response to Imminent Pressurized Thermal Shock Condition

3. NET 99-01. RCS or SG Tube Leakage Reactor Coolant System Potential Loss l.B Page 213 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: B. Inadequate Heat Removal Degradation Threat: Loss Threshold:

None.

Page 214 of 241 INFORMATION USE

EIP-ZZ-OOIOt ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: B Inadequate Heat Removal Degradation Threat: Potential Loss Threshold:

1. CSFST Heat Sink-RED path conditions met.

AND Heat sink is required.

Definition(s):

None Basis:

In combination with EC Potential Loss B.2. meeting this threshold results in a Site Area Emergency.

Critical Safety Function Status Tree CSFST) Heat Smk-RED path indicates the heat sink function is tinder 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. I ).

The phrase and heat sink recjuired precludes the need for classification for conditions in which RCS iiressure is less than SG pressure or Heat Sink-RED path entry was created through operator action directed by an EOP. For example. FR-H. I is entered from CSFST Heat Sink-Red. Step I 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 350°F or RCS pressure is greater than 360 psig. It 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 becatise 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 he 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 he met. This condition warrants a Site Area Emergency declaration because inadequate RCS heat removal may result in fuel heat-up sufficient to damage the cladding and increase RCS pressure to the point where mass will be lost from the system.

Page 215 of 241 INFORMATION USE

EIP-ZZ-0010I ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Catlaway Basis Reference(s):

CSF- 1, Critical Safety Function Status Trees Figtire 3 Heat Sink

2. FR-H.!. Response to Loss of Secondary Heat Sink

3. NE! 99-0 1. Inadequate Heat Removal RCS Loss 2.B Page 216 of 24! INFORMATION USE

EIP-ZZ-00 101 ADDENDUrvI 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 - fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: C. CMT Radiation! RCS Activity Degradation Threat: Loss Threshold:

1. Containment radiation > 59 R/hr on GT-RE-59 (591) or GT-RE-6t) (601).

Definition(s)

N one Basis:

Containment radiation monitor readings greater than 59 RJhr (ref. I) 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.. tvithin Technical Specifications) into the Containment atmosphere. Because of the very high fuel clad integrity, only small amoitnts of noble gases would be dissolved in the primary coolant.

Monitors tised for this fission prodcict barrier loss threshold are the Containment High Range Radiation Monitors GT-RE-59 (Panel RM-l I channel 591) and GT-RE-60 (Panel RM-1 I 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. I since it indicates a loss of the RCS Barrier only.

There is no Potential Loss thi-eshold associated with RCS Activity / Containment Radiation.

Callaway Basis Reference(s):

1. EPCI-l701

2. OTA-SP-RMO I I, Radiation Monitor Control Panel RM- 11

3. NEI 99-01, CMT Radiation / RCS Activity RCS Loss 3.A Page 217 of 241 INFORMATION USE

EIP-ZZ-00I0l ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: C. CMT Radiation! RCS Activity Degradation Threat: Potential Loss Threshold:

None.

Page 218 of 241 INFORMATION USE

EIP-ZZ-00J01 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: D. CMI Integrity or Bypass Degradation Threat: Loss Threshold:

None.

Page 219 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rex. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: D. CMT Integrity or Bypass Degradation Threat: Potential Loss Thresh old:

None.

Page 220 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: E. Emergency Coordinator Judgment Degradation Threat: Loss Threshold:

I. 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 deeradation exists if the degradation wilt 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 shotild 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. NEt 99-01, Emergency Director Judgment RCS Loss 6.A Page 221 of 241 INFORMATION USE

EIP-ZZ-00l0I ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: E. Emergency Coordinator Judgment Degradation Threat: Potential Loss Threshold:

1. Any condition in the opinion of the Emergency Coordinator that indicates potential loss of the RCS barrier.

Definition(s):

None Basis:

The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the RCS barrier is potentially lost. Stich a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

Imminent barrier degradation exists if the degradation will likely occtir within relatively short period of time based on a projection of cLirrent 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 Blackotit) 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 222 of 241 INFORMATION USE

EIP-ZZ-001Ol ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containtient Category: A. RCS or SG Tube Leakage Degradation Threat: Loss Threshold:

1. A leakin or RUPTURED SG is FAULTED outside of containment.

Definition(s):

f4 ULTED 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. I and Loss A. 1, respectively. This condition represents a by pass of the containment barrier.

FAULTED isa defined term within the NEJ 99-01 methodology: this determination is not necessarily dependent upon entry into, or diagnostic steps within, an FOP. 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 hound for this aspect of the containment barrier is analogous to the lower bound criteria specified in 1C SU4 for the fuel clad barrier (i.e., RCS activity saluies) 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 cootdown the plant.

orto 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 funning and being supplied by a ruptured steam generator that has not been isolated. this threshold is met. Manual Operator action can NOT be credited.

Page 223 of 241 INFORMATION USE

EIP-ZZ-00l0l ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Steam releases associated with the expected operation of a Steam Generatot 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 throcigh 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 ttihe leak or rtipture. there may be minor radiological releases through a secondary-side system component (e.g.. air ejectors, glad seal exhausters. valve packing. etc.). These types of releases do not constitute a loss or potential loss of containment bttt should be evaluated using the Recognition Category R ICs.

The ECLs resulting from primary-to-secondary leakage, with or withotit 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 I SU5. I Requires operation of a standby charin (makeup) Site Area Emereency Alert perFAl.l pump per ES 1 .1 (RCS Barrier Polenlial Loss)

Reqciires an automatic or Site Area Emergency manual ECCS (SI) actuation Alert per FAJ.l per FSI.l (RCS Barrier Loss)

There is no Potential Loss threshold associated with RCS or SG Tube Leakage.

Callaway Basis Reference(s):

E2. Faulted Steam Generator Isolation

2. 1-3, Steam Generator Tube Rupture

3. NEI 99-01. RCS or SG Tube Leakage Containment Loss I .A Page 224 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: A. RCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold:

None.

Page 225 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: B. Inadequate Heat Removal Degradation Threat: Loss ThreshoLd:

None.

Page 226 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: B. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold:

1. CSFST Core Cooling-RED path conditions met.

AND Restoration procedures not effective within 15 mm.

(Note 1)

Note 1: The Emergeney Coordinator shotild declare the event promptly upon determinine that time limit has been exceeded, or will likely he exceeded.

Definition(s):

None Basis:

Critical Safety Function Status Tree (CSFST) Core Cooling-RED path indicates significant core exit superheating and core uncoverv. The CSFSTs are normally monitored using the SPDS display on the Plant Computer (ref. I ).

The ftinction 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 he made if the effectiveness of the restoration procedures is also evaltiated. tf core exit thermocouple (TC) readings are greater than I ,200F (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.

Callawav Basis Reference(s):

1. CSF-I. Critical Safet 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. NEt 99-01. Inadeqttate Heat Removal Containment Potential Loss 2.A Page 227 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: C. CMT Radiation/RCS Activity Degradation Threat: Loss Threshold:

None.

Page 228 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 - Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: C. CMT Radiation/RCS Activity Degradation Threat: Potential Loss Threshold:

I. Containment radiation> 14,000 R/hron GI-RE-59 (591) orGT-RE-60 (601).

Definition(s):

NC) fl C Basis:

Containment radiation monitor readings greater than 14.000 R/hr (ref. 1) indicate signihcant fuel damage well in excess of that required for loss of the RCS harrier and the Ftiel Clad barrier.

The readings are higher than that specified for Fctel Clad harrier Loss C. I and RCS barrier Loss C. I Containment radiation readings at or above the Containment barrier Potential Loss threshold, therefore, signify a loss of two fission product harriers 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-I 1 channel 591) and GT-RE-60 (Panel RM-l I channel 601) (ref. 2).

The radiation monitor reading corresponds to an instantaneotts release of all reactor coolant mass into the containment, assuming that 2Ocf of the fuel cladding has failed. This level of fuel clad failtire is well above that used to determine the analogous Fuel Clad Barrier Loss and RCS Barrier Loss thresholds.

NUREG-122$, Source Estimations During Incident Response to Severe Nticlear 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 Relrence(s):

1. EPCI-1701

2. OTA-SP-RMOI 1, Radiation Monitor Control Panel RM-l I

3. NEt 99-0 1. CMT Radiation / RCS Activity Containment Potential Loss 3.A Page 229 of 241 INFORMATION USE

EIP-ZZ-00l0I ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: D. CMT Integrity or Bypass Degradation Threat: Loss Threshold:

1 . Containment isolation is required.

AND EITHER:

Containment integrity has been Lost based on Emergency Coordinator judgment.

UNISOLABLE pathway from containment to [he environment exists.

Definition(s):

UNJSOL4BLE An open or breached system line that cannot he isolated, remotely or locally.

Basis:

These thresholds address a situation where containment isolation is recluired and one of two conditions exists as discussed below. Users are reminded that there may be accident and release conditions that simultaneously meet both bul leted 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 dciring accident conditions, it is expected that the Emergency Coordinator will assess this threshold tising 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.

Page 230 of 241 INFORMATION USE

EIP-ZZ-0010I ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Second Threshold Conditions are such that there is an UNISOLABLE pathway for the migration of radioactive material from the containment atmosphere to the environment. As used here. the term environment includes the atmosphere of a room or area, outside the containment, that may. in turn, communicate with the outside-the-plant atmosphere (e.g., throtigh discharge of a ventilation system or atmospheric Leakage). Depending tipon 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 vatves remained open after a containment isolation was required tie.. containment isolation was not successful).

There is nos 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 steaimhigh 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 Atixiliary Building. The radioactivity would he 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 he met. Depending upon radiation monitor locations ind sensitivities.

this leakage cotild he detected by any of the four monitors depicted in the figure and catise the first threshold to he met as well.

Following the leakage of RCS mass into containment and a rise in containment pressttre. 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 hut should be evaluated tising the Recognition Category R ICs.

The status of the containment barrier during an event involving steam generator tube leakage is assessed tising Loss Threshold A. I Callaway Basis Reference(s):

1. NEI 99-01, CMT Integrity or Bypass Containment Loss 4.A Page 231 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 Fission Product Barrier LossfPotentia Loss Matrix and Bases Barrier: Containment Category: D. CMT Integrity or Bypass Degradation Threat: Loss Threshold:

2. Indications of RCS leakage outside of containment.

Detlnition(s):

N one Basis:

ECA-l.2 LOCA Outside Containment (ref. I) provides instructions to identify and isolate a LOCA outside of the containment. Potential RCS leak pathways outside containment include (ret. 1, 2):

Residual Heat Removal

Safety Injection

Chemical & Volume Control

RCP seals

PZR/RCS Loop sample lines Containment sump. temperattire. 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). tncreases in sump. temperature.

presstire, 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 fttel clad barrier has not been lost, radiation monitor readings otitside of containment may not increase significantly: however, other unexpected changes in sump levels, area temperatures or pressures, flow rates. etc. should he 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 I 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.l to be met as weLl.

To ensure proper escalation of the emergency classification, the RCS leakage outside of containment mtist be related to the mass loss that is causing the RCS Loss and/or Potential Loss threshold A. I to be met.

Page 232 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Callaway Basis Reference(s):

1. ECA-I.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 233 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Figure 1: Containment Integrity or Bypass Examples Thnshold Airborne

- release Effluent- _._-_ - --

Auxiliary-Building Monitor : -. .

from Inside-Reactor Ven til pdURvay

S___.

Building .?

*5 ._ - . - -

Area-.

Monitor Open -valve *5 Open-valve

- S_S 4

Dampe 1 Penetration fliieshold Airborne Airborne- -

S tlonitor r7 - -. -

L0oen.vaive ) Onen valve

- 1 I hr hjId :-

Threshold-- -

In te rfa c e leak a a e RC Sleakag Airborne- 2 4;-:.-. - RE - -

s. Process - ..

F tvloflitot

- f,_

r

- Closed-4 4_rn

- Cooling-Open-valve 1.- Open-valve Pump

- -v-- -*. 4-. ; -

RCP-Seal Cooling Page 234 of 241 INFORMATION USE

EIP-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: D. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:

1. CSFST Containment-RED path conditions met.

Detl nhtion( 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).

4$ 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 Coolatit Accident (LOCA). there exists a potential to lose the Containment Barrier. To reach this level. thei-e 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):

CSF- 1, Critical Safety Function Status Trees Containment Figure 5

2. CaIc 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 235 of 241 INFORMATION USE

E[P-ZZ-00101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: D. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:

2. Containment hydrogen concentration 4%.

Definition(s):

None Basis:

Following a design basis accident, hydrogen gas may be generated inside the containment by reactions such as zirconium metal with water, corrosion of materials of construction and racliolysis 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. I).

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 inadeqtiate 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 Fttel Clad and RCS barriers must have occurred.

With the Potential Loss of the containment barrier, the threshotd hydrogen concentration, therefore, will likely warrant declaration of a General Emergency.

Two Containment hydrogen monitors (GS Al-It) and GS Al-19) with a range of 0% to 10% provide indication on Control Room Panel RLO2O and ERF[S (ref. 3). The hydrogen monitors require a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months warmup period when starting from the OFF position and 15 minutes when starting from STANDBY (ref. 4, 5). If an actual hydrogen concentration meastirement is unavailable. CA-3 (ref. 6) may be used to estimate the Containment atmosphere hydrogen concentration.

The existence of an explosive mixttire 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.

Page 236 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Callaway Basis Reference(s):

1. ESAR, 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-0000I. Containment Hydrogen Control System

5. CaIc No. 392.2 XX-95. Callaway Containment Parameters FOP Action Values. Setpoint ID TI0] &

Ti 02

6. CA3. Hydrogen Flammability in Containment

7. NEI 99-01, CMT Integrity or Bypass Containment Potential Loss 4. B Page 237 of 241 INFORMATION USE

EIP-ZZ-0010J ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: D. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:

3. Containment pressure > 27 psig with < one full train of containment depressurization equipment operating per design for 15 mm.

(Notes 1, 9)

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

Now 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) sttpplies 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 recluirement. 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. Dtiring normal operation. all four fan units may he operating. In postaccident operation following an actuation signal. the Containment Cooling System fans are designed to start atitomaticalty 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 shotitd actuate aoL! begin performing its function. The design basis accident analyses and evaluations assume the loss of one Containment Spray System train and one Containment Cooling System train (ref. 7). Consistent with the design requirement. one full train of depressurization equipment is therefore defined to be the availability of one train of each system. If less than this equipment is operating and Containment pressure is above the actuation setpoint, the threshold is met.

This threshold describes a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design. The 15-minute criterion is incicided 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.

Page 238 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases 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) EigtLre 5, Containment

5. FR-Z. I. Response to High Containment Pressure

6. Technical Specifications. Table 3.3.2-1

7. Technical Specifications. 33.6.6

8. NE! 99-01. CMT Integrity or Bypass Containment Potential Loss 4.C Page 239 of 241 INFORMATION USE

EIP-ZZ-f)0 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: E. Emergency Coordinator Judgment Degradation Threat: Loss Threshold:

1. Any condition in the opinion of the Emergency Coordinator that indicates loss of the Containment barrier.

Definition(s):

None Basis:

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 occtir 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 harriers and likely entry to the EOPs. The Emergency Coordinator should he mindftil 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 240 of 241 INFORMATION USE

EIP-ZZ-00 101 ADDENDUM 2 Rev. 015 EMERGENCY ACTION LEVEL TECHNICAL BASES DOCUMENT Attachment 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: E. Emergency Coordinator Judgment Degradation Threat: Potential Loss Threshold:

I. 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 shotild include imminent barrier degradation, harrier monitoring capability and dominant accident sequences.

Imminent barrier decradation 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 alt fission product barriers and likely entry to the EOPs. The Emergency Coordinator should he mindful of the Loss of AC power (Station Blackotlu and ATWS EALs to assure timely emergency classification declarations.

This threshold addresses any other factors that may he used by the Emergency Coordinator in determining whether the Containment Barrier is lost.

Callaway Basis Reference(s):

1. NEt 99-01, Emergency Director Judgment PC Potential Loss 6.A Page 241 of 241 INFORMATION USE

ULNRC-06445, Revised Emergency Action Level Technical Basis Document

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