ML17268A143

Emergency Preparedness Procedures, Including Revision 048 to 13.1.1, Revision 02 to 13.1A, and Revision 037 to 13.8.1

ML17268A143
Person / Time
Site:	Columbia
Issue date:	09/12/2017
From:	Energy Northwest
To:	Office of Nuclear Reactor Regulation
References
Download: ML17268A143 (223)

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DISTRIBUTION -VOLUME 13- EMERGENCY PREPAREDNESS PROCEDURES TO: ENERGY NORTHWEST EXTERNAL CONTROLLED COPY PROCEDURE HOLDERS The following documents have been revised and are to be inserted into your controlled copy manual and the superseded revisions removed and destroyed. No receipt acknowledgement is required for the listed document(s).

Should you have questions on this distribution please contact Kim Saenz, Records and Information Supervisor at (509)377-2492 or KDSaenz@energy-northwest.com.

Distribution Date: 9/12/17 JG Procedure Number Revision Procedure Number Revision Procedure Number Revision 13.1.1 048 13.1.1A 032 13.8.1 037 EXTERNAL DISTRIBUTION Copy# Location Procedures 26 Reqion IV - US Nuclear Requlatorv Commission All 28 Region IV- US Nuclear Regulatory Commission All 52 State of Washinqton, Military Department All 55 Federal Emergency Manaqement Agency (FEMA) All 57 Benton County - Department of Emerqency Manaqement All 75 Department of Health Radiation Protection All 87 Document Control Desk - NRC All 142 Hanford EOC/SMT All 164 Oreqon State Department of Enerqy All 223 Franklin County Emergency Management -

All 224 Washinqton State Department of Health - Office of Radiation Protection All NOTE: PPM 13.1.1 is printed in color, single sided, on 11" x 17" paper, folded, 3-hold punched and stapled on the top Updated 9/12/17

CG1.1 GENERAL EMERGENCY Lon of RPV lrw91Dy effitdlng l'Uel c9d lr'llllgrttywtil oont11lrwn.... chlllf9'd

~I__l~~l-~l~*~l_6~~1-~I RPV level LT -161 in. for GE 30 min . (Note 1)

AND SITE AREA EMERC::..... 1-JCY CS1 .1 l.ol1 of RPV c.pebllty l~rery

~I__1~~1~~1~

(1) CONTAINMEN T CLOSURE AND efflctlrlg c:ore deitey hellt r.rnov&1I

• ~!_6~~

0521 established 1-~! CA1 .1 ALERT L.I_ _,l. __,l_ __,.!--"-4--'-l-6~,_I-~I (1) loss of RPV klven tory as indicated by RPV level LT-50 In.

cu1 .1 UNUSUAL EVENT

.l __,l_ __,i_~l-*~,_I~6~1~_,I (1) UNPLANN ED loss of reactor coolant results in R PV level less than a required lower Umit for GE 15 m in . (Note 1) 6DY. of the folowiig indications of containment challenge: OR OR RPV level l T -129 in .

(2) RPV level wruz1 be monitored for GE 15 min. (Note 1) (2) RPV level a.a.a21 be monitored CONTAINMENT CLOSURE w:z1 established (Note 6) OR AND AND Explosive mixture inside PC (2) CONTAINMENT C LOSURE established (H2 GE 6% and 0 2 GE 5%) UNPLANNED increase In ifl:t Table 1 sum p o r pool UNPLANN ED increase in w Table 1 sump or pool levels AND due to a loss o f RPV Inventory UNPLANNED rlse In PC pre$$U te levels due to a loss of RPV inventory RPV level l T -161 In .

RB area radiation G T IDl Maximu m Safe Operating level (PPM 5.3.1 Table 24)

CS1 .2 l~-~l~~l-~l-*~~ I ~6~~1-~I CG1 .2 I I I I 4 I 6 I I RPV level an.nm. be monlored for GE 30 min. (Note 1) 1 RPV level AND W!l21 be monitored fo r G E 30 min. (Note 1)

AND RPV Core uncovery Is Indicated by iID'.. of th e following : Table 1 S um p s/Pool Level Core uncovery is indicated by am.of the followlng :

UNPLANNED wetwell level rise GT 2 Inches

• UNPLANNED wetwell level rise GT 2 inches (PPM 5 .2 .1 entry condition) Anx Vlllld HI-HI level alarm on R-1 (PPM 5 .2 .1 entry condition) VALID indication of RB room flooding as Identified by through R-5 sumpa

• VALID indication of RB room flooding as Ident ified by high level alarms (PPM 5.3 .1 Table 25)

Observation of UNISOLABLE RCS leakage outside high level alarms (PPM 5 .3.1 Table 25)

Observation of UNISOLABLE RCS leakage outside EDRGE 25GPM FORGE 10GPM Sta rtup Transformer TR-$

primary containm ent o f sufficient magnitude to indicate primary containme nt of sufficient magnitude to Indicate core u ncovery Wetwell level rise Backup Transformer TR-8 core uncovery Observat ion of UNISOLABLE RCS Baekfeed 500 KV power through Main AND leakage Transformers (If 1trelldy 11Hgned In

&rt of the following indications of conta inm ent challenge: modes 4. 5. def only)

CONTAINMENT CLOSURE !!21 established (Note 6)

Explosive mixture inside PC Ons ltit (H 2G E 6% and 0 2GE 5 %) 001 UNPLANNED rise In PC pressure 002 RB area radiation GT iilY Maximum Safe Operating Main Gentr&IOI' via TR*N11N2 level (PPM 5.3.1 Table 24)

Lon of d ofJlllil Ind .d on1lt9 AC l)OW'el' to lt!Mrgency buMI Lon of .tl.tM:o ne AC ~ 10WC.1D9""'rv-ncy buMtfor15 for15mlnutn orlonger mlni.tnOf'lol'Qlll'

!~_,l_ __,l_~!-*~~ I ~6~~1~OE,,,F~! L.!_ _.l __,1...___,.l--"- 4--'-l-6~,_I_,.oe,,,*'-I' 2 CA2.1 Loss of ill offsite and ii onsite AC power capability to c u2.1 AC power capability, Table 2, to emergency buses SM-7 loH of emergency buses SM-7 and SM-8 for GE 15 min. (Note 1) and SM-8 reduced to a single power source for GE 15 min .

c C<>ld SDI Emergonoi AC Power (Note 1)

AND Am.,a ddltional single power source fa ilure will result in a loss of .iJ.! AC power to SAFE TY SYSTEM S RefUel Systom I

Molf\lnct.

3 Table 7 RCS fitehHt Ourltton Thre&holc:ls CA3.1

-- I

• UNPLANNED Increase ln RCS temperature to GT 200°F I 6 I c u 3.1 UNPLANNED l~* In RCS tlm!*'INN i~ ~~~...::::::;::;:::;..:;;::;::=::::;--~

I I I I 4 UN PLANNED increase In RCS tempera ture to GT 200"F I 6 I I If en RCS neat remove! system Is In operetlon whtlln this for GT Table 7 duration (Note 1)

RCS time frlme and RCS tempereture ls being reduced tne EAL ls!121.1ppllce'Ole OR T.mp. CU3.2 1

..._ _,1. __,j_ __,.l--"- * --'-I~6,__,_I_ _,!

UNPLANNED RPV pressure increase GT 10 psig Contai nm ent Heat -up RCS Statu s Clos ure Statu s Ou ration loss of I.!! RCS temperature and RPV water level Indication for GE 15 min . (Note 1)

Intact NIA 60 min.

• Lon of vitsl OC po\W(fot 15 mlnutlllor longer 4 esta blished 20 min.

• l"":'.C~U4~.~1~1;:::~1;:'._...::::;l::::::;l~4~;1~6~1;::::_,! ___,.

W lntact Loss of .o21established Omin .

VIUIDC Indicated voltage l T 108 VOC on~ 125 VDC buses OP-S1-1 and DP-51-2 for GE 15 min . (Note 1)

T*bt** Commu ntcaaon Methods C U6.1 I

-- I I Loud tJ ol'll"' or olPlh ~mmi..rtlcetioN ~pe bllltH I 4 I 6 I DEf I 5 System Onslt* ORO NRC l oss of ill Table 4 onsite communication met hods OR Loss of ill Table 4 ORO comm.Jnk:atlon methods x

"""""' Plant Public Address (PA) Syste m OR Loss of ill Table 4 NRC communication methods Plant Telephone System x x Plant Radio System Operations and x H1.1:1 rdou1 event "'9cfng 1 SAFETY SYSTEM neieded t'or tM cUl'T9ntoptr11t1r"9mO(ft Security Channels CA6.1 I -~1~~1-~1~*~1_6~~1-~I

~

Offsite calling capabitlty from t he x x Table B H1urdou1 Events The occurrence of am. Table 8 hazardous event Contro l Room via direct telephone AND 6

Sofely Long distance calling capabiHty on the commercial phone system x x Seismic event Internal or external FLOODING event High winds Tornado strike FIRE Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode AND EITHER :

Event damage has caused Indications of degraded performance l o a second train of a SAFETY SYSTEM needed for the current operating mode OR EXPLOSION Even! damage has resulted in VISIBLE DAMAGE lo a Volcanic ash fallout second train of a SAFETY SYSTEM needed fo r the Other event& w ith similar hazard current operal lng mode characteristics as determined by the Shift (Notes 9 , 10)

Ma nager 13.1.1Rev. 48 CLASSIFYING THE EMERGENCY Modes: I~_ 4~1 1 Cold Shutdown 5

Refueling I I DEF I Defueled 8'ERGY NORTHWEST COLD CONDITIONS 9/1212017 (RCS :'.S 200°F)

Initials Date Number: 13.1.1 A Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

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Classifying The Emergency - Technical Bases Page: 1 of 187 PCN#:

PLANT PROCEDURES MANUAL N/A Effective Date:

Illllll lllll llll llllll llll llllll llllll Ill llll 13.1.1A 09/12/17

e

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 2of187 DESCRIPTION OF CHANGES Implemented NEI 99-01 Rev 6 EAL scheme

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Title:

Classifying The Emergency - Technical Bases Page: 3 of 187 TABLE OF CONTENTS 1.0 PURPOSE ............................................................................................................................... 4

2.0 REFERENCES

........................................................................................................................ 5 3.0 DISCUSSION ........................................................................................................................... 6 3.1 Background .............................................................................................................................. 6 3.2 Fission Product Barriers ........................................................................................................... 6 3.3 Emergency Classification Based on Fission Product Barrier Degradation ................................ 7 3.4 EAL Organization ..................................................................................................................... 8 3.5 Technical Bases Information .................................................................................................. 10 3.6 Mode Applicability .................................................................................................................. 11 3.7 Basis ...................................................................................................................................... 11 3.8 CGS Basis Reference(s) ........................................................................................................ 11 3.9 Operating Mode Applicability ................................................................................................. 11 3.10 Storage Operations ................................................................................................................ 12 4.0 PROCEDURE ........................................................................................................................ 12 4.1 General Considerations ......................................................................................................... 12 4.2 Classification Methodology ..................................................................................................... 14 5.0 DEFINITIONS ........................................................................................................................ 18 6.0 ATTACHMENTS .................................................................................................................... 30 6.1 EAL Technical Bases ............................................................................................................. 31 6.2 Tables .................................................................................................................................. 177 6.3 Safe Operation & Shutdown Areas Table 9 Bases ............................................................... 183 6.4 Emergency Classification Chart Distribution ......................................................................... 187

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Classifying The Emergency - Technical Bases Page: 4 of 187 1.0 PURPOSE 1.1 This document provides an explanation and rationale for each Emergency Action Level (EAL) included in the EAL Upgrade Project for Columbia Generating Station (CGS). It should be used to provide historical documentation for future.reference and serve as a training aid.

Decision-makers responsible for implementation of PPM 13.1.1, Classifying the Emergency, may (though not required) use this document as a technical reference in support of EAL interpretation.

1.2 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.

1.3 Because the information in a basis document can affect emergency classification decision-making (e.g., the Emergency Director 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, some criteria/values in the CGS EALs and fission product barrier thresholds are drawn from plant AOPs and EOPs. The impact of any changes to those procedures on EAL bases must be evaluated for screening in accordance with the provisions of 10 CFR 50.54(q). This Emergency Plan Implementing Procedure is identified by reference in the Emergency Plan. Changes to the EAL Scheme (Attachments 6.1) require an LDCN since it is part of the Emergency Plan .

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2.0 REFERENCES

2.1 Developmental 2.*1.1 NEI 99-01 Revision 6, Methodology for the* Development of Emergency Action Levels for Non-Passive Reactors, ADAMS Accession Number ML12326A805 2.1.2 Technical Specifications Table 1.1-1 Modes 2.1.3 NSIR/DPR-ISG-01 Interim Staff Guidance, Emergency Planning for Nuclear Power Plants 2.1.4 10CFR 50.72 Immediate Notification Requirements for Operating Nuclear Power Reactors 2.1.5 RIS 2007-02 Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events, February 2, 2007 2.1.6 NUREG-1022 Event Reporting Guidelines: 10CFR50.72 and 50.73 2.1.7 Hl-2002444, Holtec International Final Safety Analysis Report for the HI-STORM 100 Cask System, USNRC Docket No. 72-1014, Chapter 7, Confinement

• 2.1.8 2.1.9 2.1.10 PPM 1.20.3, Outage Risk Management 10CFR 50. 73 License Event Report System M570, General Arrangement Plan - El. 572 ft - O in. and El. 606 ft - 10 1/2 in. -

Reactor Building 2.1.11 Certificate of Compliance No. 1014 Appendix A Technical Specifications for the HI-STORM 100 Cask System Section 1.1 Definitions 2.1.12 SWP-PR0-03, Procedure Writer's Manual 2.1.13 CGS Physical Security Plan 2.1.14 CGS Graphics Plant Drawing 902118-P 2.1.15 Energy Northwest Columbia Generating Station Offsite Dose Calculation Manual, Amendment 52

• 2.2 Implementing 2.2.1 PPM 13.1.1, Classifying the Emergency 2.2.2 EP-01 Emergency Plan 2.2.3 Columbia Generating Station NEI 99-01 Revision 6 EAL Comparison Matrix

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Classifying The Emergency - Technical Bases Page: 6 of 187 3.0 DISCUSSION 3.1 Background 3.1.1 EALs are the plant-specific indications, conditions or instrument readings that are utilized to classify emergency conditions defined in the CGS Emergency Plan.

3.1.2 In 1992, the NRC endorsed NUMARC/NESP-007 "Methodology for Development of Emergency Action Levels" as an alternative to NUREG-0654 EAL guidance.

3.1.3 NEI 99-01 (NUMARC/NESP-007) Revisions 4 and 5 were subsequently issued for industry implementation. Enhancements over earlier revisions included:

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

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

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

• 3.1.4 Subsequently, Revision 6 of NEI 99-01 has been issued which incorporates resolutions to numerous implementation issues including the NRC EAL Frequently Asked Questions (FAQs). Using NEI 99-01 Revision 6, "Methodology for the Development of Emergency Action Levels for Non-Passive Reactors," November 2012 (ADAMS Accession Number ML12326A805) (ref. 2.1.1 ), CGS conducted an EAL implementation upgrade project that produced the EALs discussed herein.

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

Many of the EALs derived from the NEI methodology are fission product barrier threshold based. That is, the conditions that define the EALs are based upon thresholds that represent .

the loss or potential loss of one or more of the three fission product barriers. "Loss" and "Potential Loss" signify the relative damage and threat of damage to the barrier. A "Loss" threshold means the barrier no longer assures containment of radioactive materials. A "Potential Loss" threshold implies an increased probability of barrier loss and decreased certainty of maintaining the barrier.

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• 3.2.1 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 is the reactor coblant system pressure boundary and includes the RPV and all reactor coolant system piping out to and including the isolation valves.

c. Containment (PC): The drywell, the wetwell, their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves comprise the PC barrier. Containment Barrier thresholds are used as criteria for escalation of the ECL from Alert to either a Site Area Emergency or a General Emergency using the Fission Product Barrier table.

3.3 Emergency Classification Based on Fission Product Barrier Degradation The following criteria are the bases for event classification related to fission product barrier loss or potential loss:

3.3.1 Alert

Any loss or any potential loss of either Fuel Clad or RCS barrier 3.3.2 Site Area Emergency:

Loss or potential loss of any two barriers 3.3.3 General Emergency:

Loss of any two barriers and loss or potential loss of the third barrier

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Classifying The Emergency - Technical Bases Page: 8 of 187 3.4 EAL Organization 3.4.1 The CGS EAL scheme includes the following features:

a. Division of the EAL set into three broad groups:

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

2) EALs applicable only under hot operating modes - This group would only be reviewed by the EAL-user when the plant is in Hot Shutdown, Startup, or Power Operations mode.

3) EALs applicable only under cold operating modes - This group would only be reviewed by the EAL-user when the plant is in Cold Shutdown, Refuel or Defueled mode.

3.4.2 The purpose of the groups is to avoid review of hot condition EALs when the plant is in a cold condition and avoid review of cold condition EALs when the plant is in a hot condition. This approach significantly minimizes the total number of EALs that must be reviewed by the EAL-user for a given plant condition, reduces EAL-user reading burden and, thereby, speeds identification of the EAL that applies to the emergency .

• 3.4.3 3.4.4 Within each group, assignment of EALs to categories and subcategories:

Category and subcategory titles are selected to represent conditions that are operationally significant to the EAL-user. The CGS EAL categories are aligned to and represent the NEI 99-01"Recognition Categories". Subcategories are used in the CGS scheme as necessary to further divide the EALs of a category into logical sets of possible emergency classification thresholds. The CGS EAL categories and subcategories are listed in Table 3.4-1.

3.4.5 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 bases in order to obtain additional information concerning the EALs under classification consideration. The user should consult Section 4.0 and Attachment 6.1 of this document for such information .

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Classifying The Emergency - Technical Bases Page: 9 of 187 Table 3.4-1 EAL Groups, Categories and Subcategories EAL Group/Category EAL Subcategory I

Any Operating Mode:

R - Abnormal Rad Release I Rad 1 - Radiological Effluent Effluent 2 - Irradiated Fuel Event 3 - Area Radiation Levels H - Hazards and Other Conditions 1 - Security Affecting Plant Safety 2 - Seismic Event 3 - Natural or Technological Hazard 4- Fire 5 - Hazardous Gas 6 - Control Room Evacuation 7 - Emergency Director Judgment E - Independent Spent Fuel Storage 1 - Confinement Boundary Installation (ISFSI)

• Hot Conditions:

M - System Malfunction 1 - Loss of Emergency AC Power 2 - Loss of Vital DC Power 3 - Loss of Control Room Indications 4- RCS Activity 5 - RCS Leakage

• 6 - RPS Failure 7 - Loss of Communications 8 - Hazardous Event Affecting Safety Systems F - Fission Product Barrier None Degradation Cold Conditions:

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

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Classifying The Emergency - Technical Bases Page: 10of187 3.5 Technical Bases Information 3.5.1 EAL technical bases are provided in Attachment 6.1. A summary explanation of each category and subcategory is given at the beginning of the technical bases discussions of the EALs included in the category. For each EAL, the following information is provided:

a. Category Letter & Title

b. Subcategory Number & Title

c. Initiating Condition (IC) 3.5.2 Site-specific description of the generic IC given in NEI 99-01 Rev. 6.

a. EAL Identifier (enclosed in rectangle)

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

a) First character (letter): Corresponds to the EAL category as described above (R, C, H, M, F or E) b) Second character (letter): The emergency classification (G, S, A or U)

G = General Emergency S = Site Area Emergency A= Alert U = Unusual Event c) Third character (number): Subcategory number within the given category. Subcategories are sequentially numbered beginning with the number one (1). If a category does not have a subcategory, this character is assigned the number one (1 ).

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

2) Classification (enclosed in rectangle)

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

., 3) EAL (enclosed in rectangle)

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

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4) Notes Any notes applicable to the EAL

5) Tables 3.6 Mode Applicability One or more of the following plant operating conditions comprise the mode to which each EAL is applicable: 1 - Power Operations, 2 - Startup, 3 - Hot Shutdown, 4 - Cold Shutdown, 5 - Refuel, D - Defueled, or All. Additionally, unique to the ISFSI, Storage Operations.

(See Section 2.10 for operating mode definitions).

3.7 Basis A basis section that provides CGS-relevant information concerning the EAL as well as a description of the rationale for the EAL as provided in NEI 99-01 Rev. 6.

3.8 CGS Basis Reference(s)

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

• 3.9 Operating Mode Applicability 3.9.1 Power Operations Reactor mode switch is in RUN.

3.9.2 Startup The mode switch is in STARTUP/HOT STANDBY or REFUEL with all reactor vessel head closure bolts fully tensioned.

3.9.3 Hot Shutdown The mode switch is in SHUTDOWN, with all reactor vessel head closure bolts fully tensioned, and reactor coolant temperature is GT 200°F.

3.9.4 Cold Shutdown The mode switch is in SHUTDOWN, all reactor vessel head closure bolts are fully tensioned, and reactor coolant temperature is LE 200°F.

3.9.5 Refuel The mode switch is in REFUEL or SHUTDOWN and one or more reactor vessel head closure bolts less than fully tensioned.

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Classifying The Emergency - Technical Bases Page: 12of187 3.9.6 Defueled All reactor fuel removed from RPV. (Full core off load during refueling or extended outage).

3.9.7 The plant operating mode that exists at the time that the event occurs (prior to any protective system or operator action being initiated in response to the condition) should be compared to the mode applicability of the EALs. If a lower or higher plant operating mode is reached before the emergency classification is made, the declaration shall be based on the mode that existed at the time the event occurred.

3.9.8 For events that occur in Cold Shutdown or Refuel, escalation is via EALs that have Cold Shutdown or Refuel for mode applicability, even if Hot Shutdown (or a higher mode) is entered during any subsequent heat-up. In particular, the fission product barrier EALs are applicable only to events that initiate in Hot Shutdown or higher.

3.9.9 The ISFSI related EAL EU1 .1 is applicable in the Storage Operations mode as defined in the Certificate of Compliance Appendix A Section 1.1 Definitions (ref 2.1.12).

3.10 Storage Operations

• Storage operations include all licensed activities that are performed at the ISFSI while a Spent Fuel Storage Cask (SFSC) containing spent fuel is situated within the ISFSI perimeter.

Storage Operations does not include MPC transfer between the Transfer Cask and the Overpack which begins when the MPC is lifted off the HI-TRAC bottom lid and ends when the MPC is supported from beneath by the Overpack (or the reverse).

4.0 PROCEDURE 4.1 General Considerations When making an emergency classification, the Emergency Director 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.

4.1.1 Classification Timeliness NRC regulations require the licensee to establish and maintain the capability to assess, classify, and declare an emergency condition within 15 minutes after the availability of indications to plant operators that an emergency action level has been exceeded and to promptly declare the emergency condition as soon as possible following identification of the appropriate emergency classification level. The NRC staff has provided guidance on implementing this requirement in NSIR/DPR-ISG-01, "Interim Staff Guidance, Emergency Planning for Nuclear Power Plants" (ref. 2.1.3).

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Classifying The Emergency - Technical Bases Page: 13of187 4.1.2 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 indicator operability, condition existence, or report accuracy. For example, verification could be accomplished through an instrument channel check, response on related or redundant indicators, or direct observati.on by plant personnel.

An indication, report, or condition is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to indicator operability, the condition existence, or the report accuracy is removed. Implicit in this definition is the need for timely assessment. The validation of indications should be completed in a manner that supports timely emergency declaration.

4.1.3 Imminent Conditions For ICs and EALs that have a stipulated time duration (e.g., 15 minutes, 30 minutes, etc.), the Emergency Director 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.

4.1.4 Planned vs. Unplanned Events A planned work activity that results in an expected event or condition which meets or exceeds an EAL does not warrant an emergency declaration provided that: 1) the activity proceeds as planned, and 2) the plant remains within the limits imposed by the operating license. Such activities include planned work to test, manipulate, repair, maintain or modify a system or component. In these cases, the controls associated with the planning, preparation and execution of the work will ensure that compliance is maintained with all aspects of the operating license provided that the activity proceeds and concludes as expected. Events or conditions of this type may be subject to the reporting requirements of 1O § CFR 50. 72 (ref. 2.1.4) .

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Classifying The Emergency - Technical Bases Page: 14of187 4.1.5 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).

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

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

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

For a full discussion of this timing requirement, refer to NSIR/DPR-ISG-01 (ref. 2.1.3).

4.2.1 Classification of Multiple Events and Conditions

a. When multiple emergency events or conditions are present, the user will identify all met or exceeded EALs. The highest applicable ECL identified during this review is declared. For example:

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

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

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

c. Related guidance concerning classification of rapidly escalating events or conditions is provided in Regulatory Issue Summary (RIS) 2007-02, Clarification of NRG Guidance for Emergency Notifications During Quickly Changing Events (ref. 2.1.5).

4.2.2 Consideration of Mode Changes During Classification

a. 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.

• b. For events that occur in Cold Shutdown or Refuel, escalation is via EALs that are applicable in the Cold Shutdown or Refuel 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.

4.2.3 Classification of Imminent Conditions Although EALs provide specific thresholds, the Emergency Director must remain alert to events or conditions that could lead to meeting or exceeding an EAL within a relatively short period of time (i.e., a change in the ECL is IMMINENT). If, in the judgment of the Emergency Director, meeting an EAL is IMMINENT, the emergency classification should be made as if the EAL has been met. While applicable to all emergency classification levels, this approach is particularly important at the higher emergency classification levels since it provides additional time for implementation of protective measures.

4.2.4 Emergency Classification Level Upgrading and Downgrading

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

• b. As noted above, guidance concerning classification of rapidly escalating events or conditions is provided in RIS 2007-02 (ref. 2.1.5).

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Classifying The Emergency - Technical Bases Page: 16of187 4.2.5 Classification of Short-Lived Events Event-based ICs and EALs define a variety of specific occurrences that have potential or actual safety significance. By their nature, some of these events may be short-lived and, thus, over before the emergency classification assessment can be completed. If an event occurs that meets or exceeds an EAL, the associated ECL must be declared regardless of its continued presence at the time of declaration.

Examples of such events include an earthquake or a failure of the reactor protection system to automatically scram the reactor followed by a successful manual scram.

4.2.6 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.

a. 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.

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

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

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

4.2. 7 Transitory Event Classification

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

b. In these cases, no emergency declaration is warranted; however, the guidance contained in NUREG-1022 (ref. 2.1.6) is applicable. Specifically, the event should be reported to the NRC in accordance with 10 CFR § 50.72 (ref. 2.1.5) within one hour of the discovery of the undeclared event or condition. The licensee should also notify appropriate State and local agencies in accordance with the agreed upon arrangements.

In some cases, the licensee discovers that a condition existed that met the Emergency Action Level (EAL) criteria, but no emergency was declared and the basis for the emergency classification no longer exists at the time of this discovery. This may be due to a rapidly concluded event or an oversight in the emergency classification made during the event, or it may be determined during a post event review. In these cases, in accordance with NU REG 1022, no emergency declaration is warranted.

If the licensee does not declare an emergency under these circumstances, an Emergency Notification System (ENS) notification within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of the discovery of the undeclared (or misclassified) event should be performed, in accordance with PPM 13.4.1.

If the licensee does declare an emergency, then all notifications required by PPM 13.4.1 are to be made.

4.2.8 Retraction of an Emergency Declaration

• Guidance on the retraction of an emergency declaration reported to the NRC is discussed in NUREG-1022 (ref. 2.1.6).

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Classifying The Emergency - Technical Bases Page: 18of187 5.0 DEFINITIONS 5.1 Definitions (ref. 2.1.1 except as noted)

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

5.1.1 ALERT Events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA PAG exposure levels.

5.1.2 CAN/CANNOT BE MAINTAINED ABOVE/BELOW The value of an identified parameter is/is not able to be held within the specified limit. The determination requires an evaluation of system performance and availability in relation to parameter values and trends. An instruction prescribing action when a parameter cannot be maintained above or below a specified limit neither requires nor prohibits anticipatory action-depending upon plant conditions, the action may be taken as soon as it is determined that the limit will ultimately be exceeded, or delayed until the limit is actually reached. Once the parameter does exceed the limit, however, the action must be performed; it may not be delayed while attempts are made to restore the parameter to within the desired control band.

5.1.3 CAN/CANNOT BE RESTORED ABOVE/BELOW The value of an identified parameter is/is not able to be brought within the specified limit. The determination requires an evaluation of system performance and availability in relation to parameter values and trends. An instruction prescribing action when a value cannot be restored and maintained above or below a specified limit does not require immediate action simply because the current values is outside the range, but does not permit extended operation beyond the limit; the action must be taken as soon as it is apparent that the specified range cannot be attained.

5.1.4 CONFINEMENT BOUNDARY The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the CGS ISFSI, Confinement Boundary is defined as the Multi-Purpose Canister (MPG) (ref. 2.1. 7) .

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Classifying The Emergency - Technical Bases Page: 19of187 5.1.5 CONTAINMENT CLOSURE The procedurally defined conditions or actions taken to secure Containment (Primary or Secondary) and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. A functional barrier is one which mitigates offsite release during an event.

Containment Closure requires a functional barrier (not necessarily Technical Specification Operable; the appropriate structures, systems, and components are functional) to exist at the time of an event. The site cannot rely on contingency methods to establish a functional barrier after the event has started. In Mode 4 either a functional Primary Containment or a functional Secondary Containment is sufficient to mitigate offsite release. In Mode 5, a functional Secondary Containment is sufficient to mitigate offsite release. Therefore, Containment Closure is met in Mode 4 with either a functional Primary Containment or a functional Secondary Containment. Containment Closure is met in Mode 5 with a functional Secondary Containment.

5.1.6 EPA PAGS Environmental Protection Agency Protective Action Guidelines. The EPA PAGs are expressed in terms of dose commitment: 1 Rem TEDE or 5 Rem COE Thyroid.

Actual or projected offsite exposures in excess of the EPA PAGs requires CGS to recommend protective actions for the general public to Offsite Response Organizations (ORO).

5.1.7 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.

5.1.8 EMERGENCY CLASSIFICATION LEVEL One of a set of names or titles established by the US Nuclear Regulatory Commission (NRG) 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)

• General Emergency

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Classifying The Emergency - Technical Bases Page: 20 of 187 5.1.9 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.

5.1.10 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.

5.1.11 FISSION PRODUCT BARRIER THRESHOLD A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier.

5.1.12 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.

5.1.13 GENERAL EMERGENCY Events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.

5.1.14 HOSTAGE A person(s) held as leverage against the station to ensure that demands will be met by the station .

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Classifying The Emergency - Technical Bases Page: 21 of 187 5.1.15 HOSTILE ACTION An act toward CGS or its personnel that includes the use of violent force to destroy equipment, take hostages, and/or intimidate Energy Northwest 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 CGS. 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).

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

5.1.17 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 .

• 5.1.18 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).

5.1.19 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI)

A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage.

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

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

5.1.22 NORMAL LEVELS

• As applied to radiological IC/EALs, the highest reading in the past twenty-four hours excluding the current peak value.

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Classifying The Emergency-Technical Bases Page: 22of187 5.1.23 OWNER CONTROLLED AREA The area that Energy Northwest maintains industrial and process control of (ref. 2.2.2).

5.1.24 PROJECTILE An object directed toward CGS that could cause concern for its continued operability, reliability, or personnel safety.

5.1.25 PROTECTED AREA An area located within the OWNER CONTROLLED AREA which contains the Columbia Generating Station power block and is surrounded by chain link fence (ref. 2.2.2).

5.1.26 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) .

• 5.1.27 5.1.28 REFUELING PATHWAY Reactor cavity and spent fuel pool comprise the Refuel Pathway (ref. 2.1.11 ).

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

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

a. The integrity of the reactor coolant pressure boundary;

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

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

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Classifying The Emergency - Technical Bases Page: 23-of 187 5.1.29 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.

5.1.30 SITE AREA EMERGENCY Events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; 1) toward site personnel or equipment that could lead to the likely failure of or; 2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA PAG exposure levels beyond the site boundary.

5.1.31 SITE BOUNDARY 1950-meter radius around the plant as depicted in Figure 3-1 of the CGS ODCM (ref. 2.1.16). The key-hole area between the river and this radius is not within the Site Boundary .

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

5.1.33 UNPLANNED A parameter change or an event that is not: 1) the result of an intended evolution, or

2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

5.1.34 UNUSUAL EVENT 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.

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

Implicit in this definition is the need for timely assessment.

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Classifying The Emergency - Technical Bases Page: 24 of 187 5.1.36 VISIBLE DAMAGE Damage to a SAFETY SYSTEM train that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected SAFETY SYSTEM train.

5.2 Abbreviations/Acronyms OF Degrees Fahrenheit 0

Degrees AC Alternating Current APRM Average Power Range Meter ARI Automatic Rod Insertion ATWS Anticipated Transient Without Scram BWR Boiling Water Reactor BWROG Boiling Water Reactor Owners Group COE Committed Dose Equivalent CFR Code of Federal Regulations cps counts per second OBA Design Basis Accident DC Direct Current EAL Emergency Action Level ECCS Emergency Core Cooling System ECL Emergency Classification Level EOF Emergency Operations Facility EOP Emergency Operating Procedure EPA Environmental Protection Agency EPG Emergency Procedure Guideline EPIP Emergency Plan Implementing Procedure ESF Engineered Safety Feature FAA Federal Aviation Administration FBI Federal Bureau of Investigation FEMA Federal Emergency Management Agency

• FSAR GOS Final Safety Analysis Report Graphic Display System

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Classifying The Emergency - Technical Bases Page: 25 of 187 GE Greater than or Equal to gm Gram GT Greater Than HCTL Heat Capacity Temperature Limit HPCS High Pressure Core Spray HOO NRC Headquarters Operations Officer IC Initiating Condition ISFSI Independent Spent Fuel Storage Installation Kett Effective Neutron Multiplication Factor LCO Limiting Condition of Operation LE Less than or Equal to LOCA Loss of Coolant Accident LPCS Low Pressure Core Spray LT Less Than MPC Maximum Permissible Concentration/Multi-Purpose Canister

µCi Micro Curie MSCRWL Minimum Steam Cooling RPV Water Level MSIV Main Steam Isolation Valve MSL Main Steam Line mR milliRoentgen MW Megawatt NEI Nuclear Energy Institute NESP National Environmental Studies Project NORAD North American Aerospace Defense Command NRC Nuclear Regulatory Commission NSSS Nuclear Steam Supply System QBE Operating Basis Earthquake OCA Owner Controlled Area ODCM Off-site Dose Calculation Manual ORO Offsite Response Organization PMU Panel Meter Unit

• PRA PRM Probabilistic Risk Assessment Process Radiation Monitor

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Classifying The Emergency - Technical Bases Page: 26 of 187 PSIG Pounds per Square Inch Gauge PSP Pressure. Suppression Pressure R Roentgen RB Reactor Building RCC Reactor Building Closed Cooling RCIC Reactor Core Isolation Cooling RCS Reactor Coolant System Rem Roentgen Equivalent Man RHR Residual Heat Removal RPS Reactor Protection System RPV Reactor Pressure Vessel RWCU Reactor Water Cleanup SGT Stand-By Gas Treatment SBO Station Blackout SDSP Shutdown Safety Plan

• SLC SPDS SRO SSC Standby Liquid Control Safety Parameter Display System Senior Reactor Operator Structure, System or Component SW Service Water TEA Turbine Exhaust Air TEDE Total Effective Dose Equivalent TAF Top of Active Fuel TSC Technical Support Center TSW Plant Service Water WEA Waste Exhaust Air

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Classifying The Emergency - Technical Bases Page: 27 of 187 5.3 CGS-TO-NEI 99-01 Rev. 6 EAL CROSS-REFERENCE This cross-reference is provided to facilitate association and location of a CGS EAL within the NEI 99-01 IC/EAL identification scheme. Further information regarding the development of the CGS EALs based on the NEI guidance can be found in the EAL Comparison Matrix.

NEI 99-01 Rev. 6 CGS EAL Example IC EAL CG1.1 CG1 1 CS1.1 CS1 1, 2 CA1.1 CA1 1, 2 CU1.1 CU1 1, 2 CG1.2 CG1 2 CS1.2 CS1 3

• CA2.1 CU2.1 CA3.1 CA2 CU2 CA3 1

1 1, 2 CU3.1 CU3 1 CU3.2 CU3 2 CU4.1 CU4 1 CU5.1 CU5 1, 2, 3 CA6.1 CA6 1 EU1.1 E-HU1. 1 FG1.1 FG1 1 FS1.1 FS1 1 FA1.1 FA1 1 HS1.1 HS1 1 HA1.1 HA1 1, 2 HU1.1 HU1 1, 2 3

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Classifying The Emergency - Technical Bases Page: 2S of 1S7 NEI 99-01 Rev. 6 CGS EAL Example IC EAL HU2.1 HU2 1 HU3.1 HU3 1, 5 HU3.2 HU3 2 HU3.3 HU3 3,4 HU4.1 HU4 1 HU4.2 HU4 2 HU4.3 HU4 3,4 HA5.1 HAS 1 HS6.1 HS6 1 HA6.1 HA6 1 HG7.1 HG? 1 HS7.1 HS? 1 HA7.1 HA? 1 HU7.1 HU? 1 MG1.1 SG1 1 MS1.1 SS1 1 MA1.1 SA1 1 MU1.1 SU1 1 MG1.2 SGS 1 MS2.1 SSS 1 MA3.1 SA2 1 MU3.1 SU2 1 MU4.1 SU3 1

• MU4.2 SU3 2

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Classifying The Emergency - Technical Bases Page: 29 of 187 NEI 99-01 Rev. 6 CGS EAL Example IC EAL MU5.1 SU4 1, 2, 3 MS6.1 SS5 1 MA6.1 SA5 1 MU6.1 SU5 1, 2 MU7.1 SU6 1, 2, 3 MA8.1 SA9 1 RG1.1 AG1 1, 2 RS1.1 AS1 1, 2 RA1.1 AA1 1, 2

• RU1.1 RG1.2 RS1.2 RA1.2 AU1 AG1 AS1 AA1 1, 2, 3 3

3 3

RA1.3 AA1 4 RG2.1 AG2 1 RS2.1 AS2 1 RA2.1 AA2 1 RU2.1 AU2 1 RA2.2 AA2 2 RA2.3 AA2 3 RA3.1 AA3 1, 2

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Classifying The Emergency - Technical Bases Page: 30 of 187 6.0 ATTACHMENTS 6.1 EAL Technical Bases 6.2 Tables and Notes 6.3 Table 9 Basis 6.4 Emergency Action Level Chart Distribution

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Classifying The Emergency - Technical Bases Page: 31 of 187 EAL TECHNICAL BASES Category C - Cold Shutdown I Refuel System Malfunction EAL Group: Cold Conditions (RCS temperature s 200°F);

EALs in this category are applicable only in one or more cold operating modes.

Category C EALs are directly associated with cold shutdown or Refuel system safety functions. Given the variability of plant configurations (e.g., systems out-of-service for maintenance, containment open, reduced AC power redundancy, time since shutdown) during these periods, the consequences of any given initiating event can vary greatly. For example, a loss of decay heat removal capability that occurs at the end of an extended outage has less significance than a similar loss occurring during the first week after shutdown. Compounding these events is the likelihood that instrumentation necessary for assessment may also be inoperable. The cold shutdown and Refuel 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 (4 - Cold Shutdown, 5 - Refuel, D - Defueled).

The events of this category pertain to the following subcategories:

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

• 2. Loss of Emergency AC Power Loss of emergency plant electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of onsite and offsite power sources for 4160 V emergency buses.

3. RCS Temperature Uncontrolled or inadvertent temperature or pressure increases are indicative of a potential loss of safety functions.

4. Loss of Vital DC Power Loss of vital plant electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of power to or degraded voltage on the vital125 voe 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 .

• Attachment 6.1, EAL Technical Bases

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Classifying The Emergency -Technical Bases Page: 32 of 187 Category: C - Cold Shutdown I Refuel System Malfunction Subcategory: 1 - RPV Level Initiating Condition: Loss of RPV inventory affecting fuel clad integrity with containment challenged EAL:

CG1 .1 General Emergency RPV level LT -161 in. for GE 30 min. (Note 1)

AND Any of the following indications of Containment Challenge:

• CONTAINMENT CLOSURE not established (Note 6)

• Explosive mixture inside PC (H 2 GE 6% and 02 GE 5%)

• UNPLANNED rise in PC pressure

• RB area radiation GT any Maximum Safe Operating level (PPM 5.3.1 Table 24)

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

• Mode Applicability:

Basis:

When RPV level drops to the top of active fuel (an indicated RPV level of -161 in.), core uncovery starts to occur (ref. 1, 2).

Four conditions are associated with a challenge to primary containment (PC) integrity:

• Containment Closure is defined as the Shutdown Safety Plan (SDSP) actions taken to secure primary or secondary containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions. This definition is less restrictive than Technical Specification criteria governing Primary and Secondary Containment operability. If the Technical Specification criteria are met, therefore, Containment Closure has been established. (ref. 3, 4, 5)

• Explosive (deflagration) mixtures in the primary containment are assumed to be elevated concentrations of hydrogen and oxygen. BWR industry evaluation of hydrogen generation for development of EOPs/SAGs indicates that any hydrogen concentration above minimum detectable is not to be expected within the short term. Post-LOCA hydrogen generation primarily caused by radiolysis is a slowly evolving, long-term condition. Hydrogen concentrations that rapidly develop are most likely caused by metal-water reaction. A metal-water reaction is indicative of an accident more severe than accidents considered in the plant design basis and would be indicative, therefore, of a potential threat to primary containment

• integrity. Hydrogen concentration of approximately 6% is considered the global deflagration concentration limit (ref. 6).

Attachment 6.1, EAL Technical Bases

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Title:

Classifying The Emergency - Technical Bases Page: 33 of 187 The specified values for this threshold are the minimum global deflagration concentration limits (6% hydrogen and 5% oxygen, ref 5) and readily recognizable because 6% hydrogen is well above the EOP flowchart entry condition (ref. 8). The minimum global deflagration hydrogen/oxygen concentrations (6% I 5%, respectively) require intentional primary containment venting, which is defined to be a loss of the primary containment barrier.

Atmosphere samples from a minimum of two locations inside the primary containment and one location in the suppression chamber are sequentially monitored for hydrogen and oxygen percentage levels by each of two redundant analyzer systems. The analyzers are single range (0 to 30% hydrogen and 0 to 30% oxygen). Two redundant (divisional) recorders are provided in the Main Control Room CMS-02/H2R-1 (H13-P827) and CMS-02/H2R-2 (H13-P811).

Hydrogen and oxygen concentrations can also be displayed on the plant computers (ref. 9-12)

• Any UNPLANNED rise in PC pressure in the Cold Shutdown or Refueling mode indicates Containment Closure cannot be assured and the primary containment cannot be relied upon as a barrier to fission product release.

• RB (Reactor Building) area radiation monitors should provide indication of increased release that may be indicative of a challenge to Containment Closure. The Maximum Safe Operating levels are indicative of problems in the secondary containment that are spreading. The locations into which the primary system discharge is of concern correspond to the areas addressed in Table 24 of the EOP flowcharts (ref. 13). All Table 24 Maximum Safe Operating radiation levels can be determined in the main Control Room.

• If RPV level is restored and maintained above the top of active fuel before a Containment Challenge condition occurs and subsequently a Containment Challenge condition is reached, this EAL is not met.

This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA PAG exposure levels off~ite 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 RPV level cannot be restored, fuel damage is probable.

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

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

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

• of-service, operators may use the other listed indications to assess whether or not containment is challenged.

Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: NIA

Title:

Classifying The Emergency - Technical Bases Page: 34 of 187 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.

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 Operations at Commercial Nuclear Power Plants in the United States; and NU MARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.

CGS Basis Reference(s):

1. Calculation NE-02-03-05 Attachment 3 Note 8

2. PPM 5.1.1 RPV Control

3. Technical Specifications 3.6.1.1

4. Technical Specifications 3.6.4.1

5. PPM 1.20.3 Outage Risk Management

6. BWROG EPG/SAG Revision 2, Sections PC/G

7. PPM 5.7.1 RPVand Primary Containment Flooding SAG, Table 19

8. PPM 5.2.1 Primary Containment Control

9. FSAR Section 7.5.1.5.4

10. PPM 5.0.10 Flowchart Training Manual

11. PPM 4.814.J1 814.J1 Annunciator Panel Alarms, 2-2

12. PPM 4.814.J2 814.J2 Annunciator Panel Alarms, 2-2

13. PPM 5.3.1 Secondary Containment Control

14. NEI 99-01 CG1

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A \ Use Category: REFERENCE Major Rev: 032 l Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 35 of 187 Category: C - Cold Shutdown I Refuel System Malfunction Subcategory: 1 - RPV Level Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability EAL:

CS1 .1 Site Area Emergency (1) CONTAINMENT CLOSURE not established AND RPV level LT-129 in.

OR (2) CONTAINMENT CLOSURE established AND

~PV level LT -161 in.

Mode Applicability:

• Basis:

EAL#1 The threshold RPV water level of -129 in. is the low-low-low ECCS actuation setpoint. The magnitude of this loss of water indicates that makeup systems have not been effective and may not be capable of preventing further RPV water level decrease and potential core uncovery. The inability to restore and maintain level after reaching this setpoint infers a failure of the RCS barrier and Potential Loss of the Fuel Clad barrier. (ref. 1)

EAL#2 When RPV level drops to the top of active fuel (an indicated RPV level of -161 in.), core uncovery starts to occur (ref. 2).

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 RPV 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 RPV levels of CS1 .1 (1) and CS1 .1 (2) reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment.

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 36 of 187 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 Operations at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.

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

CGS Basis Reference(s):

1. Technical Specifications Table 3.3.5.1-1, "Emergency Core Cooling System Instrumentation"

2. PPM 5.1.1 RPV Control

3. NEI 99-01 CS1

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A IUse Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 37 of 187 Category: C - Cold Shutdown I Refuel System Malfunction Subcategory: 1 - RPV Level Initiating Condition: Significant loss of RPV inventory EAL:

CA1.1 Alert (1) Loss of RPV inventory as indicated by RPV level LT -50 in.

OR (2) RPV,level cannot be monitored for GE 15 min. (Note 1)

AND UNPLANNED increase in any Table 1 sump or pool levels due to a loss of RPV inventory Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Table 1 Sumps/Pool

• Any valid Hi-Hi level alarm on R-1 through R-5 sumps

• EDR GE 25 GPM

• FDR GE 10 GPM

• Wetwell level rise

• Observation of UNISOLABLE RCS leakage Mode Applicability:

Basis:

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

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

EAL#1 The threshold RPV level of -50 in. is the low-low ECCS (HPCS) actuation setpoint (ref. 1, 2).

In Cold Shutdown mode, the RCS will normally be intact and standard RPV 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 .

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A \ Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 38 of 187 For EAL #1, a lowering of water level below -50 in. indicates that operator actions have not been successful in restoring and maintaining RPV water level. The heat-up rate of the coolant will increase as the available water inventory is reduced. A continuing decrease in water level will lead to core uncovery.

Although related, EAL #1 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 ICCA3.

EAL#2 In this EAL, all RPV level indication is unavailable and the RPV inventory loss must be detected by the leakage indications listed in Table 1. Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure they are indicative of RPV leakage. Reactor Building equipment or floor drain sump level rise may be indicative of RPV inventory losses external to the primary containment from systems connected to the RPV (ref. 3, 4).

With RHR System operating in the Shutdown Cooling mode, an unexplained rise in wetwell level could be indicative of RHR valve misalignment or leakage (ref. 5). If the make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV 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 in areas outside the primary containment that cannot be isolated could be indicative of a loss of RPV inventory .

• For EAL #2, the inability to monitor RPV level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV.

The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CS1.

CGS Basis Reference(s):

1. Technical Specifications Table 3.3.5.1-1

2. PPM 5.1.1 RPV Control

3. SOP-EDR-OPS Equipment Drain System Operation

4. SOP-FDR-OPS Floor Drain System Operation

5. SOP-RHR-SDC RHR Shutdown Cooling

6. NEI 99-01 CA1

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A IUse Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 39 of 187 Category: C - Cold Shutdown I Refuel System Malfunction Subcategory: 1 - RPV Level Initiating Condition: UNPLANNED loss of RPV inventory EAL:

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

OR (2) RPV level cannot be monitored AND UNPLANNED increase in any Table 1 sump or pool levels due to a loss of RPV inventory Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Table 1 Sumps/Pool

• Any valid Hi-Hi level alarm on R-1 through R-5 sumps

• EDR GE 25 GPM

• FDR GE 10 GPM

• Wetwell level rise

• Observation of UNISOLABLE RCS leakage Mode Applicability:

Basis:

These Cold Shutdown EALs represent the hot condition EAL MU5.1, in which RCS leakage is associated with Technical Specification limits. In Cold Shutdown, these limits are not applicable; hence, the use of RPV level as the parameter of concern in this EAL.

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

Refuel 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.

Attachment 6.1, EAL Technical Bases

Number: 13.1.1A [ Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 40 of 187 Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA 1 or CA3.

EAL#1 In Mode 4 and Mode 5, prior to flood up, RPV level is monitored from -310 in. to +400 in. to ensure adequate coverage for expected and postulated conditions of RPV level. All instruments are referenced to a benchmark at 527.5 in. above the inside bottom head of the reactor vessel. This benchmark corresponds to the bottom edge of the steam dryer skirt and is the 0 in. reference indication on the RPV level instruments (ref. 1, 2, 3).

In preparation for refueling operations, level instruments are modified to provide continuous level indication from within the RPV to the refuel floor (ref. 4, 5).

The RPV level is controlled in a designated band in the reactor vessel and 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 Refuel mode, RPV water level is normally maintained at or above the reactor vessel flange (ref 6).

EAL #1 recognizes that the minimum required RPV level can change several times during the course of a Refuel outage as different plant configurations and system lineups are implemented. This EAL is met if the minimum level, specified for the current plant conditions, cannot be maintained for 15 minutes or longer. The minimum level is typically specified in the applicable operating procedure but may be specified in another controlling document.

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

EAL#2 In this EAL, all RPV level indication is unavailable and the RPV inventory loss must be detected by the leakage indications listed in Table 1. Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure they are indicative of RPV leakage. Reactor Building equipment or floor drain sump level rise may be indicative of RPV inventory losses external to the primary containment from systems connected to the RPV (ref. 7, 8, 6).

With RHR System operating in the Shutdown Cooling mode, an unexplained rise in wetwell level could be indicative of RHR valve misalignment or leakage (ref. 10). If the make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV 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 in areas outside the primary containment that cannot be isolated could be indicative of a loss of RPV inventory.

EAL #2 addresses a condition where all means to determine RPV level have been lost. In this condition, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV .

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 41 of 187 CGS Basis Reference(s):

1. FSAR Section 7.5.1.1

2. FSAR Table 7.5-1

3. FSAR Figure 7.7-1

4. PPM 10.27.39 Refueling Reactor Vessel Level (Temporary)

5. SOP-CAVITY-FILL Reactor Cavity and Dryer Separator Pit Fill

6. Technical Specifications 3.9.6

7. FSAR Section 7.6.1.3

8. SOP-EDR-OPS Equipment Drain System Operation

9. SOP-FDR-OPS Floor Drain System Operation

10. SOP-RHR-SDC RHR Shutdown Cooling

11. NEI 99-01 CU1

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 42 of 187 Category: C - Cold Shutdown I Refuel System Malfunction Subcategory: 1 - RPV Level Initiating Condition: Loss of RPV inventory affecting fuel clad integrity with containment challenged EAL:

CG1 .2 General Emergency RPV level cannot be monitored for GE 30 min. (Note 1)

AND Core uncovery is indicated by any of the following:

• UNPLANNED wetwell level rise GT 2 inches (PPM 5.2.1 entry condition)

• Valid indication of RB room flooding as identified by high level alarms (PPM 5.3.1 Table 25)

• Observation of UNISOLABLE RCS leakage outside primary containment of sufficient magnitude to indicate core uncover AND Any of the following indication of containment challenge:

• CONTAINMENT CLOSURE not established (Note 6)

• Explosive mixture inside PC (H 2 GE 6% and 0 2 GE 5%)

• UNPLANNED rise in PC pressure

• RB area radiation GT any Maximum Safe Operating level (PPM 5.3.1 Table 24)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required Mode Applicability:

Basis:

In this EAL, all RPV level indication is unavailable and the RPV inventory loss must be detected by the leakage indications provided. Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure they are indicative of RPV leakage.

Reactor Building equipment or floor drain sump level rise may be indicative of RPV inventory losses external to the primary containment from systems connected to the RPV (ref. 1, 2). With RHR System operating in the Shutdown Cooling mode, an unexplained rise in wetwell level could be indicative of RHR valve misalignment or leakage (ref. 3). If the make-up rate to the RPV unexplainably rises above

• the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified.

Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

Classifying The Emergency - Techn_ic_a_l_B_a_se_s_ _ _ _ _ _ _ _ _ _ _ _~P_ag_e_:_4_3_o_f_1_8_7~

An UNPLANNED wetwell level increase to GT 2 inches or a VALID RB room high level alarm indicates a significant loss of RCS that could lead to core uncovery if not isolated (ref. 4, 5).

Visual observation of significant leakage from systems connected to the RCS in areas outside the primary containment that cannot be isolated could be indicative of a loss of RPV inventory sufficient to lead to core uncovery.

Four conditions are associated with a challenge to primary containment (PC) integrity:

• CONTAINMENT CLOSURE is not established.

• Explosive (deflagration) mixtures in the primary containment are assumed to be elevated concentrations of hydrogen and oxygen. BWR industry evaluation of hydrogen generation for development of EOPs/SAGs indicates that any hydrogen concentration above minimum detectable is not to be expected within the short term. Post-LOCA hydrogen generation primarily caused by radiolysis is a slowly evolving, long-term condition. Hydrogen concentrations that rapidly develop are most likely caused by metal-water reaction. A metal-water reaction is indicative of an accident more severe than accidents considered in the plant design basis and would be indicative, therefore, of a potential threat to primary containment integrity. Hydrogen concentration of approximately 6% is considered the global deflagration concentration limit (ref.

6).

• The specified values for this threshold are the minimum global deflagration concentration limits (6% hydrogen and 5% oxygen, ref. 6) and readily recognizable because 6% hydrogen is well above the EOP flowchart entry condition. (ref. 8). The minimum global deflagration hydrogen/oxygen concentrations (6%/5%, respectively) require intentional primary containment venting, which is defined to be a loss of the primary containment barrier.

Atmosphere samples from a minimum of two locations inside the primary containment and one location in the suppression chamber are sequentially monitored for hydrogen and oxygen percentage levels by each of two redundant analyzer systems. The analyzers are single range (0 to 30% hydrogen and Oto 30% oxygen). Two redundant (divisional) recorders are provided in the Main Control Room CMS 02/H2R 1 (H13 P827) and CMS 02/H2R 2 (H13 P811).

Hydrogen and oxygen concentrations can also be displayed on the plant computers (Ref. 9-12)

• Any unplanned rise in PC pressure in the Cold Shutdown or Refueling mode indicates Containment Closure cannot be assured and the primary containment cannot be relied upon as a barrier to fission product release.

• RB (Reactor Building) area radiation monitors should provide indication of increased release that may be indicative of a challenge to Containment Closure. The Maximum Safe Operating levels are indicative of problems in the secondary containment that are spreading. The locations into which the primary system discharge is of concern correspond to the areas addressed in Table 24 of the EOP flowcharts (ref.13).

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

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

Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 44 of 187 With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.

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

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

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

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

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

CGS Basis Reference(s):

1. SOP-EDR-OPS Equipment Drain System Operation

2. SOP-FDR-OPS Floor Drain System Operation

3. SOP-RHR-SDC RHR Shutdown Cooling

4. PPM 5.2.1 Primary Containment Control

5. PPM 5.3.1 Secondary Containment Control

6. BWROG EPG/SAG Revision 2, Sections PC/G

7. PPM 5.7.1 RPV and Primary Containment Flooding SAG, Table 19

8. PPM 5.2.1 Primary Containment Control

9. FSAR Section 7.5.1.5.4

10. PPM 5.0.10 Flowchart Training Manual

11. PPM 4.814.J1 814.J1 Annunciator Panel Alarms, 2-2

12. PPM 4.814.J2 814.J2 Annunciator Panel Alarms, 2-2

13. PPM 5.3.1 Secondary Containment Control

14. NEI 99-:01 CG1 Attachment 6.1, EAL Technical Bases

Number: 13.1.1A \ Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 45 of 187 Category: C - Cold Shutdown I Refuel System Malfunction Subcategory: 1 - RPV Level Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability EAL:

CS1 .2 Site Area Emergency RPV level cannot be monitored for GE 30 min. (Note 1)

AND Core uncovery is indicated by any of the following:

• UNPLANNED wetwell level rise GT 2 inches (PPM 5.2.1 entry condition)

• VALID indication of RB room flooding as identified by high level alarms (PPM 5.3.1 Table 25)

• Observation of UNISOLABLE RCS leakage outside primary containment of sufficient magnitude to indicate core uncovery Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded

• Mode Applicability:

Basis:

In this EAL, all RPV level indication is unavailable and the RPV inventory loss must be detected by the leakage indications provided. Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure they are indicative of RPV leakage.

Reactor Building equipment or floor drain sump level rise may be indicative of RPV inventory losses external to the primary containment from systems connected to the RPV (ref. 1, 2). With RHR System operating in the Shutdown Cooling mode, an unexplained rise in wetwell level could be indicative of RHR valve misalignment or leakage (ref. 3). If the make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified.

An UNPLANNED wetwell level increase to GT 2 inches or a VALID RB room high level alarm indicates a significant loss of RCS that could lead to core uncovery if not isolated (ref. 4, 5).

Visual observation of significant leakage from systems connected to the RCS in areas outside the primary containment that cannot be isolated could be indicative of a loss of RPV inventory sufficient to lead to core uncovery.

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

• Emergency declaration.

Attachment 6.1, EAL Technical Bases

Number: 13.1.1A J Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 46 of 187 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 RPV level cannot be restored, fuel damage is probable.

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

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

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

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

• 1.

2.

3.

4.

5.

SOP-EDR-OPS Equipment Drain System Operation SOP-FDR-OPS Floor Drain System Operation SOP-RHR-SDC RHR Shutdown Cooling PPM 5.2.1 Primary Containment Control PPM 5.3.1 Secondary Containment Control

6. NEI 99-01 CS1

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A \ Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 47of187 Category: C - Cold Shutdown I Refuel System Malfunction Subcategory: 2 - Loss of Emergency AC Power Initiating Condition: Loss of all offsite and fill onsite AC power to emergency buses for 15 minutes or longer EAL:

CA2.1 Alert Loss of all offsite and all onsite AC power capability to emergency buses SM-7 and SM-8 for GE 15 min. (Note 1)

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

4 5 def Basis:

This cold condition EAL is equivalent to the hot condition EAL MS1 .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, Refuel, or defueled mode, this condition is not classified as a Site Area Emergency because of the increased time available to restore an emergency bus to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition represents an actual or potential substantial degradation of the level of safety of the plant.

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

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

CGS Basis Reference{s):

1. FSAR Figure 8.1-2.1 Main One-Line Diagram - Main Buses

2. FSAR Figure 8.1-2.2 Main One-Line Diagram - Emergency Buses

3. FSAR Section 8.2

4. 01-53 Offsite Power

5. FSAR Section 8.3

6. ABN-ELEC-LOOP Loss Of All Off-Site Electrical Power

7. SOP-ELECT-BACKFEED

8. NEI 99-01 CA2

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 48 of 187 Category: C - Cold Shutdown I Refuel System Malfunction Subcategory: 2 - Loss of Emergency AC Power Initiating Condition: Loss of all but one AC power source to emergency buses for 15 minutes or longer EAL:

CU2.1 Unusual Event AC power capability, Table 2, to emergency buses SM-7 and SM-8 reduced to a single power source for GE 15 min. (Note 1)

AND Any additional single power source failure will result in a loss of all AC power to SAFETY SYSTEMS Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Table 2 AC Power Sources Offsite Startup Transformer TR-S Backup Transformer TR-B Backfeed 500 KV power throygh Main Transformers (if already align'ed in modes 4, 5, def only)

On site

• DG1

• DG2

• Main Generator via TR-N 1/N2 Mode Applicability:

4 5 def Basis:*

Table 2 provides the list of AC power sources available to power emergency buses (ref. 1, 2).

Station Startup 230KV power comes from the Ashe substation through Startup transformer TR-S. The Startup transformer usually supplies station auxiliary loads when the main generator is not available.

• Station Backup 115KV power from the Benton Substation feeder can be supplied to emergency buses SM-7 and SM-8 (ref. 3, 4).

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A J Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 49 of 187 Credit is not taken in this EAL for SM-4/DG3 crosstie capability because establishing the crosstie to SM-7 or SM-8 is assumed to require more than 15 minutes (5). SM-4 is not a site specific emergency AC buss source since SM-4 does not provide core cooling or containment cooling.

It is possible to remove Startu*p power from service and continue to supply the plant during shutdown conditions by backfeeding 500 KV power from Ashe Substation through the Main Transformers, the Normal Transformers and associated "N" breakers. This involves disconnecting the Main Generator from the Isolated Phase conductors (25 KV system) and overriding various interlocks. This action would take significantly longer than 15 minutes; therefore, backfeed must be in service to credit this source (ref. 7).

The second threshold statement in this EAL does not describe a separate condition; it is clarifying the first threshold statement.

This cold condition EAL is equivalent to the hot condition EAL MA 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 safety-related equipment.

When in the cold shutdown, Refuel, or defueled mode, this condition is not classified as an Alert because of the increased time available to restore another power source to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition is considered to be a potential

• degradation of the level of safety of the plant.

An "AC power source" is a source recognized in AOPs, and 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 one division of emergency power sources (e.g., onsite diesel generators).

• 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 division of emergency buses being back-fed from an offsite power source.

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

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

CGS Basis Reference(s):

1. FSAR Figure 8.1-2.1 Main One-Line Diagram - Main Buses

2. FSAR Figure 8.1-2.2 Main One-Line Diagram - Emergency Buses

3. FSAR Section 8.2

4. 01-53 Offsite Power

5. FSAR Section 8.3

6. ABN-ELEC-LOOP Loss Of All Off-Site Electrical Power

• 7.

8.

SOP-ELECT-BACKFEED NEI 99-01 CU2 Attachment 6.1, EAL Technical Bases

Number: 13.1.1A IUse Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 50 of 187 Category: C - Cold Shutdown I Refuel System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: Inability to maintain the plant in cold shutdown EAL:

CA3.1 Alert UNPLANNED increase in RCS temperature to GT 200°F for GT Table 7 duration (Note 1)

OR UNPLANNED RPV pressure increase GT 10 psig Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been excee ded, orw1*11 l'k I be excee ded 1 e1y Table 7 RCS Heat-up Duration Thresholds CONTAINMENT CLOSURE RCS Status Heat-up Duration Status Intact N/A 60 min.*

established 20 min.*

Not intact not established 0 min.

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

Mode Applicability:

Basis:

200°F is the Technical Specification cold shutdown temperature limit (ref. 1).

10 psi is one-half of the 20 psi minor division on the Wide Range RPV pressure instrument, RFW-Pl-605, on Main Control Room Panel H13- P603 (ref. 2). This instrument has a range of 0 to 1200 psig. This RPV pressure indication is also displayed on plant computer point B016 (ref. 3).

Recirculation suction temperature, RRC TR 650 pt 1(2), is the primary temperature measurement instrument when RPV pressure is less than 100 psig and the associated RRC pump is operating.

Monitoring of the RWCU bottom head drain temperature element, RWCU TE 21, as read on RWCU Tl 607 pt 5 (H13 P602) or MS TR 6 pt 316 (RB 522) is acceptable only if a RRC pump is operating for forced flow and RWCU flow of greater than 50 gpm exists. (ref. 4)

With flow through the RHR Heat Exchanger, the inlet temperature (TDAS pt. X045) is indicative of RRC

• system temperature. If adequate core flow cannot be provided, RPV metal temperature can be monitored on MS TR 6. (ref. 5)

Attachment 6.1, EAL Technical Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: 032

---< Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 51 of 187 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).

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

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

The RCS Heat-up Duration Thresholds table addresses an increase in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact.. The 20-minute criterion was included to allow time for operator action to address the temperature increase.

The RCS Heat-up Duration Thresholds table also addresses an increase in RCS temperature with the RCS intact. The status of CONTAINMENT CLOSURE is not crucial in this condition since the intact RCS is providing a high pressure barrier to a fission product release. The 60-minute time frame should allow sufficient time to address the temperature increase without a substantial degradation in plant safety.

Finally, in the case where there is an increase in RCS temperature, the RCS is not intact , and CONTAINMENT CLOSURE is not established, no heat-up duration is allowed (i.e., 0 minutes). This is because 1) the evaporated reactor coolant may be released directly into the Containment atmosphere and subsequently to the environment, and 2) there is reduced reactor coolant inventory above the top

• of irradiated fuel.

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

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

CGS Basis Reference(s):

1. Technical Specifications Table 1.1-1

2. Instrument Master Datasheet for EPN RFW-Pl-605

3. PPM 10.27.36 Reactor Pressure High Alarm - CC

4. OSP-RCS-C102 RPV Non-Critical Cooldown Surveillance

5. SOP-RHR-SDC RHR Shutdown Cooling

6. NEI 99-01 CA3

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 52 of 187 Category: C - Cold Shutdown I Refuel System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: UNPLANNED increase in RCS temperature EAL:

CU3.1 Unusual Event UNPLANNED increase in RCS temperature to GT 200°F Mode Applicability:

Basis:

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

The Technical Specification cold shutdown temperature limit is 200°F (ref. 1).

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

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

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

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

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

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

CGS Basis Reference(s}:

1. Technical Specifications Table 1.1-1

2. NEI 99-01 CU3

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 53 of 187 Category: C - Cold Shutdown I Refuel System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: UNPLANNED increase in RCS temperature EAL:

CU3.2 Unusual Event Loss of fill RCS temperature and RPV water level indication for GE 15 min. (Note 1)

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

Basis:

Recirculation suction temperature, RRC-TR-650 pt 1(2), is the primary temperature measurement instrument when RPV pressure is less than 100 psig and the associated RRC pump is operating.

Monitoring of the RWCU bottom head drain temperature element, RWCU-TE-21, as read on RWCU-Tl-607 pt 5 (H13 P602) or MS-TR-6 pt 316 (RB 522) is acceptable only if a RRC pump is operating for

• forced flow and RWCU flow of greater than 50 gpm exists. (ref. 4)

With flow through the RHR Heat Exchanger, the inlet temperature (TDAS pt. X045) is indicative of RRC system temperature. If adequate core flow cannot be provided, RPV metal temperature can be monitored on MS-TR-6. (ref. 5)

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

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

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

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

CGS Basis Reference(s):

1. FSAR Table 7.5-1

2. FSAR Figure 7.7-1

3. FSAR Section 7.6.1.3

4. OSP-RCS-C102 RPV Non-Critical Cooldown Surveillance

5. SOP-RHR-SDC RHR Shutdown Cooling

• 6. NEI 99-01 CU3 Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 54 of 187 Category: C - Cold Shutdown I Refuel System Malfunction Subcategory: 4 - Loss of Vital DC Power Initiating Condition: Loss of vital DC power for 15 minutes or longer EAL:

CU4.1 Unusual Event Indicated voltage LT 108 VDC on required 125 VDC buses DP-S1-1 and DP-S1-2 for GE 15 min.

(Note 1)

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

Basis:

The 125 VDC Class 1E DC power system consists of three electrically independent and separate distribution systems (S1-1, S1-2, and S1-HPCS). S1-HPCS is not included in this EAL. Each DC distribution system has a battery and a battery charger that are normally connected to the bus such that these two sources of power are operating in parallel. The charger is normally supplying system

• electrical loads with the battery on a float charge. Each battery has the necessary amp-hour discharge capacity to sustain system loads for a minimum of two hours. This capacity is specifically for a loss of power to the charger coincident with a design basis accident. The batteries have capacity to carry design load at 60°F without decreasing battery voltage below 1.81 volts/cell (or 108 VDC, ref. 1) with loss of output from the battery chargers during the specified period. Battery capacity is sufficient to provide starting currents while operating at full load. (ref. 2)

This EAL is the cold condition equivalent of the hot condition loss of DC power EAL MS2.1.

This IC addresses a loss of essential DC power which compromises the ability to monitor and control operable SAFETY SYSTEMS when the plant is in the cold shutdown or Refuel mode. In these modes, the core decay heat load has been significantly reduced, and coolant system temperatures and pressures are lower; these conditions increase the time available to restore a vital DC bus to service.

Thus, this condition is considered to be a potential degradation of the level of safety of the plant.

As used in this EAL, "required" means the essential DC buses necessary to support operation of the in-service, or operable, train or trains of SAFETY SYSTEM equipment. For example, if Division I is out-of-service (inoperable) for scheduled outage maintenance work and Division II is in-service (operable),

then a loss of essential DC power affecting Division II would require the declaration of an Unusual Event. A loss of essential DC power to Division I would not warrant an emergency classification.

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

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

CGS Basis Reference(s):

1. Calculation No. 2.05.01 Battery Sizing, Voltage Drop, and Charger Studies for Div. 1 & 2 Systems

2. FSAR Section 8.3.2

3. NEI 99-01 CU4 Attachment 6.1, EAL Technical Bases

Number: 13.1.1A IUse Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 55 of 187 Category: C - Cold Shutdown I Refuel System Malfunction Subcategory: 5 - Loss of Communications Initiating Condition: Loss of§.!! onsite or offsite communications capabilities EAL:

CU5.1 Unusual Event (1) Loss of all Table 4 onsite communication methods OR (2) Loss of§.!! Table 4 ORO communication methods OR (3) Loss of all Table 4 NRC communication methods Table 4 Communication Methods System Onsite ORO NRC Plant Public Address (PA) System x

• Plant Telephone System Plant Radio System Operations and Security Channels

• x x

x Offsite calling capability from the Control x x Room via direct telephone Long distance calling capability on the x x commercial phone system Mode Applicability:

4 5 def Basis:

Onsite and offsite (ORO and NRC) communications include one or more of the systems listed in Table 4 (ref. 1, 2).

Public Address (PA) System The public address system provides a way of contacting personnel in the various buildings of the plant and locations of the site that might be inaccessible using other means of communication. The building-wide alarm system alerts (via the public address system speakers) operating personnel to fire hazards and other trouble conditions for which plant management finds it necessary to alert plant personnel.

Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 56 of 187 Plant Telephone System This system consists of interconnections to the public telephone network (and trunks to the PBX) with individual direct lines that provide inward and outward dialing access to most plant locations.

Plant Radio System Operations and Security Channels The radio communications system is used for communications with personnel involved in maintenance and security in and around the plant complex by means of hand-held portable radio units, mobile radio units, and paging receivers. The telephone link to BPA provides a direct communication link to the BPA Dittmer Control Center. The radio communications system provides a communications link for security and emergency communications to local law enforcement agencies and emergency control centers.

Offsite calling capability from the Control Room via direct telephone and fax lines*

This communications method includes following dedicated phone networks that are available for emergency communications in addition to the normal Energy Northwest phone network:

• Energy Northwest Emergency Center Network

• Response Agency Network

• NRC Emergency Notification System Various locations such as the Control Room, Technical Support Center, Emergency Operations Facility, Joint Information Center, Department of Energy-RL, Washington State Emergency Operations Center,

• Oregon State Emergency Coordination Center and the Benton and Franklin County Emergency Operations Centers have facsimile transceivers. The facsimile transceivers enable the transmission and receipt of printed material. The facsimile system which connects the Energy Northwest emergency centers with the county and state emergency centers uses dedicated phone lines.

Long distance calling capability on the commercial phone system The Energy Northwest Richland phone system is a computer based, software controlled telephone exchange (Computerized Branch Exchange). It is equipped with redundant computerized processor units and is served by an uninterruptible power supply. The direct-dial private telephone system provides communication between the Energy Northwest facilities. The phone system is arranged such that plant telephones can reach other Energy Northwest facilities by direct-dialing and without the need of an operator.

This EAL is the cold condition equivalent of the hot condition EAL MU7.1.

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

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

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

The second EAL condition addresses a total loss of the communications methods used to notify all OROs of an emergency declaration. The OROs referred to here are Washington Stare, Benton County,

• Franklin County and DOE RL.

Attachment 6.1, EAL Technical Bases

Number: 13.1.1 A Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 57 of 187 The third EAL addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.

CGS Basis Reference(s):

1. Emergency Plan Section 6.6

2. FSAR Section 9.5.2

3. NEI 99-01 CU5

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A IUse Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 58 of 187 Category: C - Cold Shutdown I Refuel 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 8 hazardous event AND Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode AND EITHER:

Event damage has caused indications of degraded performance to a second train of a SAFETY SYSTEM needed for the current operating mode OR Event damage has resulted in VISIBLE DAMAGE to a second train of a SAFETY SYSTEM needed for the current operating mode (Notes 9, 10)

• ~---------

Note 9: If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then emergency classification is not warranted.

Note 10: If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted.

Table 8 Hazardous Events

• • Seismic event

• Internal or external FLOODING event

• High winds

• Tornado strike

• FIRE

• EXPLOSION

• Volcanic ash fallout

• Other events with similar hazard characteristics as determined by the Shift Manager

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A IUse Category: REFERENCE Major Rev: 032 Minor Rev: NIA

Title:

Classifying The Emergency-Technical Bas_e_s_ _ _ _ _ _ _ _ _ _ _ _~P_a_g_e_:_5_9_o_f_1_8_7~

Basis:

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

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

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

The significance of a seismic event is discussed under EAL HU2.1 (ref. 1).

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

Seismic Category I structures are analyzed to withstand a sustained, design wind velocity of at least 100 mph (ref. 3).

Areas containing functions and systems required for safe shutdown of the plant are identified by Fire Areas in the fire response procedure (ref. 4).

The potential for volcanic eruption exists in the Pacific Northwest. Heavy ash fall, such as that experienced at certain locations following the eruption of Mt. St. Helens in 1980, could affect operation of plant equipment if precautionary measures are not taken. The design basis ash fall is projected for a twenty hour duration. (ref. 5)

Table 5 provides a list of CGS safety system areas (ref. 6).

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

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A \ Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 60 of 187 CGS Basis Reference{s):

1. FSAR Section 3. 7 Seismic Design

2. FSAR Section 3.4.1 Flood Protection

3. CGS Calculation CALC CE-02-93-16 Evaluate PMR/BDC 98-0131-0A change from 5 min. to 15 min. averaging of 33 ft. elev. met tower wind speeds for UE and Alert declarations

4. ABN-FIRE Attachment 13.2, Fire Areas

5. ABN-ASH Ash Fall

6. FSAR Table 3.2-1 Equipment Classification

7. NEI 99-01 CA6

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 61 of 187 Category E - Independent Spent Fuel Storage Installation (ISFSI)

EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold)

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

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

A hostile security event that leads to a potential loss in the level of safety of the ISFSI is a classifiable event under Security category EAL HA 1.1 .

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A IUse Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 62 of 187 Category: E- ISFSI Sub-category: None Initiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARY EAL:

EU1.1 Unusual Event Damage to a loaded canister (MPC) CONFINEMENT BOUNDARY as indicated by measured dose rates on a loaded overpack GT EITHER:

• 20 mrem/hr (gamma + neutron) on the top of the overpack

• 100 mrem/hr (gamma + neutron) on the side of the overpack, excluding inlet and outlet ducts Mode Applicability:

Storage Operations Basis:

The Independent Spent Fuel Storage Installation utilizes the HOLTEC International (HOLTEC) HI-STORM 100 Spent Fuel Dry Storage (SFDS) system. HI-STORM overpack or storage overpack means the cask that receives and contains the sealed multi-purpose canisters containing spent nuclear

• fuel. It provides the gamma and neutron shielding, ventilation passages, missile protection, and protection against natural phenomena and accidents for the MPC. (ref. 1, 2)

The EAL threshold values represent two-times the limits specified in the ISFSI Certificate of Compliance Technical Specification Section 3.2, Radiation Protection Program (ref. 2).

CGS has casks loaded to various amendments to the Certificate of Compliance (COC) Technical Specifications. The numbers above reflect the most limiting Technical Specification (TS) values (Amendment 1).

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

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

Security-related events for ISFSls are covered under ICs HU1 and HA1 .

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A \ Use Category: REFERENCE Major Rev: .032 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 63 of 187 CGS Basis Reference(s):

1. ABN-ISFSI, ISFSI Abnormal Conditions

2. ISFSI Certificate of Compliance No. 1014 Amendment 1, Appendix A, Technical Specifications for the HI-STORM 100 Cask System, Section 3.2 Radiation Protection Program

3. NEI 99-01 E-HU1

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: 032

_, Minor Rev: NIA

Title:

Classifying The Emergency -Technical Bases Page: 64 of 187 Fission Product Barrier Matrix and Bases Introduction Table F-1 lists the threshold conditions that define the Loss and Potential Loss of the three fission product barriers (Fuel Clad, Reactor Coolant System, and Containment). The table is structured so that each of the three barriers occupies adjacent columns. Each fission product barrier column is further divided into two columns; one for Loss thresholds and one for Potential Loss thresholds.

The first column of the table (to the left of the Fuel Clad Loss column) lists the categories (types) of fission product barrier thresholds. The fission product barrier categories are:

A. RPV Water Level B. RCS Leak Rate C. PC Conditions D. PC Radiation I RCS Activity E. PC Integrity or Bypass F. Emergency Director Judgment Each category occupies a row in Table F-1 thus forming a matrix defined by the categories. The intersection of each row with each Loss/Potential Loss column forms a cell in which one or more fission product barrier thresholds appear. If NEI 99-01 does not define a threshold for a barrier Loss/Potential Loss, the word "None" is entered in the cell.

Thresholds are assigned sequential numbers within each Loss and Potential Loss column beginning with number one. In this manner, a threshold can be identified by its category title and number. For example, the first Fuel Clad barrier Loss in Category A would be assigned "FC Loss A.1," the third Containment barrier Potential Loss would be assigned "PC P-Loss B.3," etc.

If a cell in Table F-1 contains more than one numbered threshold, each of the numbered thresholds, if exceeded, signifies a Loss or Potential Loss of the barrier. It is not necessary to exceed all of the thresholds in a category before declaring a barrier Loss/Potential Loss.

Subdivision of Table F-1 by category facilitates association of plant conditions to the applicable fission product barrier Loss and Potential Loss thresholds. This structure promotes a systematic approach to assessing the classification status of the fission product barriers.

When equipped with knowledge of plant conditions related to the fission product barriers, the EAL-user first scans down the category column of Table F-1, locates the likely category and then reads across the fission product barrier Loss and Potential Loss thresholds in that category to determine if a threshold has been exceeded. If a threshold has not been exceeded, the EAL-user proceeds to the next likely category and continues review of the thresholds in the new category.

If the EAL-user determines that any threshold has been exceeded, by definition, the barrier is lost or potentially lost - even if multiple thresholds in the same barrier column are exceeded, only that one barrier is lost or potentially lost. The EAL-user must examine each of the three fission product barriers to determine if other barrier thresholds in the category are lost or potentially lost. For example, if containment radiation is sufficiently high, a Loss of the Fuel Clad and RCS barriers and a Potential Loss of the primary containment barrier can occur. Barrier Losses and Potential Losses are then applied to the algorithms given in EALs FG1 .1, FS1 .1, and FA 1.1 to determine the appropriate emergency classification.

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, ... , F. Blank sections of the table do not have basis pages.

Attachment 6.1, EAL Technical Bases

• Number: 13.1.1A

Title:

Classifying The Emergency - Technical Bases

• l Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A Page: 65 of 187 Table F-1 Fission Product Barrier Threshold Matrix FC - Fuel Clad Barrier RCS - Reactor Coolant System Barrier PC - Containment Barrier Category Loss Potential Loss Loss Potential Loss Loss Potential Loss A RPV level cannot be restored and RPV level cannot be restored and None None SAG entry required maintained GT -161 in. maintained GT -161 in. SAG entry required RPVWater or cannot be determined or cannot be determined Level UNISOLABLE break in fil!l! of the following:

Main steam lines UNISOLABLE primary system leakage UNISOLABLE primary system

. RCIC steam Line that results in exceeding EITHER:

leakage that results in exceeding EITHER:

B None None . RWCU RB area temperature alarm level (PPM 5.3.1 Table 23)

RB area maximum safe operating temperature (PPM 5.3.1 Table 23) None I

RCS Leak Rate . Feedwater OR RB area radiation alarm level OR OR RB area maximum safe operating (PPM 5.3.1 Table 24) radiation (PPM 5.3.1 Table 24)

Emergency RPV Depressurization is required PC pressure GT 45 psig UNPLANNED rapid drop in PC OR c None None PC pressure GT 1.68 psig None pressure following PC pressure rise OR Explosive mixture exists inside PC (H 2 GE 6% and 0 2 GE 5%)

PC due to RCS leakage PC pressure response not consistent OR Conditions with LOCA conditions WW temperature and RPV pressure cannot be maintained below the HCTL Containment Radiation Monitor D CMS-RIS-27E or CMS-RIS-27F Containment Radiation Monitor CMS- Containment Radiation Monitor CMS-reading GT 3,600 R/hr None None RIS-27E or CMS-RIS-27F reading PC Rad/ None RIS-27E or CMS-RIS-27F reading RCS OR GT70 R/hr GT 14,000 R/hr I Primary coolant activity GT 300 Activity 11

µCi/~m dose eouivalent 1-131 UNISOLABLE direct downstream I E pathway to the environment exists PC Integrity None None None None after PC isolation signal None I OR or Bypass Intentional PC ventina oer EOPs I II F 8ill'. condition in the opinion of the 8ill'. condition in the opinion of the 8ill'. condition in the opinion of the 8ill'. condition in the opinion of the 8ill'. condition in the opinion of the 8ill'. condition in the opinion of the Emergency Director that indicates :I Emergency Emergency Director that indicates Emergency Director that indicates Emergency Director that indicates Emergency Director that indicates Emergency Director that indicates potential loss of the Containment Director loss of the fuel clad barrier potential loss of the fuel clad barrier loss of the RCS barrier potential loss of the RCS barrier loss of the Containment barrier barrier Judgment II

!l Attachment 6.1, EAL Technical Bases

Number: 13.1.1A \ Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 66 of 187

• Category:

Subcategory:

Initiating Condition:

Fission Product Barrier Degradation N/A Loss of any two barriers and loss or potential loss of third barrier EAL:

FG1.1 General Emergency Loss of any two barriers AND Loss or potential loss of third barrier (Table F-1)

Mode Applicability:

1 2 3 Basis:

Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 (Attachment 2) lists the fission product barrier thresholds, bases and references.

At the General Emergency classification level each barrier is weighted equally. A General Emergency is therefore appropriate for any combination of the following conditions:

Loss of Fuel Clad, RCS and Containment barriers Loss of Fuel Clad and RCS barriers with potential loss of Containment barrier Loss of RCS and Containment barriers with potential loss of Fuel Clad barrier Loss of Fuel Clad and Containment barriers with potential loss of RCS barrier CGS Basis Reference(s):

1. NEI 99-01 FG1

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: 032 r--------------------~------------; Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 67 of 187

• Category:

Subcategory:

Fission Product Barrier Degradation Initiating Condition:

N/A Loss or potential loss of any two barriers EAL:

FS1.1 Site Area Emergency Loss or potential loss of any two barriers (Table F-1)

Mode Applicability:

1 2 3 Basis:

Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 (Attachment 2) lists the fission product barrier thresholds, bases and references.

At the Site Area Emergency classification level, each barrier is weighted equally. A Site Area Emergency is therefore appropriate for any combination of the following conditions:

• One barrier loss and a second barrier loss (i.e., loss - loss)

• One barrier loss and a second barrier potential loss (i.e., loss - potential loss)

• One barrier potential loss and a second barrier potential loss (i.e., potential loss - potential loss)

At the Site Area Emergency classification level, the ability to dynamically assess the proximity of present conditions with respect to the threshold for a General Emergency is important. For example, the existence of Fuel Clad and RCS Barrier loss thresholds in addition to offsite dose assessments would require continual assessments of radioactive inventory and Containment integrity in anticipation of reaching a General Emergency classification. Alternatively, if both Fuel Clad and RCS potential loss thresholds existed, the Emergency Director would have greater assurance that escalation to a General Emergency is less imminent.

CGS Basis Reference(s):

1. NEI 99-01 FS1

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 68 of 187

• Category:

Subcategory:

Initiating Condition:

Fission Product Barrier Degradation N/A Any loss or any potential loss of either Fuel Clad or RCS EAL:

FA1.1 Alert Any loss or any potential loss of EITHER Fuel Clad or RCS barrier (Table F-1)

Mode Applicability:

2 3 Basis:

Fuel Clad, RCS and Containment comprise the fission product barriers.

Table F-1 (Attachment 6.2) lists the fission product barrier thresholds, bases and references.

At the Alert classification level, Fuel Clad and RCS barriers are weighted more heavily than the Containment barrier. Unlike the Containment barrier, loss or potential loss of either the Fuel Clad or RCS barrier may result in the relocation of radioactive materials or degradation of core cooling capability. Note that the loss or potential loss of Containment barrier in combination with loss or potential loss of either Fuel Clad or RCS barrier results in declaration of a Site Area Emergency under EAL FS1.1

• CGS Basis Reference(s):

1. NEI 99-01 FA1

• Attachment 6.1, EAL Tec.hnical Bases

Number: 13.1.1A luseC~egor~REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 69 of 187

• Barrier:

Category:

Fuel Clad A. RPV Level Degradation Threat: Loss Threshold:

\ SAG entry required Basis:

EOP flowcharts provide instructions to assure adequate core cooling by restoring and maintaining RPV water level above prescribed limits, operate sufficient RPV injection sources to assure adequate core cooling, and assess the possibility of core damage when RPV level cannot be determined. The Fuel Clad Loss threshold conditions are the EOP flowchart conditions that signal a loss of adequate core cooling and a requirement to exit all EOPs and enter the SAGs (ref. 1-6).

This threshold is also a Loss of the RCS barrier (RCS Loss A) and a Potential Loss of the Containment barrier (PC P-Loss A), and therefore represents a Loss of two barriers and a Potential Loss of a third, which requires a General Emergency classification.

The Loss threshold represents the EOP requirement for entry to the Severe Accident Guidelines (SAGs) .

CGS Basis Reference(s):

1. PPM 5.1.1 RPV Control

2. PPM 5.1.2 RPV Control - A TWS

3. Calculation NE-02-03-06 Attachment 10 RPV Variables

4. PPM 5.0.10 Flowchart Training Manual

5. PPM 5.1.4 RPV Flooding

6. PPM 5.1.6 RPV Flooding-ATWS

7. NEI 99-01 RPV Water Level Fuel Clad Loss 2.A

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A luseC~ego~:REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 70 of 187

• Barrier:

Category:

Fuel Clad A. RPV Level Degradation Threat: Potential Loss Threshold:

RPV level cannot be restored and maintained GT -161 in. or cannot be determined Basis:

An RPV water level instrument reading of-161 in. indicates RPV level is at the top of active fuel (TAF)

(ref. 1, 2). When RPV level is at or above the TAF, the core is completely submerged. Core submergence is the most desirable means of core cooling. When RPV level is below TAF, the uncovered portion of the core must be cooled by less reliable means (i.e., steam cooling or spray cooling). If core uncove~ is threatened, the EOPs specify alternate, more extreme, RPV water level control measures in order to restore and maintain adequate core cooling. Since core uncove~ begins if RPV level drops to TAF, the level is indicative of a challenge to core cooling and the Fuel Clad barrier.

When RPV level cannot be determined, EOPs require RPV flooding strategies-. RPV water level indication provides the prima~ means of knowing if adequate core cooling is being maintained. When all means of determining RPV level are unavailable, the fuel clad barrier is threatened and reliance on alternate means of assuring adequate core cooling must be attempted. The instructions in PPM 5.1.4 and PPM 5.1.6 specify these means, which include emergency depressurization of the RPV and injection into the RPV at a rate needed to flood to the elevation of the main steam lines or hold RPV pressure above the Minimum Steam Cooling Pressure (in ATWS events) (ref. 3, 4). If RPV level cannot be determined with respect to the top of active fuel, a*potential loss of the fuel clad barrier exists.

This water level corresponds to the top of the active fuel and is used in the EOPs to indicate a challenge to core cooling.

The RPV water level threshold is the same as RCS barrier Loss RPV Water Level threshold .A. Thus, this threshold indicates a Potential Loss of the Fuel Clad barrier and a Loss of the RCS barrier that appropriately escalates the emergency classification level to a Site Area Emergency.

This threshold is considered to be exceeded when, as specified in the site-specific EOPs, RPV water cannot be restored and maintained above the specified level following depressurization of the RPV (either manually, automatically or by failure of the RCS barrier) or when procedural guidance or a lack of low pressure RPV injection sources preclude Emergency RPV depressurization. EOPs allow the operator a wide choice of RPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurization of the RPV in order to facilitate RPV water level control with low-pressure injection sources. In some events, elevated RPV pressure may prevent restoration of RPV water level until pressure drops below the shutoff heads of available injection sources. Therefore, this Fuel Clad barrier Potential Loss is met only after either:

1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or

2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory.

Attachment 6.1, EAL Technical Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: 032

---

< Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 71 of 187

• The term "cannot be restored and maintained above" means the value of RPV water level is not able to be brought above the specified limit (top of active fuel). The determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the current value is below the top of active fuel, but does not permit extended operation below the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained.

In high-power ATWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level to the top of active fuel in order to reduce reactor power. RPV water level is then controlled between the top of active fuel and the Minimum Steam Cooling RPV Water Level (MSCRWL). Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority. For such events, ICs MA6 or MS6 will dictate the need for emergency classification.

Since the loss of ability to determine if adequate core cooling is being provided presents a significant challenge to the fuel clad barrier, a potential loss of the fuel clad barrier is specified.

CGS Basis Reference(s):

1. Calculation NE-02-03-05 Attachment 3 Note 8

2. PPM 5.1.1 RPV Control

3. PPM 5.1.4 RPV Flooding

4. PPM 5.1.6 RPV Flooding - A TWS

5. PPM 5.1.2 RPV Control - A TWS

6. NEI 99-01 RPV Water Level Fuel Clad Potential Loss 2.A

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A j Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 72 of 187

• Barrier:

Category:

Fuel Clad D. PC Radiation I RCS Activity Degradation Threat: Loss Threshold:

Containment Radiation Monitor CMS-RIS-27E or CMS-RIS-27F reading GT 3,600 R/hr Basis:

Four high range area radiation detectors (CMS-RE-27A, B, E and F) are installed to monitor the drywell. CMS-RE-27A and -27B are located in the bioshield wall at elevations 522' and 525', azimuth 60° and 297°, respectively. CMS-RE-27E and -27F are located inside containment at elevation 515',

azimuth 290° and 51.5°, respectively. The companion containment radiation monitors (CMS-RIS-27A, B, E and F) are located on RAD Boards 22 and 23 in the Main Control Room. (ref. 1)

The threshold value was calculated assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with a concentration of 300 µCi/gm dose equivalent 1-131 (or approximately 5% clad failure) into the drywell atmosphere. Evaluation of detector location, geometry and anticipated response suggests CMS-RIS-27E or F will provide the desired response to a

• given radiation source in the drywell and are, therefore, identified as the preferred monitors for evaluating this Fuel Clad Loss threshold. (ref. 2)

The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that reactor coolant activity equals 300 µCi/gm dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.

The radiation monitor reading in this threshold is higher than that specified for RCS Barrier Loss threshold D since it indicates a loss of both the Fuel Clad Barrier and the RCS Barrier. Note that a combination of the two monitor readings appropriately escalates the emergency classification level to a Site Area Emergency.

There is no Potential Loss threshold associated with primary containment radiation.

CGS Basis Reference(s):

1. TM-2117 TSG - Core Thermal Engineer, Attachment 4.2

2. Calculation NE-02-94-57

2. NEI 99-01 Primary Containment Radiation Fuel Clad Loss 4.A

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 73 of 187

• Barrier:

Category:

Fuel Clad D. PC Radiation I RCS Activity Degradation Threat: Loss Threshold:

Primary coolant activity GT 300 µCi/gm dose equivalent 1-131 Basis:

This threshold indicates that RCS radioactivity concentration is greater than 300 µCi/gm Dose Equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.

It is recognized that sample collection and analysis of reactor coolant with highly elevated activity levels could require several hours to complete. Nonetheless, a sample-related threshold is included as a backup to other indications.

There is no Potential Loss threshold associated with RCS Activity.

• CGS Basis Reference(s):

1. NEI 99-01 RCS Activity Fuel Clad Loss 1.A

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A J Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 74of187

• Barrier:

Category:

Fuel Clad F. Emergency Director Judgment Degradation Threat: Loss.

Threshold:

Any condition in the opinion of the Emergency Director that indicates loss of the Fuel Clad barrier Basis:

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

• Imminent barrier degradation exists if the degradation will likely occur within two hours 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 Director should be mindful of the Loss of AC power (Station Blackout) and A TWS EALs to assure timely emergency classification declarations.

This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad barrier is lost.

CGS Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment Fuel Clad Loss 6.A

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 75 of 187

• Barrier:

Category:

Fuel Clad F. Emergency Director Judgment Degradation Threat: Potential Loss Threshold:

Any condition in the opinion of the Emergency Director that indicates potential loss of the Fuel Clad barrier Basis:

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

• Imminent barrier degradation exists if the degradation will likely occur within two hours 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 Director should be mindful of the Loss of AC power (Station Blackout) and A TWS EALs to assure timely emergency classification. declarations.

This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad 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.

CGS Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment Potential Fuel Clad Loss 6.A

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 76 of 187

• Barrier:

Category:

Reactor Coolant System A. RPV Water Level Degradation Threat: Loss Threshold:

RPV level cannot be restored and maintained GT -161 in. or cannot be determined Basis:

An RPV water level instrument reading of -161 in. indicates level is at the top of active fuel (TAF)

(ref. 1, 2). TAF is significantly lower than the normal operating RPV level control band. To reach this level, RPV inventory loss would have previously required isolation of the RCS and PC barriers, and initiation of all ECCS. If RPV water level cannot be maintained above TAF, ECCS and other sources of RPV injection have been ineffective or incapable of reversing the decreasing level trend. The cause of the loss of RPV inventory is therefore assumed to be a LOCA. By definition, a LOCA event is a Loss of the RCS barrier.

When RPV water level cannot be determined, EOPs require RPV flooding strategies. RPV water level indication provides the primary means of knowing if adequate core cooling is being maintained. The instructions in PPM 5.1.4 and PPM 5.1.6 specify emergency depressurization of the RPV, which is defined to be a Loss of the RCS barrier (RCS Loss B threshold #2). (ref. 3, 4)

The conditions of this threshold are also a Potential Loss of the Fuel Clad barrier (FC P-Loss A). A Loss of the RCS barrier and Potential Loss of the Fuel Clad barrier requires a Site Area Emergency classification.

This water level corresponds to the top of active fuel and is used in the EOPs to indicate challenge to core cooling.

The RPV water level threshold is the same as the Fuel Clad barrier RPV Water Level Potential Loss threshold. Thus, this threshold indicates a Loss of the RCS barrier and Potential Loss of the Fuel Clad barrier and that appropriately escalates the emergency classification level to a Site Area Emergency.

This threshold is considered to be exceeded when, as specified in the site-specific EOPs, RPV water cannot be restored and maintained above the specified level following depressurization of the RPV (either manually, automatically or by failure of the RCS barrier) or when procedural guidance or a lack of low pressure RPV injection sources preclude Emergency RPV depressurization. EOPs allow the operator a wide choice of RPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurization of the RPV in order to facilitate RPV water level control with low-pressure injection sources. In some events, elevated RPV pressure may prevent.

restoration of RPV water level until pressure drops below the shutoff heads of available injection sources. Therefore, this RCS barrier Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of

• RPV inventory .

Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 77 of 187

• The term, "cannot be restored and maintained above," means the value of RPV water level is not able to be brought above the specified limit (top of active fuel). The determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the current value is below the top of active fuel, but does not permit extended operation beyond the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained.

In high-power ATWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level to the top of active fuel in order to reduce reactor power. RPV water level is then controlled between the top of active fuel and the Minimum Steam Cooling RPV Water Level (MSCRWL). Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority. For such events, ICs MA6 or MS6 will dictate the need for emergency classification.

There is no RCS Potential Loss threshold associated with RPV Water Level.

CGS Basis Reference(s):

1. Calculation NE-02-03-05 Attachment 3 Note 8

2. PPM 5.1.1 RPV Control

3. PPM 5.1.4 RPV Flooding

4. PPM 5.1.6 RPV Flooding-.ATWS

5. PPM 5.1.2 RPV Control - ATWS

6. NEI 99-01 RPV Water Level RCS Loss 2.A

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: 032 1--~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~MinorRev: NIA

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 78 of 187

• Barrier:

Category:

Reactor Coolant System B. RCS Leak Rate Degradation Threat: Loss Threshold:

UNISOLABLE break in any of the following:

• Main steam line

• RCIC steam line

• RWCU

• . Feedwater Basis:

The conditions of this threshold include required containment isolation failures allowing a flow path to the environment. A release pathway outside primary containment exists when flow is not prevented by downstream isolations. In the case of a failure of both isolation valves to close but in which no downstream flowpath exists, emergency declaration under this threshold would not be required.

• Similarly, if the emergency response requires the normal process flow of a system outside containment (e.g., EOP requirement to bypass MSIV low RPV water level interlocks and maintain the main condenser as a heat sink using main turbine bypass valves), the threshold is not met. The combination of these threshold conditions represent the loss of both the RCS and Containment (see PC Loss E Threshold #1) barriers and justifies declaration of a Site Area Emergency (i.e., Loss or Potential Loss of any two barriers). (ref. 1-4)

Even though RWCU and Feedwater systems do not contain steam, they are included in the list because an unisolable break could result in the high-pressure discharge of fluid that is flashed to steam from relatively large volume systems directly connected to the RCS. (ref. 1)

Large high-energy lines that rupture outside primary containment can discharge significant amounts of inventory and jeopardize the pressure-retaining capability of the RCS until they are isolated. If it is determined that the ruptured line cannot be promptly isolated from the control room, the RCS barrier Loss threshold is met.

CGS Basis Reference{s):

1. FSAR Section 5.4.5

2. FSAR Section 5.4.6

3. FSAR Section 5.4.8

4. FSAR Section 10.3

5. NEI 99-01 RCS Leak Rate RCS Loss 3.A

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 79 of 187

• Barrier:

Category:

Reactor Coolant System B. RCS Leak Rate Degradation Threat: Loss Threshold:

Emergency RPV Depressurization is required Basis:

Plant symptoms requiring Emergency RPV Depressurization per the EOPs are indicative of a loss of the RCS barrier. Emergency RPV Depressurization is specified in the EOP flowcharts when symbols containing the phrase "EMERG DEPRESS REQ'D" are reached (ref. 1-7). If Emergency RPV Depressurization is required, the plant operators are directed to open safety relief valves (SRVs) and keep them open as needed to maintain adequate core cooling with available injection sources (ref. 8, 9). Even though the RCS is being vented into the suppression pool, a loss of the RCS exists due to the diminished effectiveness of the RCS pressure barrier to a release of fission products beyond its boundary.

Emergency RPV Depressurization in accordance with the EOPs is indicative of a loss of the RCS barrier. If Emergency RPV Depressurization is performed, the plant operators are directed to open safety relief valves (SRVs) and keep them open. Even though the RCS is being vented into the suppression pool, a Loss of the RCS barrier exists due to the diminished effectiveness of the RCS to retain fission products within its boundary.

CGS Basis Reference(s):

1. PPM 5.1.1 RPV Control

2. PPM 5.1.2 RPV Control - A TWS

3. PPM 5.1.4 RPV Flooding

4. PPM 5.1.6 RPV Flooding -ATWS

5. PPM 5.2.1 Primary Containment Control

6. PPM 5.3.1 Secondary Containment Control

7. PPM 5.4.1 Radioactivity Release Control

8. PPM 5.1.3 Emergency RPV Depressurization

9. PPM 5.1.5 Emergency RPV Depressurization -ATWS

10. NEI 99-01 RCS Leak Rate RCS Loss 3.8

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 80 of 187

• Barrier:

Category:

Reactor Coolant System B. RCS Leak Rate Degradation Threat: Potential Loss Threshold:

UNISOLABLE primary system leakage that results in exceeding EITHER:

RB area temperature alarm level (PPM 5.3.1 Table 23)

OR RB area radiation alarm level (PPM 5.3.1 Table 24)

Basis:

The presence of elevated general area temperatures or radiation levels in the secondary containment may be indicative of unisolable primary system leakage outside the primary containment. The PPM 5.3.1 Table 23 and Table 24 alarm levels define this RCS threshold because they are the maximum normal operating values and signify the onset of abnormal system operation. When parameters reach this level, equipment failure or misoperation may be occurring. Elevated parameters may also adversely affect the ability to gain access to or operate equipment within the affected area. The

• locations into which the primary system discharge is of concern correspond to the areas addressed in PPM 5.3.1 Tables 23 and 24 (ref. 1).

Area temperature alarms are provided by the leak detection and reactor building recirculation air (RRA) systems (ref. 2)

The ARM alarm setpoints listed in Table 24 vary due to plant operating mode and Health Physics radiation surveys. A program is established to maintain the current setpoint values in PPM 4.602.A5 for annunciator window 3-1; thus, reference is made to the annunciator response procedure in Table 24.

(ref. 2)

In general, multiple indications should be used to determine if a primary system is discharging outside primary containment. For example, a high area radiation condition does not necessarily indicate that a primary system is discharging into the secondary containment since this may be caused by radiation shine from nearby steam lines or the movement of radioactive materials. Conversely, a high area radiation condition in conjunction with other indications (e.g. room flooding, high area temperatures, reports of steam in the secondary containment, an unexpected rise in feedwater flowrate, or unexpected main turbine control valve closure) may indicate that a primary system is discharging into the secondary containment.

Potential loss of RCS based on primary system leakage outside the primary containment is determined from EOP temperature or radiation Max Normal Operating values in areas such as main steam line tunnel, RCIC, etc., which indicate a direct path from the RCS to areas outside primary containment.

A Max Normal Operating value is the highest value of the identified parameter expected to occur during normal plant operating conditions with all directly associated support and control systems functioning properly .

Attachment 6.1, EAL Technical Bases

Number: 13.1.1A IUse Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

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• The indicators reaching the threshold barriers and confirmed to be caused by RCS leakage from a primary system warrant an Alert classification. A primary system is defined to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system.

An UNISOLABLE leak which is indicated by Max Normal Operating values escalates to a Site Area .

Emergency when combined with Containment Barrier Loss threshold 3.A (after a containment isolation) and a General Emergency when the Fuel Clad Barrier criteria is also exceeded.

CGS Basis Reference(s):

1. PPM 5.3.1 Secondary Containment Control

2. PPM 5.0.10 Flowchart Training Manual

3. NEI 99-01 RCS Leak Rate RCS Potential Loss 3.A

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 82 of 187

• Barrier:

Category:

Reactor Coolant System C. PC Conditions Degradation Threat: Loss Threshold:

PC pressure GT 1.68 psig due to RCS leakage Basis:

The drywell high pressure scram setpoint is an entry condition to the EOP flowcharts: PPM 5.1.1, RPV Control, and PPM 5.2.1, Primary Containment Control (ref. 1, 2, 3). Normal primary containment (PC) pressure control functions such as operation of drywell cooling and venting through SGT are specified in PPM 5.2.1 in advance of less desirable but more effective functions such as operation of drywell or wetwell sprays.

In the CGS design basis, primary containment pressures above the drywell high pressure scram setpoint are assumed to be the result of a high-energy release into the containment for which normal pressure control systems are inadequate or incapable of reversing the increasing pressure trend.

Pressures of this magnitude, however, can be caused by non-LOCA events such as a loss of drywell

• cooling or inability to control primary containment venUpurge (ref. 3) .

The threshold phrase " ... due to RCS leakage" focuses the barrier failure on the RCS instead of the non-LOCA malfunctions that may adversely affect primary containment pressure. Primary containment pressure greater than 1.68 psig with corollary indications (e.g., elevated drywell temperature, indications of loss of RCS inventory) should therefore be considered a Loss of the RCS barrier. Loss of drywell cooling that results in pressure greater than 1.68 psig should not be considered an RCS barrier loss.

1.68 psig is the drywell high pressure setpoint which indicates a LOCA by automatically initiating the ECCS or equivalent makeup system.

There is no Potential Loss threshold associated with drywell pressure.

CGS Basis Reference(s):

1. Technical Specifications Table 3.3.5.1-1

2. PPM 5.1.1 RPV Control

3. PPM 5.2.1 Primary Containment Control

4. FSAR Section 6

5. NEI 99-01 Primary Containment Pressure RCS Loss 1.A

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A IUse Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 83 of 187

• Barrier:

Category:

Reactor Coolant System D. PC Radiation I RCS Activity Degradation Threat: Loss Threshold:

Containment Radiation Monitor CMS-RIS-27E or CMS-RIS-27F reading GT 70 R/hr Basis:

Four high range area radiation detectors (CMS-RE-27A, B, E and F) are installed in the drywell. CMS-RE-27 A and -27B are located in the bioshield wall at elevations 522' and 525', azimuth 60° and 297°,

respectively. CMS-RE-27E and -27F are located inside containment at elevation 515', azimuth 290° and 51.5°, respectively. The companion containment radiation monitors (CMS-RIS-27A, B, E and F) are located on RAD Boards 22 and 23 in the Main Control Room. (ref. 1)

The threshold value was calculated assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with normal operating concentrations (i.e., within Technical Specifications) into the drywell atmosphere. Evaluation of detector location, geometry and anticipated response suggests CMS-RIS-27E or F will provide the desired response to a given radiation

• source in the drywell and are, therefore, identified as the preferred monitors for evaluating this RCS Loss threshold. (ref. 2)

The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for Fuel Clad Barrier Loss threshold D.1 since it indicates a loss of the RCS Barrier only. -

There is no Potential Loss threshold associated with primary containment radiation.

CGS Basis Reference(s):

1. TM-2117 TSG - Core Thermal Engineer, Attachment 4.2

2. Calculation NE-02-94-57

3. NEI 99-01 Primary Containment Radiation RCS Loss 4.A

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A \ Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 84 of 187 Barrier: Reactor Coolant System Category: F. Emergency Director Judgment Degradation Threat: Loss Threshold:

Any condition in the opinion of the Emergency Director that indicates loss of the RCS barrier Basis:

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

• Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance. The term "imminent" refers to the 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 Director 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 Director in determining whether the RCS Barrier is lost.

CGS Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment RCS Loss 6.A

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 85 of 187

• Barrier:

Category:

Reactor Coolant System F. Emergency Director Judgment .

• Degradation Threat: Potential Loss Threshold:

Any condition in the opinion of the Emergency Director that indicates potential loss of the RCS barrier Basis:

The Emergency Director judgment threshold addresses any other factors relevant to determining if the RCS barrier is potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

• Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance. The term "imminent" refers to the inability to reach final safety acceptance criteria before completing 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 Director 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 Director 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.

CGS Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment RCS Potential Loss 6.A

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 86 of 187

• Barrier:

Category:

Containment A. RPV Water Level Degradation Threat: Potential Loss Threshold:

I SAG entry required Basis:

EOP flowcharts provide instructions to assure adequate core cooling by restoring and maintaining RPV water level above prescribed limits, operate sufficient RPV injection sources to assure adequate core cooling, and assess the possibility of core damage when RPV level cannot be determined. The Fuel Clad Loss threshold conditions are the EOP flowchart conditions that signal a loss of adequate core cooling and a requirement to exit all EOPs and enter the SAGs (ref. 1-6).

This threshold is also a Loss of the RCS barrier (RCS Loss A) and a Loss of the Fuel Clad barrier (FC Loss A), and therefore represents a Loss of two barriers and a Potential Loss of a third, which requires a General Emergency classification.

The Potential Loss threshold is identical to the Fuel Clad Loss RPV Water Level threshold. The Potential Loss requirement for SAG entry indicates adequate core cooling cannot be restored and maintained and that core damage is possible. BWR EPGs/SAGs specify the conditions that require SAG entry. When SAG entry is required, the EPGs are exited and SAGs are entered. Entry into SAGs is a logical escalation in response to the inability to restore and maintain adequate core cooling.

PRA studies indicate that the condition of this Potential Loss threshold could be a core melt sequence which, if not corrected, could lead to RPV failure and increased potential for primary containment failure. In conjunction with the RPV water level Loss thresholds in the Fuel Clad and RCS barrier columns, this threshold results in the declaration of a General Emergency.

CGS Basis Reference(s):

1. PPM 5.1.1 RPV Control

2. PPM 5.1.2 RPV Control - ATWS

3. Calculation NE-02-03-06 Attachment 10 RPV Variables

4. PPM 5.0.1 O Flowchart Training Manual

5. PPM 5.1.4 RPV Flooding

6. PPM 5.1.6 RPV Flooding - A TWS

7. NEI 99-01 RPV Water Level PC Potential Loss 2.A

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 87 of 187

• Barrier:

Category:

Containment B. RCS Leak Rate Degradation Threat: Loss Threshold:

UNISOLABLE primary system leakage that results in exceeding EITHER:

RB area maximum safe operating temperature (PPM 5.3.1 Table 23)

OR RB area maximum safe operating radiation (PPM 5.3.1 Table 24)

Basis:

The presence of elevated general area temperatures or radiation levels in the Reactor Building (RB) may be indicative of unisolable primary system leakage outside the primary con.tainment. The maximum safe operating values define this Containment barrier threshold because they are indicative of problems in the secondary containment that are spreading and pose a threat to achieving a safe plant shutdown. This threshold addresses problematic discharges outside primary containment that may not originate from a high-energy line break. The locations into which the primary system discharge

• is of concern correspond to the areas addressed in PPM 5.3.1 Tables 23 and 24 (ref. 1) .

RB maximum safe operating temperatures are conservatively defined by the qualification temperature of safety related equipment in the area. The equipment qualification program has proven that safety related equipment will perform satisfactorily to at least this temperature. In an area with multiple components and different qualification temperatures, the maximum safe operating temperature assigned to that area is generally the lowest of the individual temperatures. (ref. 2)

The maximum safe operating radiation value is defined to be 10,000 mR/hr in areas other than the refueling floor. This is the maximum indication bn all but the high level instruments. This value is high enough to be indicative of substantial and immediate problems yet low enough to allow time for shutdown or isolation of a leak without exceeding the total integrated dose allowable for even the most sensitive safety related equipment. No area radiation levels are defined for the refueling floor because no primary systems are routed there. (ref. 2)

In general, multiple indications should be used to determine if a primary system is discharging outside primary containment. for example, a high area radiation condition does not necessarily indicate that a primary system is discharging into the secondary containment since this may be caused by radiation shine from nearby steam lines or the movement of radioactive materials. Conversely, a high area radiation condition in conjunction with other indications (e.g. room flooding, high area temperatures, reports of steam in the secondary containment, an unexpected rise in feedwater flowrate, or unexpected main turbine control valve closure) may indicate that a primary system is discharging into the secondary containment.

The Max Safe Operating Temperature and the Max Safe Operating Radiation Level are each the highest value of these parameters at which neither: (1) equipment necessary for the safe shutdown of the plant will fail, nor (2) personnel access necessary for the safe shutdown of the plant will be precluded. EOPs utilize these temperatures and radiation levels to establish conditions under which RPV depressurization is required.

Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 88 of 187

• The temperatures and radiation levels should be confirmed to be caused by RCS leakage from a primary system. A primary system is defined to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system.

In combination with the RCS Potential Loss RCS Leak Rate threshold this threshold would result in a Site Area Emergency.

There is no Potential Loss threshold associated with primary containment isolation failure.

CGS Basis Reference(s):

1. PPM 5.3.1 Secondary Containment Control

2. PPM 5.0.1 O Flowchart Training Manual

3. NEI 99-01 RCS Leak Rate PC Loss 3.C

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A \ Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 89 of 187

• Barrier:

Category:

Containment C. PC Conditions Degradation Threat: Loss Threshold:

UNPLANNED rapid drop in PC pressure following PC pressure rise Basis:

Rapid UNPLANNED loss of primary containment pressure (i.e., not attributable to containment spray or condensation effects) following an initial pressure increase indicates a loss of primary containment integrity.

This threshold relies on operator recognition of an unexpected response for the condition and therefore a specific value is not assigned. The unexpected (UNPLANNED) response is important because it is the indicator for a containment bypass condition.

CGS Basis Reference(s):

1. NEI 99-01 Primary Containment Conditions PC Loss 1.A

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: NIA

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 90 of 187

• Barrier:

Category:

Containment C. PC Conditions Degradation Threat: Loss Threshold:

PC pressure response not consistent with LOCA conditions Basis:

This indicator is considered to be a loss of both the RCS and PC barriers.

Normal LOCA conditions are drywell pressure rising with wetwell pressure following. Primary containment or drywell pressure responses not consistent with LOCA conditions indicate a loss of the primary containment barrier. This may be noticed as a decrease in drywell pressure when no operator action (e.g., starting drywell cooling fans) has been taken. It would also include a failure of the drywell pressure to increase as expected during a LOCA. Also, a loss of suppression function in conjunction with a LOCA would indicate a loss of the primary containment barrier. Exceeding Pressure Suppression Pressure (PSP) is an indication of loss of pressure suppression function.

Primary containment pressure should increase as a result of mass and energy release into the primary containment from a LOCA. Thus, primary containment pressure not increasing under these conditions indicates a loss of primary containment integrity.

This threshold relies on operator recognition of an unexpected response for the condition and therefore a specific value is not assigned. The unexpected (UNPLANNED) response is important because it is the indicator for a containment bypass condition.

CGS Basis Reference(s):

1. FSAR Section 6.2.1.1.3.3

2. FSAR Figure 6.2-3

3. FSAR Table 6.2-5

4. FSAR Table 6.2-1

5. NEI 99-01 Primary Containment Conditions PC Loss 1.B

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 91 of 187

• Barrier:

Category:

Containment C. PC Conditions Degradation Threat: Potential Loss Threshold:

I PC pressure GT 45 psig Basis:

If this threshold is exceeded, a challenge to the primary containment structure has occurred because assumptions used in the accident analysis are no longer valid and an unanalyzed condition exists (ref. 1, 2). This constitutes a Potential Loss of the Containment barrier even if a containment breach has not occurred.

The threshold pressure is the primary containment internal design pressure. Structural acceptance testing demonstrates the capability of the primary containment to resist pressures greater than the internal design pressure. A pressure of this magnitude is greater than those expected to result from any design basis accident and, thus, represent a Potential Loss of the Containment barrier .

• CGS Basis Reference{s):

1.

2.

3.

FSAR Table 6.2-1 FSAR Section 6.2 NEI 99-01 Primary Containment Conditions PC Potential Loss 1.A

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

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• Barrier:

Category:

Containment C. PC Conditions Degradation Threat: Potential Loss Threshold:

Explosive mixture exists inside PC {H 2 GE 6% and 02 GE 5%)

Basis:

Explosive (deflagration) mixtures in the primary containment are assumed to be elevated concentrations of hydrogen and oxygen. BWR industry evaluation of hydrogen generation for development of EOPs/SAGs indicates that any hydrogen concentration above minimum detectable is not to be expected within the short term. Post-LOCA hydrogen generation primarily caused by radiolysis is a slowly evolving, long-term condition. Hydrogen concentrations that rapidly develop are most likely caused by metal-water reaction. A metal-water reaction is indicative of an accident more severe than accidents considered in the plant design basis and would be indicative, therefore, of a potential threat to primary containment integrity. Hydrogen concentration of approximately 6% is considered the global deflagration concentration limit (ref. 1).

• Except for brief periods during plant startup and shutdown, oxygen concentration in the primary containment is maintained at insignificant levels by nitrogen inerting. The specified values for this Potential Loss threshold are the minimum global deflagration concentration limits (6% hydrogen and 5% oxygen, ref. 2) and readily recognizable because 6% hydrogen is well above the EOP flowchart entry condition (ref. 3). The minimum global deflagration hydrogen/oxygen concentrations (6% I 5%,

respectively) require intentional primary containment venting, which is defined to be a Loss of Containment (PC Integrity or Bypass).

If hydrogen concentration reaches or exceeds the lower flammability limit, as defined in plant EOPs, in an oxygen rich environment, a potentially explosive mixture exists. If the combustible mixture ignites inside the primary containment, loss of the Containment barrier could occur.

CGS Basis Reference(s):

1. BWROG EPG/SAG Revision 2, Sections PC/G

2. PPM 5.7.1 RPVand Primary Containment Flooding SAG, Table 19

3. PPM 5.2.1 Primary Containment Control

4. FSAR Section 7.5.1.5.4

5. PPM 5.0.10 Flowchart Training Manual

6. PPM 4.814.J1 814.J1 Annunciator Panel Alarms, 2-2

7. PPM 4.814.J2 814.J2 Annunciator Panel Alarms, 2-2

8. NEI 99-01 Primary Containment Conditions PC Potential Loss 1.B

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A IUse Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 93 of 187

• Barrier:

Category:

Containment C. PC Conditions Degradation Threat: Potential Loss Threshold:

WW temperature and RPV pressure cannot be maintained below the HCTL Basis:

The HCTL is given in EOP flowchart Figure C (ref. 1). This is the only instance in which the threshold could be met.

• Heat Capacity Temperature Limit (HCTL) is the highest Wetwell temperature from which emergency RPV depressurization will not exceed:

• Capability of the Wetwell, and equipment within the Wetwell which may be required to operate, when the RPV is pressurized

• Pressure Limit (PCPL), while the rate of energy transfer from the RPV to the Containment is GT the capacity of the Containment vent The HCTL is a function of RPV pressure, suppression pool temperature and suppression pool water level. It is utilized to preclude failure of the containment and equipment in the containment necessary for the safe shutdown of the plant and therefore, the inability to maintain plant parameters below the limit constitutes a potential loss of containment.

CGS Basis Reference(s):

1. PPM 5.2.1 Primary Containment Control

2. NEI 99-01 Primary Containment Conditions PC Potential Loss 1.C

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 94 of 187

• Barrier:

Category:

Containment D. PC Radiation I RCS Activity Degradation Threat: Potential Loss Threshold:

Containment Radiation Monitor CMS-RIS-27E or CMS-RIS-27F reading GT 14,000 R/hr Basis:

Four high range area radiation detectors (CMS-RE-27A, B, E and F) are installed in the drywell.

CMS-RE-27A and -27B are located in the bioshield wall at elevations 522' and 525', azimuth 60° and 297°, respectively. CMS-RE-27E and -27F are located inside containment at elevation 515', azimuth 290° and 51.5°, respectively. The com*panion containment radiation monitors (CMS-RIS-27A, B, E and F) are located on RAD Boards 22 and 23 in the Main Control Room. (ref. 1)

The threshold value was calculated assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with 20% fuel clad damage into the drywell atmosphere. Evaluation of detector location, geometry and anticipated response suggests CMS-RIS-27E or F will provide the desired response to a given radiation source in the drywell and are, therefore,

• identified as the preferred monitors for evaluating this Containment barrier Potential Loss threshold .

(ref. 2)

The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that 20% of the fuel cladding has failed. This level of fuel clad failure is well above that used to determine the analogous Fuel Clad Barrier Loss and RCS Barrier Loss thresholds.

NUREG-1228, Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents, indicates the fuel clad failure must be greater than approximately 20% in order for there to be a major release of radioactivity requiring offsite protective actions. For this condition to exist there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential loss of containment which would then escalate the emergency classification level to a General Emergency.

CGS Basis Reference(s):

1. TM-2117 TSG - Core Thermal Engineer, Attachment 4.2

2. Calculation NE-02-94-57

3. NEI 99-01 Primary Containment Radiation Fuel Clad Potential Loss 1.D

• Attachment 6.1,. EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 95 of 187

• Barrier:

Category:

Containment E. PC Integrity or Bypass Degradation Threat: Loss Threshold:

UNISOLABLE direct downstream pathway to the environment exists after PC isolation signal Basis:

This thres.hold addresses failure of open isolation devices which should close upon receipt of a manual or automatic containment isolation signal resulting in a significant radiological release pathway directly to the environment. The concern is the unisolable open pathway to the environment. A failure of the ability to isolate any one line indicates a breach of containment integrity.

Leakage into a closed system is to be considered only if the closed system is breached and thereby creates a significant pathway to the environment. Examples include unisolable main steam line or RCIC steam line breaks, unisolable RWCU system breaks, and unisolable PC vent paths.

PPM 5.2.1, Primary Containment Control, may specify primary containment venting and intentional bypassing of the containment isolation valve logic, even if offsite radioactivity release rate limits are exceeded (ref. 1). Under these conditions with a valid containment isolation signal, the Containment barrier should be considered lost.

The use of the modifier "direct" in defining the release path discriminates against r.elease paths through interfacing liquid systems or minor release pathways, such as instrument lines, not protected by the Primary Containment Isolation System (PCIS).

The existence of a filter is not considered in the threshold assessment. Filters do not remove fission product noble gases. In addition, a filter could become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream.

Following the leakage of RCS mass into primary containment and a rise in primary containment pressure, there may be minor radiological releases associated with allowable primary containment leakage through various penetrations or system components. Minor releases may also occur if a primary containment isolation valve(s) fails to close but the primary containment atmosphere escapes to an enclosed system. These releases do not constitute a loss or potential loss of primary containment but should be evaluated using the Recognition Category R ICs.

CGS Basis Reference(s):

1. PPM 5.2.1 Primary Containment Control

2. NEI 99-01 Primary Containment Isolation Failure PC Loss 3.A

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A j Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

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• Barrier:

Category:

Containment E. PC Integrity or Bypass Degradation Threat: Loss Threshold:

I Intentional PC venting per EOPs Basis:

EOP flowcharts (PPM 5.2.1, Primary Containment Control) may specify primary containment venting and intentional bypassing of the containment isolation valve logic, even if offsite radioactivity release rate limits are exceeded (ref. 1). The threshold is met when the operator begins venting the primary containment in accordance with EOP Support Procedures (PPM 5.5.14 or PPM 5.5.15) or ABN-CONT-VENT, not when actions are taken to bypass interlocks prior to opening the vent valves (ref. 2, 3, 4).

Purge and vent actions specified in PPM 5.2.1 to control primary containment pressure below the drywell high pressure scram setpoint or to lower hydrogen concentration does not meet this threshold because such action is only permitted if offsite radioactivity release rates will remain below the ODCM RFO limits (ref. 1).

• EOPs may direct primary containment isolation valve logic(s) to be intentionally bypassed, even if offsite radioactivity release rate limits will be exceeded. Under these conditions with a valid primary containment isolation signal, the containment should also be considered lost if primary containment venting is actually performed.

Intentional venting of primary containment for primary containment pressure control to the secondary containment and/or the environment is a Loss of the Containment. Venting for primary containment pressure control when not in an accident situation (e.g., to control pressure below the drywell high pressure scram setpoint) does not meet the threshold condition.

CGS Basis Reference(s):

1. PPM 5.2.1 Primary Containment Control

2. PPM 5.5.14 Emergency Wetwell Venting

3. PPM 5.5.15 Emergency Drywell Venting

4. ABN-CONT-VENT

5. NEI 99-01 Primary Containment Isolation Failure PC Loss 3.B

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 97 of 187

• Barrier:

Category:

Containment F. Emergency Director Judgment Degradation Threat: Loss Threshold:

Any condition in the opinion of the Emergency Director that indicates loss of the Containment barrier Basis:

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

• Imminent barrier degradation exists if the degradation will likely occur within two hours 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 Director 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 Director in determining whether the Containment Barrier is lost.

CGS Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment PC Loss 6.A

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A luseC~ego~:REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 98 of 187

• Barrier:

Category:

Containment F. Emergency Director Judgment Degradation Threat: Potential Loss Threshold:

Any condition in the opinion of the Emergency Director that indicates potential loss of the Containment barrier Basis:

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

• Imminent barrier degradation exists if the degradation will likely occur within two hours 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 ent~

to the EOPs. The Emergency Director should be mindful of the Loss of AC power (Station Blackout) and A TWS EALs to assure timely emergency classification declarations.

This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Containment Barrier is lost.

CGS Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment PC Loss 6.A

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 99 of 187

• Category H - Hazards and Other Conditions Affecting Plant Safety EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)

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

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

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

3. Natural or 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 Gas Toxic, corrosive, asphyxiant or flammable gas leaks can affect normal plant operations or preclude access to plant areas required to safely shutdown the plant.

6. Control Room Evacuation If the Control Room must be evacuated, additional support for monitoring and controlling plant functions is necessary through the emergency response facilities.

7. Emergency Director 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 Director the latitude to classify emergency conditions consistent with the established classification criteria based upon Emergency Director judgment.

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: NIA

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 100 of 187

• Category:

Subcategory:

Initiating Condition:

H - Hazards 1 - Security HOSTILE ACTION within the Protected Area EAL:

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

I 1 I 2 I 3 I 4 I s I def I Basis:

The Security Shift Supervision is defined as either the Security Lieutenant or the Security Sergeant (ref. 1).

Note that the ISFSI Protected Area is an area separate from the Protected Area surrounding the power block.

The Safeguards Contingency Plan (Appendix C of CGS Physical Security Plan) defines the events that meet the criteria of a SECURITY CONDITION or HOSTILE ACTION (ref. 1).

This IC addresses the occurrence of a HOSTILE ACTION within the PROTECTED AREA. This event will require rapid response and assistance due to the possibility for damage to plant equipment.

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

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

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 a HOSTILE ACTION directed at an ISFSI PROTECTED AREA located outside the plant PROTECTED AREA; such an attack should be assessed using IC HA 1. It also does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73. 71 or 10 CFR § 50. 72.

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

• CGS Basis Reference(s):

1.

2.

CGS Physical Security Plan NEI 99-01 HS1 Attachment 6.1, EAL Technical Bases

Number: 13.1.1A IUse Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

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• Category:

Subcategory:

Initiating Condition:

H - Hazards 1 - Security HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes EAL:

HA1.1 Alert (1) A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the Security Sergeant or Security Lieutenant OR (2) A validated notification from NRC of an aircraft attack threat within 30 min. of the site Mode Applicability:

I 1 I 2 I 3 I 4 I s I def I Basis:

The Security Shift Supervision is defined as either the Security Lieutenant or the Security Sergeant (ref. 1).

Note that the ISFSI Protected Area is an area separate from the Protected Area surrounding the power block.

The Safeguards Contingency Plan (Appendix C of CGS Physical Security Plan) defines the events that meet the criteria of a SECURITY CONDITION or HOSTILE ACTION (ref. 1).

This IC addresses the occurrence of a HOSTILE ACTION within the OVVNER 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 Supervision and the Control Room is essential for proper classification of a security-related event.

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

As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Alert declaration will also heighten the awareness of Offsite Response Organizations (OROs), allowing them to be better prepared should it be necessary to consider further actions.

This IC does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73. 71 or 10 CFR § 50.72 .

Threshold #1 is applicable for any HOSTILE ACTION occurring, or that has occurred, in the OWNER CONTROLLED AREA. This includes any action directed against the ISFSI which is located outside the plant PROTECTED AREA.

Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: NIA

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 102 of 187

• Threshold #2 addresses the threat from the impact of an aircraft on the plant, and the anticipated arrival time is within 30 minutes. The intent of this EAL is to ensure that threat-related notifications are made in a timely manner so that plant personnel and OROs are in a heightened state of readiness. This EAL is met when the threat-related information has been validated in accordance with ABN-AIRBORNE-ATTACK (ref. 2).

The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may be provided by NORAD through the NRC.

In some cases, it may not be readily apparent if an aircraft impact within the OWNER CONTROLLED AREA was intentional (i.e., a HOSTILE ACTION). It is expected, although not certain, that notification by an appropriate Federal agency to the site would clarify this point. In this case, the appropriate federal agency is intended to be NORAD, FBI, FAA or NRC. The emergency declaration, including one based on other ICs/EALs, should not be unduly delayed while awaiting notification by a Federal agency.

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

CGS Basis Reference(s):

1. CGS Physical Security Plan

2. ABN-AIRBORNE-ATTACK

3. NEI 99-01 HA 1

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A IUse Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

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• Category:

Subcategory:

Initiating Condition:

H - Hazards 1 - Security Confirmed SECURITY CONDITION or threat EAL:

HU1.1 Unusual Event (1) A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the Security Sergeant or Security Lieutenant OR (2) Notification of a credible security threat directed at the site OR (3) A validated notification from the NRC providing information of an aircraft threat Mode Applicability:

I 1 I 2 I 3 I 4 I s I def I Basis:

The Security Shift Supervision is defined as either the Security Lieutenant or the Security Sergeant (ref. 1) .

This EAL is based on the CGS Physical Security Plan (ref. 1).

The Safeguards Contingency Plan (Appendix C of CGS Physical Security Plan) defines the events that meet the criteria of a SECURITY CONDITION or HOSTILE ACTION (ref. 1).

This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR § 73. 71 or 10 CFR § 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HA1 and HS1.

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

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

Threshold #1 references the Security Shift Supervision because these are the individuals trained to confirm that a security event is occurring or has occurred. Training on security event confirmation and classification is controlled due to the nature of Safeguards and 10 CFR § 2.39 information.

Threshold #2 addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with the CGS Physical Security Plan (ref. 1).

Threshold #3 addresses the threat from the impact of an aircraft on the plant. The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may also be provided by NORAD through the NRC. Validation of the threat is performed in accordance with ABN-AIRBORNE-ATTACK (ref. 2).

Attachment 6.1, EAL Technical Bases

Number: 13.1.1A IUse Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 104 of 187

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

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

CGS Basis Reference(s):

1. CGS Physical Security Plan

2. ABN-AIRBORNE-ATTACK

3. NEI 99-01 HU1

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 105 of 187

• Category:

Subcategory:

H - Hazards and other Conditions Affecting Plant Safety Initiating Condition:

2 - Seismic Event Seismic event GT OBE levels EAL:

HU2.1 Unusual Event Seismic event GT Operating Basis Earthquake (QBE) as indicated by H13.P851.S1.5-1 (OPERATING BASIS EARTHQUAKE EXCEEDED) activated Mode Applicability:

I 1 I 2 I 3 I 4 I s I def I Basis:

CGS seismic instrumentation consists of a Kinemetrics SMA-3 Strong Motion Accelerograph and associated sensors that are equipped with seismic triggers set to initiate recording at an acceleration equal to or exceeding 0.01 g (ref. 1, 2). This also annunciates the seismic activity alarm H13.P851.S1 .2-5 Minimum Seismic Earthquake Exceeded (ref. 2, 3, 4)

A seismic switch unit that is similar to the seismic trigger unit is also provided. The trip point of the seismic switch unit is set at the maximum acceleration corresponding to the OBE, and it provides immediate Control Room annunciation that the OBE has been exceeded requiring declaration of an Unusual Event (ref. 1, 3, 4)

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 QBE but less than a Safe Shutdown Earthquake (SSE) should have no significant impact on safety-related systems, structures and components; however, some time may be required for the plant staff to ascertain the actual post-event condition of the plant (e.g., performs walk-downs and post-event inspections). Given the time necessary to perform walk-downs and inspections, and fully understand any impacts, this event represents a potential degradation of the level of safety of the plant.

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

Earthquakes of this magnitude should be readily felt by on-site personnel and recognized as a seismic event (e.g., lateral accelerations in excess of 0.08g). The Shift Manager or Emergency may seek external verification if deemed appropriate (e.g., a call to the USGS, check internet news sources, etc.);

however, the verification action must not preclude a timely emergency declaration.

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

CGS Basis Reference(s):

1. CGS FSAR Section 3.7.4 Seismic Instrumentation

2. ISP-SEIS-M201 Seismic Systems Channel Check

3. PPM 4.851.S1 .2-5 Minimum Seismic Earthquake Exceeded

4. ABN-EARTHQUAKE Earthquake

5. NEI 99-01 HU2

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

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• Category:

Subcategory:

Initiating Condition:

H - Hazards and Other Conditions Affecting Plant Safety 3 - Natural or Technology Hazard Hazardous event EAL:

HU3.1 Unusual Event (1) A tornado strike within the PROTECTED AREA OR (2) Volcanic ash fallout requiring plant shutdown Mode Applicability:

I 1 I 2 I 3 I 4 I s I def I Basis:

If damage is confirmed visually or by other in-plant indications, the event may be escalated to an Alert under EAL CA6.1 or MA8.1.

Threshold #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. A dust devil is not a tornado.

Threshold #2 The potential for volcanic eruption exists in the Pacific Northwest. Heavy ash fall, such as that experienced at certain locations following the eruption of Mt. St. Helens in 1980, could affect operation of plant equipment if precautionary measures are not taken. The design basis ash fall is projected for a 20 hour2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> duration. Plant shutdown may be warranted, based on several individual criteria specified in ABN-ASH (ref. 1). This threshold is met when ABN-ASH requires plant shutdown.

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

Threshold #1 addresses a tornado striking (touching down) within the PROTECTED AREA.

Threshold #2 addresses a volcanic ash fallout event.

Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, M orC.

CGS Basis Reference{s):

1. ABN-ASH Ash Fall

2. NEI 99-01 HU3

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

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• Category:

Subcategory:

Initiating Condition:

H - Hazards and Other Conditions Affecting Plant Safety 3 - Natural or Technology Hazard Hazardous event EAL:

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

I 1 I 2 I 3 I 4 I s I def I Basis:

An uncontrolled flooding event may pose a direct threat to safety-related equipment. As such, the potential exists for substantial degradation of the level of safety of the plant. One indication of FLOODING is indicated by ECCS room level alarms on P601 (ref. 1, 2).

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

This EAL addresses FLOODING of a building room or area that results in operators isolating power to a

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

Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, M orC.

CGS Basis Reference(s):

1. Calculation ME 02-02-02 Reactor Building Flooding

2. Calculation ME 02-02-46, RB/RW/TB/DG Corridor Flooding

3. NEI 99-01 HU3

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A j Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

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• Category:

Subcategory:

Initiating Condition:

H - Hazards and Other Conditions Affecting Plant Safety 3 - Natural or Technology Hazard Hazardous event EAL:

HU3.3 Unusual Event (1) Movement of personnel within the PROTECTED AREA is IMPEDED due to an offsite event involving hazardous materials (e.g., an offsite chemical spill, 618-11 event or toxic gas release)

OR (2) A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles (Note 7)

Note 7: This EAL does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents Mode Applicability:

I 1 I 2 13 4 s I def I Basis:

• As used here, the term "offsite" is meant to be areas external to the PROTECTED AREA.

Threshold #1 includes an event at the 618-11 burial ground which would IMPEDE movement of personnel within the PROTECTED AREA.

Threshold #2 includes a range fire causing Hanford officials to limit vehicle access to the site. The origin of the hazardous event could be from on or off-site.

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

Threshold #1 addresses a hazardous materials event originating at an offsite location and of sufficient magnitude to impede the movement of personnel within the PROTECTED AREA.

Threshold #2 addresses a hazardous event that causes an on-site impediment to vehicle movement and significant enough to prohibit the plant staff from accessing the site using personal vehicles.

Examples of such an event include site flooding caused by a hurricane, heavy rains, up-river water releases, dam failure, etc., or an on-site train derailment blocking the access road.

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

Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, M orC.

CGS Basis Reference(s):

1. NEI 99-01 HU3 Attachment 6.1, EAL Technical Bases

Number: 13.1.1A IUse Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

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• Category:

Subcategory:

Initiating Condition:

H - Hazards and Other Conditions Affecting Plant Safety 4- Fire FIRE potentially degrading the level of safety of the plant EAL:

HU4.1 Unusual Event A FIRE is not extinguished within 15 min. of any of the following FIRE detection indications (Note 1):

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

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

• Field verification of a single fire alarm

• AND The FIRE is located within any Table 5 area Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Table 5 Safe Shutdown Areas

• Vital portions of the Rad Waste/Control Building:

467' elevation vital island 487' elevation cable spreading room Main Control Room and vertical cable chase 525' elevation HVAC area

• Reactor Building

• Vital portions of the Turbine Building DEH pressure switches RPS switches on turbine throttle valves Main steam line radiation monitors Turbine Building ventilation radiation monitors Main steam line piping up to MS-V-146 and the first stop valves

• Standby Service Water Pump Houses

• Diesel Generator Building Mode Applicability:

• 1 2 3 4 s I def 1 Attachment 6.1, EAL Technical Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: 032 1--~~~~~~~~~~~~~~~~~~-'-~~~~~~~~~~~MinorRev: N/A

Title:

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• Basis:

A fire alarm can be confirmed by multiple/redundant indications such as additional alarms on FCP-1 or FCP-2, fire pumps starting, fire suppression system discharge, fire water header pressure fluctuations or by notification by plant personnel (ref. 1).

The Table 5 Safe Shutdown Areas include those structures/areas that contain any Class 1, 2 or 3 SSC.

Table 5 includes those structures containing functions and systems required to achieve and maintain cold shutdown (including all auxiliary equipment such as AC/DC power, cooling water and instrumentation) (ref. 2).

The concept of this EAL is that a fire exists in a Table 5 area that is not extinguished within 15 minutes.

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

For EAL HU4.1 the intent of the 15-minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., smoldering waste paper basket). In addition to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation of a suppression system, etc.

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

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

CGS Basis Reference(s):

1. ABN-FIRE

2. FSAR Table 3.2-1 Equipment Classification

3. NEI 99-01 HU4

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

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• Category:

Subcategory:

Initiating Condition:

H - Hazards and Other Conditions Affecting Plant Safety 4- Fire 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 5 area AND The existence of a FIRE is not verified within 30 min. of alarm receipt (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Table 5 Safe Shutdown Areas

• Vital portions of the Rad Waste/Control Building:

467' elevation vital island 487' elevation cable spreading room Main Control Room and vertical cable chase 525' elevation HVAC area

• Reactor Building

• Vital portions of the Turbine Building DEH pressure switches RPS switches on turbine throttle valves Main steam line radiation monitors Turbine Building ventilation radiation monitors Main steam line piping up to MS-V-146 and the first stop valves

• Standby Service Water Pump Houses

• Diesel Generator Building Mode Applicability:

1 2 3 4 s I def I

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

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• Basis:

The 30 minute requirement begins upon receipt of a single valid fire detection system alarm. The alarm is to be validated using available Control Room indications or alarms to prove that it is not spurious, or by reports from the field. Actual field reports must be made within the 30 minute time limit or a classification must be made. If a fire is verified to be occurring by field report, classification shall be made based on EAL HU4.1.

A single point fire alarm, with no other indications of a fire, may be more indicative of an instrumentation issue rather than a fire in the plant.

The concept of this EAL is that there is 30 minutes to determine if a fire exists when only one fire alarm is received.

The Table 5 Safe Shutdown Areas include those structures/areas that contain any Class 1, 2 or 3 SSC.

Table 5 includes those structures containing functions and systems required to achieve and maintain cold shutdown (including all auxiliary equipment such as AC/DC power, cooling water and instrumentation) (ref. 1).

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

This EAL addresses receipt of a single fire alarm, and the existence of a FIRE is not verified (i.e.,

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

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

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

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

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

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

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

Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: NIA

Title:

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

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

CGS Basis Reference(s):

1. FSAR Table 3.2-1 Equipment Classification

2. NEI 99-01 HU4

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A IUse Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 114of187

• Category:

Subcategory:

Initiating Condition:

H - Hazards and Other Conditions Affecting Plant Safety 4- Fire FIRE potentially degrading the level of safety of the plant EAL:

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

OR (2) A FIRE within the ISFSI or plant PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Mode Applicability:

I 1 I 2 I 3 I 4 I s I def I Basis:

These thresholds reflect the potential issues that can arise from a fire in other areas of the plant for

• greater than one-hour or a fire requiring offsite fire department to actively support firefighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish.

Basis:

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

Threshold #1 In addition to a FIRE addressed by EAL HU4.1 or HU4.2, a FIRE within the plant PROTECTED AREA not extinguished within 60-minutes may also potentially degrade the level of plant safety. This basis extends to a FIRE occurring within the PROTECTED AREA of an ISFSI located outside the plant PROTECTED AREA.

Threshold #2 If a FIRE within the plant or ISFSI PROTECTED AREA is of sufficient size to require a response by an offsite firefighting agency (e.g., a local town Fire Department), then the level of plant safety is potentially degraded. The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish. Declaration is not necessary if the agency resources are placed on stand-by, or supporting post-extinguishment recovery or investigation actions.

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

CGS Basis Reference(s):

• 1. NEI 99-01 HU4 Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 115of187

• Category:

Subcategory:

Initiating Condition:

H - Hazards and Other Conditions Affecting Plant Safety 5 - Hazardous Gases Gaseous release IMPEDING access to equipment necessary for normal plant operations, cooldown or shutdown EAL:

HA5.1 Alert Release of a toxic, corrosive, asphyxiant or flammable gas into any Table 9 rooms or areas AND Entry into the room or area is prohibited or IMPEDED (Note 5)

Note 5: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted Table 9 Safe Operation & Shutdown Areas Room/Area Mode Applicability RW 467' Radwaste Control Room (RHR flush to RW tanks) 3

• RW 467' Vital Island (RHR-V-9 disconnect)

RB 422' B RHR Pump Rm (local pump temperatures)

RB 454' B RHR Pump Rm (operate RHR-V-85B) 3 3

3 Mode Applicability:

I 1 I 2 I 3 4 s I def I Basis:

The list of plant rooms or areas with entry-related mode applicability identified specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations) are not included. In addition, the list specifies the plant mode(s) during which entry would be required for each room or area (ref. 1).

This IC addresses an event involving a release of a hazardous gas that precludes or impedes access to equipment necessary to maintain normal plant operation, or required for a normal plant cooldown and shutdown. This condition represents an actual or potential substantial degradation of the level of safety of the plant.

An Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally

• required during the plant operating mode in effect at the time of the gaseous release. The emergency classification is not contingent upon whether entry is actually necessary at the time of the release.

Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 116of187

• Evaluation of the IC and EAL do not require atmospheric sampling; it only requires the Emergency Director's judgment that the gas concentration in the affected room/area is sufficient to preclude or significantly impede procedurally required access. This judgment may be based on a variety of factors including an existing job hazard analysis, report of ill effects on personnel, advice from a subject matter expert or operating experience with the same or similar hazards. Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).

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

• The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the gaseous release). For example, the plant is in Mode 1 when the gaseous release occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4.

• The gas release is a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., fire suppression system testing).

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

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

An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels. Most

• commonly, asphyxiants work by merely displacing air in an enclosed environment. This reduces the concentration of oxygen below the normal level of around 19%, which can lead to breathing difficulties, unconsciousness or even death.

This EAL does not apply to firefighting activities that automatically or manually activate a fire suppression system in an area, or to intentional inerting of containment.

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

CGS Basis Reference(s):

1. Attachment 7.4 Safe Operation & Shutdown Areas Table 9 Bases

2. NEI 99-01 HAS

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A IUse Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 117of187

• Category:

Subcategory:

Initiating Condition:

H - Hazards and Other Conditions Affecting Plant Safety 6 - Control Room Evacuation Inability to control a key safety function from outside the Control Room EAL:

HS6.1 Site Area Emergency An event has resulted in plant control being transferred from the Control Room to the Remote Shutdown Panel or Alternate Remote Shutdown Panel AND Control of any of the following key safety functions is not reestablished within 15 min. (Note 1):

• Reactivity (Modes 1 and 2 only)

• RPV water level

• RCS heat removal Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Mode Applicability:

• 1 Basis:

2 3 4 5 The Shift Manager determines if the Control Room is inoperable and requires evacuation. This determination can depend on a number of factors, including Control Room habitability, loss of safe shutdown control circuity, or a Security event (ref. 1).

For the purpose of this EAL the 15 minute classification clock starts when the last licensed operator leaves the Control Room.

This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations, and the control of a key safety function cannot be reestablished in a timely manner. The failure to gain control of a key safety function following a transfer of plant control to alternate locations is a precursor to a challenge to one or more fission product barriers within a relatively short period of time.

The determination of whether or not "control" is established at the remote safe shutdown location(s) is based on Emergency Director judgment. The Emergency Director is expected to make a reasonable, informed judgment within 15 minutes whether or not the operating staff has control of key safety functions from the remote safe shutdown location(s).

Escalation of the emergency classification level would be via IC FG1 or CG1.

CGS Basis Reference(s):

1. ABN-CR-EVAC Control Room evacuation and Remote Cooldown

2. NEI 99-01 HS6

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032

---

'--------------1 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 118of187

• Category:

Subcategory:

Initiating Condition:

H - Hazards and Other Conditions Affecting* Plant Safety 6 - Control Room Evacuation Control Room evacuation resulting in transfer of plant control to alternate locations EAL:

HAG.1 Alert An event has resulted in plant control being transferred from the Control Room to the Remote Shutdown Panel or Alternate Remote Shutdown Panel Mode Applicability:

I 1 I 2 I 3 4 s I def 1 Basis:

The Shift Manager (SM) determines if the Control Room is inoperable and requires evacuation. This determination can depend on a number of factors, including Control Room habitability, loss of safe shutdown control circuity, or a Security event (ref. 1). For the purpose of this EAL the 15 minute classification clock starts when the last licensed operator leaves the Control Room.

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

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

Following a Control Room evacuation, control of the plant will be transferred to alternate shutdown locations. The necessity to control a plant shutdown from outside the Control Room, in addition to responding to the event that required the evacuation of the Control Room, will present challenges to plant operators and other on-shift personnel. Activation of the ERO and emergency response facilities will assist in responding to these challenges.

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

CGS Basis Reference(s):

1. ABN-CR-EVAC Control Room Evacuation and Remote Cooldown

2. NEI 99-01 HA6

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 119 of 187

• Category:

Subcategory:

Initiating Condition:

H - Hazards and Other Conditions Affecting Plant Safety 7 - Emergency Director Judgment Other conditions exist which in the judgment of the Emergency Director warrant declaration of a General Emergency EAL:

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

Mode Applicability:

I 1 I 2 I 3 I 4 I s I def I Basis:

The Emergency Director is the designated onsite individual having the responsibility and authority for implementing the Emergency Response Plan. The Shift Manager(SM) initially acts in the capacity of the Emergency Director and takes actions as outlined in the Emergency Plan implementing procedures. If

• required by the emergency classification or if deemed appropriate by the Emergency Director, 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 Director to fall under the emergency classification level description for a General Emergency.

CGS Basis Reference{s):

1. NEI 99-01 HG7

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 120 of 187 Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Director Judgment Initiating Condition: Other conditions existing which in the judgment of the Emergency Director warrant declaration of a Site Area Emergency EAL:

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

Mode Applicability:

I 1 I 2 I 3 4 s I def I Basis:

The Emergency Director is the designated onsite individual having the responsibility and authority for implementing the Emergency Response Plan. The Shift Manager (SM) initially acts in the capacity of the Emergency Director and takes actions as outlined in the Emergency Plan implementing procedures.

If required by the emergency classification or if deemed appropriate by the Emergency Director, 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 Director to fall under the emergency classification level description for a Site Area Emergency.

CGS Basis Reference(s):

1. NEI 99-01 HS7

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 121 of 187

• Category:

Subcategory:

Initiating Condition:

H - Hazards and Other Conditions Affecting Plant Safety 7 - Emergency Director Judgment Other conditions exist which in the judgment of the Emergency Director warrant declaration of an Alert EAL:

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

Mode Applicability:

I 1 I 2 I3 4 s I def I Basis:

The Emergency Director is the designated onsite individual having the responsibility and authority for implementing the Emergency Response Plan. The Shift Manager (SM) initially acts in the capacity of the Emergency Director and takes actions as outlined in the Emergency Plan implementing procedures.

• If required by the emergency classification or if deemed appropriate by the Emergency Director, 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 Director to fall under the emergency classification level description for an Alert.

CGS Basis Reference(s):

1. NEI 99-01 HA7

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 122 of 187

• Category:

Subcategory:

Initiating Condition:

H - Hazards and Other Conditions Affecting Plant Safety 7 - Emergency Director Judgment Other conditions existing which in the judgment of the Emergency Director warrant declaration of a UE EAL:

HU7.1 Unusual Event Other conditions exist which, in the judgment of the Emergency Director, 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:

I 1 I 2 I 3 4 s I def I Basis:

The Emergency Director is the designated onsite individual having the responsibility and authority for implementing the Emergency Response Plan. The Shift Manager (SM) initially acts in the capacity of the Emergency Director and takes actions as outlined in the Emergency Plan implementing procedures.

• If required by the emergency classification or if deemed appropriate by the Emergency Director, 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 Director to fall under the emergency classification level description for an Unusual Event.

CGS Basis Reference(s):

1. NEI 99-01 HU7

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 123 of 187

• Category M - System Malfunction EAL Group: Hot Conditions (RCS temperature GT 200°F);

EALs in this category are applicable only in one or more hot operating modes.

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

The events of this category pertain to the following subcategories:

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

2. Loss of vital DC Power .

Loss of vital electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of power to or degraded voltage on the vital 125 VDC buses.

3. Loss of Control Room Indications

• Certain events that degrade plant operator ability to effectively assess plant conditions within the plant warrant emergency classification. Losses of indicators are in this subcategory.

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

clad failures) is indicative of fuel failures and is covered under the Fission Product Barrier Degradation category. However, lesser amounts of clad damage may result in coolant activity exceeding Technical Specification limits. These fission products will be circulated with the reactor coolant and can be detected by coolant sampling.

5. RCS Leakage The reactor pressure vessel provides a volume for the coolant that covers the reactor core. The reactor pressure vessel and associated pressure piping (reactor coolant system) together provide a barrier to limit the release of radioactive material should the reactor fuel clad integrity fail. Excessive RCS leakage greater than Technical Specification limits indicates potential pipe cracks that may propagate to an extent threatening fuel clad, RCS and Containment integrity.

6. RPS Failure This subcategory includes events related to failure of the Reactor Protection System (RPS) to initiate and complete reactor scrams. In the plant licensing basis, postulated failures of the RPS to complete a reactor scram comprise a specific set of analyzed events referred to as Anticipated Transient Without Scram (ATWS) events. For EAL classification, however, ATWS is intended to

• mean any trip failure event that does not achieve reactor shutdown. If RPS actuation fails to assure reactor shutdown, positive control of reactivity is at risk and could cause a threat to fuel clad, RCS and Containment integrity.

Attachment 6.1, EAL Technical Bases

Number: 13.1.1A \Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 124 of 187

• 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. 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 .

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: NIA

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 125 of 187

• Category:

Subcategory:

Initiating Condition:

M -System Malfunction 1 - Loss of Emergency AC Power Prolonged loss of§.!! offsite and all onsite AC power to emergency buses EAL:

MG1.1 General Emergency Loss of§.!! offsite AND all onsite AC power capability to emergency buses SM-7 and SM-8 AND EITHER:

Restoration of emergency bus SM-7 or SM-8 in LT 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not likely (Note 1)

OR RPV level cannot be restored.and maintained GT-186 in.

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

1 2 3 Basis:

Credit may be taken in this EAL for DG 3 crosstie capability provided a reasonable expectation exists that AC power can be restored to either SM-7 or SM-8 from DG3 and SM-4 within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. (ref. 1).

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

Indication of continuing core cooling degradation is manifested by the inability to restore and maintain RPV water level above the Minimum Steam Cooling Reactor Water Level (-186 in.) (ref. 3).

Core submergence is the most desirable means of core cooling, however when RPV level is below TAF, the uncovered portion of the core can be cooled by less reliable means (i.e., steam cooling or spray cooling).

This IC addresses a prolonged loss of all power sources to AC emergency buses. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A prolonged loss of these buses will lead to a loss of one or more fission product barriers. In addition, fission product barrier monitoring capabilities may be degraded under these conditions.

The EAL should require declaration of a General Emergency prior to meeting the thresholds for IC FG1.

This will allow additional time for implementation of offsite protective actions.

Escalation of the emergency classification from Site Area Emergency will occur if it is projected that power cannot be restored to at least one AC emergency bus by the end of the analyzed station blackout coping period. Beyond this time, plant responses and event trajectory are subject to greater uncertainty, and there is an increased likelihood of challenges to multiple fission product barriers.

The estimate for restoring at least one essential bus should be based on a realistic appraisal of the situation. Mitigation actions with a low probability of success should not be used as a basis for delaying a classification upgrade. The goal is to maximize the time available to prepare for, and implement, protective actions for the public.

The EAL will also require a General Emergency declaration if the loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core.

Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor.Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 126 of 187

• CGS Basis Reference(s):

1.

2.

FSAR Section 8.2 PPM 5.6.1 Station Blackout (SBO)

3. PPM 5.1.1 RPV Control

4. NEI 99-01 SG1

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 127 of 187

• Category:

Subcategory:

Initiating Condition:

M - System Malfunction 1 - Loss of Emergency AC Power Loss of fill offsite power and fill onsite AC power to emergency buses for 15 minutes or longer EAL:

MS1 .1 Site Area Emergency Loss of fill offsite and all onsite AC power capability to emergency buses SM-7 and SM-8 for GE 15 min. (Note 1)

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

1 2 3 Basis:

This hot condition EAL is equivalent to the cold condition EAL CA2.1.

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.

In addition, fission product barrier monitoring capabilities may be degraded under these conditions. This IC represents a condition that involves actual or likely major failures of plant functions needed for the protection of the public.

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

Escalation of the emergency classification level would be via ICs RG1, FG1 or MG1.

CGS Basis Reference(s):

1. PPM 5.6.1 Station Blackout (SBO)

2. NEI 99-01 SS1

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: 032

---

1 Minor Rev: NIA

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 128 of 187

• Category:

Subcategory:

Initiating Condition:

M - System Malfunction 1 - Loss of Emergency AC Power Loss of fill but one AC power source to emergency buses for 15 minutes or longer EAL:

MA1.1 Alert AC power capability, Table 2, to emergency buses SM-7 and SM-8 reduced to a single power source for GE 15 min. (Note 1)

AND Any additional single power source failure will result in a loss of all AC power to SAFETY SYSTEMS Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Table 2 AC Power Sources Offsite Startup Transformer TR-S Backup Transformer TR-B Backfeed 500 KV power through Main Transformers (if already aligned in modes 4, 5, def only)

Onsite

• DG1

• DG2

• Main Generator via TR-N 1/N2 Mode Applicability:

1 2 3 Basis:

Table 2 provides the list of AC power sources available to power emergency buses (ref. 1, 2).

Station Startup 230KV power comes from the Ashe substation through Startup transformer TR-S. The Startup transformer usually supplies station auxiliary loads when the main generator is not available.

Station Backup 115KV power from the Benton Substation feeder can be supplied to emergency buses SM-7 and SM-8 (ref. 3, 4) .

Credit is not taken in this EAL for SM-4/DG3 crosstie capability because establishing the crosstie to SM-7 or SM-8 is assumed to require more than 15 minutes (5). SM-4 is not a site specific emergency AC buss source since SM-4 does not provide core cooling or containment cooling.

Attachment 6.1, EAL Technical Bases

Number: 13.1.1A luseC~egory:REFERENCE Major Rev: 032 Minor Rev: NIA

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 129 of 187

• The second threshold statement in this EAL does not describe a separate condition, it is clarifying the first threshold statement.

This hot condition EAL is equivalent to the cold condition EAL CU2.1.

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

This IC provides an escalation path from IC MU1.

An "AC power source" is a source recognized in AOPs, 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 back-fed from an offsite power source.

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

Escalation of the emergency classification level would be via IC MS1 .

• CGS Basis Reference(s):

1.

2.

3.

FSAR Figure 8.1-2.1 Main One-Line Diagram - Main Buses FSAR Figure 8.1-2.2 Main One-Line Diagram - Emergency Buses.

FSAR Section 8.2

4. 01-53 Offsite Power

5. FSAR Section 8.3

6. ABN-ELEC-LOOP

7. NEI 99-01 SA1

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A juseC~ego~:REFERENCE Major Rev: 032 Minor Rev: NIA

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 130 of 187

• Category:

Subcategory:

Initiating Condition:

M - System Malfunction 1 - Loss of Emergency AC Power Loss of all offsite AC power capability to emergency buses for 15 minutes or longer EAL:

MU1.1 Unusual Event Loss of fill offsite AC power capability, Table 2, to emergency buses SM-7 and SM-8 for GE 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Table 2 AC Power Sources Offsite

• Startup Transformer TR-S

• Backup Transformer TR-B

• Backfeed 500 KV power through Main Transformers (if already aligned in modes 4, 5, def only)

DG1 On site

• DG2

• Main Generator via TR-N 1/N2 Mode Applicability:

1 2 3 Basis:

Table 2 provides the list of AC power sources available to power emergency buses (ref. 1, 2).

Station Startup 230KV power comes from the Ashe substation through Startup transformer TR-S. The Startup transformer usually supplies station auxilia~ loads when the main generator is not available.

Station Backup 115KV power from the Benton Substation feeder can be supplied to emergency buses SM-7 and SM-8. (ref. 3, 4)

Credit is not taken in this EAL for SM-4/DG3 crosstie capability because establishing the crosstie to SM-7 or SM-8 is assumed to require more than 15 minutes (5). SM-4 is not a site specific emergency AC buss source since SM-4 does not provide core cooling or containment cooling .

This IC addresses a prolonged loss of offsite power. The loss of offsite power sources renders the plant more vulnerable to a complete loss of power to AC emergency buses. This condition represents a potential reduction in the level of safety of the plant.

Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 131 of 187

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

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

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

CGS Basis Reference(s):

1. FSAR Figure 8.1-2.1 Main One-Line Diagram - Main Buses

2. FSAR Figure 8.1-2.2 Main One-Line Diagram - Emergency Buses

3. FSAR Section 8.2

4. Ol-53 Offsite Power

5. FSAR Section 8.3

6. NEI 99-01 SU1

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: NIA

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 132 of 187

• Category:

Subcategory:

Initiating Condition:

M -System Malfunction 1 - Loss of Essential AC Power Loss of .ill! emergency AC and vital DC power sources for 15 minutes or longer EAL:

MG1 .2 General Emergency Loss of .ill! offsite AND .ill! onsite AC power capability to emergency buses SM-7 and SM-8 for GE 15 min. (Note 1)

AND Indicated voltage is LT 108 VDC on both 125 VDC buses DP-S1-1 and DP-S1-2 for GE 15 min.

(Note 1)

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

1 2 3 Basis:

This EAL addresses operating experience from the March 2011 accident at Fukushima Daiichi.

The 125 VDC Class 1E DC power system consists of three electrically independent and separate distribution systems (S1-1, S1-2, and S1-HPCS) (ref. 2). S1-HPCS is not included in this EAL. Each DC distribution system has a battery and a battery charger that are normally connected to the bus such that these two sources of power are operating in parallel. The charger is normally supplying system electrical loads with the battery on a float charge. Each battery has the necessary amp-hour discharge capacity to sustain system loads for a minimum of two hours. This capacity is specifically for a loss of power to the charger coincident with a design basis accident. The batteries have capacity to carry design load at 60°F without decreasing battery voltage below 1.81 volts/cell (or 108 VDC, ref. 3) with loss of output from the battery chargers during the specified period. Battery capacity is sufficient to provide starting currents while operating at full load. (ref. 1, 3).

This IC addresses a concurrent and prolonged loss of both emergency AC and Vital DC power. A loss of all emergency AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A loss of vital DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both emergency AC and vital DC power will lead to multiple challenges to fission product barriers.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. The 15-minute emergency declaration clock begins at the point when both EAL thresholds are met.

CGS Basis Reference(s):

1. FSAR Section 8

2. E505 DC One Line Diagram

3. Calculation No. 2.05.01 Battery Sizing, Voltage Drop, and Charger Studies for Div. 1 & 2 Systems

• 4.

5.

PPM 5.6.1 Station Blackout (SBO)

NEI 99-01 SGS Attachment 6.1, EAL Technical Bases

Number: 13,1.1A IUse Category: REFERENCE Major Rev: 032 Minor Rev: NIA

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 133 of 187

• Category:

Subcategory:

Initiating Condition:

M - System Malfunction 2 - Loss of Vital DC Power Loss of all vital DC power for 15 minutes or longer EAL:

MS2.1 Site Area Emergency Indicated voltage is LT 108 VDC on both 125 VDC buses DP-S1-1 and DP-S1-2 for GE 15 min.

(Note 1)

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

2 3 Basis:

The 125 VDC Class 1E DC power system (ref. 1) consists of three electrically independent and separate distribution systems (S1-1, S1-2, and S1-HPCS) (ref. 2). S1-HPCS is not included in this EAL.

Each DC distribution system has a battery and a battery charger that are normally connected to the bus such that these two sources of power are operating in parallel. The charger is normally supplying system electrical loads with the battery on a float charge. Each battery has the necessary amp-hour

• discharge capacity to sustain system loads for a minimum of two hours. This capacity is specifically for a loss of power to the charger coincident with a design basis accident. The batteries have capacity to carry design load at 60°F without decreasing battery voltage below 1.81 volts/cell (or 108 VDC, ref. 2) with loss of output from the battery chargers during the specified period. Battery capacity is sufficient to provide starting currents while operating at full load. (ref. 3).

This EAL is the hot condition equivalent of the cold condition loss of DC power EAL CU4.1.

This IC addresses a loss of vital DC power which compromises the ability to monitor and control SAFETY SYSTEMS. In modes above Cold Shutdown, this condition involves a major failure of plant functions needed for the protection of the public.

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

Escalation of the emergency classification level would be via ICs RG1, FG1 or MG1.

CGS Basis Reference(s):

1. E505 DC One Line Diagram

2. Calculation No. 2.05.01 Battery Sizing, Voltage Drop, and Charger Studies for Div. 1 & 2 Systems

3. FSAR Section 8.3.2

4. NEI 99-01 SSS

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 134 of 187

• Category:

Subcategory:

Initiating Condition:

M - System Malfunction 3 - Loss of Control Room Indications UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress EAL:

MA3.1 Alert An UNPLANNED event results in the inability to monitor one or more Table 1O parameters from within the Control Room for GE 15 min. (Note 1)

AND Any Table 11 transient event in progress Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Table 10 Safety System Pa1a11111:rn:. . .,.

• Reactor power

• RPV level RPV pressure Primary containment pressure Wetwell level Wetwell temperature Table 11 Significant Transients

• Reactor scram

• Runback GT 25% thermal reactor power

• Electrical load rejection GT 25% full electrical load

• ECCS injection

• Thermal power oscillations GT 10%

Mode Applicability:

1 2 3

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: 032 1----------------------'----------------1 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 135 of 187

• Basis:

SAFETY SYSTEM parameters listed in Table 10 are monitored in the Control Room through a combination of hard control panel indicators as well as computer based information systems. The Plant Computers and Graphic Display System provide redundant parameter indications (ref. 1-4).

Significant transients are listed in Table 11 and include response to automatic or manually initiated functions such as scrams, run backs involving greater than 25% thermal power change, electrical load rejections of greater than 25% full electrical load, ECCS injections, or thermal power oscillations of 10%

or greater.

This IC addresses the difficulty associated with monitoring rapidly changing plant conditions during a transient without the ability to obtain SAFETY SYSTEM parameters from within the Control Room.

During this condition, the margin to a potential fission product barrier challenge is reduced. It thus represents a potential substantial degradation in the level of safety of the plant.

As used in this EAL, an "inability to monitor means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room.

An event involving a loss. of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRG event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal

• operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.

This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, RPV level and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for RPV water level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well.

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

Escalation of the emergency classification level would be via ICs FS1 or IC RS1 CGS Basis Reference{s):

1. FSAR Section 7.7.1

2. ABN-COMPUTER

3. SOP-COMPUTER-OPS Plant Process Computer (PPC)

4. SOP-GOS-OPS Graphics Display System

5. NEI 99-01 SA2

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 136 of 187

• Category:

Subcategory:

Initiating Condition:

M - System Malfunction 3 - Loss of Control Room Indications UNPLANNED loss of Control Room indications for 15 minutes or longer EAL:

MU3.1 Unusual Event An UNPLANNED event results in the inability to monitor one or more Table 10 parameters from within the Control Room for GE 15 min. (Note 1)

Note 1: .The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Table 10 Safety System Parameters

• Reactor power

• RPV level

• RPV pressure

• Primary containment pressure

• Wetwell level

• Wetwell temperature Mode Applicability:

2 3 Basis:

SAFETY SYSTEM parameters listed in Table 10 are monitored in the Control Room through a combination of hard control panel indicators as well as computer based information systems. The Plant Computers and Graphic Display System provide redundant parameter indications (ref. 1-4).

This IC addresses the difficulty associated with monitoring normal plant conditions without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition is a precursor to a more significant event and represents a potential degradation in the level of safety of the plant.

As used in this EAL, an "inability to monitor means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room.

An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.

Attachment 6.1, EAL Technical Bases

Number: 13.1.1A IUse Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 137 of 187

• This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, RPV level and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for RPV water level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well.

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

Escalation of the emergency classification level would be via IC MA3.

CGS Basis Reference(s):

1. FSAR Section 7.7.1

2. ABN-COMPUTER

3. SOP-COMPUTER-OPS Plant Process Computer (PPC)

4. SOP-GOS-OPS Graphics Display System

5. NEI 99-01 SU2

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A ,I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 138 of 187

• Category:

Subcategory:

Initiating Condition:

M - System Malfunction 4 - RCS Activity Reactor coolant activity greater than Technical Specification allowable limits EAL:

MU4.1 Unusual Event SJAE CONDSR OUTLET RAD HI-HI alarm (P602)

Mode Applicability:

1 2 3 Basis:

The main condenser offgas gross gamma activity rate is an initial condition of the Main Condenser Offgas System failure event. The gross gamma activity rate is controlled to ensure that during the event, the calculated offsite doses will be well within the limits of 10 CFR 50.67 (ref. 1).

SJAE CONDSR OUTLET RAD HI HI monitor and alarm, OG-RIS-612 (GE 2300 mR/hr), senses the offgas effluent and, therefore, may be one of the first indicators of degrading fuel conditions. The alarm is confirmed by verification of greater than the current alarm setpoint on Recorder OG-RIS-612 on Panel P604 or high offgas pre-treatment air activity (determined by sample results) greater than limits specified in Technical Specification .

If OG-RIS-612 and OG-RR-604 are reading off-scale high, the alarm may be confirmed by a significant increase in the Main Steam Line radiation monitors (MS-RIS-610A-D) on H13-P606 and H13-P633 (ref. 2).

This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant.

Escalation of the emergency ciassification level would be via ICs FA1 or the Recognition Category R ICs.

CGS Basis Reference(s):

1. Technical Specifications 3. 7.5

2. PPM 4.602.A5 ANNUNCIATOR RESPONSE, P602 ANNUNCIATOR A5 3-3

3. NEI 99-01 SU3

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 139 of 187

• Category:

Subcategory:

Initiating Condition:

M - System Malfunction 4 - RCS Activity Reactor coolant activity greater than Technical Specification allowable limits EAL:

MU4.2 Unusual Event Coolant activity GT 0.2 µCi/gm dose equivalent 1-131 Mode Applicability:

1 2 3 Basis:

The limits on the specific activity of the primary coolant ensure that the 2-hour thyroid and whole body doses at the SITE BOUNDARY, resulting from an Main Steam Line Break (MSLB) outside containment during steady state operation, will not exceed the dose guidelines of 10 CFR 50.67.

This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant.

Escalation of the emergency classification level would be via ICs FA 1 or the Recognition Category R

• ICs .

CGS Basis Reference(s):

1.

2.

Technical Specifications 3.4.8 NEI 99-01 SU3

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A IUse Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 140 of 187

• Category:

Subcategory:

Initiating Condition:

M - System Malfunction 5 - RCS Leakage RCS leakage for 15 minutes or longer EAL:

MU5.1 Unusual Event (1) RCS unidentified or pressure boundary leakage GE 10 gpm for GE 15 min.

OR (2) RCS identified leakage GT 25 gpm for GE 15 min.

OR (3) Leakage from the RCS to a location outside containment GT 25 gpm for GE 15 min.

(Note 1)

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

1 2 3

• Basis:

Pressure boundary leakage is defined to be leakage through a non-isolable fault in a RCS component body, pipe wall, or vessel wall.

This EAL does not apply to relief valves performing their normal design function.

Unidentified leakage is defined to be all leakage into the drywell that is not identified leakage.

Identified leakage is defined to be leakage into the drywell such as that from pump seals or valve packing, that is captured and conducted to a sump; or leakage into the drywell atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary leakage. (ref. 1)

The Leak Detection (LD) system is designed to monitor leakage from the reactor coolant pressure boundary and to isolate this leakage when limits are exceeded. Systems, or parts of systems, that are in direct communication with the reactor vessel (form part of the primary coolant pressure boundary) are provided with leakage detection systems. (ref. 2-8)

Drain flow from the drywell equipment and floor drain sumps is monitored and recorded (EDR-FRS-623) on P632. The flow rate for unidentified leakage in the EAL is equal to the full scale reading on EDR-FRS-623 Pen 1.

Leakage not explicitly identified by installed instrumentation requires analysis and declaration clock starts at completion of analysis. This includes use of alternate means.

As an alternate means, leaks within the drywell are detected by monitoring for abnormally high:

• Pressure or temperature inside the drywell

• Fill up rates of equipment and floor drain sumps

• Containment leak detection rad monitors (CMS-SR-20/21)

Attachment 6.1, EAL Technical Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: 032 1 - - - - - - - - - - - - - - - - - - - - - - ' - - - - - - - - - - - - 1 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 141of187

• Outside Containment leakage may require analysis to quantify leak rate GT 25 gpm and declaration clock starts at completion of analysis.

Examples of outside Containment leakage include:

• GT 25 gpm RWCU differential flow (RWCU-Fl-620) due to RCS leakage

• Instrument line break in the RX building with failure to isolate

• Rx Building sump fill timers due to RCS leakage RFW and RCC are not considered part of RCS leakage for this EAL.

For classification under this EAL, RCS leakage includes a broken SRV tailpipe that is discharging into the drywell or wetwell airspace. Once the SRV is closed, however, this RCS leakage path is considered isolated.

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.

Threshold #1 and threshold #2 are focused on a loss of mass from the RCS due to "unidentified leakage", "pressure boundary leakage" or "identified leakage" (as these leakage types are defined in the plant Technical Specifications). Threshold #3 addresses an RCS mass loss caused by an UNISOLABLE leak through an interfacing system. These conditions thus apply to leakage into the containment, or a location outside of containment.

The leak rate values for each threshold 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). Threshold #1 uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage.

A stuck-open Safety Relief Valve (SRV) or SRV leakage is not considered either identified or unidentified leakage by Technical Specifications and, therefore, is not applicable to this EAL.

The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible.

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

CGS Basis Reference(s):

1. Technical Specification 1.1

2. Technical Specifications 3.4.7

3. FSAR Section 5.2.5

4. FSAR Section 7.6.1

5. ABN-LEAKAGE Reactor Coolant Leakage

6. SOP-EDR-OPS Equipment Drain System Operation

7. SOP-FDR-OPS Floor Drain System Operation

8. PPM 10.27.35 Leakage Surveillance And Prevention Program

9. NEI 99-01 SU4

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 142 of 187

• Category:

Subcategory:

Initiating Condition:

M - System Malfunction 6 - RPS Failure Inability to shut down the reactor causing a challenge to RPV water level or RCS heat removal EAL:

MS6.1 Site Area Emerge,ncy An automatic OR manual scram fails to shut down the reactor AND All actions to shut down the reactor are not successful as indicated by reactor power GT 5%

AND EITHER:

RPV level cannot be restored and maintained above -186 in. or cannot be determined OR WW temperature and RPV pressure cannot be maintained below the HCTL Mode Applicability:

1 2

• Basis:

This EAL addresses the following:

• Any automatic reactor scram signal followed by a manual scram that fails to shut down the reactor to an extent the reactor is producing energy in excess of the heat load for which the safety systems were designed (EAL MA6.1 ), and

• Indications that either core cooling is extremely challenged or heat removal is extremely challenged.

Reactor shutdown achieved by use of control rod insertion methods in PPM 5.5.11 is also credited as a successful manual scram provided reactor power can be reduced below the APRM downscale trip setpoint before indications of an extreme challenge to either core cooling or heat removal exist. (ref. 1)

The APRM downscale trip setpoint (5%) is a minimum reading on the power range scale that indicates 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 the APRM downscale trip setpoint, plant response will be similar to that observed during a normal shutdown. Nuclear instrumentation (APRM) indications or other reactor parameters (steam flow, RPV pressure, wetwell temperature trend) can be used to determine if reactor power is greater than 5% power (ref. 2).

The combination of failure of both front line and backup protection systems to function in response to a plant transient, along with the continued production of heat, poses a direct threat to the Fuel Clad and RCS barriers .

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 143of187

• Indication that core cooling is extremely challenged is manifested by inability to restore and maintain RPV water level above the Minimum Steam Cooling RPV Water Level (MSCRWL) (ref. 3). The MSCRWL is the lowest RPV level at which the covered portion of the reactor core will generate sufficient steam to prevent any clad temperature in the uncovered part of the core from exceeding 1500°F. This water level is utilized in the EOPs to preclude fuel damage when RPV level is below the top of active fuel. RPV level below the MSCRWL for an extended period of time without satisfactory core spray cooling could be a precursor of a core melt sequence.

The Heat Capacity Temperature Limit (HCTL) is the highest suppression pool water temperature from which Emergency RPV Depressurization will not raise suppression pool temperature above the maximum design suppression pool temperature.

The HCTL is a function of RPV pressure and wetwell level. It is utilized to preclude failure of the containment and equipment in the containment necessary for the safe shutdown of the plant (ref. 4).

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, all subsequent operator actions to manually shutdown the reactor are unsuccessful, and continued power generation is challenging the capability to adequately remove heat from the core and/or the RCS. This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration of a Site Area Emergency.

In some instances, the emergency classification resulting from this IC/EAL may be higher than that resulting from an assessment of the plant responses and symptoms against the Recognition Category F ICs/EALs. This is appropriate in that the Recognition Category F ICs/EALs do not address the additional threat posed by a failure to shut down the reactor. The inclusion of this IC and .EAL ensures the timely declaration of a Site Area Emergency in response to prolonged failure to shutdown the reactor.

A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.

Escalation of the emergency classification level would be via IC RG1 or FG1.

CGS Basis Reference(s):

1. PPM 5.5.11 Alternate Control Rod Insertions

2. Technical Specifications Table 3.3.1.1-1

3. PPM 5.1.2 RPV Control - A TWS

4. PPM 5.2.1 Primary Containment Control,

5. NEI 99-01 SS5

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 144of187

• Category:

Subcategory:

Initiating Condition:

M - System Malfunction 6 - RPS Failure Automatic or manual scram fails to shut down tl:le reactor and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor EAL:

MA6.1 Alert An automatic OR manual scram fails to shut down the reactor AND Manual scram actions taken at the reactor control console (mode switch in shutdown, manual push buttons or ARI) are not successful in shutting down the reactor as indicated by reactor power GT 5% (Note 8)

Note 8: A manual scram action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies Mode Applicability:

1 2

• Basis:

This EAL addresses any automatic or manual reactor scram signal that fails to shut down the reactor followed by a subsequent manual scram 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.

For the purposes of emergency classification, successful manual scram actions are those which can be quickly performed from the reactor control console (i.e., mode switch in shutdown, manual push buttons or ARI). Reactor shutdown achieved by use of the alternate control rod insertion methods of PPM 5.5.11 does not constitute a successful manual scram (ref. 1).

The APRM downscale trip setpoint (5%) is a minimum reading on the power range scale that indicates 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 the APRM downscale trip setpoint, plant response will be similar to that observed during a normal shutdown. Nuclear instrumentation (APRM) indications or other reactor parameters (steam flow, BPV position or continuous SRV operation) can be used to determine if reactor power is greater than 5% power (ref. 2).

Escalation of this event is via EAL MS6.1.

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and subsequent operator manual actions taken at the reactor control consoles to shutdown the reactor are also unsuccessful. This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if the reactor is subsequently shutdown by an action taken away from the reactor control consoles since this event entails a significant failure of the RPS .

Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: NIA

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 145of187

• 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 scram). This action does not include manually driving in control rods or implementation of boron injection strategies. If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control console (e.g., locally opening breakers). Actions taken at backpanels or other locations within the control room, or any location outside the control room, are not considered to be "at the reactor control console".

Taking the reactor mode switch to shutdown is considered to be a manual scram action.

The plant response to the failure of an automatic or manual reactor scram will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If the failure to shut down the reactor is prolonged enough to cause a challenge to RPV water level or RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC MS6.

Depending upon plant responses and symptoms, escalation is also possible via IC FS1. Absent the plant conditions needed to meet either IC MS6 or FS1, an Alert declaration is appropriate for this event.

It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F ICs; however, this IC and EAL are included to ensure a timely emergency declaration.

A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.

• CGS Basis Reference(s}:

1.

2.

3.

PPM 5.5.11 Alternate Control Rod Insertions Technical Specifications Table 3.3.1.1-1 NEI 99-01 SAS

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A IUse Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 146of187

• Category:

Subcategory:

Initiating Condition:

M - System Malfunction 6 - RPS Failure Automatic or manual scram fails to shut down the reactor EAL:

MU6.1 Unusual Event An automatic OR manual scram did not shut down the reactor AND A subsequent automatic scram OR manual scram action taken at the reactor control console (mode switch in shutdown, manual push buttons or ARI) is successful in shutting down the reactor as indicated by reactor power LE 5% (APRM downscale) (Note 8)

Note 8: A manual scram action is ~ny operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies Mode Applicability:

1 2 Basis:

• This EAL addresses a failure of an automatic or manually initiated scram and a subsequent automatic or manual scram is successful in shutting down the reactor (reactor power LE 5%) (ref.1 ).

A successful scram has occurred when there is sufficient rod insertion from the trip of RPS to bring the reactor power to or below the APRM downscale trip setpoint of 5%. For the purposes of this EAL, a successful automatic initiation of ARI that reduces reactor power to or below 5% is a not a successful automatic scram. (ref. 2, 3, 4, 5)

For the purposes of emergency classification at the Unusual Event level, successful manual scram actions are those which can be quickly performed from the reactor control console (i.e., mode switch, manual scram pushbuttons, and manual ARI actuation). Reactor shutdown achieved by use of the alternate control rod insertion methods of PPM 5.5.11 does not constitute a successful manual scram (ref. 6).

Following any automatic RPS scram signal plant procedures prescribe insertion of redundant manual scram signals to back up the automatic RPS scram function and ensure reactor shutdown is achieved.

Even if the first subsequent manual scram signal inserts all control rods to the full-in position immediately after the initial failure of the automatic scram, the lowest level of classification that must be declared is an Unusual Event.

The operating mode change associated with movement of the Mode Switch, by itself, does not justify failure to declare an emergency for A TWS events.

If both subsequent automatic and subsequent manual reactor scram actions in the Control Room fail, the event escalates to an Alert under EAL MA6.1.

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram

• that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic scram 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.

Attachment 6.1, EAL Technical Bases

Number: 13.1.1A IUse Category: REFERENCE Major Rev: 032.

Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 147 of 187

• Following the failure on an automatic reactor scram, operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.

If an initial manual reactor scram is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram} using a different switch). Depending upon several factors, the initial or subsequent effort to manually scram the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor scram signal. If a subsequent manual or automatic scram is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.

A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram). This action does not include manually driving in control rods or implementation of boron injection strategies.

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

Taking the reactor mode switch to shutdown is considered to be a manual scram action.

The plant response to the failure of an automatic or manual reactor scram 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 MA6. Depending upon the plant response, escalation is also possible via IC FA1. Absent the plant conditions needed to meet either IC MA6 or FA 1, an unusual event declaration is appropriate for this event.

A reactor shutdown is determined in accordance with applicable emergency operating procedure criteria.

Should a reactor scram signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied.

• If the signal causes a plant transient that should have included an automatic reactor scram and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and should be evaluated.

• If the signal does not cause a plant transient and the scram failure is determined through other means (e.g., assessment of test results), then this IC and the EALs are not applicable and no classification is warranted.

CGS Basis Reference(s):

1. Technical Specifications Table 3.3.1.1-1

2. FSAR Section 7.2

3. FSAR Section 7.4

4. PPM 5.1.1 RPV Control

• 5.

6.

7.

PPM 5.1.2 RPV Control-ATWS PPM 5.5.11 Alternate Control Rod Insertions NEI 99-01 SU5 Attachment 6.1, EAL Technical Bases

Number: 13.1.1A \ Use Category: REFERENCE Major Rev: 032 Minor Rev: NIA

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 148 of 187

• Category:

Subcategory:

Initiating Condition:

M - System Malfunction 7 - Loss of Communications Loss of all onsite or offsite communications capabilities EAL:

MU7.1 Unusual Event (1) Loss of fill Table 4 onsite communication methods OR (2) Loss of all Table 4 ORO communication methods OR (3) Loss of all Table 4 NRC communication methods Table 4 Communication Methods System On site ORO NRC x

Plant Public Address (PA) System Plant Telephone System x x Plant Radio System Operations and x Security Channels Offsite calling capability from the Control x x Room via direct telephone Long distance calling capability on the x x commercial phone system Mode Applicability:

1 2 3 Basis:

Onsite and offsite (ORO and NRC) communications include one or more of the systems listed in Table 4 (ref. 1, 2).

Public Address (PA) System The public address system provides a way of contacting personnel in the various buildings of the plant and locations of the site that might be inaccessible using other means of communication. The building.:

wide alarm system alerts (via the public address system speakers) operating personnel to fire hazards

• and other trouble conditions for which plant management finds it necessary to alert plant personnel.

Attachment 6.1, EAL Technical Bases

Number: 13.1.1A IUse Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 149 of 187

• Plant Telephone System This system consists of interconnections to the public telephone network (and trunks to the PBX) with individual direct lines that provide inward and outward dialing access to most plant locations.

Plant Radio System Operations and Security Channels The radio communications system is used for communications with personnel involved in maintenance and security in and around the plant complex by means of hand-held portable radio units, mobile radio units, and paging receivers. The telephone link to BPA provides a direct communication link to the BPA Dittmer Control Center. The radio communications system provides a communications link for security and emergency communications to local law enforcement agencies and emergency control centers.

Offsite calling capability from the Control Room via direct telephone and fax lines This communications method includes following dedicated phone networks that are available for emergency communications in addition to the normal Energy Northwest phone network:

• Energy Northwest Emergency Center Network

• Response Agency Network

• NRC Emergency Notification System Various locations such as the Control Room, Technical Support Center, Emergency Operations Facility, Joint Information Center, Department of Energy-RL, Washington State Emergency Operations Center, Oregon State Emergency Coordination Center and the Benton and Franklin County Emergency Operations Centers have facsimile transceivers. The facsimile transceivers enable the transmission and receipt of printed material. The facsimile system which connects the Energy Northwest emergency centers with the county and state emergency centers uses dedicated phone lines.

Long distance calling capability on the commercial phone system The Energy Northwest Richland phone system is a computer based, software controlled telephone exchange (Computerized Branch Exchange). It is equipped with redundant computerized processor units and is served by an uninterruptible power supply. The direct-dial private telephone system provides communication between the Energy Northwest facilities. The phone system is arranged such that plant telephones can reach other Energy Northwest facilities by direct-dialing and without the need of an operator.

This. EAL is the hot condition equivalent of the cold condition EAL CU5.1.

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

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

Threshold #1 addresses a total loss of the communications methods used in support of routine plant operations.

Threshold #2 addresses a total loss of the communications methods used to notify all OROs of an emergency declaration. The OROs referred to here are Washington State, Benton County, Franklin County and DOE RL.

Threshold #3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A j Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 150 of 187 CGS Basis Reference(s):

1. Emergency Plan Section 6.6

2. FSAR Section 9.5.2

3. NEI 99-01 SU6

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A J Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 151of187

• Category:

Subcategory:

Initiating Condition:

M - System Malfunction 8 - Hazardous Event Affecting Safety Systems Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode EAL:

MA8.1 Alert The occurrence of any Table 8 hazardous event AND Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode AND EITHER:

Event damage has caused indications of degraded performance to a second train of a SAFETY SYSTEM needed for the current operating mode OR Event damage has resulted in VISIBLE DAMAGE to a second train of a SAFETY SYSTEM needed for the current operating mode (Notes 9, 10)

Note 9: If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then emergency classification is not warranted.

Note 10: If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted.

Table 8 Hazardous Events

• Seismic event

• Internal or external FLOODING event

• High winds

• Tornado strike

• FIRE

• EXPLOSION

• Volcanic ash fallout

• Other events with similar hazard characteristics as determined by the Shift Manager

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 152 of 187

• Mode Applicability:

1 Basis:

2 3 This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the current operating mode. In order to provide the appropriate context for consideration of an ALERT classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train, and there must be either indications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance, and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE such that the potential exists for performance issues. Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has indications of degraded performance; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events.

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

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

The significance of a seismic event is discussed under EAL HU2.1 (ref. 1).

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

Seismic Category I structures are analyzed to withstand a sustained, design wind velocity of at least 100 mph (ref. 3).

Areas containing functions and systems required for safe shutdown of the plant are identified by Fire Areas in the fire response procedure (ref. 4).

The potential for volcanic eruption exists in the Pacific Northwest. Heavy ash fall,. such as that experienced at certain locations following the eruption of Mt. St. Helens in 1980, could affect operation of plant equipment if precautionary measures are not taken. The design basis ash fall is projected for a twenty hour duration (ref. 5).

Table 5 provides a list of CGS safety system structures/areas (ref. 6). Table 8 provides a list of hazardous events.

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

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A luseC~egor~REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 153 of 187

• CGS Basis Reference(s):

1.

2.

FSAR Section 3. 7 Seismic Design FSAR Section 3.4.1 Flood Protection

3. CGS Calculation CALC CE-02-93-16 Evaluate PMR/BDC 98-0131-0A change from 5 min. to 15 min. averaging of 33 ft. elev. met twr. wind speeds for UE and Alert declarations

4. ABN-FIRE Attachment 13.2, Fire Areas

5. ABN-ASH Ash Fall

6. FSAR Table 3.2-1 Equipment Classification

7. NEI 99-01 SA9

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 154 of 187

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

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

At lower levels, abnormal radioactivity releases may be indicative of a failure of containment systems or precursors to more significant releases. At higher release rates, offsite radiological conditions may result which require offsite protective actions. Elevated area radiation levels in the plant may also be indicative of the failure of containment systems or preclude access to plant vital equipment necessary to ensure plant safety.

Events of this category pertain to the following subcategories:

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

2. Irradiated Fuel Event Conditions indicative of a loss of adequate shielding or damage to irradiated fuel may preclude

• access to vital plant areas or result in radiological releases that warrant emergency classification.

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

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 155 of 187

• Category:

Subcategory:

Initiating Condition:

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

RG1 .1 General Emergency (1) Reading on any Table 3 effluent radiation monitor GT column "GENERAL" for GE 15 min.

OR (2) Dose assessment using actual meteorology indicates doses GT 1,000 mrem TEDE or GT 5,000 mrem thyroid COE at or beyond the SITE BOUNDARY (Notes 1, 2, 3, 4)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes Note 4: The pre-calculated effluent monitor values presented in EALs RA 1.1, RS1 .1 and RG1 .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available Table 3 Effluent 1\1.lonitor Classification Thresholds Release Point Monitor General SAE Alert UE PRM-RE-1B ---- ---- ---- 3.00E+03 cps rn Reactor Building Exhaust

I 0

PRM-RE-1C 2.00E+04 cps 2.00E+03 cps 4.00E+02 cps ----

Cl) rn Turbine Building Exhaust TEA-RIS-13 8.35E-02 µCi/cc 8.35E-03 µCi/cc 8.35E-04 µCi/cc 4.22E-05 µCi/cc C'll

(!)

Radwaste Building WEA-RIS-14 3.45E-01 µCi/cc 3.45E-02 µCi/cc 3.45E-03 µCi/cc 3.98E-04 µCi/cc Exhaust Radwaste Effluent FDR-RIS-606 ---- ---- .......... 2 X HI-HI alarm "C TSW Effluent TSW-RIS-5

-- ---- ---- 3.00E-05 µCi/cc

I C" Service Water Process A SW-RIS-604 1.00E+02 cps

i ---- ---- ----

Service Water Process B SW-RIS-605 ---- ---- ---- 1.00E+02 cps

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 156 of 187

• Mode Applicability:

I 1 I 2 I3 I4 s I def I Basis:

Threshold #1 The pre-calculated effluent monitor values presented in Table 3 should only be used for emergency classification assessments until the resuits from a dose assessment using actual meteorology are available.

This EAL address gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to SITE BOUNDARY doses that exceed either (ref. 1, 2):

• 1000 mRem TEDE

• 5000 mRem COE Thyroid The column "GENERAL" gaseous effluent release values in Table 3 correspond to calculated doses of 100% of the EPA Protective Action Guidelines (TEDE or COE Thyroid) (ref. 1).

Threshold #2 Dose assessments are performed by computer-based methods (ref. 3).

This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses

• greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.

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

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

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.

CGS Basis Reference(s):

1. Calculation NE-02-09-12 Revision 3

2. FSAR Section 11.5 Process and Effluent Radiological Monitoring and Sampling

3. PPM 13.8.1 Emergency Dose Projection System Operations

4. NEI 99-01 AG1

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 157 of 187

• Category:

Subcategory:

Initiating Condition:

R - Abnormal Rad Release I Rad Effluent 1 - Radiological Effluent Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL:

RS1.1 Site Area Emergency (1) Reading on any Table 3 effluent radiation monitor GT column "SAE" for GE 15 min.

OR (2) Dose assessment using actual meteorology indicates doses GT 100 mrem TEDE or GT 500 mrem thyroid CDE at or beyond the SITE BOUNDARY (Notes 1, 2, 3, 4)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes

• Note 4: The pre-calculated effluent monitor values presented in EALs RA 1.1, RS1 .1 and RG 1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available Table 3 Effluent Monitor Classification Thresholds Release Point Monitor General SAE Alert UE PRM-RE-1B ---- ----- ---- 3.00E+03 cps Ill Reactor Building Exhaust

s 0

PRM-RE-1C 2.00E+04 cps 2.00E+03 cps 4.00E+02 cps ----

Cl>

Ill Turbine Building Exhaust TEA-RIS-13 8.35E-02 µCi/cc 8.35E-03 µCi/cc 8.35E-04 µCi/cc 4.22E-05 µCi/cc C'G C>

Radwaste Building WEA-RIS-14 3.45E-01 µCi/cc 3.45E-02 µCi/cc 3.45E-03 µCi/cc 3.98E-04 µCi/cc Exhaust Radwaste Effluent FDR-RIS-606 ---- --- ---- 2 X HI-HI alarm

,, TSW Effluent TSW-RIS-5

---

--- --- 3.00E-05 µCi/cc

s er Service Water Process A SW-RIS-604

---

---- ---- 1.00E+02 cps

i Service Water Process B SW-RIS-605 ---- ---- ---- 1.00E+02 cps

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 158 of 187

• Mode Applicability:

I 1 I 2 I3 4 s I def I Basis:

Threshold #1 The pre-calculated effluent monitor values presented in Table 3 should only be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

This EAL address gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to SITE BOUNDARY doses that exceed either (ref. 1, 2):

• 100 mRem TEDE

• 500 mRem COE Thyroid The column "SAE" gaseous effluent release values in Table 3 correspond to calculated doses of 10%

of the EPA Protective Action Guidelines (TEDE or COE Thyroid) (ref. 1).

Threshold #2 Dose assessments are performed by computer-based methods (ref. 3).

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

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

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

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

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

CGS Basis Reference{s):

1. Calculation NE-02-09-12 Revision 3

2. FSAR Section 11.5 Process and Effluent Radiological Monitoring and Sampling

3. PPM 13.8.1 Emergency Dose Projection System Operations

4. NEI 99-01 AS1

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 159 of 187

• Category:

Subcategory:

Initiating Condition:

R - Abnormal Rad Release I Rad Effluent 1 - Radiological Effluent Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid COE EAL:

RA1.1 Alert (1) Reading on any Table 3 effluent radiation monitor GT column "ALERT" for GE 15 min.

OR (2) Dose assessment using actual meteorology indicates doses GT 10 mrem TEDE or GT 50 mrem thyroid COE at or beyond the SITE BOUNDARY (Notes 1, 2, 3, 4)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes Note 4: The pre-calculated effluent monitor values presented in EALs RA 1.1, RS1 .1 and RG 1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available Table 3 Effluent Monitor Classification Thresholds Release Point Monitor General SAE Alert UE PRM-RE-1B ---- ---- ---- 3.00E+03 cps U) Reactor Building Exhaust

s 0

PRM-RE-1C 2.00E+04 cps 2.00E+03 cps 4.00E+02 cps ----

Cl)

U) ns Turbine Building Exhaust TEA-RIS-13 8.35E-02 µCi/cc 8.35E-03 µCi/cc 8.35E-04 µCi/cc 4.22E-05 µCi/cc

(!)

Radwaste Building WEA-RIS-14 3.45E-01 µCi/cc 3.45E-02 µCi/cc 3.45E-03 µCi/cc 3.98E-04 µCi/cc Exhaust Radwaste Effluent FDR-RIS-606 ---- ---- ---- 2 X HI-HI alarm "C TSW Effluent TSW-RIS-5 3.00E-05 µCi/cc

s er Service Water Process A SW-RIS-604

---

---- ---- 1.00E+02 cps

J Service Water Process B SW-RIS-605 ---- ---- ---- 1.00E+02 cps

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1 A J Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 160 of 187

• I Mode Applicability:

1 I 2 I3 4 5 I def I Basis:

Threshold #1 The pre-calculated effluent monitor values presented in Table 3 should be used for emergency classification assessments only until the results from a dose assessment using actual meteorology are available.

This EAL address gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to SITE BOUNDARY doses that exceed either (ref. 1, 2):

• 10 mRem TEDE

• 50 mRem COE Thyroid The column "ALERT" gaseous effluent release values in Table 3 correspond to calculated doses of 1%

of the EPA Protective Action Guidelines (TEDE or COE Thyroid) (ref. 1).

Threshold #2 Dose assessments are performed by computer-based methods (ref. 3).

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

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

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

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

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

CGS Basis Reference(s):

1. Calculation NE-02-09-12 Revision 3

2. FSAR Section 11.5 Process and Effluent Radiological Monitoring and Sampling

3. PPM 13.8.1 Emergency Dose Projection System Operations

4. NEI 99-01 AA 1

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 161 of 187

• Category:

Subcategory:

Initiating Condition:

R - Abnormal Rad Release I Rad Effluent 1 - Radiological Effluent Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer EAL:

RU1.1 Unusual Event (1) Reading on any Table 3 effluent radiation monitor GT column "UE" for GE 60 min.

OR (2) Sample analysis for a gaseous or liquid release indicates a concentration or release rate

> 2 x ODCM limits for GE 60 min.

(Notes 1, 2, 3)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes Table 3 Effluent Monitor Classification Thresholds Release Point Monitor General SAE Alert UE PRM-RE-1B ---- ---- ---- 3.00E+03 cps U) Reactor Building Exhaust

~

0 PRM-RE-1C 2.00E+04 cps 2.00E+03 cps 4.00E+02 cps ----

Cl>

U) ns Turbine Building Exhaust TEA-RIS-13 8.35E-02 µCi/cc 8.35E-03 µCi/cc 8.35E-04 µCi/cc 4.22E-05 µCi/cc

(!)

Radwaste Building WEA-RIS-14 3.45E-01 µCi/cc 3.45E-02 µCi/cc 3.45E-03 µCi/cc 3.98E-04 µCi/cc Exhaust Radwaste Effluent FDR-RIS-606 ---- ---- ---- 2 X HI-HI alarm TSW Effluent TSW-RIS-5

---

---- --- 3.00E-05 µCi/cc

'2

~

C" Service Water Process A SW-RIS-604 1.00E+02 cps

J ---- ---- ----

Service Water Process B SW-RIS-605 ---- ---- ---- 1.00E+02 cps Mode Applicability:

2 3 4 s I def I Attachment 6.1, EAL Technical Bases

Number: 13.1.1A luseC~ego~:REFERENCE Major Rev: 032 Minor Rev: NIA

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 162 of 187 Basis:

Per NEI 99-01, this EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways and planned.batch releases from releases from non-continuous release pathways. The column "UE" gaseous release values in Table 3 represent two times the appropriate ODCM release rate limits associated with the specified monitors (ref. 1, 2, 3, 4).

The Radwaste Effluent monitor (FDR-RIS-606) Hi-Hi alarm is established per a discharge permit and should be multiplied by 2 to determine the effluent threshold.

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

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

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

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

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

Threshold #1 - This threshold addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways as well as radioactivity releases that cause effluent radiation monitor readings to exceed 2 times the limit established by a radioactivity discharge permit.

This EAL may also be associated with planned batch releases from non-continuous release pathways (e.g., radwaste, waste gas).

Threshold #2 - This threshold addresses uncontrolled gaseous or liquid releases that are detected by sample analyses or environmental surveys, particularly on unmonitored pathways (e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.).

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

CGS Basis Reference(s):

1. CGS Offsite Dose Calculation Manual (ODCM)

2. Calculation NE-02-09-12 Revision 3

3. 16.10.1 Radioactive Liquid Waste Discharge to the River

4. FSAR Section 11.5 Process and Effluent Radiological Monitoring and Sampling

5. NEI 99-01 AU1

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A IUse Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 163 of 187

• Category:

Subcategory:

Initiating Condition:

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

RG1 .2 General Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:

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

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

(Notes 1, 2)

Note 1: The Emergency Director 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:

• I 1 I 2 I 3 I 4 I s I def 1 Basis:

Plant procedures provide guidance for emergency or post-accident radiological environmental monitoring (ref. 1).

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

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

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

CGS Basis Reference(s):

1. PPM 13.9.1 Environmental Field Monitoring Operations

2. NEI 99-01 AG1

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A luseC~ego~:REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 164 of 187

• Category:

Subcategory:

Initiating Condition:

R - Abnormal Rad Release I Rad Effluent 1 - Radiological Effluent Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid COE EAL:

RS1 .2 Site Area Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:

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

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

(Notes 1, 2)

Note 1: The Emergency Director 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:

• I 1 Basis:

I 2 I 3 I 4 I s I def I Plant procedures provide guidance for emergency or post-accident radiological environmental monitoring (ref. 1).

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

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

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

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

CGS Basis Reference(s):

1. PPM 13.9.1 Environmental Field Monitoring Operations

2. NEI 99-01 AS1

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 165 of 187

• Category:

Subcategory:

Initiating Condition:

R - Abnormal Rad Release I Rad Effluent 1 - Radiological Effluent Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid COE EAL:

RA1.2 Alert Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses GT 10 mrem TEDE or GT 50 mrem thyroid COE at or beyond the SITE BOUNDARY for 60 min. of exposure (Notes 1, 2)

Note 1: The Emergency Director 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:

I 1 I 2 I 3 4 s I def 1 Basis:

• For a radiological water release, the calculated effluent concentration from a field team sample is compared to the emergency action level (ref. 1, 2, 3).

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

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

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

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

CGS Basis Reference(s}:

1. FSAR Section 11.5 Process and Effluent Radiological Monitoring and Sampling

2. PPM 13.9.1 Environmental Field Monitoring Operations

3. PPM 13.9.5 Environmental Sample Collection

4. NEI 99-01 AA1

• Attachment 6.1, EAL Technical Bases L_________ -

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 166of187

• Category:

Subcategory:

Initiating Condition:

R - Abnormal Rad Release I Rad Effluent 1 - Radiological Effluent Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid COE EAL:

RA1.3 Alert Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:

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

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

(Notes 1, 2)

Note 1: The Emergency Director 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:

• I 1 Basis:

I 2 I 3 I 4 I s I def I Plant procedures, provides guidance for emergency or post-accident radiological environmental monitoring (ref. 1).

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

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

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

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

CGS Basis Reference(s):

1. PPM 13.9.1 Environmental Field Monitoring Operations

. 2. NEI 99-01 AA 1

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 167 of 187

• Category:

Subcategory:

Initiating Condition:

R - Abnormal Rad Release I Rad Effluent 2 - Irradiated Fuel Event Spent fuel pool level cannot be restored to at least the top of the spent fuel racks for 60 minutes or longer EAL:

RG2.1 General Emergency Spent fuel pool level cannot be restored to at least 0.5 ft for GE 60 min. (Note 1)

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

• Mode Applicability:

I 1 I 2 I 3 I 4 I s I def I Basis:

The spent fuel pool is designed to maintain the water level in the pool above the top of irradiated fuel and thus providing cooling for the fuel assemblies. SFP level can be determined by FPC-LIT-21A, FPC-LIT-21 B, FPC-Ll-21 or local indication. Instrument "reference zero" is the top of the spent fuel pool racks (ref. 1).

• The spent fuel pool is equipped with primary and backup guided wave radar probes to measure pool level. The range is continuous from the high pool level elevation to the top of the spent fuel racks. Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication (FPC-LIT-21A and FPC-LIT-21 B) capable of identifying SFP level near top of the fuel racks (Level 3: 0.4 ft [rounded to 0.5 ft]). (ref. 1).

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

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

CGS Basis Reference{s):

1. IMDS for FPC-LIT-21A/21 B

2. NEI 99-01 AG2

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A \ Use Category: REFERENCE Major Rev: 032 Minor Rev: NIA

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 168 of 187

• Category:

Subcategory:

Initiating Condition:

R - Abnormal Rad Release I Rad Effluent 2 - Irradiated Fuel Event Spent fuel pool level at the top of the fuel racks EAL:

RS2.1 Site Area Emergency Lowering of spent fuel pool level to 0.5 ft Mode Applicability:

I 1 I 2 I 3 I 4 I s I def I Basis:

The spent fuel pool is designed to maintain the water level in the pool above the top of irradiated fuel and thus providing cooling for the fuel assemblies. SFP level can be determined by FPC-LIT-21A, FPC-LIT-21 B, FPC-Ll-21 or local indication. Instrument "reference zero" is the top of the spent fuel pool racks (ref. 1).

The spent fuel pool is equipped with primary and backup guided wave radar probes to measure pool level. The range is continuous from the high pool level elevation to the top of the spent fuel racks. Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication (FPC-LIT-21A and FPC-LIT-21 B) capable of identifying SFP level near top of the fuel racks (Level 3: 0.4 ft [rounded to 0.5 ft]) (ref. 1).

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

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

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

CGS Basis Reference(s):

1. IMDS for FPC-LIT-21A/21 B

2. NEI 99-01 AS2

• Attachment 6.1, EAL Technical Bases

Number:* 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 169 of 187

• Category:

Subcategory:

Initiating Condition:

R - Abnormal Rad Release I Rad Effluent 2 - Irradiated Fuel Event Significant lowering of water level above, or damage to, irradiated fuel EAL:

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

I 1 I 2 I 3 I 4 I s I def I Basis:

The spent fuel pool is designed to maintain the water level in the pool above the top of active fuel providing cooling for the fuel bundles.

This EAL is applicable for conditions in which irradiated fuel is being transferred to and from the RPV and spent fuel pool as well as for spent fuel pool drain down events.

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

As such, they represent an actual or potential substantial degradation of the level of safety of the plant.

This EAL applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with EU1 .1.

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

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

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

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

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

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

CGS Basis Reference(s):

1. ABN-FPC-LOSS Loss of Fuel Pool Cooling

2. NEI 99-01 AA2

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A J Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 170 of 187

• Category:

Subcategory:

Initiating Condition:

R - Abnormal Rad Release I Rad Effluent 2 - Irradiated Fuel Event Unplanned loss of water level above irradiated fuel EAL:

RU2.1 Unusual Event UNPLANNED water level drop in the REFUELING PATHWAY as indicated by EITHER of the following:

• SFP level LE 22.3 ft.

• SFP low level alarm AND UNPLANNED rise in area radiation levels as indicated by any of the following radiation monitors:

• ARM-RIS-1 Reactor Building Fuel Pool Area

• ARM-RIS-2 Reactor Building Fuel Pool Area

• ARM-RIS-34 Reactor Building Elevation 606 Mode Applicability:

I 1 I 2 1. 3 I 4 I s I def I Basis:

The spent fuel pool is designed to maintain the water level in the pool above the top of active fuel providing cooling for the fuel bundles. The fuel pool low level alarm is actuated by level switch FP-LS-4A when fuel pool water level drops below 605' 5-1/2". SFP level is can be determined by FPC-Ll-21, FPC-LIT-21A, FPC-LIT-21B or local indication (ref. 1, 2, 3).

This EAL is applicable for conditions in which irradiated fuel is being transferred to and from the RPV and spent fuel pool as well as for spent fuel pool drain down events.

ARM-RIS-1 and ARM-RIS-2 are located in the fuel pool area of the 606' elevation of the Reactor Building. ARM-RIS-34 is located on the east side of the 606' elevation of the Reactor Building (ref. 4).

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

A water level decrease will be primarily determined by indications from available level instrumentation.

Other sources of level indications may include reports from plant personnel (e.g., from a Refuel 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 Refuel bridge area radiation monitor reading may increase due to planned evolutions such as lifting of the reactor vessel head or movement of.a fuel assembly. Note that this EAL is applicable only in cases where the elevated reading

• is due to an unplanned loss of water level.

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

Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: NIA

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 171 of187

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

CGS Basis Reference(s):

1. PPM 4.626.FPC1-2.2 (4.626.FPC2-2.2) Fuel Pool Level High/Low

2. PPM 4.627.FPC2-2.2 (4.627.FPC2-2.2) Fuel Pool Level High/Low

3. ABN-FPC-LOSS Loss of Fuel Pool Cooling

4. FSAR Table 12.3-1 Area Monitors

5. NEI 99-01 AU2

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 172 of 187

• Category:

Subcategory:

Initiating Condition:

R - Abnormal Rad Release I Rad Effluent 2 - Irradiated Fuel Event Significant lowering of water level above, or damage to, irradiated fuel EAL:

RA2.2 Alert Damage to irradiated fuel resulting in a release of radioactivity AND High alarm on any of the following radiation monitors:

• ARM-RIS-1 Reactor Building Fuel Pool Area

• ARM-RIS-2 Reactor Building Fuel Pool Area

• ARM-RIS-34 Reactor Building Elevation 606

• REA-RIS-609A-D Rx Bldg Vent Mode Applicability:

I 1 I 2 I 3 I 4 I s I def I Basis:

ARM-RIS-1 and ARM-RIS-2 are located in the fuel pool area of the 606' elevation of the Reactor Building. ARM-RIS-34 is located on the east side of the 606' elevation of the Reactor Building (Ref. 1).

The ARM alarm setpoints are controlled by procedure.

REA-RIS-609A-D are the Reactor Building Exhaust Plenum radiation monitors. This system monitors the radiation level of the reactor building ventilation system exhaust plenum prior to its discharge from the building into the elevated release duct. A high radioactivity level in the exhaust system could be due to fission gases from damaged or leaking spent fuel or an accident (ref. 2). Actuation of the High-High alarm actuates a Secondary Containment isolation and starts SGT (ref. 3).

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

As such, they represent an actual or potential substantial degradation of the level of safety of the plant.

This EAL applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with EU1 .1.

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

This EAL addresses a release of radioactive material caused by mechanical damage to irradiated fuel.

Damaging events may include the dropping, bumping or binding of an assembly, or dropping a heavy load onto an assembly. A rise in readings on radiation monitors should be considered in conjunction with in-plant reports or observations of a potential fuel damaging event (e.g., a fuel handling accident).

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

Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: NIA

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 173 of 187

• CGS Basis Reference{s):

1.

2.

CGS FSAR Table 12.3-1 Area Monitors FSAR Section 11.5.2.1.2 Reactor Building Exhaust Plenum Radiation Monitoring System

3. PPM 4.602.A5-1.4 Reactor Building Exh Plenum Rad Hi-Hi

4. NEI 99-01 AA2

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1 A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 174of187

• Category:

Subcategory:

Initiating Condition:

R - Abnormal Rad Release I Rad Effluent 2 - Irradiated Fuel Event Significant lowering of water level above, or damage to, irradiated fuel EAL:

RA2.3 Alert Lowering of spent fuel pool level to 10 ft Mode Applicability:

I 1 I 2 I 3 I 4 I s I def I Basis:

The spent fuel pool is designed to maintain the water level in the pool above the top of irradiated fuel and thus providing cooling for the fuel assemblies. SFP level can be determined by FPC-LIT-21A, FPC-LIT-218, FPC-Ll-21 or local indication. Instrument "reference zero" is the top of the spent fuel pool racks (ref. 1).

The spent fuel pool is equipped with primary and backup guided wave radar probes to measure pool level. The range is continuous from the high pool level elevation to the top of the spent fuel racks. Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication (FPC-LIT-21A and FPC-LIT-21 B) capable of identifying SFP level providing personnel shielding (Level 2: 9.8 ft [rounded to 10 ft.]) (ref. 1).

This EAL 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.

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 Refuel modes.

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

Escalation of the emergency classification level would be via ICs RS 1or RS2).

CGS Basis Reference(s):

1. IMDS for FPC-LIT-21A/218

2. NEI 99-01 AA2

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 175 of 187

• Category:

Subcategory:

Initiating Condition:

R - Abnormal Rad Release I Rad Effluent 3 - Area Radiation Levels Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown EAL:

RA3.1 Alert (1) Dose rates GT 15 mR/hr in Control Room (ARM-RIS-19) or CAS (by survey)

OR (2) An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to any Table 9 rooms or areas (Note 5)

Note 5: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted Table 9 Safe Operation & Shutdown Areas Room/Area Mode Applicability RW 467' Radwaste Control Room (RHR flush to RW tanks) 3 RW 467' Vital Island (RHR-V-9 disconnect) 3 RB 422' B RHR Pump Rm (local pump temperatures) 3 RB 454' B RHR Pump Rm (operate RHR-V-85B) 3 Mode Applicability:

I 1 I 2 I 3 I 4 s I def I Basis:

Threshold #1 The CGS Control Room requires continuous occupancy because of its importance to assure safe plant operations and control of site security functions (Central Alarm Station).

Control Room ARM (ARM-RIS-19) measures area radiation in a range of 1 to 104 mR/hr (ref. 1).

Threshold #2 The list of plant rooms or areas with entry-related mode applicability identified specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations) are not included. In addition, the list specifies the plant mode(s) during which entry would be required for each room or area (ref. 2).

Attachment 6.1, EAL Technical Bases

Number: 13.1.1A IUse Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 176 of 187

• 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 Director should consider the cause of the increased radiation levels and determine if another IC may be applicable.

For threshold #2, an Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect at the time of the elevated radiation levels. The emergency classification is not contingent upon whether entry is actually necessary at the time of the increased radiation levels. Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., installing temporary shielding, requiring use of non-routine protective equipment, requesting an extension in dose limits beyond normal administrative limits).

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

• The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the elevated radiation levels). For example, the plant is in Mode 1 when the radiation increase occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4.

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

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

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

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

CGS Basis Reference(s):

1. FSAR Table 12.3-1 Area Monitors

2. Attachment 7.4 Safe Operation & Shutdown Rooms/Areas Tables 9 Bases

3. NEI 99-01 AA3

• END Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 177 of 187

• TABLES AND NOTES Table 1 Sumps/Pool

• Any valid Hi-Hi level alarm on R-1 through R-5 sumps

• EDR GE 25 GPM

• FDR GE 10 GPM

• Wetwell level rise

• Observation of UNISOLABLE RCS leakage Table 2 AC Power Sources Offsite

• Startup Transformer TR-S

• Backup Transformer TR-B

• Backfeed 500 KV power through Main Transformers (if already aligned in modes 4, 5, def only)

Onsite

• DG1

• DG2

• Main Generator via TR-N1/N2

• Attachment 6.2, Tables and Notes L__ _____ - --

Number: 13.1.1A j Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 178 of 187

• Release Point Table 3 Monitor Effluent Monitor Classification Thresholds General SAE Alert UE PRM-RE-1B ---- ---- .......... 3.00E+03 cps Ill Reactor Building Exhaust

s 0

PRM-RE-1C 2.00E+04 cps 2.00E+03 cps 4.00E+02 cps ----

Q)

Ill Turbine Building Exhaust TEA-RIS-13 8.35E-02 µCi/cc 8.35E-03 µCi/cc 8.35E-04 µCi/cc 4.22E-05 µCi/cc ns

(!)

Radwaste Building WEA-RIS-14 3.45E-01 µCi/cc 3.45E-02 µCi/cc 3.45E-03 µCi/cc 3.98E-04 µCi/cc Exhaust Radwaste Effluent FDR-RIS-606 ---- ---- ---- 2 X HI-HI alarm TSW Effluent TSW-RIS-5

---

---- ---- 3.00E-05 µCi/cc

'5!

s er Service Water Process A SW-RIS-604

---

---- ---- 1.00E+02 cps

J Service Water Process B SW-RIS-605 ---- ---- ---- 1.00E+02 cps Table 4 Communication Methods

• System Plant Public Address (PA) System Plant Telephone System Plant Radio System Operations and x

x On site x

ORO x

NRC Security Channels Offsite calling capability from the Control x x Room via direct telephone Long distance calling capability on the x x commercial phone svstem

• Attachment 6.2, Tables and Notes

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 179 of 187

• Table 5 Safe Shutdown Areas

• Vital portions of the Rad Waste/Control Building:

467' elevation vital island 487' elevation cable spreading room Main Control Room and vertical cable chase 525' elevation HVAC area

• Reactor Building

• Vital portions of the Turbine Building DEH pressure switches RPS switches on turbine throttle valves Main steam line radiation monitors Turbine Building ventilation radiation monitors Main steam line piping up to MS-V-146 and the first stop valves Standby Service Water Pump Houses Diesel Generator Building Table 7 RCS Heat-up Duration Thresholds CONTAINMENT CLOSURE RCS Status Heat-up Duration Status Intact N/A 60 min.*

established 20 min.*

Not intact not established 0 min.

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

• Attachment 6.2; Tables and Notes

Number: 13.1.1A IUse Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 180 of 187

• Table 8 Hazardous Events

• Seismic event

• Internal or external FLOODING event

• High winds

• Tornado strike

• FIRE

• EXPLOSION.

• Volcanic ash fallout

• Other events with similar hazard characteristics as determined by the Shift Manager Table 9 Safe Operation & Shutdown Areas

• Room/Area RW 467' Radwaste Control Room (RHR flush to RW tanks)

RW 467' Vital Island (RHR-V-9 disconnect)

Mode Applicability 3

3 RB 422' B RHR Pump Rm (local pump temperatures) 3 RB 454' B RHR Pump Rm (operate RHR-V-858) 3 Table 10 Safety System Parameter~"' "

• Reactor power

• RPV level

• RPV pressure

• Primary containment pressure

• Wetwell level

• Wetwell temperature

• Attachment 6.2, Tables and Notes

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032

- Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 181 of 187

• Table 11 Significant Transients

• Reactor scram

• Runback GT 25% thermal reactor power

• Electrical load rejection GT 25% full electrical load

• ECCS injection

• Thermal power oscillations GT 10%

• Attachment 6.2, Tables and Notes

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 182 of 187

• Table 12 Notes Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes Note 4: The pre-calculated effluent monitor values presented in EALs RA 1.1, RS 1.1 and RG1 .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available Note 5: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required

• Note 7: This EAL does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents Note 8: A manual scram action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies Note 9: If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then emergency classification is not warranted.

Note 10:1f the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted .

• END Attachment 6.2, Tables and Notes

Number: 13.1.1A IUse Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 183 of 187 SAFE OPERATION & SHUTDOWN AREAS TABLE 9 BASES

Background

NEI 99-01 Revision 6 ICs AA3 and HAS prescribe declaration of an Alert based on impeded access to rooms or areas (due to either area radiation levels or hazardous gas concentrations) where equipment necessary for normal plant operations, cooldown or shutdown is located. These areas are intended to be plant operating mode dependent. Specifically the Developers Notes for AA3 and HAS states:

The "site-specific list of plant rooms or areas with entry-related mode applicability identified" should specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, coo/down and shutdown. Do not include rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations). In addition, the list should specify the plant mode(s) during which entry would be required for each room or area.

The list should not include rooms or areas for which entry is required solely to perform actions of an administrative or record keeping nature (e.g., normal rounds or routine inspections).

Further, as specified in IC HAS:

The list need not include the Control Room if adequate engineered safety/design features are in place to preclude a Control Room evacuation due to the release of a hazardous gas. Such features may include, but are not limited to, capability to draw air from multiple air intakes at different and separate locations, inner and outer atmospheric boundaries, or the capability to acquire and maintain positive pressure within the Control Room envelope .

• Attachment 6.3, Safe Operation & Shutdown Areas Table 9 Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 184 of 187 The following table lists the locations into which an operator may be dispatched in order to safely shut down the reactor and reach cold shutdown conditions in accordance with plant procedures. The reason for these in-plant actions has been evaluated and a determination made whether or not the actions, if not performed, would prevent achieving cold shutdown. The minimum set of in-plant actions, associated locations, and operating modes to shut down and cool down the reactor are identified as "yes". These comprise the rooms/areas to be included in EAL Table 9.

If not performed, prevents Building Elevation Room Modes Reason cool down/shutdown?

TG 441 Booster pump 1,3,4 Condensate Booster Pump No area SID per SOP-COND-SHUTDOWN RFT Area 1,3 RFT SID per SOP-RFT- No SHUTDOWN IR-9 Area 1 Verify Desuperheater pressure No per SOP-MT-SHUTDOWN Mech Vacuum 3 Mech Vacuum Pmp Start per No, can break vacuum and cool Pump Room SOP-AR-SHUTDOWN down with SRVs Mech Vacuum 3 Mech Vacuum Pmp Stop per No Pump Room SOP-AR-START OG Preheater 3 OG System SID per SOP-OG- No Room SHUTDOWN Gland Exh 3 OG System SID per SOP-OG- No Condenser Area SHUTDOWN H2 valve station 1,3,4 H2 makeup to Mn Generator No per SOP-H21C02-0PS 501 MT Turning 1 Place MT on Turning Gear per No Gear Area SOP-MT-START CW Pump nla CW Pump Area 1 CW Pmp SID per SOP-CW- No House SHUTDOWN Towers and CW 1 Monitor water level per SOP- No Basin CW-SHUTDOWN

• Attachment 6.3, Safe Operation & Shutdown Areas Table 9 Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 185 of 187 If not performed, Building Elevation Room Modes Reason prevents cooldown/shutdown?

RW 467 Radwaste 1,3 Remove CFDs from No Control Room service per SOP-CFO-SHUTDOWN 3 Align RW tanks to receive Yes, RWCR operator RHR water per SOP-RHR- will need to align soc Radwaste tanks to accept RHR SDC flush water.

Vital Island 3 Close disc for RHR-V-9 per Yes, Disconnect for SOP-RHR-SDC RHR-V-9 is normally left open during power operations.

525 Communication 4 Check Oscillograph per No Room PPM 3.2.1 TMU n/a TMU Pump 1 TMU Pmp Shutdown per No Area SOP-TMU-SHUTDOWN

• Switchyard Rx Bldg n/a 422 500KV MODs B RHR Pump Room 1

3 Open MODs per SOP-MT-SHUTDOWN RHR Pump local temperature reading per SOP-RHR-SDC No Yes, local readings of RHR pump taken prior to and during flush to ensure minimal delta-T is established 441 Railroad Bay 1 CIA N2 Bottle Change out No, Many installed per SOP-CIA-OPS bottles, infrequent task 454 B RHR Pump 3 Cycle RHR-V-85B for flush Yes, valve must be Room per SOP-RHR-SDC cycled to perform RHR SDC line flush 501 HCU Area 1 HCU Charging per SOP-. No, infrequent task CRD-HCU 548 B RHR Valve 3 Vent RHR system post No, vent not Room flush per SOP-RHR-SDC necessary to enter soc 572 B RHR HX 3 Vent RHR system post No, vent not Room flush per SOP-RHR-SDC necessary to enter

• Attachment 6.3, Safe Operation & Shutdown Areas Table 9 Bases soc

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: NIA

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 186 of 187 Table 9 Results Table 9 Safe Operation & Shutdown Areas Room/Area Mode Applicability RW 467' Radwaste Control Room (RHR flush to RW tanks) 3 RW 467' Vital Island (RHR-V-9 disconnect) 3 RB 422' B RHR Pump Rm (local pump temperatures) 3 RB 454' B RHR Pump Rm (operate RHR-V-85B) 3 Plant Operating Procedures Reviewed

1. PPM 3.2.1 NORMAL PLANT SHUTDOWN 13. SOP-MT-SHUTDOWN

• 2. SOP-FWH-SHUTDOWN

3. SOP-MSR-OPS

4. SOP-CW-SHUTDOWN

5. SOP-COND-SHUTDOWN

6. SOP-CFO-SHUTDOWN

14. SOP-CW-OPS

15. SOP-CG-SHUTDOWN

16. SOP-AR-START

17. SOP-MT-START

18. OSP-RHR-M102

7. SOP-TMU-SHUTDOWN 19. SOP-RHR-SDC

8. SOP-AS-START 20. SOP-RCIC-SHUTDOWN

9. SOP-SS-OPS 21. SOP-SS-SHUTDOWN

10. SOP-RFT-SHUTDOWN 22. SOP-H2/C02-0PS

11. SOP-RFT-OPS 23. SOP-CIA-OPS

12. SOP-AR-SHUTDOWN 24. SOP-CRD-HCU

• END Attachment 6.3, Safe Operation & Shutdown Areas Table 9 Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 032 Minor Rev: N/A

Title:

CLASSIFYING THE EMERGENCY - TECHNICAL BASES Page: 187 of 187 EMERGENCY CLASSIFICATION CHART DISTRIBUTION NOTE: The Emergency Classification Chart is provided in a separate, controlled distribution to the following locations:

Location No. Of Copies Control Room (MCR) 2 half size Control Room Simulator 2 half size Technical Support Center (TSC) 2 half size, 1 full size Alternate TSC 2 half size Emergency Operations Facility (EOF) 2 half size, 2 full size Alternate EOF 2 half size, 1 full size Joint Information Center (JIC) 1 half size Remote Shutdown Room 1 half size Simulator Remote S/D Room 1 half size NOTE: Information Only charts should be provided to the following locations:

• Benton County EOC Franklin County EOC Washington State EOC Grant County EOC Adams County EOC 1 half size 1 half size 1 half size 1 half size 1 half size Yakima County EOC 1 half size

• END Attachment 6.4, Emergency Classification Chart Distribution l

Initials Date Number: 13.8.1 Major Rev: 037 r------------------~---------------j Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 1of32 PCN#:

PLANT PROCEDURES MANUAL N/A Effective Date:

Illllll lllll llll llllll llll llllll llll llll 13.8.1 09/12/17

Number: 13.8.1 I Use Category: REFERENCE Major Rev: 037 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 2 of 32 DESCRIPTION OF CHANGES t..JU~ti.tic~H6~.(t~t1ufrnci***rijr*hlaj6fr~visi~nr************. ********r>************

Revised procedure to align with new EAL scheme per NEI 99-01 Rev 6

'Pag¢(s) >> .* Description (including summary,J~asonJ initiati(lg ctqc.ume11J, ifappli(;aple)

• . :* **.**.-... ... -*:.* .-*.* **.*. . . . . **  :**. . .:'*.,., .. *.. *::=:.* .. :.-:*:.':":*.:.*:::*. :... ::*,::.: . *.***:* .:*:: . . *. **.. **.***:*** .. : .:*.:.*-:-.*:.*** ::. . . : .* .:: .. : .. :.** .. *... * ..,-*

6 Step 4.1.1, added direction to not use URI for UE classification Several Revised procedure number from 13.1.1 to 13.1.1 A 29 Revised Engineering calc # Table 3 changes to UE values 9,11,19 Changed "GT UE" to "Greater Than or Equal To Alert" 10 & 12 Deleted reference to "Table 4". This table no longer exists.

31 Updated applicable ARMs for new EAL scheme

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __J

Number: 13.8.1 I Use Category: REFERENCE Major Rev: 037 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 3 of 32 TABLE OF CONTENTS Page 1.0 PURPOSE ............................................................................................................................... 4 2.0 DEFINITIONS .......................................................................................................................... 4 3.0 RESPONSIBILITIES ................................................................................................................ 5 4.0 PROCEDURE .......................................................................................................................... 6

5.0 REFERENCES

...................................................................................................................... 14 6.0 DOCUMENTATION ............................................................................................................... 14 7.0 ATTACHMENTS .................................................................................................................... 15 7.1 URI User Guidance ................................................................................................................ 16 7.2 Air Sampling Worksheet Calculation ...................................................................................... 24 7.3 Obtaining Alternate Met Data ................................................................................................. 25 7.4 Computer Points Used in the eDNA View Radiological Status Screen ................................... 28 7.5 Alternate Method for PRM-RE-1 Band PRM-RE-1C ............................................................... 29 7.6 Alternate Method forTEA-RIS-13 orWEA-RIS-14 ................................................................. 30 7.7 Alternate Method for ARMS or OG-RIS-612 ........................................................................... 31

Number: 13.8.1 j Use Category: REFERENCE Major Rev: 037 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 4of32 1.0 PURPOSE This procedure provides instructions for the use of the computerized dose projection system Unified RASCAL Interface (URI) to predict offsite dose rates, integrated doses and radioactive material deposition for locations within the 10-mile Plume Emergency Planning Zone (EPZ) and the 50-mile Ingestion EPZ. Actual manipulation of system display terminals is described in document MAN-URl-01 referred to as the URI User's Manual.

The dose projection system is used for estimating the whole body (TEDE) and thyroid (COE) doses of onsite and offsite persons in the event of potential or actual accidental release of radioactivity to the environment. The dose projection system used at Columbia consists of a computer software program that relies on pre-calculated, real time, site-specific relationships between effluent monitor readings (or sample results) and on-site and offsite dose rates.

The dose projection system is available in the Control Room, TSC, EOF, and Alternate EOF.

Having the dose projection system software loaded on multiple, stand-alone computers located in the various emergency centers maximizes dose projection capability. Field Team data may be used to calculate dose projections or validate previous projections.

The software program supports a rapid version of dose projection using limited pathways for a rapid evaluation of the release out to 10 miles. The software also supports a detailed dose projection based on more detailed pathways with dose projections from the site boundary out to 1O or 50 miles .

• 2.0 State and county organizations will have access to this system in the EOF or by transmission of output information to their emergency centers.

DEFINITIONS 2.1 COE Thyroid - Committed Dose Equivalent to the thyroid.

2.2 Delta T - The temperature difference between two sensors located at different elevations on a meteorological tower.

2.3 EDE to TEDE Ratio (EDE/TEDE) - Ratios computed by URI that are used for determining the Emergency Worker Dose Adjustment Factor. The EDE is the external gamma dose that is normally monitored by emergency workers through their self-reading dosimeters (electronic dosimeters or pocket ion chambers); and TEDE is the total dose from both EDE and the internal dose from inhaled iodines and particulates.

2.4 Elevated Release - An effluent release point model that assumes that the release point is from a discreet true elevated (tall) stack.

2.5 Emergency Worker Dose Adjustment Factor (EWDAF) - An adjustment factor determined from the EDE/TEDE Ratio that is used by offsite emergency workers to monitor their TEDE whole body dose based on their EDE doses as measured on their self-reading dosimeters. The adjustment factor is the inverse of the EDE/TEDE Ratio reported by URI (for example, an EDE/TEDE Ratio of 0.15 would mean that the EWDAF is 1 / 0.15 = 6.7) .

IUse Category: REFERENCE Number: 13.8.1 Major Rev: 037 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 5 of 32 2.6 Radioactive Release - A radioactive release is in progress when:

Effluent monitors indicate an increase in radiation levels from normal readings for plant operating conditions, or Field teams detect environmental radiation 1O times greater than normal background.

AND The increased levels are attributable to the emergency event.

2.7 Radioactive Release Termination -A radioactive release is terminated when the following criteria have been met:

1. The source of the release has been isolated;

2. The effluent monitors are trending downward (if available);

3. Environmental Field Team surveys indicate a decrease in radiation levels or airborne radioactivity.

2.8 RASCAL - NRC supported and distributed computer software for determining source term, atmospheric dispersion, and dose consequences.

2.9 Site Boundary (SB) - Closest distance between owner controlled area boundary and core, set

• 2.10 2.11 at 1.2 miles .

Source Term - The quantity and radionuclide makeup of the material in the release. The source term used in URI is based on NUREG-1228.

Stability Class - Values from A to G representing ranges of Delta T which in turn represent atmospheric mixing estimations. The NRC definitions of these ranges are used to define the stability classes used in URI.

2.12 Stack Door Monitor (SOM) -A portable radiation survey meter (e.g., Teletector) or radiation monitoring instrument (e.g., AMP-100) used for monitoring radiation levels at the exterior and center of the Elevated Release Stack Access Door (R-DOOR-RS 15). This provides an alternate means for monitoring Reactor Building post-accident effluent releases when the installed Stack Monitors (PRM-RE-1B and/or PRM-RE-1C) are inoperable or unavailable.

SOM readings (in mR/hr) are available from HP; or, when set up for remote monitoring, through eDNA Real-Time Client or eDNA Trend from eDNA Point IDs EP99M (mid-range SOM) and EP99H (high-range SOM) in the ENW.WRM Service.

2.13 TEDE-Total Effective Dose Equivalent (TEDE)-The sum of the Deep Dose Equivalent (ODE) and the Committed Effective Dose Equivalent (CEDE).

2.14 Unified RASCAL Interface (URI) - Computer software which replaces the NRC issued RASCAL user interface for user input of dose assessment parameters and interpretation of results that interfaces to the RASCAL meteorological and dose processor modules.

3.0 RESPONSIBILITIES 3.1 STA/ Incident Advisor Responsible for performing dose assessments until relieved by an on-call ERO Member.

Number: 13.8.1 j Use Category: REFERENCE Major Rev: 037 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 6 of 32 3.2 Dose Projection HP Once the EOF takes control of offsite radiological assessment the Dose Projection HP is responsible for performing dose assessments until the event is terminated.

3.3 Chemistry Effluent Manager Once the TSC is manned, the Chemistry Effluent Manager is responsible for performing dose assessments if required for the TSC.

4.0 PROCEDURE 4.1 General Instructions 4.1.1 URI is not to be used for UE classification.

4.1.2 If in a declared emergency and an offsite dose or dose rate projection is needed, or if so directed, use URI to perform offsite dose calculations.

4.1.3 Access the Plant Data Information System (POIS) through the electronic Distributed Network Architecture (eDNA) software to obtain and monitor key radiation monitor, meteorological, and Plant effluent data.

a. At a LAN supported computer:

• 1) 2)

Double click on the "eDNA View" icon on the Desktop, if available. If no "eDNA View" icon is available, open eDNA through the network folder and select eDNA View.

Select CGS, then "POIS", and "Rad Status" to obtain the "Radiological Status" screen. *

• A screen print of the "Radiological Status" screen may be used to capture the current values of radiation monitor, meteorological, and effluent data for tracking changes (setup printer to Landscape mode, and use the "Print Direct" function rather than "Print" from the "File" drop-down menu on the eDNA View "Radiological Status" screen or if the file menu is not available, right click the eDNA screen and select "Print Direct").

• Access other eDNA View screens of POIS data by selecting the appropriate "eDNA View (*.rtv)" in the Application Service Utility window, as desired.

• Graphical trends of the plant parameters displayed on the eDNA View screens can be displayed by clicking in the appropriate plant parameter display box. The time span of the graphical trend may be modified by selecting "Modify Graph Parameters" in the "Graph" pull-down menu, selecting the "Graph" tab, and entering the desired date and time range (click to check the "Y-Axis Auto Scale" box for a clearer graph).

3) Radiation monitor, effluent monitor, SOM, and meteorological data are also available through other eDNA applications (e.g., eDNA Real-Time Client or eDNA Trend) from the Start menu/All Programs/eDNA folder and selecting the appropriate ENW Service (e.g., ENW.CGS, ENW.WRM, etc.).

Number: 13.8.1 \ Use Category: REFERENCE Major Rev: 037 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 7 of 32 4.1.4 Use URI to estimate doses within 15 minutes of a start of a release or as information becomes available. Also use URI to estimate offsite dose during rapidly changing meteorological conditions or release conditions as appropriate.

4.1.5 Start URI by double-clicking on the appropriate icon on the Desktop or start from the "C" drive.

4.1.6 If necessary, refer to Attachment 7.1, URI User Guidance, which provides detailed guidance on using URI.

4.1.7 Review dose projection printouts, note any qualifying factors as appropriate and brief the RPM or REM on the dose projection.

4.1.8 In the EOF/TSC, if any data is suspect, request the Radiation Detection System Engineer or the EOF/TSC Information Coordinator to verify the data.

4.1.9 In the event of unmonitored release paths or if instrumentation (including alternate instrumentation) is out of service or off-scale, Field Team results are used to calculate dose projections. Use one of the following processes to assess Field Team results:

a. Air sample Excel spreadsheet calculator found on the Window Desktop

1) Enter the cartridge and background readings, and press the tab key to perform the calculation.

b. Attachment 7.2, Air Sampling Worksheet Calculation

1) Enter sample and background count rate and sample volume into calculations to determine micro curies/cc.

c. Air Sample Calculator in URI Detailed assessment (preferred)

1) See Attachment 7.1 for details on Air Sample Calculator use.

NOTE: GPS coordinates for the center of the Reactor Pressure Vessel are 119.33278 longitude and 46.47167 latitude

Number: 13.8.1 I Use Category: REFERENCE Major Rev: 037 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 8 of 32 4.1.10 Provide the Emergency Worker Dose Adjustment Factor to the REM and Field Team Coordinator for their use in establishing field team exposure limits.

a. If the Dose Adjustment Factor is 5 or greater, use Dose Adjustment Factor of 5.

b. On page three of the detailed Dose Assessment Report invert the EDE/TEDE ratios to acquire the Dose Adjustment Factor. Use the EDE/TE DE Ratio with Iodine. If respiratory protection or a thyroid blocking agent (i.e., potassium iodide) is being used, then it may be appropriate to use the EDE/TEDE Ratio without Iodine.

Example:

EDE/TEDE ratio with iodine at 2 miles is 0.90 the Emergency Worker Dose Adjustment Factor would be 1.1. (1 divided by 0.9)

c. If the EDE/TEDE ratio is less than or equal to 0.2, use a Dose Adjustment Factor of 5.

4.1.11 If intentionally venting the primary containment, perform dose projection assessment using Reactor Building Exhaust radiation monitors or field team data as applicable.

Venting is a puff release and thus release duration for intentional venting is one hour.

IUse Category: REFERENCE Number: 13.8.1 Major Rev: 037

---

'-----------------! Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 9 of 32 4.2 Dose Estimation Using Rapid URI NOTE: Use of Rapid URI requires that the release pathway matches one of the limited options in the Rapid URI software.

4.2.1 The following steps will provide guidance on performing a Rapid URI dose assessment with normal met tower and radiation monitors available. See Attachment 7.1 for URI imp lementation when normal plant indications are not available.

a. Start the URI software with icon (software on "C" drive if icon missing)

b. If not a real event then select "This is a drill"

c. Select Rapid - LIGHTNING SOLT icon at top left of startup page

d. Select Fuel Clad Damage "Yes", if EAL chart PPM 13.1.1 Table 3 Effluent Monitor reading is Greater Than or Equal to Alert value, otherwise select NO. I

e. If reactor power is LE 1% (shutdown), select box for Reactor Shutdown

f. Verify shutdown date I time - change if needed

g. Select CGS 33ft Tower (Channel A primary, Channel 8 backup) and enter Met data, otherwise use FFTF #9

• h.

i.

j.

k.

Enter Release Duration - for ongoing release round up to next hour and add two Select Release Point Pathway - NOTE ensure correct release point pathway is selected; might have to go to "Detailed" URI Select a Effluent Monitor that has valid and on-scale data for selected pathway Enter the monitor reading for the selected monitor I. Enter the Release Point Flow Rate, if SGT is running and Reactor Building Exhaust Fans are shutdown; use the total of the running SGT train flowrates.

m. Select "Process Assessment" button. A green progress bar will be displayed.

Typical calculation times will be less than 20 seconds.

n. When complete, a graphic of affected areas that may require protective action will be displayed

0. Double click on the protective action display to show a pop-up of the dose results table .

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Emergency Dose Projection System Operations Page: 1O of 32 NOTE: To perform an additional calculation, change any or all input values as needed and select the Process Assessment button again. If only changing one field, it may be necessary to click out of that field to get the Process Assessment button to appear.

4.2.2 See Attachment 7.1, URI User Guidance, for URI use when normal plant indications are not available.

4.2.3 Select Print to Default Printer (printer icon, upper left corner of page) to produce a paper output.

4.2.4 URI has the capability of summing dose assessments from multiple release points.

a. Clicking on the sigma icon in the upper left of the screen will bring up the Assessment Summations table. Minimize this screen for later use.

b. Clicking on the plus icon in the upper right of the screen will add the latest assessment to the summations table.

c. Run a new dose assessment and click on the plus icon to sum the two dose projections. Assessments performed earlier can be added to the summations table by clicking on the browse button in summations table.

• 4.2.5 4.2.6

d. A combined dose assessment report and evacuation area map appears at the bottom of the summations table.

Have the REM or RPM compare URI output at 1.2 miles for EALs per PPM 13.1.1A and for potential protective action recommendations beyond 10 miles per PPM 13.2.2.

Have ED, RPM, or REM sign printed data for distribution.

I

a. Forward to the Emergency Director for approval prior to releasing data for distribution.

b. In the Control Room the Shift Manager as Emergency Director has approval authority.

c. The Washington Senior State Official approves release data for distribution during the ingestion phase.

4.2.7 Distribution of Maps and Data

a. Any dose projection maps or data printouts selected for distribution to offsite agencies shall have REM and Emergency Director review and approval.

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Title:

Emergency Dose Projection System Operations Page: 11 of 32 4.3 Dose Estimation Using Detailed URI 4.3.1 The following steps will provide guidance on performing a Detailed URI dose assessment with normal met tower and radiation monitors available. See Attachment 7.1 for URI implementation when normal plant indications are not available. If in Rapid URI, you must close Rapid URI to get to Detailed.

a. Start URI with icon (software on "C" drive if icon missing).

b. If not a real event then select "This is a drill".

c. Select Detailed Assessment - CLOUD icon top left of startup page.

d. Select fuel condition - Normal Coolant or Reactor Core Accident or Spent Fuel Accident. If Normal Coolant is selected, select Spiking Factor if Reactor Bldg process radiation monitors spike following a plant transient.

e. If Reactor Core Accident- select core condition; CLAD (PPM 13.1.1 Table 3 Effluent Monitor reading is Greater Than or Equal to Alert value) or MELT (SAGs entered).

f. If Spent Fuel Accident, specify age based on last refueling outage shutdown.

g. Selecting 'Spent Fuel Accident' changes the 'Reactor Shutdown' date/time to

'last irradiated/time and sets a default irradiation date/time, which does not necessarily correspond to last refueling outage shutdown .

h. Select CGS 33ft Tower unless data is not available (Channel A primary, Channel B backup), then select FFTF #9 ..

i. Enter Met data.

j. If reactor power is LE 1% (shutdown), select box for Reactor Shutdown.

k. Verify shutdown date I time - change if needed.

I. Enter Release Duration - for ongoing release round up to next hour and add two.

m. Double click yellow Pathway bar to show pathways selection screen.

n. Select best matching pathway.

0. Review the process reduction factor selections on the bottom of the pathway page. Defaults are specified but these may be changed as needed to better represent the plant status.

p. If pathway is through SGT, change Rx Bldg HUT to 2-24 Hours.

Factor groupings that are grey do not apply to the selected pathway.

q. Select "Accept" on the pathways page.

r. Select Monitored Release TAB center top.

NOTE: If TAB cannot be selected, then the pathway selected does not support monitored release assessment. Select a monitor that has valid and on-scale data for selected pathway .

s. Enter the Monitor reading for the selected monitor.

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Emergency Dose Projection System Operations Page: 12 of 32

t. Enter the Release Point Flow Rate, if SGT is running and Reactor Building Exhaust Fans are shutdown; use the total of the running SGT train flowrates.

u. Select the 10 miles or 50 miles Process Assessment button. - A green progress bar will be displayed. Typical calculation times will be less than 20 seconds.

v. Results for dose assessment by distance will be displayed.

w. Select "View the Evacuation Area Graphic" (GLOBE) icon below "Monitored" TAB for PAR graphic.

NOTE: To perform an additional calculation, change any or all input values as needed and select the Process Assessment button again. If only changing one field, it may be necessary to click out of that field to get the Process Assessment button to appear.

4.3.2 Select Print Preview Options (magnifying glass icon directly below the "Monitored" TAB) to view the Dose Assessment report. The Receptor Point Report is not used at CGS.

4.3.3 Select Print Options (printer icon directly below the "Monitored" TAB) to print out a copy of the Dose Assessment report.

4.3.4 URI has the capability of summing dose assessments from multiple release points.

a. Selecting the sigma icon in the upper right of the screen will bring up the

• b.

c.

Assessment Summations table. Minimize this screen for later use .

Selecting the plus icon in the upper right of the screen will add the latest assessment to the summations table.

Run a new dose assessment and click on the plus icon to sum the two dose projections. Assessments performed earlier can be added to the summations table by clicking on the browse button in summations table.

d. A combined dose assessment report and evacuation area map appears at the bottom of the summations table.

4.3.5 Have the REM or RPM compare URI output at 1.2 miles for EALs per PPM 13.1.1A and for potential protective action recommendations beyond 1O miles per PPM 13.2.2.

4.3.6 Have RPM or REM sign printed data for distribution.

a. Forward to the Emergency Director for approval prior to releasing data for distribution.

b. In the Control Room, the Shift Manager as Emergency Director has approval authority.

c. The Washington Senior State Official approves release data for distribution during the ingestion phase .

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Title:

Emergency Dose Projection System Operations Page: 13 of 32 4.3.7 Distribution of Maps and Data

a. Any dose projection maps or data printouts selected for distribution to offsite agencies shall have REM and Emergency Director review and approval.

b. Maps selected for distribution should always be accompanied by the data. This is very important because the plume projected on the map is not closed and without the data sheet, the plume may be misinterpreted.

4.3.8 For Plume map, select "View Receptor Point Locations" (World) icon in the upper left of screen under "View"

a. Recommend selecting RASCAL puff grids and Show Balloons.

b: Hover above grid point with cursor to obtain grid projected dose.

c. Print option is in lower right hand corner of map screen .

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Emergency Dose Projection System Operations Page: 14 of 32 I

5.0 REFERENCES

5.1 Documents 5.1.1 10 CFR 50 .47(b) Planning Standard 5.1.2 10 CFR 50, Appendix E, Emergency Plan 5.1.3 812-03-020, Elimination of Requirements for Post-Accident Sampling System 5.1.4 NUREG-0654, FEMA REP-1, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants 5.1.5 NUREG 1228, Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents 5.1.6 AED CAL NE-02-12-05, Calculation for Alternate Post-Accident Radiation Monitor At Elevated Release Stack Access Door 5.1.7 RSCS TSD-13-035 (CVI 1057-00,131) TEA and WEA replacement 5.1.8 NE-02-09-12 Table 3 (CMR 10715)

• 5.1.9 5.1.10 5.1.11 5.1.12 NE-02-10-05 URI FW-SOFT-COTS-RASCAL and FW-SOFT-COTS-URI CR 244578, CMR changed EAL classification value without adequate 50.54q SQA SOD SDD-PDIS-01, Plant Data Information System Software Design Description 5.1.13 Radiation Protection Calculation 15-04, Calculation of Iodine Air Sampling Cartridge Efficiencies 5.2 Procedures 5.2.1 PPM 13.1.1 A, Classifying the Emergency 5.2.2 PPM 13.2.1, Emergency Exposure Levels/Protective Action Guides 5.2.3 PPM 13.2.2, Determining Protective Action Recommendations 5.2.4 PPM 13.14.11, EP Equipment 6.0 DOCUMENTATION All logs, forms and records completed as the result of implementing this procedure during an actual declared event shall be retained as permanent plant records. Transmit documents to the Permanent Plant File under DIC 2304.2.

A sub-set of documents generated during drills shall be maintained in the Emergency

• Preparedness Department files, as necessary, to support completion of drill/exercise commitments.

Number: 13.8.1 I Use Category: REFERENCE Major Rev: 037

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Emergency Dose Projection System Operations Page: 15 of 32 7.0 ATTACHMENTS 7.1 URI User Guidance 7.2 Air Sampling Worksheet Calculation 7.3 Alternate Method for Obtaining Met Data 7.4 Computer Points Used in the eDNA View Radiological Status Screen 7.5 Alternate Method for PRM-RE-18 and PRM-RE-1C 7.6 Alternate Method for TEA-RIS-13 or WEA-RIS-14 7.7 Alternate Method for ARM or OG-RIS-612

• ~.

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Title:

Emergency Dose Projection System Operations Page: 16 of 32 URI User Guidance 1.0 OVERVIEW 1.1 For ease of explanation, the different sections of the URI screen will be described as "groupings" with the name in the upper left of the grouping as the title of that grouping.

1.2 Options used in the Source Term Grouping will determine which options will be available in succeeding grouping.

• 1.3 The red octagons with the exclamation point inside are warnings that an option needs to be selected or information filled in to before the calculation can be performed. A message also appears in the lower left hand corner of the page stating that all errors must be resolved to complete the calculation. If the cursor is placed over the octagon, a message is provided that will describe the range of the input variable.

1.4 Holding the cursor over a box will often result in the program providing information about the contents of the box.

1.5 Options used in the Release Point Pathway Grouping will determine which inputs to the Process Reduction Factor will become available.

1.6 Options used in the Release Point Pathway Grouping will determine which options will be available in the Assessment Methodologies Grouping .

1.7 Notes

Stack Door Monitor (SOM) is a direct monitored input option. FFTF delta-Tis a direct input for stability class. CGS has committed to performing ground release assessments; so only ten meter Met conditions should be selected.*

1.8 During the ingestion phase, determine manual contour lines on the 10-50 mile map by selecting grid and balloon options to determine the projected 500 µR (relocation boundary),

20 µRand 0.4 µR (food control boundary).

2.0 Rapid URI Activate Rapid URI by selecting the Rapid option (upper right hand corner) on the URI main screen. The Rapid Assessment page is now available for data input.

2.1.1 Source Term Grouping (this grouping provides input on what type of damage to the fuel has occurred)

a. In the Source Term box, answer the initial question concerning Fuel Clad Damage. If damage to the core is suspected (PPM 13.1.1A Table 3 Effluent Monitor reading is GT UE value), select "yes". This causes the assessment to include isotopic mix and inventory for fuel clad damage.

b. Selecting Fuel Clad Damage "No" causes selection of normal coolant concentrations as the basis of the source term. This also activates the Conditions for Coolant Spiking question. Select "Yes" if Reactor Bldg process radiation monitors spike following a plant transient; otherwise select "No".

c. Check the "Reactor Shutdown" box if appropriate and enter date of shutdown using the pull down calendar. Overwrite the indicated time as necessary.

Attachment 7.1, URI User Guidance

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Title:

Emergency Dose Projection System Operations Page: 17 of 32

d. Select Damaged Spent Fuel Assembly for any spent fuel handing accidents This will change the Reactor Shutdown date/time box to "Last Irradiated" an d allow input of when the spent fuel was last irradiated. This should be 5/13/17 (CGS does not have the ability to update software.)

2.1.2 Meteorological Data Grouping (this grouping allows input of meteorological data pertinent to the release)

NOTE: If Primary Met Tower information is not available, see attachment 7.3 for obtaini ng alternate met tower information. {5.1.4}

NOTE: Met Data is normally taken from the eDNA View "Radiological Status" or POIS Rad Status display. Met Channel A is primary input, Channel Bis backup input. IF any parameter cannot be obtained from eDNA View or POIS THEN refer to Attachm ent 7.3, Obtaining Alternate Met Data, to determine appropriate Stability Class.

NOTE: If the wind direction value is greater than 360° subtract 360 from value before entering it into URI program.

a. Select one of the four choices of met tower input to be used. MeteorologicaI parameters from the primary met tower are normally available on the Radiological Parameters screen. If the primary met tower (default) is not selected, the program will provide a notification of this.

b. Input Wind Speed and Wind Direction data. Wind Speed range is GE 0 to LE60 mph. Wind Direction range is GE 0 and LT 360

c. Input delta Tor select appropriate Stability Class from the pull down menu.

Delta T range is GT -10 to LE 10.

d. Use the pull down menu to select the appropriate level of precipitation.

2.1.3 Release Duration Grouping (this grouping allows input on how long the release has been in progress)

a. Provide release duration in hours and minutes. This function defaults to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. If End of Release is not known, a default value of the time of the rele ase is rounded up to the next hour plus two hours should be used.

b. EXAMPLE: Release has lasted for 25 minutes. Round 25 minutes up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and add 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to give a release duration of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> .

• Attachment 7.1, URI User Guidance

IUse Category: REFERENCE Number: 13.8.1 Major Rev: 037 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 18 of 32 2.1.4 Release Point Pathway Grouping (this grouping allows input on where the release originates from and its path for leaving the plant)

NOTE: A Radwaste building release pathway is only available in Detailed URI

a. Select the appropriate release path from the available list. A description of the pathway will appear as the cursor is held over the individual pathways. A list of available monitors will also be displayed. Choose the pathway that best matches the conditions of the release in progress.

b. Selections made in the Source Term Grouping will determine which release paths are available.

c. The release path selected will determine which effluent monitors will be available.

d. The release path selected will also determine the Process Reduction Factor (PRF) applied to the calculation of the offsite dose. IF SGT is running the Rx Bldg hold up time needs to be changed to 2-24 hours.

e. The pathway and assumptions for the input to the PRF are displayed in the lower right portion of the screen .

• NOTE: In the event of a loss of SM-7, the stack monitors will be out of service, alternate indications of Reactor Building effluents can be obtained from the Stack Door Monitor. Use the readings from the Stack Door Monitor to input directly into URI.

2.1.5 Release Point Information Grouping (this grouping allows input on the condition of the monitors at the release point)

a. Select "Yes" if effluent monitors are available. Depending on the Pathway selected, Turbine or Reactor Building monitors will be made available for selection in the Monitor Grouping.

1) Select the monitor to be used

2) Provide the appropriate reading in the Reading box, ensuring the units are correct

3) Verify the Release Point Flow Rate is correct or revise default value to accurately reflect plant conditions. If SGT is running and Reactor Building Exhaust Fans are shutdown; use the total of the running SGT train flowrates. If building (RB, RW, TB) exhaust flow indication is lost, the Control Room should be contacted to determine flow rate via ODCM 6.1.2.D method.

4) Click on the Process Assessment button to run a Rapid Assessment based on the supplied input data .

• Attachment 7 .1, URI User Guidance

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Title:

Emergency Dose Projection System Operations Page: 19 of 32

b. Select "No" if effluent monitors are not available. This action will bring up several options for input.

a) If RCS leakage is suspected, select "Estimated RCS Leak Rate", fill in a known leak rate or select the "I Don't Know" option. Click on the Process Assessment button to run a Rapid Assessment based on the supplied input data.

b) If "Containment Leakage" is selected (only available if a Pathway from the Drywell has been selected), select either containment high rad monitor and provide an appropriate reading in the R/hr box or select "No HRA Available or Applicable". Click on the Process Assessment button to run a Rapid Assessment based on the supplied input data.

c. If Damaged Spent Fuel Assembly is selected in the Source Term Grouping and no effluent monitors are available, proceed as follows.

a) Select "Unmonitored Damaged Spent Fuel Assembly". Click on the Process Assessment button to run a Rapid Assessment based on the supplied input data. '

3.0 Detailed URI Activate Detailed URI by selecting the Detailed option (upper right hand corner) on the URI main screen. The Detailed Assessment page is now available for data input.

3.1.1 Source Term Grouping (this grouping provides input on what type of damage to fuel has occurred)

a. In the Source Term box, select "Normal Coolant" if there is no core damage (No PPM 13.1.1A Table 3 Effluent Monitor reading is GreaterThan or Equal to Alert value). Select Spiking Factor if Reactor Bldg process radiation monitors spike following a plant transient. The numerical Spiking Factor defaults to "30".

b. Reactor Core Accident is selected if there is actual or suspected damage to fuel assemblies. Technical support or core damage procedures should be used to estimate the extent of core damage.

1) The Clad damage option is selected if the core conditions have caused the fuel pin cladding to fail but the core temperature has not become sufficiently high to cause melting of the ceramic fuel matrix.

2) The Melt option is selected if SAGs are entered and core has been uncovered for greater 30 minutes.

c. The Spent Fuel Accident option is selected if the incident involves damage to spent fuel in a depressurized condition.

1) Select "Old" for Fuel Age. "New" is not a valid option for CGS.

2) "Fuel Status" and "Amount of Spent Fuel Damage" will be provided by the software when the Selected Pathway Option is selected. This will also provide the estimated % damage based on the water level in the Spent Fuel Pool.

Attachment 7 .1, URI User Guidance

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3) Checking the "Unmonitored Spent Fuel Accident with No other method applicable" box is only done when there is no other option for determining spent fuel source term due to loss of building integrity, loss of monitors, or other major failure making source term assessment unavailable.

3.1.2 Meteorological Data Grouping (this grouping allows input of meteorological data pertinent to the release)

NOTE: If Primary Met Tower information is not available, see attachment 7.3 for obtaining alternate met tower information. {5.1.4}

a. Select one of the four choices of "site" met tower input to be used. The two FFTF towers are considered "site" met towers. Meteorological parameters from the primary met tower are normally available on the Radiological Parameters screen. If the primary met tower (default) is not selected, the program will provide a notification of this.

b. Input Wind Speed and Wind Direction data. Wind Speed range is GE Oto LE 60 mph. Wind Direction range is GE 0 and LT 360. Double clicking on the Wind Direction box will bring up a depiction of a compass that may assist in selecting wind direction.

• 3.1.3 c.

d.

Input delta Tor select appropriate Stability Class from the pull down menu .

Delta T range is GT-10 to LE 10. Attachment 7.3 has additional information in determining Stability Class.

Use the pull down menu to select the appropriate level of precipitation.

Reactor Status Grouping (this grouping allows input of reactor status or time since the last time the spent fuel was irradiated)

NOTE: For ATWS Conditions:

IF reactor power is GT 1%: leave the Time Since Reactor Shutdown value set to zero unless the Main Control Room (MCR) states the reactor is shutdown.

IF reactor power is LE 1% : Contact the MCR, use the amount of time from when the MCR declares reactor shutdown.

a. When "Normal Coolant" or "Reactor Core Accident" is selected in the Source Term Grouping, "Reactor Status" will be shown next to the check box. If the Reactor is shutdown, click in this box. Change the date/time box to reflect actual shutdown if needed. The "TAS" box will update automatically. Time after shutdown can be entered directly into the "TAS" box by double clicking in the box.

b. Select Damaged Spent Fuel Assembly for any spent fuel handing accidents.

This will change the Reactor Shutdown date/time box to "Last Irradiated" and allow input of when the spent fuel was last irradiated. This should be 5/13/17.

(CGS does not have the ability to update software.)

Attachment 7 .1, URI User Guidance

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Emergency Dose Projection System Operations Page: 21 of 32 3.1.4 Release Duration Grouping (this grouping allows input of the duration of the release)

a. Provide release duration in hours and minutes. This function defaults to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. If End of Release is not known, a default value of the time of the release is rounded up to the next hour plus two hours should be used. Release duration must be input in 15 minute increments. Range is 15 min to LE 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

EXAMPLE: Release has lasted for 25 minutes. Round 25 minutes up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and add 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to give a release duration of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.

3.1.5 Selected Pathway Options Grouping (allows input as to the release flow path and inputs to the Process Reduction Factor)

a. To view the list of available pathways, click on the small yellow box to the right of the "Pathway" box. This action will take you to the Pathways screen.

Available pathway options will depend on selections made in the Source Term Grouping. A description of the pathway will appear as the cursor is held over the "Path" box to the right of the pathway to be selected. A list of available assessment methodologies will also be displayed. Choose the pathway that matches the conditions of the release in progress.

b. The pathway chosen will determine which "Process Reduction Factors" (PRF) will be used in the dose calculation. Hold-up times can be changed from the default values. Other options include use of Drywell sprays, condition of the Suppression Pool, and whether SGBT is working (operable with no high moisture alarms). If SGT is running, Rx Bldg hold up time needs to be changed to 2-24hrs.

c. When pathway has been selected and PRF options have been made, click on the "Accept" button. This action will return you to the main Detailed Assessment screen. The pathway chosen will be displayed in the "Pathway" box. The PRF will be displayed under the "Pathway" box along with inputs to the PRF.

NOTE: In the event of a loss of SM-7, the PRM stack monitors will be out of service, alternate indications of Reactor Building effluents can be obtained from the Stack Door Monitor. Use the readings from the Stack Door Monitor to input directly into URI. TEA and WEA monitors will also require sample carts for alternate methods.

3.1.6 Assessment Methodologies Grouping (allows input to the dose assessment calculation depending on available monitors, leakage rates, samples taken, or field team results, depending on available information)

a. Monitored TAB (available when chosen Pathway is monitored by installed or temporary plant radiation monitor)

1) Release Point: The release point is based on the Selected Pathway chosen. No input required

• 2) Monitor: Select the radiation monitor you wish to use or select the building as appropriate.

Attachment 7.1, URI User Guidance

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Title:

Emergency Dose Projection System Operations Page: 22 of 32

3) Release Point Flow Rate: Input real time data from plant computer system/instrument readings. Range for RB and RW bldgs. is GE O to LE 2E6 CFM, TB bldg is GE 0 to LE 5E6 CFM. IF SGT is running and Reactor Building Exhaust Fans are shutdown; use the total of the running SGT train flowrates. If building (RB, RW, TB) exhaust flow indication is lost, the Control Room should be contacted to determine flow rate via ODCM 6.1.2.D method.

4) Monitor Reading: Input real time data from plant computer system or instrument readings. Ensure units match those in the program. Range is GE 0 to LE 1E6 CPS for Stack Monitors, GE 0 to LE 1E3 uCi/cc for TB and RWbldgs.

b. Containment Leakage TAB (available when chosen Pathway originates inside containment)

1) Method a) Normal Coolant /w Spiking Factor b)  % Fuel Damage range is GE 0 to LE 100%

c) Containment Radiation Monitors range is GT O to LE 1E7 R/hr

2) Release Mode a) Leakage range is GT 0 to LE 100%

b) Failure to Isolate c) Catastrophic Failure d) Calculated Containment Leak Rate range is GT 0 cfm

c. RCS Leakage TAB (available for all chosen Pathways except Spent Fuel Pool) is used when the release pathway does not include normal plant effluent monitors (or they are unavailable).

1) Method a) Normal Coolant /w Spiking Factor b)  % Fuel Damage range is GE 1 to LE 100%, used when the Core Damage option is selected in the Source Term Grouping. The amount of core damage should be entered as obtained from TSC Core Thermal Engineer or STA.

2) Release Mode a) Unknown Release Mode b) Calculated RCS Leak Rate range is GE 1 to LT 1E6 gpm

• Attachment 7.1, URI User Guidance

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Title:

Emergency Dose Projection System Operations Page: 23 of 32

d. Release Pt. Sample TAB (available for all chosen Pathways except those directly releasing to the environment) is used when there is a valid sample collected from the effluent release point.

1) Release Point Data, no discernable range in SCFM, enter actual release point flow rate from the time of sample. If SGT is running and Reactor Building Exhaust Fans are shutdown, use the total of the running SGT train flowrates. If building (RB, RW, TB) exhaust flow indication is lost, the Control Room should be contacted to determine flow rate via ODCM 6.1.2.D method.

2) Isotope Box range is GE O to 1E4 uCi/cc, enter the observed concentrations of each identified isotope

e. Field Team TAB (available for all chosen Pathways) is used to estimate the source term and to calculate a complete dose projection based on readings obtained from monitoring teams in the field.

1) Analysis Basis a) Downwind (Miles) range .25 to 10 miles (1.2 miles is SB) b) Exp. Rate (mR/hr) range 0.1 to 99999.9 data must be based on observations in the centerline of the plume

• c) 1-131 Cone. (uCi/cc) range 1E-11 to 1E11 uCi/cc, data must be based on observations in the centerline of the plume. This field is auto filled in when using the Air Sample Calculation. To use the Air Sample Calculation:

(1) Bring up Air Sample Calculation screen by selecting it from the Calculations TAB in the upper left of the URI Detailed screen.

(2) Select Count Rate Meter from the Iodine Cartridge Instrument pulldown menu. *

(3) Enter Background and Cartridge Gross count rate (4) Enter Sample Flow Rate and Sample Collection Time (5) Press Enter and then the "Transfer to Field Team Cale" button (6) The Iodine concentration is transferred to the Field Team Tab for use in the assessment.

d) Survey Date/Time, enter date and time of sample taking

2) Travel Information a) "Travel Time" is derived from the "Downwind (Miles)" input and the Wind Speed used in the Meteorological Data Grouping section. No input required.

b) "Release Time" is derived from "Travel Time" and the date/time used in the "Survey Date/Time" box. No input required .

• END Attachment 7 .1, URI User Guidance

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Title:

Emergency Dose Projection System Operations Page: 24 of 32 Air Sampling Worksheet Calculation

1. Cartridge Filter: AgZ Filter I NOTE: 1.89 x 108 = 0.003 (eff) x 2.83 x 104 cc/ft3 x 2.22 x10 6 dpm/µCi (Sample CPM: - - - - ) - (Background CPM: - - - - ) =Net CPM - - - -

Net CPM

~~~~~~~~~~~-=

µCi I cc I Activity (1.89 x 10 8

) x (sample volume fi 3 ) ----

Particulate Filter 2.

I NOTE: 5.65 x 109 =0.09 (eff) x 2.83 x 104 cc/ft3 x 2.22 x 106 dpm/µCi.

(Sample CPM: - - - - ) - (Background CPM: - - - - ) = Net CPM - - - -

Net CPM

µCi I cc Particulate Activity (5.65 x 10 9

) x (sample volume fi 3 ) ----

END

• Attachment 7.2, Air Sampling Worksheet Calculation

Number: 13.8.1 I Use Category: REFERENCE Major Rev: 037 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 25 of 32 Alternate Method for Obtaining Met Data

1. IF Met Channel A instrument data is not available through the eDNA View "Radiological Status" screen, then the corresponding Met Channel B data should be used. If neither Met Channel A nor Met Channel B data is available, then alternate indications should be obtained from the following sources in the order given: {5.1.4}

• Hanford Internet Site Weather Page (primary alternate):

Select the Hanford weather icon to access the FFTF (400 Area, Station #9) meteorological information via the Internet.

IF the icon is not available, THEN start Internet Explorer and enter the following address:

http://www.hanford.gov/page.cfm/hms When the icon is selected on the desktop, either a Hanford site map or the data for FFTF will be displayed. IF the Hanford site map is displayed, THEN select the 400 Area (Station # 9) to view the FFTF data. (Use Station #11 if #9 is not available and

1. 13 last if necessary. These 3 Stations have delta T, which can be used to determine Stability Class.)

Use the wind speed and direction for the 1O meter height since a ground level release is assumed. Take 60m minus 10m temperature reading and enter into FFTF delta-Tat 10 meter tower selection.

PNNL Weather Forecaster (secondary alternate) at 373-2710 Request wind speed, direction, and differential temperature for the FFTF met tower. If this information is not available from the PNNL forecaster, contact the National Weather Service.

• Telephone the National Weather Service Forecaster (tertiary alternate) at one of the following locations:

1-541-276-8234 Pendleton, Oregon (Primary) 1-206-526-6083 Seattle, Washington (Secondary)

Request the following met data for the Hanford weather station: Wind speed (in MPH),

wind direction, and atmospheric stability. The National Weather Service does not provide a temperature differential. The NWS will describe the stability category as neutral, moderately stable, per step 2a.

• National Weather Service website: http://www.nws.noaa.gov/

• Attachment 7.3, Alternate Method for Obtaining Met Data

l Number: 13. 8.1 Use Category: REFERENCE Major Rev: 037 Minor Rev: N/A

Title:

. Emergency Dose Projection System Operations Page: 26 of 32

2. IF the Stability Class is not shown on the eDNA View "Radiological Status" screen or the POIS Rad Status screen, THEN determine Stability Class as follows:

a. IF the ~T can be obtained from PN H13-P823 Board L- Met System located in the Control Room via the Information Coordinator, THEN input~T into URI to obtain Stability Class.

Note: The following table represents CGS ~ T vs Stability Class.

Main Control Room Board L Operator Aid will require updating if the below table is revised Stability Class vs. Temperature Change for Station Met Tower Temperature Change With NRC Categories Height Stability Classification Stability (°F/212 ft- 75m -10m)

Extremely unstable A ~T $; -2.2

• Moderately unstable Slightly unstable Neutral B

c D

-2.2 < ~T S-2.0

-2.0 <~TS-1.7

-1.7 < ~T S-0.6 Slightly stable E -0.6 <~TS 1.7 Moderately stable F 1.7 <~TS 4.7 Extremely stable G ~T > 4.7

• Attachment 7.3, Alternate Method for Obtaining Met Data

Number: 13.8.1 \ Use Category: REFERENCE Major Rev: 037 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 27 of 32

b. IF no ~Tis available, THEN use the following table to determine Stability Class:

Daytime Solar Radiation (For moderate cloud cover move one column to Nighttime Conditions Surface the right)

Wind Speed Thin (mph) Spring/Fall Heavy < 3/8 Heavy Summer overcast Winter Overcast cloud Overcast Clear Sky Clear Sky (>1/2 cloud Rain cover Rain cover)

<4.5 A B B D F F D to 6.7 A B c D E F D to 11.0 B c c D D E D to 13 B D D D D D D

• > 13 c D D D Table developed using guidance in EPA-454/R-99-005 (2000).

c D D END

• Attachment 7.3, Alternate Method for Obtaining Met Data

• Number: 13.8.1

• I Use Category: REFERENCE Major Rev: 037 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 28 of 32 Computer Points Used In The eDNA View Radiological Status Screen Radiological Status ENW.CGS.TIMET I Reactor Power 0012 % I Met. Channel.A Met. Channel B Delta T META.69 deg F METS69 deg F Stability Clas.s META70 Q:Z24A !v1ET~70 Q2248 Wind Velocity META5a niph MeT85S mph Wind Direction Frqrn META60 deg Az MeTBGO deg Az Rx:Exh x21acfm ARM19 Main CR 0o~2 rnRlhr T0Exh x100 cfm Stack Lo X406cps** tN 1 606 F11el Pooi Q093 mRlhr NG LC)W X409cpm Stack Int X407cps .E"34 606 D/SEP QOSS R/hr NG Mid X394pmu.

Stack Hi Q027 cps NB 572SGTS oosemR/hr Plr1mA X079. rnRlhr

• .. S 6 572ROA Q096rnRlhr .RWExh X3i;;6cfm Plnrn c XO!t4 niR/hr E457'2.CRD Otl94mRJhr NG Low X40Scpm W5572CRD 0095 rnRlhr NG Mid X3!i!3pmu CntmtE X4:32 R/hr NW'"33 501. Entry 0084 Rlhf:

CntmtF X399 R/hr NE 7 501 TIP Q097mRlht CntintNG X46gUCilcc DWPress 0022 psig NE"32 471 Elev#i OOS3 R/hr 1 Part. X470UCi/cc DWTernp 0047 degF W 24 471 Valve q~9mRlhr 2NG xsssuci/cc FDRSUrnJ> 0041 gprn W9422RHR-A Cios1 mRlhr. 2Part X41$UCilcc EDRSump qo39 gpm W 10 422 RHRaB QOBSmR/hr N 11 422 RHR.:.C Q069mRlhr Treatment SGTSA1 X4G6 cfrn S23422CRD oossmRlhr Pre-Treat X073 mRlhr SGTSA2 X452 cfrii N12422RCIC 0.090 mRlhr *Post~Treat xoss uCifcc SGTSB1 Xa513cfm E 13422 HPCS 0091 mRlhr SGTSB2 X371 cfm END Attachment 7.4, Computer Points Used in the eDNA View Radiological Status Screen

IUse Category: REFERENCE Number: 13.8.1 Major Rev: 037 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 29 of 32 Alternate Method for PRM-RE-1B and PRM-RE-1C A portable radiation survey meter or an installed radiation monitoring instrument measuring radiation levels at the exterior center of the Elevated Release Stack Access Door (R-DOOR-R515) provides an alternate means of monitoring radioactive effluent releases through the Reactor Building Stack when the installed post-accident effluent monitors PRM-RE-1 B and PRM-RE-1 C are unavailable. The radiation level readings (in mR/hr) from this Stack Door Monitor (SOM) can be converted to an equivalent PRM-RE-1 B (RB Stack Intermediate Range) or PRM-RE-1C (RB Stack High Range) reading (in cps) that can be input into the Monitor Reading field in URI to perform a dose projection.

The technical basis for this SOM to PRM equivalency can be found in Engineering Calculation NE 09-12. The use of the table may be used for Table 3 equivalency if needed.

Data from eDNA Points EP99M (mid-range SOM, 5 mR/hr to 1E6 mR/hr) and EP99H (high-range SOM, 1000 mR/hrto 1E7 mR/hr) is updated approximately every minute and is displayed in mR/hr.

Always round to the nearest whole number (e.g., 23.645 mR/hr should be rounded to 24 mR/hr).

• When Stack Door Monitor net reading is (mR/hr) 8 Table 3 Equivalent (cps) 3000 (B)

Action Notes (based on filtered releases)

PPM 13.1.1A Table 3 Unusual Event EAL threshold*

12, 100 400 (C) PPM 13.1.1A Table 3 Alert EAL threshold*

PPM 13.1.1A Table 3 Site Area Emergency EAL 1.14E5 2,000 (C) threshold*

1.14E6 20,000 (C) PPM 13.1.1A Table 3 General Emergency EAL threshold*

• If URI is not available.

END

• Attachment 7.5, Alternate Method for PRM-RE-18 and PRM-RE-1C

IUse Category: REFERENCE Number: 13.8.1 Major Rev: 037 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 30 of 32 Alternate Method for TEA-RIS-13 or WEA-RIS-14 Compensatory Noble Gas Monitor (CNGM) sample cart(s) will be available to measure' radiation levels as a compensatory measure when the normal sample racks for TEA-RIS-13 and/or WEA-RIS-14 are unavailable. There is one cart for TEA, and a different cart for WEA readings. While the equations below apply directly to the readings from either cart, it is critical to ensure that the WEA cart readings are only used for WEA dose assessments and TEA readings be used for TEA dose assessments. IF both WEA and TEA are non-functional, then both carts will be in operation and the dose assessments must be derived from the respective cart.

The technical basis for this compensatory measure equivalency can be found in RSCS TSD-13-035 (CVI 1057-00, 131). The conversion equations below are used for calculating equivalent inputs for URI. Data from the CNGM will be available to the Main Control Room for monitoring a release during this compensatory measure. There is a slide at the frisker which is inserted for normal readings LTE 400,000 cpm and withdrawn for readings GT 400,000 cpm.

1) !E URI is unavailable, THEN use Table 3 of PPM 13.1.1A to classify.

2) IF frisker reading is GT 400,000 cpm, THEN direct Chemistry to withdraw the CNGM slide.

3) If calculating TEA values to input into URI, then use the following conversion factors, otherwise N/A this step:

• !E TEA frisker reading is LTE 400,000 cpm, slide inserted (Low Range Input) THEN TEA URI INPUT (µCi/cc)= [Frisker output (cpm)] / 1.01E7 cpm/µCi/cc Example: the frisker reading obtained from the sample cart is 80,000 cpm. The equivalent URI input for either TEA or WEA would be:

TEA 80000 cpm divided by 1.01 E7 cpm/µCi/cc =7.92e-3 µCi/cc Enter this value into URI for the TEA monitor.

• !E frisker slide is withdrawn, THEN TEA URI INPUT (µCi/cc)= [Frisker output (cpm)] I 3.87E6 cpm/µCi/cc

4) If calculating WEA values to input into URI, then use the following conversion factors, otherwise N/A this step:

• !E WEA frisker reading is LTE 400,000 cpm, slide inserted (Low Range Input) THEN WEA URI INPUT (µCi/cc)= [Frisker output (cpm)] / 1.01E7 cpm/µCi/cc

• !E frisker slide is withdrawn, THEN URI INPUT (µCi/cc)= [Frisker output (cpm)] / 3.98E6 cpm/µCi/cc END Attachment 7.6, Alternate Method for TEA-RIS-13 or WEA-RIS-14

Number: 13.8.1 I Use Category: REFERENCE Major Rev: 037 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 31 of 32 Alternate Method for ARMs or OG-RIS-612 This attachment will provide direction to establish and maintain alternate method(s) for one or more inoperable ARMs or OG-RIS-612.

The following ARMs are used for EAL classification per PPM 13.1.1A: ARM-RIS-1, ARM-RIS-2, ARM-RIS-4 thru 13, ARM-RIS-19, ARM-RIS-23, ARM-RIS-24, ARM-RIS-30, ARM-RIS-32 thru 34. HP will install the alternate methods, in accordance with the requirements of PPM 11.2.9.32 and the tables within this instruction. One disabled ARM power supply can disable multiple ARMs.

OG-RIS-612 is used for EAL classification per PPM 13.1.1A. HP will establish the alternate method, in accordance with ODCM 6.1.2.1.

1.0 PORTABLE ARMS:

The wireless remote radiation monitoring devices to be used are PAM-TRX and AMP-100.

NOTE: Wireless Radiation monitors cannot be used in the Main Control Room. In the event ARM-RIS-19 is disabled, notify HP to install a non-transmitting portable area radiation monitor with an alarm set-point of 15 mr/hr.

NOTE: In the event of unplanned ARM failures, an hourly H.P. tour of the affected areas should be established until the measures of this attachment are established .

2.0 CONTROL ROOM PERSONNEL:

2.1 Direct HP to establish the applicable compensatory measures as per this attachment.

2.2 During this evolution, area radiation levels will be provided as follows:

• HP will have established wireless remote monitors into each of the areas affected, in accordance with the requirements of PPM 11.2.9.32. These monitors will provide indication to a remote monitor in the HP access area.

• In the event an area radiation level meet or exceed an ARM set-point identified by PPM 4.602.AS, the Control Room will be notified immediately. These indications should be treated as actual ARM indications. Consider directing continuous monitoring. The Control Room will be notified immediately of instrument failure.

• In the event of an emergency classification of ALERT or higher, this information will be available by request of the TSC. Continuous monitoring is not required.

• In the event that OG-RIS-612 has been replaced with an area radiation monitor, EAL classification can be made based on area radiation monitor indication that is equivalent to OG Pretreatment Hi-Hi alarm. Refer to ODCM 6.1.2.1 .

• Attachment 7.7, Alternate Method for ARMS or OG-RIS-612

• Number: 13.8.1 I Use Category: REFERENCE Major Rev: 037 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 32 of 32 3.0 HP DEPARTMENT STAFF:

3.1 For ARMs, establish wireless remote monitors, in accordance with the requirements of PPM 11.2.9.32, in each of the affected areas, as directed by the Control Room.

3.2 For OG-RIS-612, establish the compensatory measures, in accordance with ODCM 6.1.2.1.

3.3 Ensure a knowledgeable individual is stationed at a monitor to observe the read-outs from each of the wireless remote monitors. Also ensure telephone communication is established with the Control Room.

3.4 Provide direction to the knowledgeable individual to ensure the following is clearly understood:

• The radiation indications to be monitored.

• Until released by the Control Room, remain available. Check the remote monitor radiation levels at least every 15 minutes. This may be more frequent or continuous, as directed by the Control Room.

o Notify the Control Room if radiation levels on any monitor increase above an ARM set-point or 1O R/hr.

o Provide the detector locations and the radiation levels observed.

o Make an HP log entry of the notification and information transmitted .

o If a monitor indication fails (excluding "LOST CONTACT" for less than 10 minutes),

IMMEDIATELY notify the Control Room.

4.0 ARMs by Power Supply ARMs 1 thru 10 powered by ARM-E/S-603A ARMs 11 thru 20 powered by ARM-E/S-6038 ARMs 21 thru 30 powered by ARM-E/S-603C ARMs 32 thru 34 powered by E-CP-H13/P614 END

• Attachment 7.7, Alternate Method for ARMS or OG-RIS-612