ML993360177

Bowns Ferry Nuclear Plant (BFN) - Units 1, 2, and 3 Emergency Plan Implementing Procedure (EPIP) Revisions

ML993360177
Person / Time
Site:	Browns Ferry
Issue date:	11/15/1999
From:	Abney T Tennessee Valley Authority
To:	NRC/OCIO/IMD/RMB
References
-RFPFR
Download: ML993360177 (239)

Text

Tennessee Valley Authority, Post Office Box 2000, Decatur, Alabama 35609-2000 November 15, 1999 10 CFR Part 50, APP.E U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Gentleman:

In the Matter of Docket Nos. 50-259 Tennessee Valley Authority 50-260 50-296 BROWNS FERRY NUCLEAR PLANT (BFN) - UNITS 1, 2, and 3 EMERGENCY PLAN IMPLEMENTING PROCEDURE (EPIP) REVISIONS TVA is submitting this notification in accordance with the requirements of 10 CFR Part 50, Appendix E, Section V, to provide NRC with the following EPIP revisions: (1) EPIP Index, (2) EPIP-1, Revision 28, (3) EPIP-13, Revision 7, and(4) EPIP-14, Revision 14. The revision date for these EPIPs is November 4, 1999.

The enclosed information is being sent by certified mail.

The signed receipt signifies that you have received this information. TVA requests that the NRC copy of the EPIPs be updated, and the superseded material be destroyed by December 3, 1999.

A .g5 I C) C

'D "'

A 4"' r0l11 t9 0- Printed on recycled paper

U.S. Nuclear Regulatory Commission Page 2 November 15, 1999 If you have any questions, please telephone me at (256) 729-2636.

Enclobures cc (En sures NRC Resident Inspector (Enclosures provided by Browns Ferry Nuclear Plant BFN Document Control 10833 Shaw Road Unit)

Athens, Alabama 35611 Mr. P. E. Fredrickson (2 Enclosures)

U.S. Nuclear Regulatory Commission Region II 61 Forsyth Street S.W.

Suite 23T85 Atlanta, Georgia 30303 Mr. W. 0. Long, Project Manager (w/o Enclosures)

U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 208!

ENCLOSURES TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, AND 3 EMERGENCY PLAN IMPLEMENTING PROCEDURES (EPIP)

EPIPs -1, 13, and 14 SEE ATTACHED

GENERAL REVISIONS GENERIC FILING INSTRUCTIONS FILE DOCUMENTS AS FOLLOWS:

PAGES TO BE REMOVED PAGES TO BE INSERTED EPIP INDEX (ALL) EPIP INDEX (ALL)

EPIP-1 REVISION 27 (ALL) EPIP-1 REVISION 28 (ALL)

EPIP-13 REVISION 6 (ALL) EPIP-13 REVISION 7 (ALL)

EPIP-14 REVISION 13 (ALL) EPIP-14 REVISION 14 (ALL)

Browns Ferry Nuclear Pianit Page 1 Curator Procedure DOCUIr Status=ACTIVE ProcT7 Sorted by Typ,

?I)

SC)REEN P

pgl02 11/04/99 Unit Proc Type Proc Number Title Doc Sta Group Section Remarks Reason Chg Effect Dt Rev Lvl 0 EPIP EPIP-1 EMERGENCY PLAN ACTIVE OTHER REP TONY FELTMAN Performers 11/04/1999 028 100 CLASSIFICATION LOGIC 3666 - ISSUE Change PER T. FELTMAN Request 0 EPIP EPIP-10 MEDICAL EMERGENCY ACTIVE OTHER REP Performers 07/09/1998 019 100 PROCEDURE Change Request 0 EPIP EPIP-11 SECURITY AND ACCESS ACTIVE OTHER REP 09/15/1995 007 100 CONTROL 0 EPIP EPIP-13 RADIOCHEMICAL LABORATORY ACTIVE OTHER REP Performers 11/04/1999 007 100 PROCEDURE Change Request 0 EPIP EPIP-14 RADIOLOGICAL CONTROL ACTIVE OTHER REP Performers 11/04/1999 014 100 PROCEDURES Change Request 0 EPIP EPIP-14/AUDI RADIOLOGICAL CONTROL ACTIVE OTHER REP Performers 11/04/1999 014 100 T PROCEDURES Change Request 0 EPIP EPIP-14/PED RADIOLOGICAL CONTROL ACTIVE OTHER REP Performers 08/05/1999 014 100 PROCEDURES Change Request 0 EPIP EPIP-15 EMERGENCY EXPOSURE ACTIVE OTHER REP 06/26/1996 005 100 0 EPIP EPIP-16 TERMINATION AND RECOVEPY ACTIVE OTHER REP 11/01/1995 003 100 PROCEDURE 0 EPIP EPIP-17 EMERGENCY EQUIPMENT AND ACTIVE OTHER REP PER PORC, PAGE Performers 07/14/1998 022 100 SUPPLIES (INVENTORY AND 15 CHANGED Change OPERABILITY PROCEDURE) SUPER. RADCON Request TO SUPER.

FIREPROTECTION 0 EPIP EPIP-2 NOTIFICATION OF UNUSUAL ACTIVE OTHER REP TONY FELTMAN Performers 05/12/1997 020 100 EVENT EXT. 3666 Change Request 0 EPIP EPIP-20 PLANT DATA ACTIVE OTHER REP 06/14/1996 009 100

Browns Ferry Nuclear Plant Page 2 Curator Procedure DOCU SCREEN pgi 02 Status=ACTIVE ProcT ?IP 11/04/99 (

Sorted by Typj Unit Proc Type Proc Number Title Doc Sta Group Section Remarks Reason Chg Effect Dt Rev Lv1 0 EPIP EPIP-21 FIRE EMERGENCY PROCEDURE ACTIVE OTHER REP PER 06/25/1997 002 100 Corrective Action 0 EPIP EPIP-3 ALERT ACTIVE OTHER REP TONY FELTMAN, Performers 05/12/1997 023 100 3666, 3/28/97 Change Request 0 EPIP EPIP-4 SITE AREA EMERGENCY ACTIVE OTHER REP TONY FELTMAN, Performers 05/12/1997 022 100 X3666, Change Request 0 EPIP EPIP-5 GENERAL EMERGENCY ACTIVE OTHER REP TONY FELTMAN, Performers 05/01/1999 027 100 X3666 Change Request.

0 EPIP EPIP-6 ACTIVATION AND OPERATION ACTIVE OTHER REP TONY FELTMAN, Performers 10/07/1999 019 100 OF THE TECHNICAL SUPPORT X3666 Change CENTER (TSC) Request 0 EPIP EPIP-7 ACTIVATION AND OPERATION ACTIVE OTHER REP Tony Feltman @ Performers 11/20/1996 017 100 OF THE OPERATIONS SUPPORT ext. 3666 Change CENTER (OSC) Request 0 EPIP EPIP-8 PERSONNEL ACCOUNTABILITY ACTIVE OTHER REP Performers 11/17/1997 010 100 AND EVACUATION Change Request Total records selected: 19

• *

• E N D O F R E P O R T * *

• TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURE EPIP-1 EMERGENCY CLASSIFICATION PROCEDURE REVISION 28 PREPARED BY: TONY FELTMAN PHONE: 3666 RESPONSIBLE ORGANIZATION: EMERGENCY PREPAREDNESS APPROVED BY: GILBERT V. LITTLE DATE: 11/03/99 EFFECTIVE DATE: 11/04/99 LEVEL OF USE: REFERENCE USE VALIDATION DATE: NOT REQUIRED QUALITY-RELATED

REVISION LOG Procedure Number: EPIP-1 Revision Number 28 Pages Affected: 25,31,53,170,172 Description of Change:

IC REVISION CHANGES TABLE 3.1 MAXIMUM SAFE OPERATING AREA TEMPERATURE LIMITS TO COINCIDE WITH APPLICABLE BFN-EOI'S.

IC GENERAL REVISION REGARDING THE CONVERSION OF EPIP 1 FROM PAGEMAKER TO MS WORD FORMAT FOR ENTRY INTO CURATOR. OTHER REVISION CHANGES WERE INCORPERATE THE STANDARD TECHNICAL SPECIFICATION INPLEMENTATION. CONDUCTED TO PRIMARILY CHANGES INVOLVED THE MODE OF OPERATIONS AND NEW REFERENCES.

EC 30 - During conversion of EPIP 1, one typographical problem was discovered.

EAL 4.I.S should be written 4.2.S onpage 35. EAL 8.1-S should be written 8.2-S onpage 69.

IC EPIP 1 is being revised to incorporate changes to graphical information resulting from the Unit 3 power up-grade project and revisions in the EOI associated curves and tables. Additionally one typographical mistake was corrected on page 47.

IC EPIP 1 is being revised to incorporate changes made to the EOI's per the EPG SAG Guidelines, Rev. I as implemented by the BWR owners group.

IC-33 (2/99) Procedure is being revised to incorporate changes made to engineering calculations that support this procedure. This revision deals with those calculations that primarily affect drywell radiation values. This revision will additional be used to complete PER 99-001406-000. This revision will also update calculation reference numbers.

EC-34 (2/99) Changed a typographical error (page 25).

IC-35 (4/99) Procedure is being revised to incorporate changes made to engineering calculations that support this procedure. This revisions deals with changes effecting drywell radiation values due to upgrades to Unit 2 instrumentation and others effects regarding the U2C10 Refueling Outage.

IC-36 (10/99) Revision is being conducted to correct reference regarding toxic gases.

The change involves changing Lower Toxicity Limit to Permissible Exposure Limit.

EPIP-1 EMERGENCY CLASSIFICATION PROCEDURE TABLE OF CONTENTS PAGE NUMBER TABLE OF CONTENTS ........................................ 1 SECTION I INTRODUCTION CLASSIFICATION INSTRUCTIONS ........................................ 3 GLOSSARY..................................................................................................................5 EVENT CLASSIFICATION INDEX ........................................ 11 SECTION II EVENT CLASSIFICATION MATRIX 1.0 REACTOR ....................................... 13 2.0 PRIMARY CONTAINMENT ....................................... 21 3.0 SECONDARY CONTAINMENT ....................................... 29 4.0 RADIOACTIVITY RELEASES ....................................... 33 5.0 LOSS OF POWER ....................................... 39 6.0 HAZARDS ......... 45 7.0 NATURAL EVENTS ....................................... 61 8.0 EMERGENCY DIRECTOR JUDGEMENT ....................................... 67 SECTION III BASIS 1.0 REACTOR ................................. 75 2.0 PRIMARY CONTAINMENT................................................................................

.97 3.0 SECONDARY CONTAINMENT ................................. 116 4.0 RADIOACTIVITY RELEASE ................................. 126 5.0 LOSS OF POWER ................................. 139 6.0 HAZARDS ................................. 155 7.0 NATURAL EVENTS ................................. 183 8.0 EMERGENCY DIRECTOR JUDGEMENT ................................. 190 PAGE 1 OF 207 REVISION 28

EPIP-1 EMERGENCY CLASSIFICATION PROCEDURE M

THIS PAGE INTENTIONALLY BLANK REVISION 28 PAGE 2 OF 207

EMERGENCY EPIP-1 CLASSIFICATION SECTION I CLASSIFICATION PROCEDURE INTRODUCTION INSTRUCTIONS CLASSIFICATION INSTRUCTIONS 1.0 PURPOSE Provide guidance to the Shift Manager or Site Emergency Director (SED) for proper declaration and classification of emergencies and ensure emergency classifications are consistent with those used by state and local governments and the Nuclear Regulatory Commission (NRC).

2.0 SCOPE This procedure applies to site events that constitute an emergency consistent with those site specific events outlined in NUMARC/NESP-007 August 1992. The Shift Manager and the SED are the only persons authorized to make the emergency classification determination.

3.0 INSTRUCTION 3.1 Following plant events or transients, review EPIP-1 Section II, 1.0 through 8.0 and determine if an event should be classified as an emergency.

Note: (1) If an emergency action level for a higher classification was exceeded, but the present situation indicates a lower classification, the fact that the higher classification occurred shall be reported to the NRC and the CECC, if staffed, or ODS if the CECC is not staffed. The higher classification should not be declared.

(2) If an emergency action level was met but the emergency has been totally resolved, the emergency'class that was appropriate shall be reported to the ODS and the NRC but should not be declared.

3.1.1 EPIP-1 Section II, 1.0 through 8.0 captures events in eight major categories as listed on the event classification index.

3.1.2 Each actual condition in a category is given a numeric designator indicating the section followed by the numeric designator for the specific EAL within the section and an alpha numeric designator for the event class.

Example: 5.2-U These designators provide for cross-reference between the specific EALs and the basis document which provides technical supporting information for the EALs and may aid the Shift Manager/SED in classifying events.

PAGE 3 OF 207 REVISION 28

EPIP-1 EMERGENCY CLASSIFICATION SECTION I CLASSIFICATION INSTRUCTIONS INTRODUCTION PROCEDURE (3.1 Continued)

Notes, curves, or tables contained in the Event Classification Matrix, are identified by a flag in the event classification window. The window contains an appropriate symbol to alert the user that a corresponding note, curve, or table applies to the step.

Example: K K] K Notes, curves, or tables contained in the Event Classification Matrix, that arp identified by a flag that contains an asterisk shall alert the user that the corresponding note, curve, or table is unit specific. The user must insure that the information being applied to the EAL is associated with the applicable unit.

Example: Kt K]

3.2 If the event is determined to be one of the four emergency classifications, the Shift Manager assumes the responsibility of SED until relieved by the Plant Manager or designee.

3.2.1 Implement one of the following procedures as applicable:

EPIP-2 Notification of Unusual Event EPIP-3 Alert EPIP4 Site Area Emergency EPIP-5 General Emergency 3.2.2 Continue to review the emergency conditions in the event classification matrix and escalate, terminate, or implement recovery as appropriate. Refer to EPIP-16 for termination or recovery.

3.3 If the event is determined not to be one of the four event classifications, continue to monitor plant conditions for possible changes that could result in reaching an event classification.

END OF TEXT REVISION 28 PAGE 4 OF 207 CLASSIFICATION INSTRUCTIONS

EMERGENCY EPI[P-1 CLASSIFICATION SECTION I PROCEDURE INTRODUCTION GLOSSARY ABBREVIATIONS, ACRONYMS, AND DEFINITIONS The following is a list of terms and phrases found in EPIP-1. Each term or phrase is provided with a meaning, to ensure consistent use and understanding.

TERMIPHRASE MEANING/DEFINITION ADS Automatic Depressurization System AO0 Abnormal Operating Instruction ARI Alternate Rod Insertion ARM Area Radiation Monitor ARP Alarm Response Procedure ATWS Anticipated Transient Without Scram Auto Automatic Bomb An explosive device BWR Boiling Water Reactor Can/Cannot be determined The current value or status of an identified parameter relative to that specified in the instruction can/cannot be ascertained using all available indications (direct and indirect, singly or in combination).

Can/Cannot be Maintained The value of the identified parameter(s) is/is not able to be kept Above/Below above/below specified limits. This definition includes making an evaluation that considers both current and future system performance in relation to the current value and trend of the parameter(s). "Cannot" does not imply that the actual value of the parameter must first exceed the specified limit.

Can/Cannot be Restored The value of the identified parameter(s) is/is not able to be returned to Above/Below above/below specified limits within a reasonable time after having exceeded the specified limits. This determination includes making an evaluation that considers both current and future system performance in relation to the current value and trend of the parameter(s).

GLOSSARY PAGE 5 OF 207 REVISION 28

EPIP-1 EMERGENCY SECTION I CLASSIFICATION GLOSSARY INTRODUCTION [ PROCnlT.DrjF Pflneirnrini' (CONTINUED)

TERM/PHRASE MEANING/DEFINITION CAD Containment Atmosphere Dilution CAS Central Alarm Station CDE Committed Dose Equivalent CECC Central Emergency Control Center Ci Curie Civil Disturbance A group of 20 or more persons violently protesting station operations or activities at the site.

cm, Cubic Centimeters CS Core Spray deg Degrees DG Diesel Generator Drywell The upper portion of the Primary Containment which encloses the Reactor Pressure Vessel.

EAL Emergency Action Level ECCS Emergency Core Cooling System ECL Effluent Concentration Limit EPA Environmental Protection Agency EPIP Emergency Plan Implementing Procedure EQ Environmental Qualification Event Assessment of an EVENT commences when recognition is made that one or more of the conditions associated with the event exists. Implicit in this definition is the need for timely assessment, i.e. within 15 minutes.

REVISION 28 PAGE 6 OF 207 GLOSSARY

EMERGENCY EPIP-1 CLASSIFICATION SECTION I PROCEDURE INTRODUCTION GLOSSARY M

(CONTINUED)

TERMIPHRASE MEANING/DEFINITION Explosion A rapid, violent, unconfined combustion, or a catastrophic failure of pressurized equipment that imparts energy of sufficient force to potentially damage permanent structures required for safe operation.

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

Flammable Gas Combustible gasses maintained at concentrations less than the lower explosive limit. Will not explode due to ignition.

GOI General Operating Instruction gm Gram HCLL Heat Capacity Level Limit HCTL Heat Capacity Temperature Limit Hostage A person(s) held as leverage against the station to ensure that demands will be met by the station.

HPCI High Pressure Coolant Injection HR Hour IN. Inches KV Kilovolt LCO Limiting Condition for Operation LOCA Loss Of Coolant Accident LPCI Low Pressure Coolant Injection MRFP Minimum RPV Flooding Pressure GLOSSARY PAGE 7 OF 207 REVISION 28

EPIP-1 EMERGENCY SECTION I CLASSIFICATION GLOSSARY INTRODUCTION PRnrrnTT~r~

(CONTINUED)

TERM/PHRASE MEANING/DEFINITION MCUTL Maximum Core Uncovery Time Limit MIN Minute MPH Miles per Hour mrem Millirem MSIV Main Steam Isolation Valve MSL Main Steam Line MSRV Main Steam Relief Valve NESP National Environmental Studies Project NUMARC Nuclear Management and Resources Council ODS Operations Duty Specialist 0I Operating Instruction OSC Operations Support Center PCIS Primary Containment Isolation System Primary Containment The drywell, the vent system, and the torus.

Primary System Primary systems comprise the pipes, valves and other equipment connected to the RPV such that a reduction in RPV pressure will effect a decrease in the flow of steam or water being discharged through an unisolable break in the system.

Projectile An object ejected, thrown, or launched towards a plant structure. The source of a projectile may be offsite or onsite. Damage is sufficient to cause concern regarding the integrity of the affected structure or the operability or reliability of safety equipment contained therein.

Protected Area Protected Area encompasses all areas within the security protected area fence.

REVISION 28 PAGE 8 OF 207 GLOSSARY

EMERGENCY EPIP-1 CLASSIFICATION SECTION I PROCEDURE INTRnODFir'rnW noy ncc A "X TN'r~nnTI-T~nW J'T CC'AMWI7 (CONTINUED)

TERM/PHRASE MEANING/DEFINITION PSIG Pounds Per Square Inch Gauge R Rad RCIC Reactor Core Isolation Cooling RCS Reactor Coolant System REP Radiological Emergency Plan RHR Residual Heat Removal RPS Reactor Protection System RPV Reactor Pressure Vessel Sabotage Deliberate damage, misalignment, misoperation of plant equipment with the intent to render equipment inoperable.

SEC Second Secondary The spaces immediately adjacent to, or surrounding, the primary Containment containment from which the Reactor Building Ventilation System and the Standby Gas Treatment System provides a filtered elevated release.

SED Site Emergency Director SGTS Standby Gas Treatment System Significant Transient An unplanned event involving one ormnore of the following: (1) Electrical load reduction of greater than 25% rated: (2) Reactor scram: (3) Valid ECCS initiation.

Si Surveillance Instruction Site Boundary The Site Boundary is that line beyond which the land or property is not owned, leased, or otherwise controlled by TVA.

GLOSSARY PAGE 9 OF 207 REVISION 28

EPIP-1 EMERGENCY SECTION I CLASSIFICATION GLOSSARY INTRODUCTION PROCEDUIRE PRCF)TR1 (CONTINUED)

TERM/PHRASE MEANING/DEFINITION Subcritical Reactor power below the heating range and not trending upward.

Suppression Pool The water volume contained in the suppression chamber intended to condense steam from a primary system break inside the drywell.

Suppression Chamber The structure enclosing the suppression pool water and the atmosphere above it.

TAF Top of Active Fuel TEDE Total Effective Dose Equivalent Torus The lower portion of the primary containment which encloses the suppression pool.

Toxic Gas A gas that is dangerous to life or limb by reason of inhalation or skin contact.

TSC Technical Support Center Valid An indication, report, or condition is considered to be valid when it is conclusively verified by redundant indicators or actual observation by plant personnel.

Visible Damage Damage to equipment that is readily observable without measurements, testing, or analysis. Damage is sufficient enough to cause concern regarding the continued operability or reliability of affected safety structure, system, or component.

Vital Area Any area designated as a vital area in the site Physical Security Plan.

yr Year END OF TEXT REVISION 28 PAGE 10 OF 207 GLOSSARY

EMERGENCY EPIP-1 CLASSIFICATION SECTION I CLASSIFICATION PROCEDURE INTRODUCTION INDEX INDEX EVENT CLASSIFICATION INDEX SECTION 1.0 REACTOR 1.1 WATER LEVEL 1.2 SCRAM FAILURE 1.3 REACTOR COOLANT ACTIVITY 1.4 MSL/OFFGAS RADIATION 1.5 LOSS OF DECAY HEAT REMOVAL SECTION 2.0 PRIMARY 2.1 PRIMARY CONTAINMENT PRESSURE CONTAINMENT 2.2 PRIMARY CONTAINMENT HYDROGEN 2.3 DRYWELL RADIATION 2.4 DRYWELL INTERNAL LEAKAGE 2.5 LOSS OF PRIMARY CONTAINMENT SECTION 3.0 SECONDARY 3.1 SECONDARY CONTAINMENT CONTAINMENT TEMPERATURE 3.2 SECONDARY CONTAINMENT RADIATION SECTION 4.0 RADIOACTIVITY 4.1 GASEOUS EFFLUENT RELEASES 4.2 MAIN STEAM LINE BREAK 4.3 LIQUID EFFLUENT SECTION 5.0 LOSS OF POWER 5.1 LOSS OF AC POWER 5.2 LOSS OF DC POWER SECTION 6.0 HAZARDS 6.1 RADIOLOGICAL 6.2 CONTROL ROOM EVACUATION 6.3 TURBINE FAILURE 6.4 FIRE/EXPLOSION 6.5 TOXIC GASES 6.6 FLAMMABLE GASES 6.7 SECURITY 6.8 VEHICLE CRASH SECTION 7.0 NATURAL EVENTS 7.1 EARTHQUAKE 7.2 TORNADO/HIGH WINDS 7.3 FLOOD SECTION 8.0 EMERGENCY 8.1 TECHNICAL SPECIFICATIONS DIRECTOR 8.2 LOSS OF COMMUNICATION JUDGEMENT 8.3 LOSS OF ASSESSMENT CAPABILITY 8.4 OTHER CLASSIFICATION PAGE 11 OF 207 REVISION 28 INDEX

EPIP-1 EMERGENCY SECTION I CLASSIFICATION INTrRODIjTSTTnl

_YTmvUZl _l c THIS PAGE INTENTIONALLY BLANK REVISION 28 PAGE 12 OF 207

EMERGENCY EPIP-1 CLASSIFICATION SECTION II PROCEDURE ENVENT CLASSIFICATION MATRIX 1.0 REACTOR REACTOR 1.0 1.0 REACTOR PAGE 13 OF 207 REVISION 28

EPIP-1 EMERGENCY SECTION II CLASSIFICATION 1;0 REACTOR -EVENT

~ --- CL.AR.RtAT~rnw %4Arwry

-~A LI rnVILMIJUKE NOTES:

1.l-Ul/1.1-Al Applicable when the Reactor Head is removed and the Reactor Cavity is flooded.

1.1-Si Applicable in Mode 5 when the Reactor Head is installed.

1.1-G2 The reactor will remain subcritical under all conditions without boron when:

• All control rods are inserted to or beyond position 02

• All control rods except one are inserted to or beyond position 00

• Determined by reactor engineering CURVES/TABLES:

REVISION 28 PAGE 14 OF 207 1.0 REACTOR

EMERGENCY EPIP-1 CLASSIFICATION SECTION II PROCEDURE EVENT CLASSIFICATION MATRIX 1.0 REACTOR a EV_

DESCRIPTION _ .

DESCRIPTION I 1.1-Ul h Uncontrolled water level decrease in Reactor Cavity Uncontrolled water level decrease in Spent Fuel Pool with irradiated fuel assemblies expected to remain with irradiated fuel assemblies expected to remain covered by water. covered by water.

riA OPERATING CONDmON: OPERATING CONDmON:

- Mode 5 - All

_ .I l 1.1-Al 1.1-A2l Uncontrolled water level decrease in Reactor Cavity Uncontrolled water level decrease in Spent Fuel Pool expected to result in irradiated Fuel assemblies being expected to result in irradiated fuel assemblies being uncovered. uncovered.

OPERATING CONDITION: OPERATING CONDmON:

- Mode 5 - All I~I-S1 IJ1.1-5 Reactor water level CANNOT be maintained above Reactor water level CANNOT be determined.

-162 IN.

OPERATING CONDITION: OPERATING CONDITION:

-All -Mode 1 -Mode 3

-Mode2 l~l-Gl l l 1.1-G2 0 1_

Reactor water level CANNOT be restored and Reactor water level CANNOT be determined maintained above -190 IN. AND Z EITHER of the following conditions exists:

eThe reactor will reain subcritical w1o boron under all conditions and Less than 4 MSRVs can be opened, or Reactor pressure CANNOT be restored and maintained at least 65 PSI M above Suppxession ChamberpressurM.

• It has NOT been determined that the reactor will emai subcritical w/o boron under all conditions and ble to estoe and maintain MARFP in Table 1.I-G2.

OPERATING CONDmON: OPERATING CONDITION:

-Mode 1 -Mode 3 -Mode I -Mode3

- Mode 2 - Mode 2 1.0 REACTOR PAGE 15 OF 207 REVISION 28

EPIP-1 EMERGENCY SECTION II CLASSIFICATION 1.0 REACTOR EVENT CLASSIFICATION MATR1Y D~PRV"{TT1li NOTES:

1.2 Subcritical is defined as Reactor power below the heating range and not trending upward.

CURVES/TABLES:

CURVE 1.2-G HEAT CAPACITY TEMP LIMIT 2250 .L.=

.i^

... ..I SFE WHEN RX PREsSSE¢

0. PV Press. 700 ...

W 210 R5 300 F- 1 8 0 .**. . .**... ................

150 SUPPR PL LVL (FT) ggACTION REQUIRED IFABOVE CURVE FOR EXISTING RX-REVISION 28 PAGE 16 OF 207 1.0 REACTOR

EMERGENCY EPIP-1 CLASSIFICATION SECTION II PROCEDURE EVENT CLASSIFICATION MATRIX 1.0 REACTOR

~~fUMAI UWAU RE NUWUK]kISS ~~

DESCRIPTION DESCRIPTION i ________________ I 1.3-Ul Reactor coolant activity exceeds 26 flCi/gm dose -4 equivalent I-131 (Technical Specification Limit) as cl!i rA determined by chemistry sample.

11 M

4 4

OPERATING CONDmON: *q

- ALL

_ _ _ _ _ _ II .l 1.2-A I 1.3-A Failure of automatic scram functions to bring the Reactor coolant activity exceeds 300 pCi/gm dose Reactor subcritical equivalent Iodine-131 as determined by chemistry AND sample.

Manual scram or ARI was successful.

OPERATING CONDmON: OPERATING CONDITION:

- Mode I -Mode I - Mode 3

- Mode 2 - Mode 2

_ I 1.2-S I CA Failure of automatic scram, manual scram, and ARI to 03 bring the Reactor subcritical.

M C)

OPERATING CONDITION:

- Mode I I.

1.2-G IN Failure of automatic scram, manual scram, and ARI.

Reactor power >3% z AND EITHER of the following conditions exists:

• Suppression Pool temp exceeds HCTL.

Refer to Curve 1.2-G.

• Reactor water level CANNOT be restored and maintained at or above -190 IN.

M OPERATING CONDITION:

- Mode I I

1.0 REACTOR PAGE 17 OF 207 REVISION 28

EPIP-1 EMERGENCY SECTION II CLASSIFICATION 1.0 REACTOR EVENT CLASSRIRCATTMN MAThID DDnrl'%4lMTT~lM

~~ AV% %*savnmlUK 1

NOTES:

CURVES/TABLES:

CURVE 1.5-S HEAT CAPACITY TEMP LIMIT 240IS BELOW 65 PSIG 230 210... ... . ..... ...... ..... ........... .. .....

220PP _ U L (FT) 240 Pres.....

2 I R0 FA C F S a_ 200- PV Prss 5 190 'V Press. 70S0E o150 R Prs. ..1 11.5 12 12.5 13 13.5 14 14.5' 15 15.5 16 16.5 17. 17,5 18 18.5 19 SUPPIR PL LVL (FT)

ACTION REQUIRED IFABOVE CURVE FOR EXISTING RX REVISION 28 PAGE 18 OF 207 1.0 REACTOR

EMERGENCY EPIP-1 CLASSIFICATION SECTION II PROCEDURE EVENT CLASSIFICATION MATRIX 1.0 REACTOR I ~ ~ ~ .-.

I_ LIV E I I DESCKIPTION DESCRIPTION 1.4-U I Valid MAIN STEAM LINE RADIATION HIGH-HIGH alarm, RA-90-135C OR Valid OG PRETREATMENT RADIATION HIGH alarm, RA-90-157A.

OPERATING CONDITION:

-Mode l -Mode3

- Mode 2 1.5A Reactor moderator temperature CANNOT be maintained below212 0 F whenever Technical Specifications require Mode 4 conditions or during operations in Mode 5. M OPERATING CONDmON:

-Mode4

-Mode 5 1.5-S K Suppression Pool temperature, level and RPV pressure CANNOT be maintained in the safe area of Curve 1.5-S. M 0

OPERATING CONDITION:

-Mode I -Mode 3

-Mode2 1.0 REACTOR PAGE 19 OF 207 REVISION 28

EPIP-1 EMERGENCY SECTION II CLASSIFICATION 1.0 REACTOR EVENT CLASSIFICATION MATRIX PROCEDURE THIS PAGE INTENTIONALLY BLANK REVISION 28 PAGE 20 OF 207 1.0 REACTOR

EMERGENCY EPIP-1 CLASSIFICATION SECTION II 2.0 PRIMARY PROCEDURE EVENT CLASSIFICATION MATRIX VNVTAfWW71.T

-Th s hffsTT PRIMARY C ONTAINMENT 2 .0 2.0 PRIMARY PAGE 21 OF 207 REVISION 28 CONTAINMENT

EPTP-1 EMERGENCY 2.0 PRIMARY SECTION II CLASSIFICATION CONTAINMENT EVENT CLASSIFICATION MATR1Y vP~RvtlnlTrnh P1fr-fT~

NOTES:

I l-ul~~xLUfN4I 4N~N[J-I I T-%T)XTll MT T VT - A- - - -- - ---- . -- - _____j "R 1 W I'"L rLuuKl LIUKIN bufvP PUMP EXCESSIVE OPERATION DRYWELL CAM ACTIVITY INCREASING DRYWELL TEMPERATURE HIGH ALARM CHEMISTRY SAMPLE RADIONUCLIDE COMPARSION TO RX WATER CURVE 2.1-S PRESS SUPPR PRESS 03030 _ _ _

c 2 REQUIRE 250 uS Col a

20

....... ...i.. . . .

111 2 13 A .S SUPP ....

111.512 13 4 5 16 17 18 19 20 SUPPR PL LVL (FT)

REVISION 28 PAGE 22 OF 207 2.0 PRIMARY CONTAINMENT

EMERGENCY EPIP-1 CLASSIFICATION SECTION II 2.0 PRIMARY PROCEDURE EVENT CLASSIFICATION MATRIX CONTAINMENT M J W IN KY KI 111

___W e__I_-km____

II______LI_

I_

DESCRIPTION IDESCRIPTION C~

CA ARC M

3 2.1-A K Drywell pressure at or above 2.45 PSIG AND Indications of Primary System leakage into Primary Containment. Refer to Table 2.1-A.

OPERATING CONDITION:

- Mode I -Mode 3

-Mode 2 2.1-S 2.2-S Suppression Chamber Pressure CANNOT be Drywell or Suppression Chamber hydrogen maintained in the safe area of Curve 2.1 -S. concentration at or above 4%

AND Drywell or Suppression Chamber oxygen concentration at or above 5%.

OPERATING CONDITION: OPERATING CONDITION: Z

-Mode I -Mode 3 -Mode I -Mode 3

-Mode 2 -Mode 2 2.1-G l2.2-G Suppression Chamber Pressure CANNOT be Drywell or Suppression Chamber hydrogen maintained below 55 PSIG. concentration at or above 6%

AND Drywell or Suppression Chamber oxygen concentration at or above 5%. .

OPERATING CONDITION: OPERATING CONDTION:

-Mode I -Mode 3 -Mode I -Mode33

- Mode 2 -Mode 2 2.0 PRIMARY PAGE 23 OF 207 REVISION 28 CONTAINMENT

EPIP-1 EMERGENCY 2.0 PRIMARY SECTION II CLASSIFICATION CONTAINMENT EVE NT CLAS S~frATrONW MATURY NOTES:

CURVES/TABLES:

0.,. ;.-.0..

f..

. . ;.:; .. f...

1 0 9 6 ... .. ... . -...-.. ......-

UNEXPLAINED LOSS OF PRESSIJRE EXCEEDING SI-4.7.A.2.a LIMITS INABILITY TO ISOLATE ANY LINE EXITING CONTAINMENT WHEN ISOLATION IS REQUIRED VENTING IRRESPECTIVE OF OFFSITE RELEASE RATES PER EOls/SAMGs REVISION 28 PAGE 24 OF 207 2.0 PRIMARY CONTAINMENT

EMERGENCY EPIP-1 CLASSIFICATION SECTION II 2.0 PRIMARY PROCEDURE EVENT CLASSIFICATION MATRIX CONTAINMENT I DESCRIPTION DESCRIPTION C~

32 Ut 2.3-A lg II Drywell radiation levels at or above the values listed in Table 2.3-A/2:3-S2, with the RCS barrier intact.  ;

OPERATING CONDITION:

-Mode -Mode 3

- Mode 2 2.3-Si J 2.3-S2 jgj Drywell radiation levels at or above the values listed '

Drywell radiation levels at or above 4880 R/HR in Table 2.3-A/2.3-S2, with the RCS barrier intact.

with the RCS barrier not intact. AND Either of the following exists:

• Indications of loss of Primary Containent. Refer to Table 2.3/2.5-U.

• Primary Containent integrity CANNOT be maintained.

OPERATING CONDITION: OPERATING CONDITION:

-Mode I -Mode 3 -Mode I -Mode3

-Mode 2 -Mode 2 2.3-G1 2.3-G2 Drywell radiation levels at or above 4880 RIFHR with Z Drywell radiation levels at or above 19500 R/HR the RCS barrier not intact. M with the RCS barrier not intact. AND Either of the following exists:

• Indications of loss of Primary Containment.

Refer to Table 2.3/2.5-U.

• Primary Containment integrity CANNOT be maintained.

OPERATING CONDITION: OPERATING CONDITION: - Z

-Mode I -Mode 3 -Mode I -Mode 3

-Mode 2 -Mode2 2.0 PRIMARY PAGE 25 OF 207 REVISION 28 CONTAINMENT

EPIP-1 EMERGENCY 2.0 PRIMARY SECTION II CLASSIFICATION CONTAINMENT EVENT ClARSRtATlrtfl MLAsTVn

- -- - -' .. .5-'-i-' ~ULLJUt NOTES:

CURVES/TABLES:

UNEXPLAINED LOSS OF PRESSURE EXCEEDING SI-4.7.A.2.a LIMITS INABILITY TO ISOLATE ANY LINE EXITING CONTAINMENT WHEN ISOLATION IS REQUIRED VENTING IRRESPECTIVE OF OFFSITE RELEASE RATES PER EOIs REVISION 28 PAGE 26 OF 207 2.0 PRIMARY CONTAINMENT

EMERGENCY EPIP-1 CLASSIFICATION SECTION II 2.0 PRIMARY PROCEDURE EVENT CLASSIFICATION MATRIX CONTAINMENT DESCRIPTION DESCRIPTION I I I _ 1 2.4-U 2.5-U M ci Drywell unidentified leakage exceeds 10 GPM Inability to maintain Primary Containment pressure 4 OR boundary. Refer to Table 2.3/2.5-U. rA Drywell identified leakage exceeds 40 GPM.

t_1 ti OPERATING CONDITION: OPERATING CONDITION: 4

-Mode I -Mode 3 -Mode I -Mode 3 12

- Mode 2 0-3

-Mode2 Drywell unidentified leakage exceeds 50 GPM.

OPERATING CONDITION:

- Mode I - Mode 3

- Mode 2 03 3

M C) 0z i

0 M

34 I __ __________ ___ ____ ___ ___I 2.0 PRIMARY PAGE 27 OF 207 REVISION 28 CONTAINMENT

EPIP-1 EMERGENCY 2.0 PRIMARY SECTION II CLASSIFICATION CONTAINMENT EVENT CLASSIFICATION MATRIX PROrVINTURF.

PRflCFflTrn1' THIS PAGE INTENTIONALLY BLANK REVISION 28 PAGE 28 OF 207 2.0 PRIMARY CONTAINMENT

EMERGENCY EPIP-1 CLASSIFICATION SECTION II 3.0 SECONDARY PROCEDURE EVENT CLASSIFICATION MATRIX CONTAINMENT SECONDARY C ONTAINMENT 3.0 3.0 SECONDARY PAGE 29 OF 207 REVISION 28 CONTAINMENT

EPIP-I EMERGENCY 3.0 SECONDARY SECTION II PROCEDURE CONTAINMENT EVENT CLASSIFICATION MATRIY CT.AR~~fArATdl%

(- - - - -T -l NOTES:

CURVES/TABLES:

APPLICABLE PANEL 9-21 AREA TEMPERATURE ELEMENTS MAX SAFE OPERATING VALUE 0F (UNLESS OTHERWISE NOTED) UNIT 2 UNIT 3 RHR A/C PUMP ROOM 74-95A 150 155 RHR B/D PUMP ROOM 74-95B 210 215 HPCI TURBINE AREA 73-55A 270 270 RCIC TURBINE AREA 71-41A 190 190 CS ROOM HIGH HUMIDITY OF TEMP HIGH (XA-55-3E-29) PANEL 9-3 TI-75-69B 140 150 RCIC STEAM SUPPLY AREA 71-41B, 41C, 41D 200 250 HPCI STEAM SUPPLY AREA 73-55B, 55C, 55D 240 240 RHR A/C PUMP SUPPLY AREA 74-95H 240 240 RHR B/D PUMP SUPPLY AREA 74-95G 240 240 MAIN STEAM LEAK DETECTION HIGH . (XA-55-3D-24) PANEL 9-3 TIS-1-60A 315 315 RHR VALVE ROOM 74-95E 170 175 RWCU ISOL LOGIC CHANNEL A/B TEMP HIGH (XA-55-5B-32/33) PANEL 9-5 170 175 69-835A, B, C, D AUX INST ROOM RWCU OUTBD ISOL VLV AREA 69-29F 220 220 RWCU HX AREA 69-29G 220 220 RWCU HX EXH DUCT 69-29H 220 220 RWCU RECIRC PUMP A AREA 69-29D 215 205 RWCU RECIRC PUMP B AREA 69-29E 215 205 RHR A/C HX ROOM 74-95C 195 200 RHR B/D HX ROOM 74-95D 195 200 FPC HX AREA 74-95F 150 155 l ... -.. ....... TAB-LE MAXIMIJM SAFE 0 .....

t; !...

........... AREA RAIAIN OP.ERATIN . . . . . .,, IT AREA RAD MONITOR MAX SAFE VALUE MR/HR RHR WEST ROOM 90-25A 1000 RHR EAST ROOM 90-28A 1000 HPCI ROOM 90-24A 1000 CS/RCIC ROOM 90-26A 1000 CORE SPRAY ROOM 90-27A 1000 SUPPR POOL AREA 90-29A 1000 CRD-HCU WEST AREA 90-20A 1000 CRD-HCU EAST AREA 90-21 A 1000 TIP DRIVE AREA 90-23A 1000 NORTH RWCU SYSTEM AREA 90-13A 1000 SOUTH RWCU SYSTEM AREA 90-14A 1000 RWCU SYSTEM AREA 90-9A 1000 MG SET AREA 90-4A 1000 FUEL POOL AREA 90-IA 1000 SERVICE FLR AREA 90-2A 1000 NEW FUEL STORAGE 90-IA 1000 TABLE~3jG/3 2-G DRYWELL RADIATION UNIT 2 DRYWELL RADIATION UNIT 3 2-RE-90-272A 2-RE-90-273A

> 345 R/HR 3-RE-90-272A [ > 106 R/HR

> 164 R/HR 3-RE-90-273A > 164 R/HR REACTOR COOLANT ACTIVITY Ž> 300 fLCI/gm DOSE REACTOR COOLANT ACTIVITY > 30e PCI/gm DOSE EQUILAVENT IODINE-131 EQUILAVENT IODINE-131 REVISION 28 PAGE 30 OF 207 3.CSECONDARY CONTAINMENT

EMERGENCY EPIP-1 CLASSIFICATION SECTION II 3.0 SECONDARY PROCEDURE EVENT CLASSIFICATION MATRIX CONTAINMENT I 1 W 11WA;ku I WIN .1 ci 4

C rA 11 M

4 4

0q i I _ _

3.2-A Any ofthe followng high radiation alamns on Panel 9-3:

• RA-904IA, FuelPool FloorArea

• RA-90-250A, Reactor, Turbine, Refuel Exhaust

• RA-90-142A, ReactorZone Exhaust

• RA-90-140A, Refieling Zone Exhaust AND Confirmation by Refuel Floor personnel that irradiated fuel damage may have occurred.

OPERATING CONDITION:

-All 3.1-S K '

3.2-S An unisolable Primary System leak is discharging into An unisolable Primary System leak is discharging into P Secondary Containment Secondary Containment ANDAN Any area temperature exceeds the Maximum Safe ANDab Operating Temperature limit listed in Table 3. 1. Any area radiation level at or above the Maximum

.. Safe Operating area Radiation limit listed in Table 3.2. n OPERATING CONDION: OPERATING CONDITION:

- - e Mode3 - Mode I - Mode 3

- Mode 2 - Mode 2 3.1-G 3.2-G K An unisolable Primary System leak is discharging into An unisolable Primary System leak is discharging into Z Secondary Containment Secondary Containment AND AND Any area temperature exceeds the Maximum Safe Any area radiation level at or above the Maximum Operating Temperature limit listed in Table 3.1. Safe Operating area Radiation limit listed in Table 3.2 t AND AND Any indication of potential or significant fuel failure Any indication of potential or significant fuel failure exists. Refer to Table 3.1l-G/3.2-G. exists. Refer to Table 3.1 -G/3.2-G.

OPERATING CONDITION: OPERATING CONDITION: Z

- Mode I - Mode 3 - Mode I - Mode 3 TV

- Mode 2 - Mode 2 <

3.0 SECONDARY PAGE 31 OF 207 REVISION 28 CONTAINMENT

EPIP-1 EMERGENCY 3.0 SECONDARY SECTION II PROCEDURE CONTAINMENT EVENT Cl.ASShATlrNW MATiRv rW A CC VY,"A nTJ'%s

%f¶JZ -A r At 'I JfJl THIS PAGE INTENTIONALLY BLANK REVISION 28 PAGE 32 OF 207 3.0 SECONDARY CONTAINMENT

EMERGENCY EPIP-1 CLASSIFICATION SECTION H 4.0 RADIACTIVITY PROCEDURE EVENT CLASSFICATION MATRIX RIF.,IFAqll RADIOACTIVITY" RELEASES 40 4.0 RADIACTIVITY PAGE 33 OF 207 REVISION 28 RELEASE

EPIP-1 EMERGENCY 4.0 RADIOACTIVITY SECTION II CLASSIFICATION RELEASE EVENT CLASSIFICMATTrnW MATRDw EVENT CLA%_mATanN MATDT

- -- 1~.J~~J~f.

NOTES:

NOTE 4.1-U Prior to making this emgency classification based upon the WRGERMS indication, assess the release by either of the folloaing:

1.Actlal field m =easuiments exceed the limits in Table 4.1-U

2. SI4..B.I.a.1 ReleaseFractionexceeds2.0 Ineitfhease t canibexilxthin 60unilmtes thedbreatminniutbenr mthevalidWRGERMS g NOTE 4.1-A Priorto making this emergency classification based upon the WRGERMS indication, assess the release by either of the following 1.Actual field measurnents exceed the lnits in Table 4.1 -A

2. SI4.8B. L al Release Fracion exceeds 200 Ifeiftf 1as t canbea cdxtedwiti 15 n tstn tedatinn bemade onthe valid WRGERMS ng NOTE 4.1-S Prior to malkng this emergency classification based upon the Gaseous Release Rate indication, assess the release by either of the following methods:

I. Actual field meastrments exceed the limits in Table 4.1-S.

2. Projected or Actual Dose Assessments exceed 100 mrurem TEDE or 500 nurem CDE.

Ifneither assessment can be conducted within 15 minutes then the declaration must be made based on the valid WRGERMS reading.

NOTE 4.1-G Prior to making this emergency classification based upon the Gaseous Release Rate indication, ass the release by either of the following methods:

1.Actual field measumens exceed the limits in Table 4.1-G.

2. Projected orAcual DoseAssessments exceed 1000mremTEDE or 5000meunCDE.

Ifneitherassnen canbe conductedwithm 15 mmutesthentedeclrationmustbe madebased mnthe vaidWRGERMS reading CURVES/TABLES:

T. ..........

TYPE MONITORING METHOD LIMIT DURATION GASEOUS RELEASE RATE STACK NOBLE GAS (WRGERMS) 2.88 X 10' Ci/sec I HOUR GASEOUS RELEASE RATE SI 4.8.B.a I.a. RELEASE FRACTION 2.0 SITE BOUNDARY RADIATION READING I HOUR FIELD ASSESSMENT TEAM 0.10MREM/HR y-13 I HOUR TYPE MONITORING METHOD LIMIT DURATION GASEOUS RELEASE RATE STACK NOBLE GAS (WRGERMS) 2.88 X 10 9wCi/sec 15 MINUTES GASEOUS RELEASE RATE Sl14.8.B. I.a.1I RELEASE FRACTION 200 15 MINUTES-SITE BOUNDARY RADIATION READING ELAE MIT ASSESSMENT FIELD VO S TEAM ARE

-IT' -rE(EC100MREM/HR y - - 15____

MINUTES TYPE TYPE -MONITORING

.MONITORING METHOD LIMIT METHOD -LIMIT lDURATION lDURATION GASEOUS RELEASE RATE STACK NOBLE GAS (WRGERMS) 1.3 X10"'p Ci/sec 15 MINUTES SITE BOUNDARY RADIATION READING FIELD ASSESSMET rR . TE XAI X__ - __

--- - - r-L.LI ftMaA3ryN II tEAUM I10OOMREM/HRy-p I I HOUR SITEBOUNDARYIODINE-131 FIELDASSESSMENTTEAM J3.9X I0 6 3 CI/cm l IHOUR REVISION 28 PAGE 34 OF 207 4.0 RADIOACTIVITY RELEASE

EMERGENCY EPIP-1 CLASSIFICATION SECTION II 4.0 RADIOACTIVITY PROCEDURE EVENT CLASSIFICATION MAT=lJX RELEASE

_9 _

DESCRIPTION DESCRIPTION I ~It- lt_ II 4.1-U INN 4.2-U C

Gaseous release exceeds ANY limit and duration in Main Steam Line break outside Primary Containment 2 Table 4. 1-U. C with isolation. rA ci t*

M OPERATING CONDITION: 4 OPERATING CONDITION: -Mode 1 -Mode3 2

-All -Mode2 i

4.1-A Kti Gaseous release exceeds ANY limit and duration in Table 4.1-A.

OPERATING CONDITION:

-All 4.1-S KiT 4.2-S EITHIER of the following conditions exists: Unisolable Main Steam Line break outside Primary M

• Gaseous release exceeds or is expected to exceed Contaiunent.

ANY limit and duration in Table 4. 1-S.

• Dose assessment indicates actual or projected dose consequences above 100 mrem TEDE or 500 mrem thyroid CDE. OPERATING CONDITION:

-Mode I Mode 3 OPERATING CONDITION: --Mode2

-All 4.1-G K -

EITIIER of the following conditions exists:

• Gaseous release exceeds or is expected to exceed ANY limit and duration in Table 4.1 -G.

• Dose assessment indicates actual or projected dose consequences above 1000 inrem TEDE or 5000 mrem thyroid CDE.

C OPERATING CONDITION: Q

-All 4.0 RADIOACTIVITY PAGE 35 OF 207 REVISION 28 RELEASE

EPIP-1 EMERGENCY 4.0 RADIOACTIVITY SECTION II CLASSIFICATION RELEASE EVENT CLASSIFICATION MATRIX PROCEDURE NOTES:

CURVES/TABLES:

REVISION 28 PAGE 36 OF 207 4.0 RADIOACTIVITY RELEASE

EMERGENCY EPIP-1 CLASSIFICATION SECTION II 4.0 RADIOACTIVITY PROCEDURE EVENT CLASSIFICATION MATRiX RELEASE I juhCKI1' I .

Liquid release rate exceeds 20 times ECL as determined by chemistry sample AND Release duration exceeds or will exceed 60 minutes. rA t-M OPERATING CONDITION:

- All I I ,

4.3-A Liquid release rate exceeds 2000 times ECL as detennined by chemistry sample AND Release duration exceeds or will exceed 15 minutes.

OPERATING CONDmON:

-All lJ CA "i

M z

l z

2 z4 M

I I 4.0 RADIOACTIVITY PAGE 37 OF 207 REVISION 28 RELEASE

EPIP-l EMERGENCY 4.0 RADIOACTIVITY SECTION II CLASSIFICATION RELEASE EVENT CLASSIFICATION MATRTY DR~nf'WnT}TT0

..----- At%.PtAtJnFL AJnr, THIS PAGE INTENTIONALLY BLANK REVISION 28 PAGE 38 OF 207 4.0 RADIOACTIVITY RELEASE

EMERGENCY EP[P-1 CLASSIFICATION SECTION II 5.0 LOSS OF PROCEDURE EVENT CLASSIFICATION MATRIX i POWER

~~POWER LOSS OF POWER 5.0 5.0 LOSS OF PAGE 39 OF 207 REVISION 28 POWER

EPIP-1 EMERGENCY 5.0 LOSS OF SECTION II CLASSIFICATION POWER EVENT CLASSrFTCATTON MATRIX iPvnrl;nTnTRT'

- -** f*lLUfALJV NOTES:

5.1-U Loss of normal and alternate supply voltage implies inability to restore voltage from any qualified source to normal or alternate feeder for at least one of the unit specific boards within 15 minutes. At least two boards must be energized from Diesel power to meet this classification. If only one board can be energized and that board has only one source of power then refer to 5.1-A.

5.1-Al Only one source of power (Diesel or Offsite) is available to any one of the listed unit specific 4KV Shutdown Boards. No power is available to the three remaining boards.

5.1-A2 Loss of voltage to all unit specific 4KV Shutdown Boards applies to those boards which normally supply emergency AC power to the affected unit only. Determination of the event classification depends on the affected unit operating mode. For units in operation 5.1-S would apply.

5.1-S Loss of voltage to all unit specific 4KV Shutdown Boards applies to those boards which normally supply emergency AC power to the affected unit only. Determination of the event classification depends on the affected unit operating mode. For units in Shutdown or Refuel 5.1-A2 would apply.

5.1 -G Loss of voltage to all unit specific 4KV Shutdown Boards applies to those boards which normally supply emergency AC power to the affected unit only.

CURVES/TABLES:

T: 4KV.-: iSH'UTDOWN B~R PLCB~T APPLICABLE UNIT APPLICABLE 4KV SHUTDOWN BOARDS UNIT 1 A, B, C, and D UNIT 2 A, B, C, and D UNIT 3 . 3A, 3B, 3C, and 3D REVISION 28 PAGE 40 OF 207 5.0 LOSS OF POWER

EMERGENCY EPIP-1 CLASSIFICATION SECTION I 5.0 LOSS OF PROCEDURE EVENT CLASSIFICATION MATRIX POWER

• I "lEb lKi UIN DESCRIPTION

_ ,________________________ £ 5.1-U IT I 11t Loss of normal and alternate supply voltage to ALL unit specific 4KV shutdown boards from Table 5.1 for greater than 15 minutes AND At least two Diesel Generators supplying power to unit specific 4KV Shutdown Boards listed in Table 5.1.

OPERATING CONDmON:

-All 5.1-Al 5.1-A2 Khli Loss of voltage to ANY THREE unit specific 4KV Loss of voltage to ALL unit specific 4KV Shutdown Shutdown Boards from Table 5.1 for greater than 15 Boards from Table 5.1 for greater than 15 minutes.

minutes AND Only one source of AC power available to the remaining board.

OPERATING CONDM ON: OPERATING COM ON:

-Model -Mode3 -Mode4 -Deflelod

- Mode2 -Mode5 5.1-S JII.1; Loss of voltage to ALL unit Specific 4KV Shutdown .

Boards from Table 5.1 for greater than 15 minutes.

OPERATING CONDITION: Z

-Mode I -Mode 3

- Mode 2 5.1-G Loss of voltage to ALL unit specific 4KV Shutdown Boards From Table 5.1 AND EITHER of the following conditions exists:

• Restoration of at least one 4KV Shutdown Board is NOT likely within three hours

• AduatecwecmoloigCANNOT be amzed OPERATING CONDM ON:

-Mode I -Mode 3

- Mode 2 5.0 LOSS OF PAGE 41 OF 207 REVISION 28 POWER

EPIP-1 EMERGENCY 5.0 LOSS OF SECTION II CLASSIFICATION POWER EFVENT MAQ.YLVA's~nlgr Ax X IMsTVn

-~A V Kd rImUJLLIJUmBL NOTES:

5.2 250V DC bus voltage of less than 248 volts on any feeder to any referenced board constitutes a loss of voltage for that feeder; thus, a loss of DC control voltage to the referenced board. The voltage readings are obtained at the 250V Battery Boards feeding the referenced boards.

CURVES/TABLES:

APPLICABLE UNIT APPLICABLE 4KV SHUTDOWN BOARDS UNIT I A, B, C, and D UNIT 2 A, B, C, and D UNIT 3 3A, 3B, 3C, and 3D

,....C..

COMBINATION CRITICAL 250V DC POWER ESSENTIAL SYSTEMS

_ NIT SPECIFIC UNLESS OTHERWISE NOTF. I I 4KV UNIT BD, A, B, and C CONTROL POWER i MAIN CONDENSER

- AND AND 480V UNIT BOARD A and B CONTROL POWER EHC PUMPS AND AND PANEL 9-9 CABINET 1 REACTOR FEED PUNTS II LJUVv L)ts KMUV tit) A R FEED PUMPS ITMP'T III F 'bn cflx r LJVi V UL r -. - I --

KIVIu V Ii)

+

L RriTCT IV 250V DC RMOV BDs A, B, and C > 4 MSRVs AND AND 4KV SHUTDOWN BDs A, B, C, and D CONTROL POWER I RHR PUMP (3A, 3B, 3C, and 3D FOR UNIT 3) OR 1 CORE SPRAY PUMP REVISION 28 PAGE 42 OF 207 5.0 LOSS OF POWER

EMERGENCY EPIP-1 CLASSMIICATION SECTION II 5.0 LOSS OF PROCEDURE EVENT CLASSIFICATION MAT=I POWER k61K@J U IIU MS11IN DESCRIPTION 5.2-U jŽIKMi Unplanned loss of250V DC contol powm to ALL umit specific 4KV Shutdo-n Boards from table 5.2-U for greter than 15 minutes rA OR Unplanned loss of250 DC control powe to unit specific 480V Shutdon Boards A and B for geteran 15 minute.

OPERATING CONDITION:

• 3

- Mode 4

-Mode 5 iI 5.2-S simi Loss of 250V DC power to ALL combinations of essential systems from Table 5.2-S for greater than 15 minutes.

OPERATING CONDITION:

-Mode -Mode 3

- Mode 2 5.0 LOSS OF PAGE 43 OF 207 REVISION 28 POWER

EPIP-1 EMERGENCY 5.0 LOSS -OF SECTION II CLASSIFICATION POWER EVE.NT CL.AqQ.SrA'rnV4As MrsTv DD .- _ID1

- - Yn~jI..,LIunL THIS PAGE INTENTIONALLY BLANK REVISION 28 PAGE 44 OF 207 5.0 LOSS OF POWER

EMERGENCY EPIP-1 CLASSIFICATION SECTION II PROCEDURE EVENT CLASSIFICATION MATRIX 6.0 HAZARDS HAZARD 1S 6.0 6.0 HAZARDS PAGE 45 OF 207 REVISION 28

EPIP-1 EMERGENCY SECTION II CLASSIFICATION 6.0 HAZARDS MIATDw

- - - - ClA.RRFTfVATrnV EVENT -- rILm Avm jrL~jI.lI.DmlL NOTES:

CURVES/TABLES:

REVISION 28 PAGE 46 OF 207 6.0 HAZARDS

EMERGENCY EPIP-1 CLASSIFICATION SECTION II PROCEDURE EVENT CLASSIFICATION MATRIX 6.0 HAZARDS Th'rQ('Dyh'PT I'T I a "JuotAr A turn11 U15miNCAIPTIN' I

O

-u; .'.II I

6.1-U Valid, unexpected increase of ANY in plant ARM z4 reading to 1000 mremn/r (except TIP Room). C~

UD OPERATING CONDITION:

-All -3 6.1-Al l 6.1-A2 Valid, unexpected increase of ANY in plant ARM Control Room radiation levels > 15 mrem/hr.

reading to 1000 irntr/hr (except TIP Room).

AND Personnel actions required in the affected area(s).

OPERATING CONDITION: OPERATING CONDITION:

-All -All r3 0"

i I 0

m

4 m

m 9

9 m

1;4 r) 0-I __ _ _ _ _ _ _ _ _ _ _ _ _I 6.0 HAZARDS PAGE 47 OF 207 REVISION 28

EPIP-1 EMERGENCY SECTION II CLASSIFICATION 6.0 HAZARDS EVENT CLASSIFICATION MATRIY

- AV- ~J,"%

ulant-vlm.Tlmow vr NOTES:

CURVES/TABLES:

REVISION 28 PAGE 48 OF 207 6.0 HAZARDS

EMERGENCY EPIP-1 CLASSIFICATION SECTION II PROCEDURE EVENT CLASSIFICATION MATRIX 6.0 HAZARDS 6n L 6lr 66qTJk I IEtL.JLr I 1JUJ Dlbl:u'T1ION DE~,jU K~i'I 6.3-U Turbine failure resulting in casing penetration OR Significant damage to turbine or generator seals during operation.

C2~

OPERATING CONDmON:

-Mode l

-Mode2 6.2-A 6.3-A Control Room Abandonment from entry into AOI-1 00- Turbine failure resulting in visible structural damage to 2 or SSI-1 6 for ANY Unit Control Room. or penetration of ANY of the following structures from missiles:

• Reactor Building

• Diesel Generator Building

• Intake Structures OPERATING CONDITION: Control Bay

-All OPERATING CONDITION:

-Mode I

-Mode2 6.2-S ControlRoomAbandonment from entry ntoAOI-100-2 or M SSI-16 forANYUnitCoirolRoom.

AND Control of Reactor water level, Reactor pressure, and Reactor power (for Modes 1,2, &3) or decay heat removal (for Modes 4 & 5)perAOI-100-2 or SSI-16 as applicable CANNOT be established within 20 minutes after evacuation is initiated CZ OPERATING CONDITION:

-All z

6.0 HAZARDS PAGE 49 OF 207 REVISION 28

EPIP-1 EMERGENCY SECTION II CLASSIFICATION 6.0 HAZARDS EVENT CLASq.RMVATaNV MATDnr DM~brl7.lrWT"-

- - -xkrnJmLUpUklcl NOTES:

CURVES/TABLES:

.... T .)T

....~X...

.N P.........

MnzAq I ulc bUlLLlN REFUEL FLOOR 4KV SHUTDOWN BOARD ROOMS 4KV SHUTDOWN BOARD BATT'lERY ROOMS 480V SHUTDOWN BOARD ROOMS 3A and 3B RMOV BOARD ROOMS 4KV BUS TIE BOARD ROOM CONTROL BAY ELEVATION 593', 606' and 617' DIESEL GENERATOR BUILDINGS (ALL ELEVATIONS)

TURBINE BUILDING (ALL ELEVATIONS IN OR ADJACENT TO AREAS CONTAINING SAFE SHUTDOWN EQUIPMENT INTAKE PUMPING STATION (ALL ELEVATIONS)_

RADWASTE BUILDING (ALL ELEVATIONS)

CABLE TUNNEL (INTAKE TO TURBINE BUILDING)

STANDBY GAS TREATMENT BUILDING REACTOR BUILDING REFUEL FLOOR 4KV SHUTDOWN BOARD ROOMS 4KV SHUTDOWN BOARD BATTERY ROOMS 480V SHUTDOWN BOARD ROOMS 3A and 3B RMOV BOARD ROOMS 4KV BUS TIE BOARD ROOM CONTROL BAY ELEVATION 593', 606' and 617' DIESEL GENERATOR BUILDINGS (ALL ELEVATI INTAKE PUMPING STATION (ALL ELEVATIONS)

RADWASTE BUILDING (ALL ELEVATIONS)

CABLE TUNNEL (INTAKE TO TURBINE BUILDIN STANDBY GAS TREATMENT BUILDING REVISION 28 PAGE 50 OF 207 6.0 HAZARDS

EMERGENCY EPIP-1 CLASSIFICATION SECTION II PROCEDURE EVENT CLASSIFICATION MATRIX 6.0 HAZARDS "Wor'"T~nh'U'TI'.T m m r Iamm

- gW ~ mg. u _

"M0%!,iL 1I .LVU DhbUKW,1,1Ur IIk- I I -

6.4-Ul I l 6.4-U2 Confirmed fire in ANY plant area listed in Table 6.4-UI Unanticipated explosion within the protected area AND resulting in visible damage to ANY permanent NOT extinguished within 15 minutes. structure or equipment. rA OPERATING CONDITION: M OPERATING CONDITION:

-All -All 03 6.4-A F Fire or explosion in ANY plant area listed in Table 6.4-A affecting safety system performance OR Fire or explosion causing visible damage to permanent structures or safety systems in ANY area listed in Table.

6.4-A.

OPERATING CONDITION:

-All 6.0 HAZARDS PAGE 51 OF 207 REVISION 28

EPIP-1 EMERGENCY SECTION II CLASSIFICATION 6.0 HAZARDS EVENT CLASSIFICATION MATRIX PRorr"InTUPi P-m I'lTDi NOTES:

CURVES/TABLES:

...... .A P AB E ' A .A E REACTOR BUILDINGS REFUEL FLOOR CONTROL BAY DIESEL GENERATOR BUILDINGS TURBINE BUILDING INTAKE PUMPING STATION RADWASTE BUILDING CABLE TUNNEL (INTAKE TO TURBINE BUILDING)

STANDBY GAS TREATMENT BUILDING REVISION 28 PAGE 52 OF 207 6.0 HAZARDS

EMERGENCY EPIP-1 CLASSIFICATION SECTION II PROCEDURE EVENT CLASSIFICATION MATRIX 6.0 HAZARDS TV__l~m nVV1~

"Mal3u I Jluq f l ;bUtP'T1ON _I I

6.5-U EITHER of the following conditions exists: 4

• Normal operations impeded due to access restrictions caused by toxic gas concentrations within any cl!i A

building or structure listed in Table 6.5/6.6.

• Confirmed report by Local, County, or State Officials that a large offsite toxic gas release has occurred within one mile of the site with potential to enter the site boundary in concentrations at or t-1 above the Permissible Exposure Limit (PEL) causing an evacuation of any site personnel. M 4

OPERATING CONDmTION: 4

-All 3 6.5-A 1l ALL of the following conditions exists:

Plantperomdiqogsrt oic gas within any building or stcture listed in Table 6.5/6.6.

• Plant personnel report severe adverse health reactions due to toxic gas (i.e., burning eyes, throat, or dizziness)

OR

• Sampling results by Fire Protection or Industrial Safety personnel indicate levels above t the Permissible Exposure Limit (PEL).

• Determination by the Site Emergency Director that plant personnel would be unable to perform actions necessary to establish and maintain cold shutdown conditions while utilizing appropriate personnel protective equipment.

OPERATING CONDITION:

ALL 03 r) i 0

M 114 M

r" M

A 9M 4

r) 0-I 6.0 HAZARDS PAGE 53 OF 207 REVISION 28

EPIP-1 EMERGENCY SECTION II CLASSIFICATION 6.0 HAZARDS EVENT I5.AQQ.r Aw ry% 1am rn F WN- CTA- TT axFlM lyIAX-IL t wVEU NOTES:

CURVES/TABLES:

..'..,.',',~.'.. ..'. .. '..'.."

REACTOR BUILDINGS REFUEL FLOOR CONTROL BAY DIESEL GENERATOR BUILDINGS TURBINE BUILDING INTAKE PUMPING STATION RADWASTE BUILDING CABLE TUNNEL (INTAKE TO TURBINE BUI STANDBY GAS TREATMENT BUILDING REVISION 28 PAGE 54 OF 207 6.0 HAZARDS

EMERGENCY EPIP-1 CLASSIFICATION SECTION II PROCEDURE EVENT CLASSIFICATION MATRIX 6.0 HAZARDS

- .k~.L~

m I - mm- _y.

ILL,%KLr I luJrJ IMNLAUT1ON I ___I I 6.6-U an EITHER of the following conditions exists:

• Release of flammable gas within the site boundary in concentrations at or above 25% of the Lower Explosive Limit (LEL) for any three readings obtained in a 10 ft. triangular area as indicated by Fire Protection or Industrial Safety personnel using appropriate monitoring instrumentation.

• Confirmed report by Local, County, or State Officials that a large offsite flammable gas release has occurred within one mile of the site with potential to enter the site boundary in concentrations at or above 25% of the Lower Explosive Limit (LEL).

OPERATING CONDmON:

-All 6.6-A Release of flammable gases within any building or structure listed in Table 6.5/6.6 in concentrations at or above 25% of the Lower Explosive Limit (LEL) for any three readings obtained in a 10 ft. triangular area as indicated by Fire Protection or Industrial Safety personnel using appropriate monitoring instrumentation.

OPERATING CONDITION:

-All

• q M

EM A

z0 i

z z

I 6.0 HAZARDS PAGE 55 OF 207 REVISION 28

EPIP-1 EMERGENCY SECTION II CLASSIFICATION 6;0 HAZARDS EVENT CLASSIFICATION MATRTY VRnrW"nTDV m u I NOTES:

CURVES/TABLES:

REVISION 28 PAGE 56 OF 207 6.0 HAZARDS

EMERGENCY EPIP-1 CLASSIFICATION SECTION II PROCEDURE EVENT CLASSIFICATION MATRIX 6.0 HAZARDS I T~'Q J~~~i~.r

(~Tu'rTFThjI L~a I "JV0%_Lllr I lull1 DEbSCuRATIN

_4 ANY of the following conditions exist:

• Bomb device discovered within the plant protected area but NOT within a vital area

• Attempted or imminent attempt by a hostile force to penetrate the plant protected area barrier C,)

• Civil disturbance ongoing on the owner controlled property outside the protected area that threatens to interrupt plant operations

• Hostage/Extortion situation that threatens to interrupt plant operations.

OPERATING CONDITION:

-All 6.7-A Bomb device discovered within ANY plant vital area OR Actual intrusion into the plant protected area by a hostile force.

OPERATING CONDITION:

-All 6.7-S Intrusion into ANY plant vital area by a hostile force.

OPERATING CONDITION: Z

-All 6.7-G M Intrusion by a hostile force into Control Rooms, backup control areas, or plant vital areas which results in a loss of physical control of equipment or functions required to reach and maintain safe shutdown or remove decay heat from any unit.

OPERATING CONDITION:

-All Sac 6.0 HAZARDS PAGE 57 OF 207 REVISION 28

EPIP-1 EMERGENCY SECTION II CLASSIFICATION 6.0 HAZARDS EVENT CLASSIFTCATrON MATRIY DRRnf~W"TT12i' D AvrprTv,"rv,

---

=

NOTES:

CURVES/TABLES:

REVISION 28 PAGE 58 OF 207 6.0 HAZARDS

EMERGENCY EPJP-1 CLASSIFICATION SECTION II PROCEDURE EVENT CLASSIFICATION MATRIX 6.0 HAZARDS

'A I flflr.nnra..n, - --

DLMAvtUtJllUIN DEISCRIPTION

- D UIT r I 6.8-U C

Vehicle crash (for example; aircraft or barge) into 4 C

plant structures or systems within the protected n area boundary. C t_1 M

4 OPERATING CONDmON: 3

-All 6.8-A Vehicle crash (for example; aircraft or barge) into ANY Plant vital area.

OPERATING CONDITION:

-All CQ

-3 M

+ L 0

2 zt M

C]

I _ _ _ _ _ _ __ _ _ _ _ _I 6.0 HAZARDS PAGE 59 OF 207 REVISION 28

EPIP-1 EMERGENCY SECTION II CLASSIFICATION 6.0 HAZARDS EVENT CLASSIFICATION MATRIX PROCEDURE THIS PAGE INTENTIONALLY BLANK REVISION 28 PAGE 60 OF 207 6.0 HAZARDS

EMERGENCY EPIP-1 CLASSIFICATION SECTION II 7.0 NATURAL PROCEDURE EVENT CLASSIFICATION MATRIX EVENTSP 1VRNT NATUAL EENT 70 7.0 NATURAL PAGE 61 OF 207 REVISION 28 EVENTS

EPIP-1 EMERGENCY 7.0 NATURAL SECTION II CLASSIFICATION

• EVENTS FVENT CTLA.R4Z1rCA nA n1 M1rATD1 lrane T -T OT-E-S-:-- x rXiAJLLpUil NOTES:

CURVES/TABLES:

REVISION 28 PAGE 62 OF 207 7.0 NATURAL EVENTS

EMERGENCY EPIP-1 CLASSIFICATION SECTION II 7.0 NATURAL PROCEDURE EVENT CLASSIFICATION MATRIX EVENTS DELuSCOlUN DErSqRyPT^n

.- I 7.1-U 7.2-UJ Valid annunciation in Unit One Control Room, Panel Report by plant personnel of Tornado striking within I-XA-554B, Window 29, START OF STRONG the protected area boundary.

MOTION ACCELEROGRAPH AND Assessment by Unit One and Two Control Room personnel that an earthquake has occurred.

OPERATING CONDmION: OPERATING CONDITION:

-All -All A7.2-A Any ofthe following annunciations in Unit One Confrol Tornado striking plant vital area Room, Panel I-XA-554B: - OR

-Window 22, SEISMIC TRIGGER A Onsite wind speed above 90 MPH as indicated using t

,Window 23, SEISMIC TRIGGER B the meteorological data screen of the Integrated

-Window 30, SEISMIC TRIGGER C Computer System (ICS).

AND Assessment by Unit One and Two Control Room personnel that an earthquake has occurred. OPERATING CONDITION:

OPERATING CONDITION: -All

-All CA 3-

.5-tm

!2h L ____

__ ____ ___ ____ ____ ___I 7.0 NATURAL PAGE 63 OF 207 REVISION 28 EVENTS

EPIP-1 EMERGENCY 7.0 NATURAL SECTION II CLASSIFICATION EVENTS EVENT CLASRFICATInN MATUrTY MDd'%T.WTTTDT.'

_*=- - -. _---l L NOTES:

CURVES/TABLES:

REVISION 28 PAGE 64 OF 207 7.0 NATURAL EVENTS

EMERGENCY EPIP-1 CLASSIFICATION SECTION II 7.0 NATURAL PROCEDURE EVENT CLASSIFICATION MATRIX EVENTS I U KSIS]IUE DESCRIPTION 73U (t

Wheeler Lake level greater than elevation 565 FT. 112 AND C rA Water entering permanent plant structures due to flooding.

t*

ti 9

OPERATING CONDITION: 2

-All Oq 7.3-Al Wheeler Lake level greater than elevation 565 FT.

AND Either of the following conditions exists:

• Breech or failure of any water-tight structure causing flooding of the structure.

• Affecting equipment required for safe shutdown.

OPERATING CONDITION:

-All rA 0-4

• q M

M A

PC 0

M r) 0<

i I r

C4 z

PAE 6I O 0 RE ISI N 2 7.0 NATURAL PAGE 65 OF 207 REVISION 28 EVENTS

EPIP-1 EMERGENCY 7.0 NATURAL SECTION II CLASSIFICATION EVENTS EVENT CLASSIFICATION MATRiX ppnR~tnlTR.~

PR 'I1TD THIS PAGE INTENTIONALLY BLANK REVISION 28 PAGE 66 OF 207 7.0 NATURAL EVENTS

EMERGENCY EPIP-1 CLASSIFICATION SECTION II PROCEDURE EVENT CLASSIFICATION MATRIX 8.0 OTHER EMERGENCY DIRECTOR JUDGEMENT 8.0 8.0 OTHER PAGE 67 OF 207 REVISION 28

EPIP-1 EMERGENCY SECTION II CLASSIFICATION 8;0 OTHER EVENT CLASSIFICATION MATRIX DPDAfvn.MT~rl NOTES:

mm CURVES/TABLES:

St-'t0~,0,, t f ............ ....... ...... ........... ..

ONSITE COMMUNICATION OFFSITE COMMUNICATIONS PLANT PHONE SYSTEM NODE 1 BELL LINES TWO WAY RADIO (CH Fl, F2, F3, F4, and F5) DIGITAL MICROWAVE SOUND POWER PHONES NRC (FTS-2000)

CELLULAR PHONES (IF AVAILABLE)_

REVISION 28 PAGE 68 OF 207 8.0 OTHER

EMERGENCY EPIP-1 CLASSIFICATION SECTION II PROCEDURE EVENT CLASSIFICATION MATRIX 8.0 OTHER

-S.

I 5 S A 5 DESCRIPTION d DESCRIPTION 8.1-U i 8.2-U I Inability to reach required shutdown condition within Unplanned loss of onsite communication listed in Technical Specification Limiting Conditions for Table 8.2-U that defeats the Plant Operations Operation (LCO) limits. Staffs ability to perform routine operations r~

OR Unplanned loss of ALL offsite communication listed in 317 Table 8.2-U.

OPERATING CONDITION:

-Mode I - Mode 3 OPERATING CONDITION:

- Mode 2 -All I .

i IL 4 4 (7)

M 4

tm t"

M 9

0 P

4 q

A - -

8.0 OTHIER PAGE 69 OF 207 REVISION 28

EPLP-1 EMERGENCY SECTION II CLASSIFICATION 8.0 OTHER EVENT r1.A1.R.RarATrNq msTDyv

- - - - -~x x' JrIJ&%%LPUKII EVI~T CA~ww'r~'-T ATD NOTES:

8.3 Significant transient includes response to automatic or normally initiated functions such as scrams or runbacks involving greater than 25% core thermal power change, Emergency Core Cooling System (ECCS) injections, or thermal power oscillations of 10% or greater.

CURVES/TABLES:

REVISION 28 PAGE 70 OF 207 8.0 OTHER

EMERGENCY EPIP-1 CLASSIFICATION SECTION II PROCEDURE EVENT CLASSIFICATION MATRIX 8.0 OTHER I k6JLI6] WLI 3LIk1 INI KWA Wi HIKE Unplanned loss of most or all safety system annunciators or indicators which causes a significant loss Z of plant assessment capability for greater than 15 minutes AND Compensatory non-alarming safety system indications are available (SPDS, ICS)

AND In the opinion of the Shift Manager, increased surveillance is required to safely operate the plant.

OPERATING CONDITION:

-Mode I -Mode3

-Mode 2 8.3-A K Unplanned loss of most or all safety system annunciators or indicators which causes significant loss of plant assessment capability for greater than 15 minutes AND In the opinion of the Shift Manager, increased surveillance is required to safely operate the plant AND EITHER of the following conditions exists:

• Compensatory non-alarming safety system indications are NOT available (SPDS, ICS)

• A significant transient is in progress.

OPERATING CONDITION:

-Mode I -Mode 3

- Mode 2 8.3-S Kt i Loss o most or all annunciators associated with safety systems 3 AND Compensatory non-alarming safety system indications are NOT available (SPDS, ICS)

AND Indications needed to monitor safety functions are NOT available (Refer to Table 8.3-S)

AND t A significant transient is in progress Z OPERATING CONDITION:

-Mode -Mode 3

- Mode 2 Cz M

8.0 OTHER PAGE 71 OF 207 REVISION 28

EPIP-1 EMERGENCY SECTION II CLASSIFICATION 8.0 OTHER EVENT CLASSIFICATION MATRIX PROCEDURE NOTES:

8.4-U Table 8.4-U Mains only example eve E j Urmayj sutEvea Thisevn daafit isnteIddto adma a)rcstekmrt expcitlyadaciseddwhaxe butwmwbanoofanantidagaryf crifes eils wihhtheEm mDi~ iewtorlbdie lmurnieUalEvatCaificabn Additllhis EAL shioldbe cnmiddm malkng cLaificnosnudig dvilges t fiionpnditbanrieos ntWecifi*

elsewmmtheEALmatnx 8.4-A This ev dassificaui is dtoento addimuen~ipxittd o 1ddaA4 dthat bit Ivnwant dedaamionnof nmyb~m cx ediichfite hs tEnxrDiietorbev to fillunlerth A]t clashficatio AddruallytisEAL dbe con s nmmakngkongcrgaaficaos egading challngs tofi prodixanimris ol ckaddlyes hd4eeinteEAL Iatrix 8.4-S This emt dashatio ismired to addrkss conditios e'qilicitly addesd elsewhr bt that wan'nn dc~lamtion of ana geiybecause codits xtwichiidthe Site EnryDimedorbelieves to fil underthe Site Area Engencyclassficaton AdditsaE EAL shdbei cnsndeinninakmaig edycicatos gudig challenge to fisnptdiophtbamimono tiTficallyidzi dein theEALmatnx 8.4-G This evem clas ion ismkned to ais ma ed tont elctly add eswlxit thatan vanan cdarattonof an enr geeybe auditiios AfichthetmX Sieg En yDDtorbyetofab unlerthe GoaI Enxrnydass o AdldiinathisEAL sxxldbe med inmmakig mergudasaficns teguding hallngesto fipnodLdutbarier not ibcuyadessdetintheEAL matix CURVES/TABLES:

EXAMPLE UNUSUAL EVENTS PLANT TRANSIENT RESPONSE UNEXPECTED OR NOT UNDERSTOOD UNANALYZED SAFETY SYSTEM SONFIGURATION AFFECTING, THREATENING SAFE SHUTDOWN INADEQUATE PERSONNEL TO ACHIEVE OR MAINTAIN SAFE SHUTDOWN DEGRADED PLANT CONDITIONS BEYOND LICENSE BASIS THREATENING SAFE OPERATION OR SAFE SHUTDOWN EMERGENCY PROCEDURES NOT ADEQUATE TO MAINTAIN SAFE OPERATION OR ACHIEVE SAFE SHUTDOWN REVISION 28 PAGE 72 OF 207 8.0 OTHER

EMERGENCY EPIP-1 CLASSIFICATION SECTION II PROCEDURE EVENT CLASSIFICATION MATRIX 8.0 OTHER

&s I f- I 8.4-U INN TC~

Other events are in process or have ocrredwhichindicate a pontal degradationmihe level ofsafety ofthe plant No C radioactive releases are exected which require offaite respanse. Refer to Table 8.4-U.

OR Any loss or any potential loss of containment OPERATING CONDmON:

q

-All 8.4-A X Other events are in process or have occurred winchinvolve an actual orpotential substantial degradationin the level of safety ofthe plant Radioactive releases are expeted to be within a small fiaction ofthe EPA guidelines.

OR Any loss or potential loss of fuel cladding or RCS pressure boundary OPERATING CONDITION:.

-All 8.4-S K Other events are in process or have occurred which involve actual or likely major failures of plant functions needed for protection of the public. Radioactive releases are NOT expected to result in exposure levels that exceed EPA guidelines except near the site boundary.

OR Any loss or potential loss of both fuel cladding and RCS pressure boundary OR Potential loss of either fuel cladding or RCS pressure boundary and loss of any additional barrier Z OPERATING CONDmON:

-ALL 8.4-G KI Other events are in process or have occurred which involve actual or imminent substantial core degradation or melting with potential for loss of containment integrity. Radioactive releases are expected to exceed EPA guidelines for exposure levels offsite beyond the site boundary.

OR Loss of any two barriers and potential loss of third barrier OPERATING CONDITION: -

-All 8.0 OTHER PAGE 73 OF 207 REVISION 28

EPIP-1 EMERGENCY SECTION II CLASSIFICATION 8.0 OTHER EVENT 'TARWRMATrnnN MIAT1DY IMIMXA'WTTTli

- - . -- , A rn'JLLJJUIlt THIS PAGE INTENTIONALLY BLANK REVISION 28 PAGE 74 OF 207 8.0 OTHER

EMERGENCY EPIP-1 CLASSIFICATION SECTION -E PROCEDURE TECHNICAL BASIS 1.0 REACTOR M

REACTOR 1.0 1.0 REACTOR PAGE 75 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 1.0 REACTOR UNUSUAL EVENT Uncontrolled water level decrease in Reactor Cavity with irradiated fuel assemblies expected to remain covered by water.

OPERATING - Mode 5 CONDITION BASIS This event classification only applies during Mode 5 when the Reactor Head is removed. For the purposes of this event classification the Reactor Cavity includes the cavity and the Reactor Vessel.

This event classification is anticipatory to 1.1-Al and should only be considered if, in the opinion of the Site Emergency Director, the water level decrease is substantial enough to ultimately result in increased dose rates in the area of the Reactor Cavity due to loss of shielding by water covering irradiated fuel.

Uncontrolled water level decrease during Mode 5 is indicative of valve manipulation error or failure of equipment in such a manner as to cause uncontrolled drainage of the Reactor Cavity. Uncontrolled water level decrease may be detected by the presence of the low level alarm in the spent fuel storage pool, visual observation, increased radiation levels or various other symptoms that the Site Emergency Director considers valid indicators of the event.

The degraded status of safety systems designed to makeup water to the Reactor Vessel is of particular concern during Mode 5 although plant Technical Specifications require minimum makeup systems be operable except with the spent fuel storage gates removed and water level > 22 feet over the top of the reactor pressure vessel flange. These events tend to have long lead times relative to potential for release outside the site boundary, thus impact to public health and safety is very low. Classification as Unusual Event is warranted as a precursor to a more serious event.

Escalation to Alert is by actual uncovery of irradiated fuel assemblies.

1.0 REACTOR PAGE 76 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION m PROCEDURE TECHNICAL BASIS 1.0 REACTOR UNUSUAL EVENT (CONTINUED)

REFERENCES Reg Guide 1.101 Rev. 3, (NUMARC-AU2 example -1)

Technical Specifications 3.5.2 NOTES NOTE 1.1-UI/1.1-A1 Applicable when the Reactor Head is removed and the

• Reactor Cavity is flooded.

1.0 REACTOR PAGE 77 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION mI[

PROCEDURE TECHINICAL BASIS 1.0 REACTOR UNUSUAL EVENT Uncontrolled water level decrease in Spent Fuel Storage Pool with irradiated fuel assemblies expected to remain covered by water.

OPERATING - All CONDITION BASIS This event classification is anticipatory to 1.1 -A2 and should only be considered if, in the opinion of the Site Emergency Director, the water level decrease is substantial enough to ultimately result in increased dose rates in the area of the Spent Fuel Storage Pool due to loss of shielding by water covering irradiated fuel.

Uncontrolled water level decrease may be detected by the presence of the low level alarm in the spent fuel storage pool, visual observation, increased radiation levels or various other symptoms that the Site Emergency Director considers valid indicators of the event.

Uncontrolled water level decrease in Spent Fuel Storage Pools is indicative of failure of equipment in such a manner as to cause uncontrolled drainage. These events tend to have long lead times relative to potential for release outside the site boundary, thus impact to public health and safety is very low. Classification as Unusual Event is warranted as a precursor to a more serious event.

Escalation to Alert is by actual uncovery of irradiated fuel assemblies.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-AU2 example-2) 1.0 REACTOR PAGE 78 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECIINICAL BASIS 1.0 REACTOR ALERT Uncontrolled water level decrease in Reactor Cavity expected to result in irradiated fuel assemblies being uncovered.

OPERATING - Mode 5 CONDITION BASIS This event classification only applies during Mode 5 when the Reactor Head is removed. For the purposes of this event classification the Reactor Cavity includes the cavity and the Reactor Vessel.

Uncontrolled water level decrease during Mode 5 is indicative of valve manipulation error or failure of equipment in such a manner as to cause uncontrolled drainage of the Reactor Cavity. The degraded status of safety systems designed to makeup water to the Reactor Vessel is of particular concern during Mode 5 although plant Technical Specifications require minimum makeup systems be operable except with the spent fuel storage gates removed and water level > 22 feet over the top of the reactor pressure vessel flange.

Uncontrolled water level decrease may be detected by visual observation, increased radiation levels or various other symptoms that the Site Emergency Director considers valid indicators of the event. Expected fuel uncovery may be detected by increased radiation levels, Visual observation, RPV level instrumentation expected to drop below -162 inches, or best judgement of the Site Emergency Director based on present and past events and trends.

Due to the long lead times associated with these events there is time available to take corrective actions, and there is little potential for substantial fuel damage.

Significant exposures to onsite personnel is likely during these events and it is probable that additional personnel will be needed onsite; therefore the Alert classification is warranted.

Escalation is by Radiological Release event classifications.

1.0 REACTOR PAGE 79 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION mII PROCEDURE TECHNICAL BASIS 1.i0 RFACTOR i'AUMMaE1'A. Ewi ALERT (CONTINUED)

REFERENCES - Reg Guide 1.101 Rev. 3, (NULvIARC-AA2 example-3)

Technical Specifications 3.5.2 NOTES NOTE 1.1-Ul/1.1-Al Applicable when the Reactor Head is removed and the Reactor Cavity is flooded.

1.0 REACTOR PAGE 80 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 1.0 REACTOR ALERT Uncontrolled water level decrease in Spent Fuel Storage Pool expected to result in irradiated fuel assemblies being uncovered.

OPERATING - All CONDITION BASIS Uncontrolled water level decrease in Spent Fuel Storage Pools is indicative of failure of equipment in such a manner as to cause uncontrolled drainage. These events tend to have long lead times relative to potential for release outside the site boundary, thus impact to public health and safety is very low.

Uncontrolled water level decrease may be detected by visual observation, increased radiation levels or various other symptoms that the Site Emergency Director considers valid indicators of the event. Expected fuel uncovery may be detected by increased radiation levels, Visual observation, or best judgement of the Site Emergency Director based on present and past events and trends.

There is time available to take corrective actions, and there is little potential for substantial fuel damage. Offsite exposures are expected to remain below the Environmental Protection Agency's Protective Action Guidelines; however, exposures to onsite personnel is of particular concern during this event; therefore the Alert classification is warranted.

Escalation is by Radiological Release event classifications.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUIMARC-AA2 example-4) 1.0 REACTOR PAGE 81 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 1.0 REACTOR A A ___

SITE AREA EMERGENCY Reactor water level cannot be maintained above -162 in. (TAF).

OPERATING - ALL BASIS If Reactor water level cannot be maintained above TAF the potential exist for fuel cladding damage. Events most likely to result in coolant inventory loss to this extent are RCS boundary degradation events or station blackout events. For this event to be declared, RPV water level must have decreased or be trending to a value that, in the opinion of the Site Emergency Director, has resulted in or will result in some actual core uncovery. Additionally, the Site Emergency Director must have evidence that Reactor level has been or can be recovered to above TAF.

This event classification also applies in Mode 5 when the Reactor Vessel head is installed. Inadvertent draining of the Reactor Vessel is possible under these conditions due to valving errors associated with the RHR system or failures associated with isolation valves during alignment changes of systems connected to the Reactor Vessel below the normal water level.

The fact that the transient was severe enough to result in inability to maintain RPV level coupled with the anticipatory nature of this event classification as a precursor to more serious event warrants the Site Area Emergency event classification.

For events that occur during operation, escalation to General Emergency is based on inability to assure adequate core cooling by restoring and maintaining RPV water level following transients that have resulted in extreme RPV water level decrease. For events that occur during shutdown or Mode 5, escalation is by radioactive release event classifications.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-FS, SS5, SS4, example-i)

- EOI Program Manual Section VI-J NOTES NOTE 1.1-SI Applicable in Mode 5 when the Reactor Head is installed.

1.0 REACTOR PAGE 82 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION HI PROCEDURE TECHNICAL BASIS 1.0 REACTOR TECHNICAL BASIS 1.0 REACTOR SITE AREA EMERGENCY Reactor water level cannot be determined.

OPERATING - Mode 1 CONDITION - Mode 2

- Mode 3 BASIS Inability to determine Reactor water level during operation may be due to boiling in the reference or variable instrument legs, instrument power failures, or conflicting information from uncontrolled indicator oscillations.

This condition requires Reactor flooding following emergency depressurization.

Adequate core cooling is assured by these measures. Due to the severity of these actions and the uncertainty of Reactor status it is appropriate to treat this as a potential loss for Reactor Coolant System and Fuel Cladding integrity; therefore, this event is appropriate for the Site Area Emergency classification.

Escalation to General Emergency is based on inability to assure adequate core cooling in this mode.

REFERENCES - Reg Guide 1.101 Rev. 3, (NJMARC-FS)

- EOI Program Manual Section VI-J I

1.0 REACTOR PAGE 83 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 1.0 REACTOR GENERAL EMERGENCY Reactor water level CANNOT be restored and maintained above -190 IN.

OPERATING - Mode l CONDITION - Mode 2

- Mode 3 BASIS If Reactor water level cannot be restored and maintained above -190 inches

[Minimum Steam Cooling Reactor Water Level (MSCRWL)], core damage is possible due to inadequate steam generation, by the covered portion of the Reactor core, to remove decay heat and prevent cladding heatup to a point that results in clad failure.

For either of the above conditions to be met, the control room operators should have progressed in the execution of the EOIs to the point that all high pressure and all low pressure systems that are available within a reasonable time frame have been attempted and are unsuccessful in reversing the adverse RPV water level trend.

Events most likely to result in coolant inventory loss or loss of makeup capability to this extent are RCS boundary degradation events or events resulting from loss of multiple systems such as station blackout. During such transients or accidents the potential for Primary Containment failure increases substantially; therefore, the General Emergency classification is appropriate.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-FG).

- -EOI Program Manual Section VI-J 1.0 REACTOR PAGE 84 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION mII PROCEDURE TECIINICAL BASIS 1.0 REACTOR A

GENERAL EMERGENCY Reactor water level CANNOT be determined AND EITHER of the following conditions exists:

• The reactor will remain subcritical w/o boron under all conditions and: Less than 4 MSRVs can be opened, or Reactor pressure CANNOT be restored and maintained at least 65 PSI above Suppression Chamber pressure.

• It has NOT been determined that the reactor will remain subcritical w/o boron under all conditions and unable to restore and maintain MARFP in Table 1.1-G2.

OPERATING - Mode 1 CONDITION - Mode 2

- Mode 3 BASIS Inability to determine Reactor water level during operation may be due to boiling in the reference or variable instrument legs, instrument power failures, or conflicting information from uncontrolled indicator oscillations. This condition requires Reactor flooding following emergency depressurization. It the reactor will remain subcritical without (w/o) boron under all conditions, adequate core cooling is assured only if at least 4 MSRVs are opened and Minimum Reactor Flooding Pressure (MRFP) is maintained with Reactor pressure at least 65 PSI above Suppression Chamber pressure. If it has not been determined that the reactor will remain subcritical without (w/o) boron under all conditions, adequate core cooling can only be assured when the Minimum Alternate Reactor Flooding Pressure (MARFP) is restored and maintained. If adequate core cooling is not assured core damage is probable under this scenario due to the extreme nature of the plant conditions that resulted in the inability to determine Reactor level (i.e.,

high containment temperatures, loss of multiple power supplies, etc.). Primary Containment integrity cannot be assured under all these conditions; therefore, the General Emergency classification is appropriate.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-FG)

- EOI Program Manual Section VI-J 1.0 REACTOR PAGE 85 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION Im[

PROCEDURE TECHNICAL BASIS 1.0 REACTOR TECHNICAL BASIS 10 REACTOR GENERAL EMERGENCY (CONTINUED)

CURVES/TABLES NUMBER OF OPEN MSRVs MARFP (PSIG)

.. I 6 or More 180 I 5 220 4 280 NOTES NOTE 1.1-G2 The reactor will remain subcritical under all conditions w/o boron when:

• All control rods are inserted to or beyond position 02

• All control rods except one are inserted to or beyond position 00

• Determined by reactor engineering 1.0 REACTOR PAGE 86 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 1.0 REACTOR ALERT Failure of automatic scram functions to bring the Reactor subcritical AND Manual scram or Alternate Rod Insertion (ARI) was successful.

OPERATING - Mode 1 -

CONDITION - Mode 2 BASIS A manual scram is any set of actions by the Reactor Operator(s) at the Reactor Control Console which causes control rods to be rapidly inserted into the core and brings the Reactor subcritical.

This event classification indicates failure of the RPS to automatically scram the Reactor. This condition is more than a potential degradation of a safety system in that a front line automatic protection system did not function in response to a plant transient and thus plant safety has been compromised, and design limits of the fuel may have been exceeded.

An Alert is indicated because conditions exist that lead to potential loss of fuel clad or RCS barrier. Any set of actions by the Reactor Operator at Panel 9-5 that cause control rods to rapidly insert into the core and bring the Reactor subcritical is considered a manual scram.

Escalation to Site Area Emergency is based on fuel clad barrier or RCS barrier event classifications.

REFERENCE - Reg Guide 1.101 Rev. 3, (NUMARC-SA2)

NOTES NOTE 1.2 Subcritical is defined as Reactor power below the heating range and not trending upward.

1.0 REACTOR PAGE 87 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECffi'ICAL BASIS 1.0 TEHIA _AI _REACTOR_ 1. REACTOR SITE AREA EMERGENCY Failure of automatic scram, manual scram, and ARI to bring the Reactor subcritical.

OPERATING - Mode 1 CONDITION BASIS Manual scram, and ARI are not considered successful if action away from the Reactor Control Console (Panel 9-5) was required to scram the Reactor.

A failure of the automatic and manual scram systems may result in the Reactor producing more heat than the maximum decay heat load for which the safety systems are designed. A Site Area Emergency classification is appropriate because conditions exist that lead to potential loss of both fuel clad and Reactor Coolant System (RCS) barriers. Therefore, this event classification ensures timely emergency response to the event before actual barriers loss has taken place.

Escalation to General Emergency is based upon inability to bring Reactor power within decay heat removal capability before Suppression Pool temperature reaches the Heat Capacity Temperature Limit (HCTL).

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-SS2, SS4 example -1)

NOTES NOTE 1.2 Subcritical is defined as Reactor power below the heating range and not trending upward.

1.0 REACTOR PAGE 88 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 1.0 REACTOR GENERAL EMERGENCY Failure of automatic scram, manual scram, and ARI. Reactor power > 3%.

AND EITHER of the following conditions exists:

• Suppression Pool temperature exceeds HCTL.

Refer to curve 1.2-G.

• Reactor water level CANNOT be restored and maintained at or above -190 in.

OPERATING - Mode l CONDITION BASIS Automatic scram, manual scram, and ARI are not considered successful if action away from the Reactor Control Console was required to scram the Reactor.

Under these conditions all efforts, including boron injection, have been unsuccessful in bringing Reactor power within the decay heat removal capability of the Emergency Core Cooling Systems (ECCS). Additionally, an extreme challenge to the ability to cool the Reactor Core exist if Reactor Pressure Vessel (RPV) water level cannot be maintained sufficient to ensure adequate core cooling.

Another consideration is the inability to remove heat using the Main Condenser or Suppression Pool. In the event that neither heat sink is effective and Reactor power remains above this level, then a core melt sequence exists. In this situation, core degradation can occur rapidly; therefore, a General Emergency classification is appropriate in anticipation of degradation of multiple fission product barriers.

REFERENCES - Reg Guide 1.101 Rev. 3,(NUMARC-SG2)

- EOI Program Manual Section V-K and Section V-D 1.0 REACTOR PAGE 89 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION m PROCEDURE TECEINICAL BASIS 1.0 REACTOR I GENERAL EMERGENCY (CONTINUED)

CURVES/TABLES CURVE 1.2-G HEAT CAPACITY TEMP LIMIT 2 6 04 ... ........ ........... m__

250 WHEN RX_-i:i.SAE

....... PRESS... ....

240

~~ ~ ~ ~ ~..................

~ . .S~ ........ .........................................

230.

°220 0l g 2 180 R. RPV Press. 900 170 R :PV1135 Press 160 _l..........

15 0 - . -..-........

_ , . . .. ..... . .... I 11.5 12 12.5 13 13.5 14 14.5 15 15.5 16 16.5 17 17.5 18 18.5 19 SUPPR PL LVL (FT)

ACTION REQUIRED IFABOVE CURVE FOR EXISTING RX 1.0 REACTOR PAGE 90 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION m -

PROCEDURE TECHNICAL BASIS 1.0 REACTOR

- 4 arm n ueiu wu arm iii u' mn.

UNUSUAL EVENT Reactor coolant activity exceeds 26 #Ci/gm dose equivalent 1-131 (Technical Specification limit) as determined by chemistry sample.

OPERATIN( vI All CONDITIOIN BASIS Reactor coolant activity samples exceeding Technical Specification limits for Iodine spikes are representative of fuel clad degradation. An Unusual Event is declared because of potential degradation in the level of safety of the plant. Iodine levels exceeding Technical Specification limits are a potential precursor of more serious problems.

Escalation to Alert would be based on higher Reactor coolant activity values indicative of significant fuel failure.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-SU4 example-2)

- Technical Specification 3.4.6 1.0 REACTOR PAGE 91 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION Ell PROCEDURE TECHNICAL BASIS 1.0 REACTOR

- I WAU KS] USIS] WiIM ETU UII UI'm BDi ALERT Reactor coolant activity exceeds 300 pCi/gm dose equivalent Iodine-131 as determined by Chemistry sample.

OPERATING - Mode l CONDITION - Mode 2

- Mode 3 BASIS Reactor coolant activity samples exceeding 300 gCi/gm dose equivalent Iodine-131 are well above those expected for Iodine spikes and represent a significant loss of the fuel clad barrier. Any loss or potential loss of the fuel clad barrier warrants the declaration of an Alert.

Escalation to Site Area Emergency would be based on the conditions given above coupled with a loss or potential loss of either the Primary Containment or Reactor Coolant System barrier or Radiological Releases.

REFERENCE - Reg Guide 1.101 Rev. 3, (NTJMARC-FA)

- RIMS L36 921201 806 1.0 REACTOR PAGE 92 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 1.0 REACTOR Li 11 WISJU KffLI3 WA I] FAI U[Ok EB1 UNUSUAL EVENT Valid MAIN STEAM LINE RADIATION HIGH-HIGH alarm, RA-90-135C OR Valid OG PRETREATMENT RADIATION HIGH alarm, RA-90-157A.

OPERATING - Mode 1 CONDITION - Mode 2

- Mode 3 BASIS Main Steam Line radiation high high or offgas radiation high is indicative of fuel cladding leakage.

The Main Steam Line radiation high high alarm setpoint is normally set at 3 times normal full power background. 3 times normal full power background is in excess of any spikes expected from operational transients that do not result in cladding failure. This alarm setpoint is substantially above that which would be indicative of fuel cladding damage above Technical Specification allowable limits; however, the presence of a valid alarm warrants declaration of an Unusual Event and consideration of other symptoms and event classifications for possible upgrade of the event based on fission product barrier loss.

The offgas pretreatment radiation high alarm setpoint is set at a value that is indicative of the ODCM allowable limits for radiation release.

Either of these conditions is considered a potential degradation in the level of safety of the plant and a potential precursor of a more serious problem.

Escalation to the Alert is based on either Reactor coolant samples exceeding 300 giCi/gm or Drywell radiation levels indicative of loss of the fuel cladding barrier.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-SU4 example-i) 1.0 REACTOR PAGE 93 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION HI PROCEDURE TECHNICAL BASIS 1.0 REACTOR KOIXEOJ U *]kWIU Ii DII U U NI [OVAl - Ki ALERT Reactor moderator temperature CANNOT be maintained below 212'F whenever Technical Specifications require Mode 4 conditions or during operations in Mode 5.

OPERATING - Mode 4 CONDITION - Mode 5 BASIS This event classification addresses loss of decay heat removal functions when Mode 4 is required or during Mode 5. Loss of decay heat removal capability can result in more serious consequences depending upon whether Primary Containment is in tact and Emergency Core Cooling System (ECCS) equipment status. In any condition where Mode 4 is required, loss of decay heat removal capability represents a significant degradation in plant conditions that can lead to fuel cladding damage or RCS degradation. In order to maintain anticipatory philosophy the Alert classification is appropriate for this event.

Escalation to Site Area Emergency or General Emergency is by loss of Reactor water level that has or will uncover the fuel or Radiological Release Event classification.

REFERENCES Reg Guide 1.101 Rev. 3, (NIJMARC-SA3) 1.0 REACTOR PAGE 94 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 1.0 REACTOR

- KOJ¶SJ El] KWAU 11011 U U NkI [S1'A - K' SITE AREA EMERGENCY Suppression Pool temperature, level and RPV pressure CANNOT be maintained in the safe area of Curve 1.5-S (Heat Capacity Temperature Limit)

OPERATING - Mode 1 CONDITION - Mode 2

- Mode 3 BASIS Suppression Pool temperature is limited by Curve 1.5-S as a function of suppression pool level and reactor pressure in order to preclude failure of Primary Containment or equipment necessary for the safe shutdown of the plant following emergency depressurization. When Suppression Pool temperature cannot be maintained below the limits of the curve corresponding to existing suppression pool level and reactor pressure, emergency depressurization is required and continued decay heat removal at operating temperature and pressure is no longer permissible.

Suppression Pool level is limited by Curve 1.5-S to the range of 11.5 feet to 19 feet in order to preclude failure of Primary Containment or equipment necessary for the safe shutdown of the plant and preserve the pressure suppression function of the containment for possible future emergency depressurization. When Suppression Pool level cannot be maintained within the limits of the curve, continued decay heat removal at operating pressures and temperatures is no longer permissible and emergency depressurization is required.

Exceeding the limits of Curve 1.5-S represents a loss of heat sink for decay heat removal and inability to maintain Mode 3. Under these conditions there is an actual failure of systems intended for protection of the public; therefore, Site Area Emergency is warranted. Escalation to General Emergency is by Abnormal Rad levels, Radiological Release or Primary Containment failure events.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-SS4)

- EOI Program Manual Sections VI-C and VI-F 1.0 REACTOR PAGE 95 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION mI -

PROCEDURE TECHNCAL BASIS 1.0 REACTOR SITE AREA EMERGENCY (CONTINUED)

CURVES/TABLES CURVE 1.5-S HEAT CAPACITY TEMP LIMIT 260 -.-  ;  ;

250 WHE .XPRESSBt 23180 RPV Press. 900 l S .

~*****. IS BELO 4 <. * - .65 PSIG 0 170 RPV Press 1135 .

160 11.5 12 12.5 13 13.5 14 14.5 15 15.5 16 16.5 17 17.5 18 18.5 19 SUPPR PL LVL (FT a- ACTION REQUIRED F VE CURVE FOR EXISTING RX 1.0 REACTOR PAGE 96 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION m 2.0 PRIMARY PROCEDURE TECHNICAL BASIS CONTAINMENT PRIMARY C ONTAINMENT 2.0 2.0 PRIMARY PAGE 97 OF 207 CONTAINMENT REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III 2.0 PRIMARY PROCEDURE TECHNICAL BASIS CONTAINMENT CONTAINMENT

- U4 ILl FA I'K&6]M W..I IILi I NI Eli I IIfiU I mE Wj ALERT Drywell pressure at or above 2.45 PSIG AND Indications of Primary System leakage into Primary Containment.

Refer to Table 2.1-A.

OPERATING - Mode l CONDITION - Mode 2

- Mode 3 BASIS If Drywell pressure reaches the ECCS initiation and Reactor scram setpoint (2.45 PSIG) there is clear indication that a Primary System leak of sufficient magnitude exists that could result in break propagation leading to significantly larger loss of Reactor coolant inventory.

Efforts to reduce Drywell pressure by additional cooling or Primary Containment venting have been unsuccessful either due to equipment malfunction or the magnitude of the leak. This condition represents a degraded level of safety of the plant due to Reactor Coolant System (RCS) degradation and warrants the Alert classification.

Escalation to Site Area Emergency is by exceeding the Pressure Suppression Pressure Limit or inability to maintain Reactor water level above TA.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-FA )

2.0 PRIMARY PAGE 98 OF 207 SI=

CONTAINMENT REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION m 2.0 PRIMARY PROCEDURE TECHNICAL BASIS CONTAINMENT CONTAINMENT

- UI IL' Ui I'KUi]M KI IIL1 I NU U UI OL'11U I 0mW 5i ALERT (CONTINUED)

CURVES/TABLES I .. . . . . . .......... ....... . . .-. - ...... .?::.. .?.:.::::

.::::?-:'t ubt JX PRIMARY CONTAINMENT PRESSURE HF DRYWELL FLOOR DRAIN SUMP PUMP E UKY WELL CAM ACTIVITY INCREASING DRYWELL TEMPERATURE HIGH ALARM CHEMISTRY SAMPLE RADIONUCLIDE COMPARSION TO RX WATER 2.0 PRIMARY PAGE 99 OF 207 CONTAINMENT REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION HI 2.0 PRIMARY PROCEDURE TECHNICAL BASIS CONTAINMENT

- UI Iki LI I'Lk6k1 WA killI ULU U U4 DRItU 4 inu SITE AREA EMERGENCY Suppression Chamber pressure CANNOT be maintained in the safe area of Curve 2.1-S (Pressure Suppression Pressure Curve).

OPERATING - Mode l CONDITION - Mode 2

- Mode 3 BASIS The inability to maintain Primary Containment pressure in the safe region of the Pressure Suppression Pressure curve indicates that Drywell and Suppression Chamber sprays cannot be initiated or are not effective in reversing an increasing trend in Primary Containment pressure.

Primary Containment pressure and Suppression Pool water level outside the safe region of Curve 2.1 -S represents loss of ultimate heat sink and inability to maintain hot shutdown.

Escalation to General Emergency is based on Primary Containment pressure reaching 55 PSIG or event classifications resulting from loss or potential loss of the fuel clad and RCS Barriers.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-SS4)

- EOI Program Manual Section V-D 2.0 PRIMARY PAGE 100 OF 207 I.

CONTAINMENT REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III 2.0 PRIMARY PROCEDURE TECHNICAL BASIS CONTAINMENT SITE AREA EMERGENCY (CONTINUED)

CURVES/TABLES CURVE 2.1-S PRESS SUPPR PRESS 30 0 REQUIRE

° 25 .. ...............

CO tr~~~~~~ ~ 20 L-60t0 . ............ <tto<n m - m i:  ::~~~~~~~.:-.I.............. ..,';'"

I~~~~~~~ ~ {.t.....""y .0 . '

50:2I. .......

Ai' -.' SAFE ............................... ...

co O5 ...:, ,:,.............

,': . .. . . .. .. .. . H 0..... ..

0.... ...

1111.512 13 14 15 16 17 18 19 20 SUPPR PL LVL (FT) 2.0 PRIMARY PAGE 101 OF 207 CONTAINMENT REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III 2.0 PRIMARY PROCEDURE TECHNICAL BASIS CONTAINMENT

- UU ILl M I'EKSJM WAI IILl I NI U UU OLIIJ U oinu k GENERAL EMERGENCY Suppression Chamber pressure CANNOT be maintained below 55 PSIG.

OPERATING - Mode 1 CONDITION - Mode 2

- Mode 3 BASIS Primary Containment pressures of this magnitude represent loss of the RCS barrier and require emergency venting of Primary Containment irrespective of offsite release rates. Fuel cladding integrity is threatened either directly due to loss of Reactor coolant inventory or potentially due to direction in the Emergency Operating Instructions (EOIs) to spray Primary Containment irrespective of whether adequate core cooling is assured. Under these conditions, potential loss of fuel cladding integrity should be assumed; therefore, the General Emergency classification is appropriate.

REFERENCES - Reg Guide 1.101 Rev. 3,(NUMARC-FG)

- EOI Program Manual Section V-D 2.0 PRIMARY PAGE 102 OF 207 CONTAINMENT REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION HI 2.0 PRIMARY PROCEDURE TECHNICAL BASIS CONTAINMENT CONTAINMENT

- U4 IL' LI I'kUIOkI Vi kiLl I NI U Iii I] (SIfl Mm#fl SITE AREA EMERGENCY Drywell or Suppression Chamber hydrogen concentration at or above 4%

AND Drywell or Suppression Chamber oxygen concentration at or above 5%.

OPERATING - Mode 1 CONDITION - Mode 2

- Mode 3 BASIS Hydrogen and oxygen concentrations in this range are an indication that an event has occurred more severe than any that have been analyzed in the FSAR or that systems installed for control of hydrogen and oxygen have been unsuccessful in stopping an upward trend in these concentrations. Concentrations of this magnitude indicate severe fuel degradation and are approaching the lower deflagration limits for combustible mixture. Venting through the Standby Gas Treatment System to the elevated release path (Stack) is required. Dilution is accomplished by use of Containment Atmosphere Dilution (CAD) system to control hydrogen and oxygen. Releases can be expected to approach levels associated with the Site Area Emergency Radiological Release event classification.

Escalation to General Emergency is based on higher concentrations of hydrogen which directly threaten Primary Containment integrity and require emergency venting through large unfiltered pathway or through Radioactive Release event classifications.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-FS)

- EOI Program Manual Section V-D 2.0 PRIMARY PAGE 103 OF 207 CONTAINMENT REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION m11 2.0 PRIEMARY PROCEDURE TECHNICAL BASIS CONTAINMENT CONTAINMENT

- UI ILl Kt I'KUIOJI WI WILl I MU U I hi U ISIH Nink#EI GENERAL EMERGENCY Drywell or Suppression Chamber hydrogen concentration at or above 6%

AND Dryweil or Suppression Chamber oxygen concentration at or above 5%.

OPERATING - Mode 1 CONDITION - Mode 2

- Mode 3 BASIS Hydrogen and oxygen concentrations in this range are an indication that an event has occurred more severe than any that have been analyzed in the FSAR or that systems installed for control of hydrogen and oxygen have been unsuccessful in stopping an upward trend in these concentrations. Concentrations of this magnitude indicate severe fuel degradation in conjunction with RCS barrier failure and have reached the lower deflagration limits for combustible mixture. This event constitutes a potential loss of the Primary Containment barrier because Ignition of the mixture could result in Primary Containment failure. Procedures require venting and purging through the large unfiltered pathway for control of hydrogen and oxygen representing a loss of Primary Containment; therefore the General Emergency classification is appropriate.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-FG)

- EOI Program Manual Section V-D 2.0 PRIMARY PAGE 104 OF 207 CONTAINMENT REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III 2.0 PRIMARY PROCEDURE TECHNICAL BASIS CONTAINMENT CONTAINMENT

• ]I'&ii NEU ii 5] VI U[6]in4N ALERT Drywell radiation levels at or above the values listed in Table 2.3-A/2.3-S2 with the RCS barrier intact.

OPERATING - Mode 1 CONDITION - Mode 2

- Mode 3 BASIS Drywell radiation monitor readings as indicated in Table 2.3-A/2.3-S2 are indicative of significant fuel failure. These values are different for Unit 2 & 3 due to detector geometry, and relative shielding. These values have been calculated assuming an Iodine inventory associated with a concentration of 300 pCi/gm dose equivalent Iodine-131 in Reactor coolant with the RCS barrier intact. Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations (including Iodine spikes) allowed within Technical Specifications. This level indicates a loss of the fuel clad barrier; therefore, the Alert classification is warranted.

Escalation to Site Area Emergency is based upon higher Drywell radiation levels indicative of loss of the Reactor Coolant System pressure boundary in conjunction with significant fuel failure or loss or potential loss of Primary Containment in conjunction with significant fuel failure.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-FA)

Calculation ND-N0090-930055 R3 (Unit 2/3)

Technical, Specifications 3.4.6 CURVES/TABLES U. '3 RAD MONITOR RIHR RAD MONITOR R/HR 2-RE-90-272A 2-RE-90-273A 345 164 3-RE-90-272A 3-RE-90-273A

=

t 106 164 2.0 PRIMARY PAGE 105 OF 207 CONTAINMENT REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III 2.0 PRIMARY PROCEDURE TECHNICAL BASIS CONTAINMENT

• ]I'&1I NEU WA I] LI USJin5i -

SITE AREA EMERGENCY Drywell radiation levels at or above 4880 R/HR with the RCS barrier not intact.

OPERATING - Mode 1 CONDITION - Mode 2

- Mode 3 BASIS A Drywell radiation monitor reading of 4880 R/HR is indicative of Reactor Coolant System pressure boundary failure in conjunction with significant fuel failures. The value has been calculated assuming instantaneous release and dispersal of Reactor coolant noble gas and Iodine inventory associated with a concentration of 300 jiCi/gm dose equivalent Iodine-13 1 into the Drywell atmosphere. This value is significantly higher than that specified for fuel clad barrier degradation; therefore, this level indicates either a loss of both the fuel clad barrier and RCS barrier or severe cladding degradation.

Escalation to General Emergency is based upon either loss or potential loss of the Primary Containment barrier or significantly higher radiation levels indicative of gross amounts of fission products in Primary Containment.

REFERENCE - Reg Guide 1.101 Rev. 3, (NUMARC-FS)

- Calculation ND-N0090-930050 R2 (Unit 2/3) 2.0 PRIMARY PAGE 106 OF 207 CONTAINMENT REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III 2.0 PRIMARY PROCEDURE TECHNICAL BASIS CONTAINMENT

• ] tI'&i'A HUE WU] FA fl [*] a War. -

SITE AREA EMERGENCY Drywell radiation levels at or above the values listed in Table 2.3-A/2.3-S2 with the RCS barrier intact.

AND Either of the following exists:

• Indications of loss of Primary Containment. Refer to Table 2.3/2.5-U.

• Primary Containment integrity CANNOT be maintained.

OPERATING - Mode 1 CONDITION - Mode 2

- Mode 3 BASIS Drywell radiation monitor readings as indicated in Table 2.3-A/2.3-S2 are indicative of significant fuel failure. These values are different for Unit 2 & 3 due to detector geometry, and relative shielding. These values have been calculated assuming an Iodine inventory associated with a concentration of 300 ICi/gm dose equivalent Iodine-131 in Reactor coolant with the RCS barrier intact. Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations (including Iodine spikes) allowed within Technical Specifications. This level indicates a loss of the fuel clad barrier. Table 2.3/2.5U contains indications of loss of primary containment. Primary Containment cannot be maintained encompasses any condition or set of conditions based on current trends or anticipated circumstances that, in the judgement of the Site Emergency Director, will result in inability to maintain the Primary Containment pressure barrier. Additionally, potential loss of Primary Containment should be considered when the Site Emergency Director can determine that a substantial threat exists that may result Primary Containment failure. Loss or potential loss of the Primary Containment barrier in conjunction with these levels of radiation represents loss of two of the three fission product barriers; therefore, the Site Area Emergency classification is warranted.

Escalation to General Emergency is based upon loss or potential loss of the Reactor Coolant System pressure barrier as indicated by higher levels of Drywell radiation.

2.0 PRIMARY PAGE 107 OF 207 CONTAINMENT REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION mH 2.0 PRIMARY PROCEDURE TECHNICAL BASIS CONTAINMENT I] I'&i'A HUE WI 5] VI H[6Jins SITE AREA EMERGENCY (CONTINUED)

REFERENCE - Reg Guide 1.101 Rev. 3, (NIJMARC-FS)

- Calculation ND-N0090-930055 R3 (Unit 2/3)

CURVES/TABLES UINrMti-UINrj LUbb ur 'isf EXCEEDING SI-4.7.A.2.a LIM]

INABILITY TO ISOLATE ANY LINE EXITING CONTAINMENT WHEN ISOLATION IS REQUIRED VENTING IRRESPECTIVE OF OFFSITE RELEASE RATES PER EOIs/SAMGs

...f.f.

0.. ............ .

............... ..... .2 . . ... ....-

UNIT 2 UNIT 3 RAD MONITOR R/HR RAD MONITOR RIHR 2-RE-90-272A 345 3 -RE-90-272A 106 2-RE-90-273A 164 3-RE-90-273A 164 2.0 PRMLARY PAGE 108 OF 207 CONTAINMENT REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III 2.0 PRIMARY PROCEDURE TECHNICAL BASIS CONTAIrNMENT CONTAINM1NT GENERAL EMERGENCY Drywell radiation levels at or above 19500 R/HR with the RCS barrier not intact.

OPERATING - Mode l CONDITION - Mode 2

- Mode 3 BASIS A Drywell radiation monitor reading of 19500 R/HR corresponds to approximately 20% fuel clad damage with 100% coolant release into containment.

A major release of radioactivity requiring offsite protection actions from core damage is not possible unless a major failure of fuel cladding allows radioactive material to be released from the core into the Reactor coolant. Regardless of whether containment is challenged, this amount of activity in containment, if released, could have such severe consequences that it is prudent to treat this as a potential loss of Primary Containment, such that General Emergency declaration is warranted. NUREG-1228, "Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents," indicates that such conditions do not exist when the amount of clad damage is less than 20%.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-FG)

- Calculation ND-N0090-930050 R2 (Unit 2/3)

- NUREG- 1228 "Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents" 2.0 PRIMARY PAGE 109 OF 207 CONTAINMENT I REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III 2.0 PRIMARY PROCEDURE TECHNICAL BASIS CONTAINMENT

• I'pii DIEM 7i I] FU U[SJMKP GENERAL EMERGENCY Drywell radiation levels at or above 4880 R/HR with the RCS not intact.

AND Either of the following exists:

• Indications of loss of Primary Containment. Refer to Table 2.3/2.5-U.

- Primary Containment integrity CANNOT be maintained.

OPERATING - Mode 1 CONDITION - Mode 2

- Mode 3 BASIS A Drywell radiation monitor reading of 4880 RFIR is indicative of Reactor Coolant System pressure boundary failure in conjunction with significant fuel failures. The value has been calculated assuming instantaneous release and dispersal of Reactor coolant noble gas and Iodine inventory associated with a concentration of 300 pLCi/gm dose equivalent Iodine-131 into the Drywell atmosphere. This value is higher than that specified for fuel clad barrier degradation; therefore, this level indicates a loss of both the fuel clad barrier and RCS barrier. Table 2.3/2.5U contains indications of loss of primary containment.

Primary Containment cannot be maintained encompasses any condition or set of conditions based on current trends or anticipated circumstances that, in the judgment of the Site Emergency Director, will result in inability to maintain the Primary Containment pressure barrier. Additionally, potential loss of Primary Containment should be considered when the Site Emergency Director can determine that a substantial threat exists that may result Primary Containment failure. Loss or potential loss of the Primary Containment barrier, in conjunction with these levels of radiation, represents loss of all three fission product barriers; therefore the General Emergency classification is warranted.

2.0 PRIMARY PAGE 110 OF 207 CONTAINMENT REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION mII 2.0 PRIMARY PROCEDURE TECHNICAL BASIS CONTAINMENT CONTAINMENT I] I'&i'A HUE WiI I] EU U[@Jinfl GENERAL EMERGENCY (CONTINUED)

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-FG)

- Calculation ND-N0090-930050 R2 (Unit 2/3)

CURVES/TABLES

...-......I...................... ............................. -% - ..... - . ...... -1;R.:.: .I...............wv~x UNEXPLAINED LOSS OF PRESSURE EXCEEDING SI-4.7.A.2.a LIMITS INABILITY TO ISOLATE ANY LINE EXITING CONTAINMENT WHEN ISOLATION IS REQUIRED VENTING IRRESPECTIVE OF OFFSITE RELEASE RATES PER EOIs/SAMGs 2.0 PRIMARY PAGE 111 OF 207 CONTAINMENT REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION m 2.0 PRIMARY PROCEDURE TECHNICAL BASIS CONTAINMENT CONTAiNMENT UNUSUAL EVENT Drywell unidentified leakage exceeds 10 gpm OR Drywell identified leakage exceeds 40 gpm.

OPERATING - Mode 1 CONDITION - Mode 2

- Mode 3 BASIS This event classification is included as an Unusual Event because it may be a precursor to more serious conditions (e.g., pipe break or equipment degradation) and is considered to be a potential degradation in the level of safety of the plant.

The 10 gpm value for unidentified leakage is two times the Technical Specification limit, indicating an increase beyond the licensed operating value. 10 gpm is also observable using control room.instrumentation and Surveillance Instructions.

The 40 gpm value for identified leakage is conservatively below two times the licensed operating value of 30 GPM but within the capacity of the sump pumps if only one pump were operating. 40 GPM is also observable using control room instrumentation and compatible with Surveillance Instructions. Identified leakage above 40 gpm indicates significant equipment degradation and could represent a degradation in the level of safety of the plant.

Escalation to Alert is based on 50 gpm unidentified leakage into the Drywell or Drywell pressure at or above 2.45 PSIG.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-SU5)

- Technical Specifications 3.4.4 2.0 PRIMARY PAGE 112 OF 207 CONTAINMENT REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION HI 2.0 PRIMARY PROCEDURE TECHNICAL BASIS CONTAINMENT CONTAINMENT ALERT Drywell unidentified leakage exceeds 50 gpm.

OPERATING - Mode 1 CONDITION - Mode 2

- Mode 3 BASIS The potential loss of Reactor Coolant System pressure boundary is based on RCS leakage and determined at a level indicative of a small breech of the Reactor Coolant System but well within the makeup capability of the normal or emergency high pressure makeup systems, (i.e., CRD, HPCI, RCIC). Core uncovery is not a significant concern for a leak of this magnitude; however, break propagation leading to significantly larger loss of inventory is possible. 50 gpm is within the capacity of the Drywell Floor Drain Sump Pumps and measurable using control room instrumentation. Leakage of this magnitude, if not detected early, will result in isolation of the Primary Containment Isolation Valves on the sump pump discharge and Reactor scram at 2.45 PSIG Drywell pressure.

Escalation to Site Area Emergency is by Suppression Chamber pressure exceeding Pressure Suppression Pressure limits or inability to maintain Reactor water level above TAF.

REFERENCE - Reg Guide 1.101 Rev. 3, (NUMARC-FA)

- Calculation ND-N0999-930077 R2 2.0 PRIMARY PAGE 113 OF 207 CONTAINMENT REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION HI 2.0 PRIMARY PROCEDURE TECHNICAL BASIS CONTAINMENT

- k@I'f6JE U4 ELi LI I'EU?6]I FAt IILi INU wmu UNUSUAL EVENT Inability to maintain Primary Containment pressure boundary.

Refer to Table 2.3/2.5-U.

OPERATING - Mode l CONDITION - Mode 2

- Mode 3 BASIS This event classification is intended to address loss or potential loss of the Primary Containment barrier. Use of the Technical Specification definition for Primary Containment, by itself, is not appropriate for determining if the barrier is lost or threatened because in many cases the Technical Specifications could be exceeded with the barrier still intact. Indications for entry into this event classification are those indications that are consistent with actual breech of the pressure boundary.

Some examples of pressure boundary breech are:

• Unexplained loss of pressure in the Drywell or Suppression Chamber.

• Leakage in excess of Technical Requirement Manual 3.6.5.

• Inability to isolate any line exiting containment when isolation is required.

• Intentional venting during EOI execution irrespective of offsite release rates.

Escalation to Site Area Emergency is based upon loss of the fuel clad barrier or RCS barrier in conjunction with loss of the primary containment barrier. There is no Alert classification based on loss of only the primary containment barrier.

REFERENCE - Reg Guide 1.101 Rev. 3, (NUMARC-FU)

- Technical Requirement Manual 3.6.5 2.0 PRIMARY PAGE 114 OF 207 SZ~

CONTAINMENT REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION HI 2.0 PRIMARY PROCEDURE TECHNICAL BASIS CONTAIMElFNT CANTATNMFNT KSJKS] U UI II'A LI 'I'K&SkI El IIL1 I MLU g*

UNUSUAL EVENT (CONTINUED)

CURVES/TABLES UiNt:ArLA1NtLJ LUNS UO PRESSURE EXCEEDING SI4.7.A.2.a LIMITS INABILITY TO ISOLATE ANY LINE EXITING CONTAINMENT WHEN ISOLATION IS REQUIRED VENTING IRRESPECTIVE OF OFFSITE RELEASE RATES PER EOIsISAMGs 2.0 PRIMARY PAGE 115 OF 207 CONTAINMENT REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III 3.0 SECONDARY PROCEDURE TECHNICAL BASIS CONTAINMENT

-W-1 SECONDARY CONTAINMENT 30 3.0 SECONDARY PAGE 116 OF 207 CONTAINMENT REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION m 3.0 SECONDARY PROCEDURE TECHNICAL BASIS CONTAINMENT CONTAINMENT EI k&OXI IYA I'KKSkM WI IM I NM U UMLI I ilK WiU III H i3BU SITE AREA EMERGENCY An unisolable primary system leak is discharging into Secondary Containment AND Any area temperature exceeds the Maximum Safe Operating temperature limit listed in Table 3.1.

OPERATING - Mode 1 CONDITION - Mode 2

- Mode 3 BASIS The Maximum Safe Operating Temperatures of table 3.1 are based on the Browns Ferry Environmental Qualification (EQ) program for safety related equipment. EQ program assumptions include single failure criteria for pipe break that isolates as required. Temperatures in excess of those in Table 3.1 are indicative of pipe breaks that fail to isolate as required. Secondary Containment temperatures of this magnitude resulting from primary system leakage are indicative of significant loss of both the RCS pressure boundary and the Primary Containment pressure boundary. The Site Area Emergency classification is appropriate based upon loss of any two barriers.

Escalation to General Emergency is based on loss or potential loss of the fuel cladding barrier or Radioactivity Release event classifications.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-FS)

- EOI Program Manual, Section V-E t----

3.0 SECONDARY PAGE 117 OF 207 CONTAINMENT REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION I 3.0 SECONDARY PROCEDURE TECIINICAL BASIS CONTAINMENT CONTAINMENT EI KUSkI 571 I'EI&OJI WI IILl I NM U UNil Iii NWAI 1K I U E'U SITE AREA EMERGENCY (CONTINUED)

CURVES/TABLES APPLICABLE PANEL 9-21 AREA TEMPERATURE ELEMENTS MAX SAFE OPERATING VALUE OF (UNLESS OTHERWISE NOTED) UNIT 2 UNIT 3 RHR A/C PUMP ROOM 74-95A 150 155 RHR B/D PUMP ROOM 74-95B 210 215 HPCI TURBINE AREA 73-55A 270 270 RCIC TURBINE AREA 71-41A 190 190 CS ROOM HIGH HUMIDITY OF TEMP HIGH (XA-55-3E-29) PANEL 9-3 TI-75-69B 140 150 RCIC STEAM SUPPLY AREA 71-41B, 41C, 41D 200 250 HPCI STEAM SUPPLY AREA 73-551B, 55C, 55D 240 240 RHR A/C PUMP SUPPLY AREA 74-95H 240 240 RHR B/D PUMP SUPPLY AREA 74-95G 240 240 MAIN STEAM LEAK DETECTION HIGH (XA-55-3D-24) PANEL 9-3 TIS-I -60A 315 315 RHR VALVE ROOM 74-95E 170 175 RWCU ISOL LOGIC CHANNEL A/B TEMP HIGH (XA-55-5B-32/33) PANEL 9-5 170 175 69-835A, B, C, D AUX INST ROOM RWCU OUTBD ISOL VLV AREA 69-29F 220 220 RWCU HX AREA 69-29G 220 220 RWCU HX EXH DUCT 69-29H 220 220 RWCU RECIRC PUMP A AREA 69-29D 215 205 RWCU RECIRC PUMP B AREA 69-29E 215 205 RHR A/C HX ROOM 74-95C 195 200 RHR B/D HX ROOM 74-95D 195 200 FPC HX AREA 74-95F 150 155 3.0 SECONDARY PAGE 118 OF 207 CONTAINMENT REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III 3.0 SECONDARY PROCEDURE TECHNICAL BASIS CONTAINMENT CONTA1NM1NT EI &S]I 51.1 I'US]I Fi WM I NU U HNL'I UN 1iI 1K I iEKfl GENERAL EMERGENCY An unisolable primary system leak is discharging into Secondary Containment AND Any area temperature exceeds the Maximum Safe Operating temperature limit listed in Table 3.1 AND Any indication of potential or significant fuel failure exists.

Refer to Table 3.1-G/3.2-G.

OPERATING - Mode 1 CONDITION - Mode 2

- Mode 3 BASIS The Maximum Safe Operating Temperatures of table 3.1 are based on the Browns Ferry Environmental Qualification (EQ) program for safety related equipment. EQ program assumptions include single failure criteria for pipe break that isolates as required. Temperatures in excess of those in Table 3.1 are indicative of pipe breaks that fail to isolate as required. Secondary Containment temperatures of this magnitude resulting from primary system leakage are indicative of significant loss of both the RCS pressure boundary and the Primary Containment pressure boundary. Table 3.1-G/3.2-G provides guidance for determining if significant fuel failure should be assumed.

This event classification represents loss of all three barriers designed to contain fission products during accidents; therefore, the General Emergency classification is appropriate.

3.0 SECONDARY PAGE 119 OF 207 CONTAINMENT REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION m 3.0 SECONDARY PROCEDURE TECHNICAL BASIS CONTAINMENT E1 kkSJI 17.1 I'K&O]U W. ILILA I UI U UNL'J liii WAI Iii I U BIN GENERAL EMERGENCY (CONTINUED)

REFERENCES Reg Guide 1.101 Rev. 3, (NUMARC-FG)

EOI Program Manual, Section V-E Calculation ND-N0090-930055 R3 (Unit 2/3)

Calculation ND-N0090-930050 R2 (Unit 2/3)

CURVES/TABLES APPLICABLE PANEL 9-21 AREA TEMPERATURE ELEMENTS MAX SAFE OPERATING VALUE OF S1NTLESS OTTHERWTISE NOTERD UINIT 2 TINTT A RHR A/C PUMP ROOM 7495A 150 155 RHR BID PUMP ROOM 74-95B 210 215 HPCI TURBINE AREA 73-55A 270 270 RCIC TURBINE AREA 71-41A 190 190 CS ROOM HIGH HUMIDITY OF TEMP HIGH (XA-55-3E-29) PANEL 9-3 TI-75-69B 140 150 RCIC STEAM SUPPLY AREA 71-41B, 41C, 41D 200 250 HPCI STEAM SUPPLY AREA 73-55B, 55C, 55D 240 240 RHR A/C PUMP SUPPLY AREA 74-95H 240 240 RHR BID PUMP SUPPLY AREA 74-95G 240 240 MAIN STEAM LEAK DETECTION HIGH (XA-55-3D-24) PANEL 9-3 TIS-1-60A 315 315 RHR VALVE ROOM 74-95E 170 175 RWCU ISOL LOGIC CHANNEL A/B TEMP HIGH (XA-55-5B-32/33) PANEL 9-5 170 175 69-835A, B, C, D AUX INST ROOM RWCU OUTBD ISOL VLV AREA 69-29F 220 220 RWCU HX AREA 69-29G 220 220 RWCU HX EXH DUCT 69-29H 220 220 RWCU RECIRC PUMP A AREA 69-29D 215 205 RWCU RECIRC PUMP B AREA 69-29E 215 205 RHR A/C HX ROOM 74-95C 195 200 RHR B/D HX ROOM 74-95D 195 200 FPC HX AREA 74-95F 150 155 figl~~~ll TABR 34Gi1-Gj 2b~i=^~ ~

DRYWELL RAJ)IATION UNIT 2 DRWL _ AITONUI 2-RE-90-272A > 345 R/HR 3-RE-90-272A l> 106 R/HR 2-RE-90-273A > 164 R/HR 3-RE-90-273A l >164 RIHR REACTOR COOLANT ACTIVITY > 300 jiCllg DOSE RECO COLN _ ACIV* >30 I_*_DOSE EQUILAVENT IODINE-131 EQUILAVENT IODlNE-I131 3.0 SECONDARY PAGE 120 OF 207 CONTAINMENT REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION mI 3.0 SECONDARY PROCEDURE TECHNICAL BASIS CONTAINMENT COTIMN J kkS]I 57.1 I'EKS]LU WI kiLl I NM U WI 5] FI U[S]iniWWAU ALERT Any of the following high radiation alarms on Panel 9-3:

• RA-90-1A, Fuel Pool Floor Area

• RA-90-250A, Reactor, Turbine, Refuel Exhaust

• RA-90-142A, Reactor Zone Exhaust

• RA-90-140A, Refueling Zone Exhaust AND Confirmation by Refuel Floor personnel that irradiated fuel damage may have occurred.

OPERATING

• All CONDITION BASIS This event is indicative of irradiated fuel damage caused by a dropped fuel bundle or other heavy solid objects into the Reactor Cavity or Spent Fuel Storage Pools.

The second part of this event classification associates the listed alarms with events that could result in actual irradiated fuel damage versus increased background from other possible sources. Compared to core damage that can occur from full power operating conditions, there is little potential for substantial fuel damage; however, Significant exposures to onsite personnel are possible and protective actions for site personnel may be necessary. For these reasons the Alert classification is warranted.

Escalation is by Radiological Release event classifications.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUTMARC-AA2-example-1) 3.0 SECONDARY PAGE 121 OF 207 CONTAINMENT REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION m 3.0 SECONDARY PROCEDURE TECHBICAL BASIS CONTAINMENT CONTAiNMENT I KkSJI Iii IK&SJM W. kiliI MU U WAI] FAI H[*Xi5U SITE AREA EMERGENCY An unisolable primary system leak is discharging into Secondary Containment AND Any area radiation level at or above the Maximum Safe Operating Area Radiation limit listed in Table 3.2.

OPERATING - Mode l CONDITION - Mode 2

- Mode 3 BASIS Secondary Containment radiation levels of this magnitude are indicative of significant inability to contain or control radioactive materials within the Primary System and Primary Containment. If the Primary System is the source then Site Area Emergency is warranted based upon loss of any two fission product barriers.

Escalation to General Emergency is based on loss or potential loss of the fuel cladding barrier or Radioactive Release event classifications.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-FS)

- EOI Program Manual, Section V-E 3.0 SECONDARY CONTAINMENT PAGE 122 OF 207 S

REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III 3.0 SECONDARY PROCEDURE TECHNICAL BASIS CONTAINMENT CONAINEN 1kffS]I 17.1 I'K&S]LU WA IhILI I NU U W1 I] VI U[SJ&'U SITE AREA EMERGENCY (CONTIINUED))

CURVES/TABLES RHR WEST ROOM RHR EAST ROOM 1000 HPCI ROOM 7U- 1000 CSIRCIC ROOM I 1000 I

CORE SPRAY ROOM 1000 4

SUfYK POOL AREA 1000 CRKI-ICU WEST AREA 1000 CRJJ-HCU EAST AREA -21A 1000 TIP DRIVE AREA -23A 1000 NORTH RWCU SYSTEM AREA Hi.-

-13A .000 SOUTH RWCU SYSTEM AREA ....

-14A .000 RWCU SYSTEM AREA A.

AA.QA I ,

iX -D:D -7. I

.000 MU br, I BA~ O04AA FUEL POOL AREA 90-lA 1000 SERVICE FLR AREA 90-2A 1000 NEW FUEL STORAGE 90-IA 1000 3.0 SECONDARY PAGE 123 OF 207 CONTAINMENT REVISION 28

EMERGENCY EPJP-1 CLASSIFICATION SECTION m1l 3.0 SECONDARY PROCEDURE TECHNICAL BASIS CONTAINMENT CONTAINMENT I kUSJI SY I'EkSXI WinI kiLl I NI U 1iII] El H[S]EI6U GENERAL EMERGENCY An unisolable primary system leak is discharging into Secondary Containment AND Any area radiation level at or above the Maximum Safe Operating Area Radiation limit listed in Table 3.2 AND Any indication of potential or significant fuel failure exists.

Refer to Table 3.1-G/3.2-G.

OPERATING - Mode 1 CONDITION - Mode 2

- Mode 3 BASIS Secondary Containment radiation levels of this magnitude are indicative of significant inability to contain or control radioactive materials within the primary system and Primary Containment. If the primary system is the source then these indications represent loss of RCS pressure boundary and Primary Containment pressure boundary. Table 3.1-G/3.2-G provides guidance for determining if significant fuel failure should be assumed.

This event classification represents loss or potential loss of all three barriers designed to contain fission products during accidents; therefore, the General Emergency classification is appropriate.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-FG)

EOI Program Manual, Section V-E Calculation ND-N0090-930055 R3 (Unit 2/3)

Calculation ND-N0090-930050 R2 (Unit 2/3) 3.0 SECONDARY PAGE 124 OF 207 CONTAINMENT REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION mI 3.0 SECONDARY PROCEDURE TECINICAL BASIS CONTAINMENT CONTAINMENT GENERAL EMERGENCY (CONTINUED)

CURVES/TABLES RHR WEST ROOM V2 KHK EAS'1 KURM I UUU HPCI ROOM I--

90. 1000 CS/RCIC ROOM 90. 1000 CORE SPRAY ROOM 1000 SUPPR POOL AREA 1000 kAW-tHLU WESTI AREA 1000 UCR)-IICU EASR AREA __ __

__ _ ___,1 0 90- i nnn TIP DRIVE AREA 90-23A I000 NORTH RWCU SYSTEM AREA 90-13A 1000 SOUTH RWCU SYSTEM AREA 90- 14A 1000 RWCU SYSTEM AREA 90-9A 1000 MG SET AREA 90-4A 1000 E90-1A 1000 SE C AA 90-2A 1000

_NEW FUEL STORAGE 90-IA 1000 0- i j ti;;;0:;ti iK~iIKI1000O DRYWELL RADIATION- UNIT 2 DRYWLL RADIATION UNIT 3 2-RE-90-272A > 345 R/HR 3-RE-90-272A > 106 R/HR 2-RE-90-273A > 164 RiHR 3-RE-90-273A > 164 R/HR REACTOR COOLANT ACTIVITY '> 300 MC/r DOSE REACTOR COOLANT ACTIVITY > 300 pCI~g DOSE EQUILAVENT IODINE- 131 EQUILAVENT ICDINE- 131 3.0 SECONDARY PAGE 125 OF 207 CONTAINMENT REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION m 4.0 RADIOACTIVITY PROCEDURE TECHNICAL BASIS RELEASE I

RADIOACTIVITY RELEASE 4.0 4.0 RADIOACTIVITY PAGE 126 OF 207 RELEASE REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION 11I 4.0 RADIOACTIVITY PROCEDURE TECHNICAL BASIS RELEASE RELEASE A S _______

UNUSUAL EVENT Gaseous release exceeds ANY limit and duration in Table 4.1-U.

OPERATING - All CONDITION BASIS Unplanned radioactivity releases that exceed Table 4.1-U limits and continue for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or longer represent an uncontrolled situation and potential degradation in the level of safety of the plant. The Offsite Dose Calculation Manual (ODCM) contains the site specific release limits and appropriate surveillance requirements which normally monitor these limits. Table 4.1-U is based on 2 times the ODCM limit. The release should not be averaged over 60 minutes. For example, a release of 4 times ODCM limits for 30 minutes does not meet the requirements for this event classification. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> time period allows sufficient time to isolate any release after exceeding ODCM limits. Release continuing for more than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> represents inability to isolate or control the release. The Site Emergency Director should declare the event as soon as it is determined that the release duration has or will likely exceed 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The value of 0.10 mrem/hr at the site boundary is based on a proration of twice the 500 mrem/yr ODCM instantaneous release rate limit.

Utilize Radiological Control for obtaining site boundary assessments.

Escalation to Alert is based on radiation release rate which exceeds 200 times the ODCM limit.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-AU1 example-1)

RIMS L63 950502 800 IOCFR20 4.0 RADIOACTIVITY RELEASE PAGE 127 OF 207 t

REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION m 4.0 RADIOACTIVITY PROCEDURE TECHNICAL BASIS RELEASE RELEASE ff3I KS) SRI DIIERU NhU E BI UNUSUAL EVENT (CONTINUED)

CURVES/TABLES Il I rr! [ETHOD LIMIT GASEOUS RELEASE RATE WRGERMS) 2.88 X 10 'C/se GASEOUS RELEASE RATE RELEASE FRACTIO1 SITE BOUNDARY RADIATION READING TEAM 0.10 MREM/HR y - 3 NOTES NOTE 4. 1-U Prior to making this emergency classification based upon the WRGERMS indication, assess the release by either of the following:

1. Actual field measurements exceed the limits in Table 4. 1-A

2. SI 4.8.B.1.a.1 Release Fraction 2.0 If neither assessment can be conducted within 60 minutes then the declaration must be made on the valid WRGERMS reading.

4.0 RADIOACTIVITY PAGE 128 OF 207 RELEASE REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III 4.0 RADIOACTIVITY PROCEDURE TECHNICAL BASIS RELEASE N~ IA WXI kOliRi NUWEE NhI ALERT Gaseous release exceeds ANY limit and duration in Table 4.1-A.

OPERATIN( - All CONDITIOD BASIS This event escalates from Unusual Event by increasing the magnitude of the release by a factor of 100. The release limit is equivalent to 200 times the Offsite Dose Calculation Manual (ODCM) limit. The value of 10 mrem/hr at the site boundary is based on a proration of the 500 mrem/yr criteria for both time (8766 hr/yr) and the 200 multiplier. The required release duration is reduced to 15 minutes in recognition of the increased severity. Table 4.1-A contains the Alert limits and appropriate monitoring points for the releases.

Utilize Radiological Control for obtaining site boundary assessments.

Escalation to Site Area Emergency is based on radiation release which will yield a dose to a member of the public which exceeds I OCFR20 limits.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-AA1 example-l)

- RIMS L63 950502 800

- 1IOCFR20

- EPA 400 4.0 RADIOACTIVITY RELEASE PAGE 129 OF 207 S

REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION m 4.0 RADIOACTIVITY PROCEDURE TECENICAL BASIS RELEASE M

ffiI k61IIU DIEM Eli EM E WA ALERT-(CONTINUED)

CURVES/TABLES TYPE GASEOUS RELEASE RATE GASEOUS RELEASE RATE SITE BOUNDARY RADIATION READING NOTES NOTE 4. 1-A Prior to making this emergency classification based upon the WRGERMS indication, assess the release by either of the following:

1. Actual field measurements exceed the limits in Table 4.1-A

2. SI 4.8.B. 1.a.1 Release Fraction 200 If neither assessment can be conducted within 15 minutes then the declaration must be made on the valid WRGERMS reading.

4.0 RADIOACTIVITY PAGE 130 OF 207 RELEASE REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III 4.0 RADIOACTIVITY PROCEDURE TECHNICAL BASIS RELEASE ffi11 KSIURU NUN RU NI 3 E' SITE AREA EMERGENCY EITHER of the following conditions exists:

• Gaseous release exceeds or is expected to exceed ANY limit and duration in Table 4.1-S.

• Dose assessment indicates actual or projected dose consequences above 100 mrem TEDE or 500 mrem thyroid CDE OPERATING - All CONDITION BASIS The limits in this event classification are based on 10 percent of the EPA Protective Action Guidelines or the IOCFR20 dose limit for a member of the public. These limits also provide a desirable gradient between Alert, Site Area Emergency, and General Emergency.

Table 4.1 -S limits for stack and field surveys measurements are consistent with 10 percent of the EPA Protective Action Guidelines or the 10CFR20 dose limit for a member of the public. Stack Noble Gas Release Rates of 1.3 X 1010 pCi/sec for 60 minutes, site boundary radiation readings of 100 mrem./hr for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and Iodine-131 concentration of 3.9 X 10-7 gCi/cm3 for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> are indicative of dose consequences consistent with the limits described previously. The durations in Table 4.1-S are consistent with NUMARC recommendations and industry standards. If analyses indicated a longer or shorter duration for this period in which the substantial portion of the activity is released these dose rates should be adjusted.

Utilize Radiological Control for obtaining site boundary. Dose projection assessments should be requested through the CECC by the implementation of CECC EPIP-8, if the CECC is not staffed utilize site Radiological Control for dose projection assessments through the implementation of BFN EPIP-14.

4.0 RADIOACTIVITY PAGE 131 OF 207 RELEASE REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION I 4.0 RADIOACTIVITY PROCEDURE TECHNICAL BASIS RELEASE A

SITE AREA EMERGENCY (CONTINUED)

The 500 mrem thyroid CDE limit was established in consideration of the 1 to 5 ratio of the EPA Protective Action Guidelines for TEDE and thyroid CDE.

Escalation to General Emergency is based on actual or projected dose exceeding 1000 mrem TEDE or 5000 mrem thyroid CDE.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-ASI example-1)

- RIMS L63 950502 800

- 10CFR20

- EPA 400 CURVES/TABLES I I rr .LimLit GASEOUS RELEASE RATE I.3X10' 0iiCi/SeC 1.

SITE BOUNDARY RADIATION READING I100 MREM/HRV - 1 SITE BOUNDARY IODINE-131 3.9 X10 -9 CI /c 3 I NOTES NOTE 4. 1-S Prior to making this emergency classification based upon the Gaseous Release Rate indication, assess the release by either of the following methods:

1. Actual field measurements exceed the limits in Table 4.1-S.

2. Projected or Actual Dose Assessments exceed 100 mrem

-TEDE or 500 mrem CDE.

If neither assessment can be conducted within 15 minutes then the declaration must be made based on the valid WRGERMS reading.

4.0 RADIOACTIVITY PAGE 132 OF 207 RELEASE REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTIONIH 4.0 RADIOACTIVITY PROCEDURE TECHNICAL BASIS RELEASE A 4 GENERAL EMERGENCY EITHER of the following conditions exists:

• Gaseous release exceeds or is expected to exceed ANY limit and duration in Table 4.1-G.

• Dose assessment indicates actual or projected dose consequences above 1000 mrem TEDE or 5000 mrem thyroid CDE OPERATING - All CONDITION BASIS The limits in this event classification are based on the EPA Protective Action Guidelines which require public protective actions if dose consequences of 1000 mrem TEDE or 5000 mrem thyroid CDE are indicated. These limits also provide a desirable gradient between Alert, Site Area Emergency, and General Emergency and represent the upper level of the gradient.

Table 4.1 -G limits for stack and field surveys measurements are consistent with the EPA Protective Action Guidelines for dose limits requiring public protective actions. Stack Noble Gas Release Rates of 1.3 X 1011 gCi/sec for 60 minutes, site boundary radiation readings of 1000 mrem/hr for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and Iodine-131 concentration of 3.9 X 10 6Ci/cm 3 for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> are indicative of dose consequences consistent with the limits described previously. The durations in Table 4.1-G are consistent with NUMARC recommendations and industry standards. If analyses indicated a longer or shorter duration for this period in which the substantial portion of the activity is released these dose rates should be adjusted.

Utilize Radiological Control for obtaining site boundary assessments. Dose projection assessments should be requested through the CECC by the implementation of CECC EPIP-8, if the CECC is not staffed utilize site Radiological Control for dose projection assessments through the implementation of BFN EPIP- 14.

The 5000 mrem thyroid CDE limit was established in consideration of the 1 to 5 ratio of the EPA Protective Action Guidelines for TEDE and thyroid CDE. Actual meteorology is used in dose assessment calculations to achieve_ the most accurate dose assessment possible.

4.0 RADIOACTIVITY PAGE 133 OF 207 RELEASE REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION m11 4.0 RADIOACTIVITY PROCEDURE TECHNICAL BASIS RELEASE ffi kO1IIU NEHRU NU 3K GENERAL EMERGENCY (CONTINUED)

REFERENCES - Reg Guide 1.101 Rev. 3, (NUJMARC-as 1 example- 1)

- RIMS L63 950502 800

- IOCFR20 CURVES/TABLES RMS) 1.3 X 10 1" Ci/sec 1000 MREM/HR y-3.9 X 10 -6ACI /C3 NOTES NOTE 4. I-G Prior to making this emergency classification based upon the Gaseous Release Rate indication, assess the release by either of the following methods:

1. Actual field measurements exceed the limits in Table 4.1 -G.

2. Projected or Actual Dose Assessmnets exceed 1000 mrem TEDE or 5000 mrem CDE.

If neither assessment can be conducted within 15 minutes then the declaration must be made based on the valid WRGERMS reading.

4.0 RADIOACTIVITY PAGE 134 OF 207 RELEASE REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III 4.0 RADIOACTIVITY PROCEDURE TECHNICAL BASIS RELEASE RELEASE UNUSUAL EVENT Main Steam Line break outside Primary Containment with isolation.

OPERATING - Mode 1 CONDITION - Mode 2

- Mode 3 BASIS This event classification is intended to address the puff release associated with a Main Steam Line break outside Primary that isolates by PCIS Logic as required or can be isolated from the Main Control Room. Regardlessly of whether the break is in the Turbine Building or the Reactor Building a ground level release should be anpicipated due to the blowout panels between the two buildings. Design basis analysis shows that even if MSIV closure occurs within design limits, dose consequences from a "puff' release should be expected. Thus this event classification is included due to the posibility of offsite exposures from the "puff' release.

This event is detected by instrumentation which inputs to the PCIS Logic circuitry.

Main Steam Line high flow, Reactor low pressure with the mode switch in "Run",

And Turbine Building Main Steam Space high temperature are all symptoms of the event and should be evaluated to determine if an actual break has occurred.

Escalation to Area is based on radiation release rate event classifications.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-FU) 4.0 RADIOACTIVITY PAGE 135 OF 207 RELEASE REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III 4.0 RADIOACTIVITY PROCEDURE TECHNICAL BASIS RELEASE RELEASE SITE AREA EMERGENCY Unisolable Main Steam Line break outside Primary Containment.

OPERATING - Mode 1 CONDITION - Mode 2

- Mode 3 BASIS This event classification applies to Main Steam Line Break that cannot be isolated by PCIS Logic or from the Main Control Room. Regardless of whether the break is in the Turbine Building or the Reactor Building a ground level release is expected due to the blowout panels between the two buildings. This event classification represents a loss of two of the three fission product barriers.

Main Steam Line high flow, Reactor low pressure with the mode switch in "Run",

and Turbine Building Main Steam Space high temperature are all symptoms of the event. This event is anticipatory to 4.1-S and the threshold for leakage outside Primary and Secondary Containment should be considered to be any continuous discharge of steam through the break that, in the opinion of the Site Emergency Director, could result in exceeding the limits outlined in 4.1-S.

Escalation to General Emergency is based on loss of the Fuel Clad barrier or radioactivity release event classifications.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-FS) 4.0 RADIOACTIVITY PAGE 136 OF 207 RELEASE REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III 4.0 RADIOACTIVITY PROCEDURE TECHNICAL BASIS RELEASE

• [S1UIIINIEEUNM UNUSUAL EVENT Liquid release rate exceeding 20 times ECL as determined by chemistry sample AND Release duration exceeds or will exceed 60 minutes.

OPERATING - All CONDITION BASIS Liquid release rates are determined using Surveillance Instructions which utilize liquid samples rather than instrument readings for activity determination. Effluent Concentration Limits (ECL) are those annual concentrations given in 10CFR20 Appendix B, Table 2, Column 2. 10 times ECL is equivalent to the instantaneous ODCM limit. Unplanned radioactivity releases that exceed 20 times. ECL (2 times ODCM limit) and continue for 60 minutes or longer represent an uncontrolled situation and potential degradation in the level of safety of the plant. The release should not be averaged over 60 minutes. For example, a release of 40 times ECL for 30 minutes does not meet the requirements of this event classification. The 60 minute time period allows sufficient time to isolate any release after exceeding ECL. Greater than 60 minutes represents inability to isolate or control the release.

The Site Emergency Director should declare the event as soon as it is determined that the release duration has or will likely exceed 60 minutes. The Chemistry Department determines the magnitude of the release by sample procedure for any release as required by initiating procedures (i.e., SI, ARP, AOI, EOI). The sample results are reported to the Site Emergency Director as a fraction or multiple of ECL.

Escalation to Alert is based on release in excess of 2000 times ECL for greater than 15 minutes.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-AUI example-2)

- RIMS L63 950502 800

- 10CFR20 4.0 RADIOACTIVITY PAGE 137 OF 207 RELEASE REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III 4.0 RADIOACTIVITY PROCEDURE TECHNICAL BASIS RELEASE

• [@1UhINNEURUDII ALERT Liquid release rate exceeding 2000 times ECL as determined by chemistry sample AND Release duration exceeds or will exceed 15 minutes.

OPERATING - All CONDITION BASIS This event escalates from Unusual Event by increasing the magnitude of the release by a factor of 100. The required release duration is reduced to 15 minutes in recognition of the increased severity. The Chemistry Department determines the magnitude of the release by sample procedure for any release as required by initiating procedures (e.g., SI, ARP, AO, EOI). The sample results are reported to the Site Emergency Director as a fraction or multiple of ECL. 10 times ECL is equal to the ODCM limit; therefore, 200 times the ODCM limit is equivalent to 2000 times ECL.

Escalation to Site Area Emergency is based on event classifications indicative of.

failure of the Reactor Coolant System pressure boundary and Primary Containment barrier.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-AA1 example-2)

- RIMS L63 950502 800 4.0 RADIOACTIVITY PAGE 138 OF 207 RELEASE REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION m 5.0 LOSS OF PROCEDURE TECHNICAL BASIS POWER LOS OF POWER 50 5.0 LOSS OF PAGE 139 OF 207 POWER REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION m 5.0 LOSS OF PROCEDURE TECHNICAL BASIS POWER POWER K61ISJ arm inii ml N UNUSUAL EVENT Loss of normal and alternate supply voltage to ALL unit specific 4KV shutdown boards from Table 5.1 for greater than 15 minutes AND At least two Diesel Generators supplying power to unit specific 4KV Shutdown Boards listed in Table 5.1.

OPERATINM - All CONDITIOIS BASIS Prolonged loss normal and alternate AC power (offsite) reduces required redundancy and potentially degrades the level of safety of the plant by rendering the plant more vulnerable to a complete loss of AC power (station blackout).

Note 5.1-U specifies that power sources, being taken credit for in this event classification, must be qualified. Qualified in this case means that the power source is capable of supplying the required shutdown loads under accident conditions. A threshold of 15 minutes is provided to exclude transient or momentary power losses. Table 5.1 list the unit specific shutdown boards applicable to Unit 1, Unit 2, or Unit 3.

Escalation to Alert is based on additional loss of onsite power (Diesel Generators) such that one additional single failure would result in complete loss of voltage to all unit specific shutdown boards from Table 5.1.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-SU1)

TVA Drawings 0-15E500-1, 0-15E500-2, and 3-15E500-3 A.

5.0 LOSS OF PAGE 140 OF 207 POWER REVISION 28

EMERGENCY EPLP-1 CLASSIFICATION SECTION m 5.0 LOSS OF PROCEDURE TECHNICAL BASIS POWER POWE KSJXIiJ WirU 61ii DI U B1 UNUSUAL EVENT (CONTINUED)

CURVES/TABLES APPLICABLE UNIT WN BOARDS UNIT I A, B, C, and D UNIT 2 i A, B, C, and D UNIT 3 .

3A, 3B, 3CG and 31D NOTES NOTE 5. 1-U Loss of normal and alternate supply voltage implies inability to restore voltage from any qualified source to normal or alternate feeder for at least one of the unit specific boards within 15 minutes. At least two boards must be energized from Diesel power to meet this classification. If only one board can be energized and that board has only one source of power then refer to 5.1 -A.

5.0 LOSS OF PAGE 141 OF 207 POWER REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION m 5.0 LOSS OF PROCEDURE TECHNICAL BASIS POWER K@JI6J EirU MSLkII DI I KU -

ALERT Loss of voltage to ANY THREE unit specific 4KV shutdown boards from Table 5.1 for greater than 15 minutes AND Only one source of power available to the remaining board.

OPERATING - Mode l CONDITION - Mode 2

- Mode 3 BASIS This event classification provides an escalation from classification 5.1-U. The condition indicated by this event classification is the degradation of the offsite and onsite power systems such that any additional single failure would result in a

.station blackout . This condition is indicative of loss of voltage to all but one unit specific 4KV shutdown board and only one power supply available to that remaining 4KV shutdown board from either offsite or onsite power and inability to restore any additional source within 15 minutes using any combination of feeders or sources available to provide redundancy. Credit must only be taken for those Power sources that are analyzed as creditable in the plant design analysis (FSAR).

Escalation to Site Area Emergency is based on loss of the remaining single power source such that a station blackout exists for the affected unit.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-SA5 )

5.0 LOSS OF PAGE 142 OF 207 POWER REVISION 28

EMERGENCY EPIP-l CLASSIFICATION SECTION III 5.0 LOSS OF PROCEDURE TECHNICAL BASIS POWER KSI&'EJ] EjYm i611'A DI I EU -

ALERT (CONTINUED)

CURVES/TABLES NOTES NOTE 5.1-Al Only one source of power (Diesel or offsite) is available to any one of the listed Unit specific 4KV shutdown boards. No power is available to the three remaining boards.

5.0 LOSS OF PAGE 143 OF 207 POWER REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION m 5.0 LOSS OF PROCEDURE TECHNICAL BASIS POWER I

ALERT Loss of voltage to ALL unit specific 4KV shutdown boards from Table 5.1 for greater than 15 minutes.

OPERATING - Mode 4 CONDITION - Mode 5

- Defueled BASIS Loss of all AC power compromises all plant safety systems requiring electric power including Residual Heat Removal (RHR), Emergency Core Cooling Systems (ECCS), spent fuel heat removal, and the ultimate heat sink. When in Mode 4, Mode 5, or defueled the event can be classified as an Alert because of the significantly reduced decay heat and moderator temperature. The time required to restore one of the boards is not as critical in relation to preventing core damage or reducing other significant threats to the public due to degraded plant conditions.

Escalation to Site Area Emergency is based on abnormal radiation level, radioactive release, or reactor water level event classifications. Refer also to basis for 1.5-A loss of decay heat removal capability.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-SA )

F-J:H1 5.0 LOSS OF PAGE 144 OF 207 POWER REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION m , 5.0 LOSS OF PROCEDURE TECHNICAL BASIS POWER POWER kSI¶0] wirU 11011101 N ALERT (CONTINUED)

CURVES/TABLES F

AtrIUARULE UINIT 'NBOARDS UNIT I A, B, C, and D UNIT 2 A, B, C, and D I

UNIT 3 3A, 3B, 3C, and 3D I I

NOTES NOTE 5. 1-A2 Loss of voltage to all unit specific 4KV shutdown boards applies to those boards which normally supply emergency AC power to the affected unit only. Determination of the event classification depends on the affected unit operating mode. For units in operation 5.1-S would apply.

5.0 LOSS OF PAGE 145 OF 207 POWER REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION 111 5.0 LOSS OF PROCEDURE TECHNICAL BASIS POWER PflW1lR SITE AREA EMERGENCY Loss of voltage to ALL unit specific 4KV shutdown boards from Table 5.1 for greater than 15 minutes.

OPERATING - Mode 1 CONDITION - Mode 2

- Mode 3 BASIS Loss of all AC power compromises all plant safety systems requiring electric power including Residual Heat Removal (RHR), Emergency Core Cooling Systems (ECCS), containment heat removal, and the ultimate heat sink. Prolonged loss of all AC power will cause core uncovery and loss of containment integrity; therefore, this event may lead to General Emergency. Fifteen minutes allows restoration following momentary losses and excludes transient or momentary losses.

Escalation to General Emergency is based on prolonged loss of all AC power (Station Blackout) for greater than the time specified in the site specific station blackout coping analysis.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-S1 )

5.0 LOSS OF POWER PAGE 146 OF 207 S

REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION mII 5.0 LOSS OF PROCEDURE TECHNICAL BASTIS vOWv.R I wlJM SITE AREA EMERGENCY (CONTINUED)

CURVES/TABLES Prr1A1%-DL, ulA5 AY1IUABLEI 4KV SHUTDOWN BOARDS UNIT I A, B, C, and D UNIT 2 A, B, C, and D UNIT 3 3A, 3B, 3C, and 3D NOTES NOTE 5. 1-S Loss of voltage to all unit specific 4KV shutdown board applies to those boards which normally supply emergency AC power to the affected unit only. Determination of the event classification depends on the affected unit operating mode. For units in shutdown or refuel

5. 1-A2 would apply.

5.0 LOSS OF PAGE 147 OF 207 POWER REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTIONIm 5.0 LOSS OF PROCEDURE TECHNICAL BASIS POWER K6J&ES1 iirm UIII'A N U#

GENERAL EMERGENCY Loss of voltage to ALL unit specific 4KV shutdown boards from Table 5.1 AND EITHER of the following conditions exists:

• Restoration of at least one 4KV shutdown board is NOT likely within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />

• Adequate core cooling CANNOT be assured.

OPERATING - Mode 1 CONDITION - Mode 2

- Mode 3 BASIS Loss of all AC power compromises all plant safety systems requiring electric power including Residual Heat Removal (RHR), Emergency Core Cooling Systems (ECCS), containment heat removal, and the ultimate heat sink. Prolonged loss of all AC power will cause core uncovery, loss of fuel clad, and loss of containment integrity. Adequate core cooling can be assumed to exist when the conditions are met in the EGIs that assure adequate core cooling (Refer to E0I Program Manual, Section I-C). The three-hour time limit is based on the Browns Ferry Station Blackout (SBO) Evaluation (RIMS L44-8904118 814) for station blackout (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) minus 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to ensure timely recognition and emergency response. This event classification is redundant to fission product barrier event classification, but is intended to be an earlier indication of imminent release and allow for more timely emergency response.

The likelihood of restoring at least one 4KV shutdown board should be based on a realistic appraisal of conditions and high probability of restoration. Delaying upgrade from Site Area Emergency based on only a chance of restoration could result in a loss of valuable time in preparation and implementation of public protective actions .

5.0 LOSS OF PAGE 148 OF 207 POWER REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION Im 5.0 LOSS OF PROCEDURE TECHNICAL BASIS POWER POWER KSI¶6] wru JSIII N N k GENERAL EMERGENCY (CONTINUED)

In addition, under these conditions, fission product barrier monitoring capability may be degraded. Indications of continuing core cooling degradation must be assessed and Site Emergency Director judgement must be utilized relative to imminent loss or potential loss of fission product barriers and degraded ability to monitor fission product barriers. Refer to event classification 8.4-G.

REFERENCES Reg Guide 1.101 Rev. 3, (NUMARC-SG1)

RIMS B22 91 1216 102 Station Blackout (SBO) Evaluation (RIMS L44 890418 814)

CURVES/TABLES I AErIIlAILL 4KV SHUTDOWN BOA I UNIT 1 A, B, C, and D UNIT 2 A, B, C, and D UNIT 3 3A, 3B, 3C, and 3D NOTES NOTE 5. 1-G Loss of voltage to all unit specific 4KV shutdown boards applies to those boards which normally supply emergency AC power to the affected unit only.

5.0 LOSS OF PAGE 149 OF 207 POWER REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III 5.0 LOSS OF PROCEDURE TECHNICAL BASIS POWER POWER K@I'f SI AIIkU IIU US1i'i N EI UNUSUAL EVENT Unplanned loss of 250V DC control power to ALL unit specific 4KV shutdown boards from table 5.2-U for greater than 15 minutes.

OR Unplanned loss of 250V DC control power to unit specific 480V shutdown boards A and B for greater than 15 minutes.

OPERATING - Mode 4 CONDITION - Mode 5 BASIS The purpose of this event classification is to recognize a loss of 250V DC power compromising the ability to monitor and control the removal of decay heat during Mode 4 or Mode 5 operations. This event classification is anticipatory to other event classifications in as much as the operator may not have necessary control of equipment to respond to the loss. This event classification is a precursor to loss of decay heat removal capability events.

Loss of control power to the applicable boards prevents operation of vital systems (RHR and Core Spray) which are either required by Technical Specifications or required for decay heat removal. The 15 minute time period is provided to exclude transient or momentary power losses.

Note 5.2 specifies the minimum bus voltage necessary for operation of safety related equipment consideration a reserve capacity for at least 30 minutes of operation.

Escalation to Alert is based on loss of decay heat removal capability and inability to maintain Reactor moderator temperature below 212'F.

REFERENCE - Reg Guide 1.l01 Rev. 3, (NUMARC-SU7)

Technical Specifications 3.8.4 FSAR 8.6.3 5.0 LOSS OF PAGE 150 OF 207

!fli >

POWER REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION mII 5.0 LOSS OF PROCEDURE TECHNICAL BASIS POWER POWER KSJ&¶SJ 4IIkU IIU U?@)A'i DI 81 UNUSUAL EVENT CURVES/TABLES APPLICABLE UNIT SADLI; 4AKV UIDU'ID WIN BUAKIN UNIT 1 A, B, C, and D UNIT 2 I A, B, C, and D UNIT 3

, 3A. 3B. 3C. and 3D NOTES NOTE 5.2 250V DC bus voltage of less than 248 volts on any feeder to any referenced board constitutes a loss of voltage for that feeder; thus, a loss of DC control voltage to the referenced board. The voltage readings are obtained at the 250V Battery Boards feeding the referenced boards.

5.0 LOSS OF PAGE 151 OF 207 POWER REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III 5.0 LOSS OF PROCEDURE TECHNICAL BASIS POWER kSJ¶6] I4IIkU IiU MIhi'A N SITE AREA EMERGENCY Loss of 250V DC power to ALL combinations of essential systems from Table 5.2-S for greater than 15 minutes.

OPERATING - Mode 1 CONDITION - Mode 2

- Mode 3 BASIS Loss of DC power to all essential systems will lead to core uncovery and loss of Primary Containment integrity when there is significant decay heat and sensible heat in the Reactor System. The combinations of systems listed in Table 5.2-S were chosen based on the individual systems ability to supply makeup water to the Reactor Pressure Vessel. At rated temperature and pressure the Main Steam Relief Valves (MSRVs) are utilized lifting at their respective setpoints to remove decay heat from the Reactor and control Reactor pressure. DC power is only required for MSRV operation in the manual blowdown mode. Four MSRVs is the minimum number of MSRVs required for emergency depressurization. DC power supplies listed in Table 5.2-U are only those DC power sources that will render the listed systems unavailable. Other combinations of DC power sources in combination with AC electrical transients could render these systems unavailable.

These other transients are covered under other event classifications (e.g., Reactor water level or loss of decay heat removal events) which are caused by loss of ECCS.

DC power supplies to instrumentation (e.g., ECCS inverters) are covered under loss of control room annunciator and instrumentation events. DC control power to Diesel Generators is covered under loss of onsite AC power because loss of Diesel Generator DC power sources renders emergency Diesel Generators inoperable.

5.0 LOSS OF PAGE 152 OF 207 POWER REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION m 5.0 LOSS OF PROCEDURE TECHNICAL BASIS POWER POWER KIF'IS] 4IIkU IIU US11'A N IW SITE AREA EMERGENCY (CONTINUED)

Heat removal from Primary Containment is covered under Suppression Pool temperature and Primary Containment pressure events. Loss of 250V DC power is a precursor to Reactor pressure vessel low level, fission product barrier, radiological release, and radiological effluent events.

This event classification provides redundancy to those events but is warranted because it is anticipatory and allows more time for required actions. The 15 minute time period is provided to exclude transient or momentary power losses.

Note 5.2 specifies the minimum bus voltage necessary for operation of safety related equipment consideration a reserve capacity for at least 30 minutes of operation.

Escalation to General Emergency is based on radiological release, radiological effluent, fission product barrier, or Site Emergency Director judgement event classification.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-SS3)

- TVA Drawings 45E701-1, 45E702-1, 45E703-1, and associated listed reference drawings

- Technical Specifications 3.8.4

- FSAR 8.6.3 5.0 LOSS OF PAGE 153 OF 207 POWER REVISION 28

EMERGENCY EPIP-1

.CLASSIFICATION SECTION III 5.0 LOSS OF PROCEDURE TECHNICAL BASIS POWER POWE KSI6J 41IkU IIU iiSIA'i N SITE AREA EMERGENCY (CONTINUED)

CURVES/TABLES w-AulItiLL ~u V us-.. ru WVK (IJNffT R"1V1TF TINT .R..AT nr4?1WDUTJQ VnT1r1Y%

I 4KV UNIT BD, A, B, and C CONTROL POWER MAIN CONDENSER AND AND 480V UNIT BOARD A and B CONTROL POWER EHC PUMPS AND AND PANEL 9-9 CABINET 1 REACTOR FEED PUMPS II 250V DC RMOV BD A HPCI III 250V DC RMOV BD C RCIC IV 250V DC RMOV BDs A, B, and C > 4 MSRVs AND AND 4KV SHUTDOWN BDs A, B, C, and D CONTROL POWER 1 RHR PUMP (3A, 3B, 3C, and 3D FOR UNIT 3) OR 1 CORE SPRAY PUMP NOTES NOTE 5.2 250V DC bus voltage of less than 248 volts on any feeder to any referenced board constitutes a loss of voltage for that feeder; thus, a loss of DC control voltage to the referenced board. The voltage readings are obtained at the 250V Battery Boards feeding the referenced boards.

5.0 LOSS OF PAGE 154 OF 207 POWER REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION I -

PROCEDURE TECHNICAL BASIS 6.0 HAZARDS HAZARD S 6.0 6.0 HAZARDS PAGE 155 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION mII PROCEDURE TECHNICAL BASIS 6.0 HAZARDS UNUSUAL EVENT Valid, unexpected increase of ANY in plant Area Radiation Monitor (ARM) reading to 1000 mrem/hr (except TIP Room).

OPERATING - All CONDITION BASIS This event classification addresses unexpected increases in plant radiation levels that represent a degradation in the control of radioactive material, and constitute a potential degradation in the level of safety of the plant.

The term valid means that the instrument reading can be confirmed by other plant instrumentation indications, is consistent with an on-going transient or unplanned event, or that the condition is verified by RADCON. The term unexpected implies an increase not attributable to an anticipated transient such as a radwaste resin transfer, radiography, calibration activity, etc.

The Control Room is not included because it is an area which requires continuous occupancy and is covered under event classification 6. 1-A.2. The TIP Room is not included because high ARM readings may not necessarily indicate a degradation in the control of radioactive materials and no personnel actions are required in the TIP area to safely shutdown the plant or maintain safe shutdown.

Escalation to Alert is based on operations required in the affected area(s) to safely shutdown the plant or maintain safe shutdown.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-AU2 EXAMPLE 4)

- RIMS R38 940916 970

- EOI Program Manual, Section V-E 6.0 HAZARDS PAGE 156 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 6.0 HAZARDS

'AIS

• A A ALERT Valid, unexpected increase of ANY in plant Area Radiation Monitor (ARM) reading to 1000 mrem/hr (except TIP Room)

AND Personnel actions required in the affected area(s).

OPERATING - All CONDITION BASIS This event classification addresses increased radiation levels that may impede necessary access to critical operating areas when conditions necessitate equipment operation or maintenance in those areas in order to maintain safe operation or achieve safe shutdown. The impaired ability to perform inplant operations represents an actual or potential degradation in the level of safety of the plant. It is appropriate to ensure that additional personnel are onsite to perform necessary operations and maintenance and provide proper approvals, surveys, and radiation protection; therefore, the alert classification is justified.

The term valid means that the instrument reading can be confirmed by other plant instrumentation indications or is verified by RADCON. The term unexpected implies an increase not attributable to an anticipated transient such as a radwaste resin transfer, radiography, calibration activity, etc.

The Control Room is not included because it is an area which requires continuous occupancy and is covered under event classification 6. 1-A.2. The TIP Room is not included because no personnel actions are required in the area to safely shutdown the plant or maintain safe shutdown.

Escalation to Site Area Emergency or General Emergency is based on Secondary Containment radiation, radiological release, or fission product barrier event classifications .

'V 6.0 HAZARDS PAGE 157 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION mlI PROCEDURE TECHNICAL BASIS 6.0 HAZARDS WA I] [0] KOhl [VAt

• ai -

ALERT (CONTINUED)

REFERENCES Reg Guide 1.101 Rev. 3, (NUMARC-AA3 EXAMPLE 2)

RIMS R38 940916 970 EOI Program Manual, Section V-E 6.0 HAZARDS PAGE 158 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION mII PROCEDURE TECHNICAL BASIS 6.0 HAZARDS 1iI I] [61 KOhl [Wit W P ALERT Control Room radiation levels >15 mrem/hr.

OPERATING - All CONDITION BASIS This event classification applies to areas requiring continuous occupancy to maintain safe operation or safely shutdown the plant. At Browns Ferry the Control Room is the only area within the plant that should require continuous manning under these conditions. Other areas (e.g., TSC, OSC) can be relocated to an area of lower dose. Radwaste operations can be suspended and the Radwaste Control Room can be evacuated. The Central Alarm Station (CAS) is not included because of the location at the plant entrance away from the main building. It would not be possible to achieve these levels at the CAS without already reaching alert classification through radiological release.

This event classification may be redundant to fission product barrier and radiological release event classifications; however the cause of the event is not a concern. This event classification is only intended to address exposures to personnel who must be present at required operating stations for long term operation.

Escalation to Site Area Emergency or General Emergency is based on Secondary Containment radiation, radiological release, or fission product barrier event classifications.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-AA3 EXAMPLE 1) 6.0 HAARDS PAGE 159 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION m PROCEDURE TECHNICAL BASIS 6.0 HAZARDS UiJM I {@J U (SISIIL1 UOh'AU lAIN ISJWW ALERT Control Room Abandonment from entry into AOI-100-2 or SSI-16 for ANY Unit Control Room.

OPERATING All CONDITION BASIS With the control room evacuated, additional support, monitoring, and direction through the Technical Support Center and/or other Emergency Operations Centers may be necessary. The Alert declaration ensures centers are manned to provide the necessary additional support.

Escalation to Site Area Emergency is based on inability to establish plant control from outside the control room within 20 minutes.

REFERENCES Reg Guide 1.101 Rev. 3, (NUMARC-HA5) 6.0 HAZARDS PAGE 160 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 6.0 HAZARDS

&SXU I (@] U RSIO]hIE OMA!UUAI I O]W SITE AREA EMERGENCY Control Room Abandonment from entry into AOI-100-2 or SSI-16 for ANY Unit Control Room.

AND Control of Reactor water level, Reactor pressure, and Reactor power (for Modes 1,2, & 3) or decay heat removal (for Modes 4 & 5) per AOI-100-2 or SSI-16, as applicable, CANNOT be established within 20 minutes after evacuation is initiated.

OPERATING - All CONDITION BASIS This event classification is intended to recognize loss of control of critical parameters either by failure of equipment designed to automatically initiate for control of the parameter or failure to expeditiously transfer safety system control to the backup controls. Fission product barrier damage may not yet be indicated but should be considered by assessing available parameters versus the status of safety systems and the ability to control critical parameters. In Mode 4 and Mode 5 operator concern should be directed towards, maintaining core cooling using decay heat removal systems. In power operation, hot standby, and hot shutdown operator concern is primarily directed toward maintaining critical parameters, (i.e.,

level, pressure, power, and heat sink) and thereby assuring fission product barrier integrity.

The 20 minute time period is based on time required for personnel to leave the control room, arrive at the appropriate backup control station, and take control of critical parameters before core uncovery or core damage has occurred. This timeframe has been projected within the Tennessee Valley Authority, Browns Ferry Nuclear Plant, Fire Protection Report - Unit 2 and 3. During execution of procedures and transfer of equipment control, the listed critical parameters may be considered as being controlled if the parameters can be verified as being maintained within safe value ranges by appropriate equipment and automatic initiation functions designed to control the parameter (example: HPCI auto initiated and raised RPV level to a value above the initiation setpoint.).

Escalation to General Emergency is by fission product barrier degradation radioactivity release, or Emergency Director Judgement event classification.

6A7 6.0 HAZARDS PAGE 161 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION m PROCEDURE TECHNICAL BASIS 6.0 HAZARDS WIXI I ROJ U (OIOk'J UmI'AruaI U[S]&'

SITE AREA EMERGENCY (CONTINUED)

REFERENCES Reg Guide 1.101 Rev. 3, (NUIMARC-HS2)

Fire Protection Report Units 2 and 3 6.0 HAZARDS PAGE 162 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 6.0 HAZARDS EUVIHIL1EamEIvI UNUSUAL EVENT Turbine failure resulting in casing penetration OR Significant damage to Turbine or Generator seals during operation.

OPERATING - Mode 1 CONDITION - Mode 2 BASIS This event classification is intended to address Main Turbine rotating component failures of sufficient magnitude to cause observable damages to the Turbine casing or to the seals of the Turbine or Generator. Of major concern is the potential for leakage of combustible fluids (lubricating oil) and gases (hydrogen) into the plant and environs. Actual fires and flammable gas buildup are classified under other event classification. This event classification is consistent with unusual event while maintaining the anticipatory nature desired and recognizing the risk to nonsafety related equipment.

Escalation to higher event classification is based on potential damage to safety related equipment from missiles generated by the failure or by radioactivity release.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-HU1 EXAMPLE 6) 6.0 HAZARDS PAGE 163 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION mll PROCEDURE TECHNICAL BASIS 6.0 HAZARDS IIM4IIIbIUaIRhJu ALERT Turbine failure resulting in visible structural damage to or penetration of ANY of the following structures from missiles:

• Reactor Building

• Diesel Generator Building

• Intake Structures

. Control Bay OPERATING - Mode 1 CONDITION - Mode 2 BASIS This event classification is intended to address the threat to safety-related equipment imposed by missiles generated by Main Turbine rotating component failures. Areas included are plant areas containing safety related equipment required to safely operate or safely shutdown the plant. The Alert classification assures adequate personnel are available to perform thorough assessment of damage to structures and equipment in the affected areas.

Escalation to higher event classification is based on system malfunction, fission product barrier degradation, radioactivity release, or Emergency Director Judgement event classifications.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-HA I EXAMPLE 6) 6.0 HAZARDS PAGE 164 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION m PROCEDURE TECHNICAL BASIS 6.0 HAZARDS UNUSUAL EVENT Confirmed fire in ANY plant area listed in Table 6.4-Ul AND NOT extinguished within 15 minutes.

OPERATING - All CONDITION BASIS The purpose of this event classification is to address the magnitude and extent of fires that may be potentially significant precursors to damage to safety systems.

This excludes such items as fires within administration buildings, waste basket fires, and other small fires with no safety significance. This event classification applies to plant vital areas, buildings and areas contiguous to plant vital areas, or areas which have the potential to cause significant release of radioactive material such as Radwaste. These areas are included in Table 6.4-U1.

Confirmation of fire includes those actions listed in the appropriate Alarm Response Procedure (ARP) to verify the alarm is not spurious or by visual observation by personnel in the field. If confirmation cannot be positively ascertained within 15 minutes and symptoms indicative of a fire persists then confirmation should be assumed and this event classification declared.

Allowance of fifteen minutes for extinguishment is provided to exclude small fires, easily extinguishable, with no significant safety consequence.

Escalation to Alert is based on fire affecting the operability of plant safety systems required for the current operating condition.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-HU2)

- Browns Ferry Nuclear Plant Safe Shutdown Programn 6.0~ HAAD AE 6 F27 EIIN2 6.0 ELAZARDS PAGE 165 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 6.0 HAZARDS UIEIkMK@I[eX W3h1 UNUSUAL EVENT (CONTINuED)

CURVES/TABLES IKtAulIUK BUILDING REFUEL FLOOR 4KV SHUTDOWN BOARD ROOMS 4KV SHUTDOWN BOARD BATTERY ROOMS 480V SHUTDOWN BOARD ROOMS 3A and 3B RMOV BOARD ROOMS 4KV BUS TIE BOARD ROOM CONTROL BAY ELEVATION 593', 606' and 617' DIESEL GENERATOR BUILDINGS (ALL ELEVATIONS)

TURBINE BUILDING (ALL ELEVATIONS IN OR ADJACENT TO AREAS CONTAINING SAFE SHUTDOWN EQUIPMENT INTAKE PUMPING STATION (ALL ELEVATIONS)

RADWASTE BUILDING (ALL ELEVATIONS)

CABLE TUNNEL (INTAKE TO TURBINE BUILDING)

STANDBY GAS TREATMENT BUILDING 6.0 HAZARDS PAGE 166 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION mII PROCEDURE TECHNICAL BASIS 6.0 HAZARDS EIIWiIMMK*OX WBU UNUSUAL EVENT Unanticipated explosion within the protected area resulting in visible damage to ANY permanent structure or equipment.

OPERATING All CONDITION BASIS The purpose of this event classification is to recognize only those explosions of sufficient force to damage permanent structures or equipment within the protected area. An explosion is defined as rapid, violent, unconfined combustion, or a catastrophic failure of pressurized equipment that potentially imparts significant energy to nearby structures and materials. This event classification makes no attempt to assess the actual magnitude of the damage. The occurrence of the explosion with reports of evidence of damage (e.g., deformation and scorching) is sufficient for declaration. The Site Emergency Director also needs to consider any security aspects of the explosion if applicable.

Escalation to Alert is based on explosion affecting safety system performance or causing visible damage to structures or equipment required for safe shutdown.

REFERENCES Reg Guide 1.101 Rev. 3, (NUMARC-HUI EXAMPLE 5) 6.0 HAZARDS PAGE 167 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION mII PROCEDURE TECHNICAL BASIS 6.0 HAZARDS UIIE4MKS)ILSJL ALERT Fire or explosion in ANY plant area listed in Table 6.4-A affecting safety system performance OR Fire or explosion causing visible damage to permanent structures or safety systems in ANY area listed in Table 6.4-A.

OPERATING - All CONDITION BASIS This event classification is intended to address the magnitude and extent of fires that potentially or actually affect one or more redundant trains of safety systems or structures containing safety systems. Areas listed in Table 6.4-A are those plant areas that contain systems or functions required for the safe shutdown of the plant.

With regard to explosions, only those explosions of sufficient force to damage permanent structures or equipment required for safe operation within the identified plant area should be considered. An explosion is defined as a rapid, violent, unconfined combustion, or a catastrophic failure of pressurized equipment that potentially imparts significant energy to nearby structures and materials. The occurrence of an explosion with reports of visible damage is sufficient evidence for determination of this event. No attempt should be made to perform a detailed assessment before declaration is considered. Declaration of Alert with subsequent manning of support personnel will provide adequate personnel to make a detailed assessment. The Site Emergency Director should also consider any security aspects of the explosion.

Escalation to Site Area Emergency is based on system malfunction, fission product barrier degradation, radioactivity release, or Emergency Director Judgement .

6.0 HAZARDS PAGE 168 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION m -

PROCEDURE TECHNICAL BASIS 6.0 HAZARDS m

ALERT (CONTINUED)

REFERENCES Reg Guide 1.101 Rev. 3, (NTUMARC-HA2)

Browns Ferry Nuclear Plant Safe Shutdown Program CURVES/TABLES REACTOR BUILDING REFUEL FLOOR 4KV SHUTDOWN BOARD ROOMS 4KV SHUTDOWN BOARD BATTERY ROOMS 480V SHUTDOWN BOARD ROOMS 3A and 3B RMOV BOARD ROOMS 4KV BUS TIE BOARD ROOM CONTROL BAY ELEVATION 593', 606' and 617' DIESEL GENERATOR BUILDINGS (ALL ELEVATIONS)

INTAKE PUMPING STATION (ALL ELEVATIONS)

RADWASTE BUILDING (ALL ELEVATIONS)

CABLE TUNNEL (INTAKE TO TURBINE BUILDING)

STANDBY GAS TREATMENT BUILDING 6.0 HAZARDS PAGE 169 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION mII PROCEDURE TECHNICAL BASIS 6.0 HAZARDS UNUSUAL EVENT EITHER of the following conditions.exists:

• Normal operations impeded due to access restrictions caused by toxic gas concentrations within any building or structure listed in Table 6.5/6.6.

• Confirmed report by Local, County, or State Officials that a large offsite toxic gas release has occurred within one mile of the site with potential to enter the site boundary in concentrations at or above the Permissible Exposure Limit (PEL) causing an evacuation of any site personnel.

OPERATING - All CONDITION BASIS This event classification is based on significant releases of toxic gases that could affect the health and safety of plant personnel or affect the safe operation of the plant with the plant being within the evacuation area of an offsite event (i.e., tanker truck or barge accident releasing toxic or flammable gas, etc.). The evacuation area is determined from the Department of Transportation (DOT) Evacuation Tables for selected hazardous materials, in the DOT Emergency Response Guide for hazardous materials.

Table 6.5/6.6 contains plant vital areas, buildings and areas contiguous to plant vital areas, and other significant plant buildings or structures where operations may be required to assure safe operation of the plant.

Escalation to Alert is based on gases entering plant structures and affecting safe operation of the plant.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-HU3) 6.0 HAZARDS PAGE 170 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION HI-PROCEDURE TECHNICAL BASIS 6.0 HAZARDS

-Ok -- - 5-.--- -

UNUSUAL EVENT (CONTINUED)

CURVES/TABLES ItDJX I VJI¶ DUlIJllNvIJ REFUEL FLOOR CONTROL BAY DIESEL GENERATOR BUILDINGS TURBINE BUILDING INTAKE PUMPING STATION RADWASTE BUILDING CABLE TUNNEL (INTAKE TO TURBINE BUILDING)

STANDBY GAS TREATMlENT BUILDING 6.0 HAZARDS PAGE 171 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 6.0 HAZARDS ISkl[UWXIV ___

ALERT ALL of the following conditions exists:

• Plant personnel report toxic gas within any building or structure listed in Table 6.5/6.6.

• Plant personnel report severe adverse health reactions due to toxic gas (i.e.j burning eyes, throat, or dizziness)

OR Sampling results by Fire Protection or Industrial Safety personnel indicate levels above the Permissible Exposure Limit (PEL).

• Determination by the Site Emergency Director that plant personnel would be unable to perform actions necessary to establish and maintain Mode 4 conditions while utilizing appropriate personnel protective equipment.

OPERATING - All CONDITION BASIS This event classification is based on toxic gases that have entered a plant structure affecting safe operation of the plant. This event classification applies to buildings and areas contiguous to plant vital areas or other significant buildings or areas (i.e.,

intake and Standby Gas Treatment buildings). The intent of this event classification is not to include buildings (i.e., warehouses and administration buildings) or other areas that are not contiguous or immediately adjacent to plant vital areas. It is appropriate that increased monitoring be performed to ascertain whether consequential damage has occurred.

Escalation to higher emergency class is based on system malfunction, fission product barrier degradation, radioactivity release, or Emergency Director Judgement event classification.

6.0 HAZARDS PAGE 172 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION m PROCEDURE TECHNICAL BASIS 6.0 HAZARDS

-gy A- A ALERT (CONTINUED)

REFERENCES Reg Guide 1.101 Rev. 3, (NUMARC-HA3)

CURVES/TABLES CONTROL BAY DIESEL GENERATOR BUILDINGS TURBINE BUILDING INTAKE PUMPING STATION RADWASTE BUILDING CABLE TUNNEL (INTAKE TO TURBINE BUILDING)

STANDBY GAS TREATMENT BUILDING 6.0 HAZARDS PAGE 173 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION HI-PROCEDURE TECHNICAL BASIS 6.0 HAZARDS UWAWJ IL!JMH UUK WA11 IL UNUSUAL EVENT EITHER of the following conditions exists:

• Release of flammable gas within the site boundary in concentrations at or above 25% of the Lower Explosive Limit (LEL) for any three readings obtained in a 10 ft. triangular area as indicated by Fire Protection or Industrial Safety personnel using appropriate monitoring instrumentation.

• Confinred report by Local, County, or State Officials that a large offsite flammable gas release has occurred within one mile of the site with potential to enter the site boundary in concentrations at or above 25% of the Lower Explosive Limit (LEL).

OPERATING - All CONDITION BASIS This event classification is based on significant releases of flammable gases that could affect the safe operation of the plant with the plant being within the evacuation area of an offsite event (i.e., tanker truck or barge accident releasing toxic or flammable gas, etc.). The evacuation area is determined from the Department of Transportation (DOT) Evacuation Tables for selected hazardous materials, in the DOT Emergency Response Guide for hazardous materials.

Escalation to Alert is based on flammable gases entering plant structures and affecting safe operation of the plant.

REFERENCES - Reg Guide 1. 101 Rev. 3, (NUNMARC-HU3) 6.0 HAZARDS PAGE 174 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION m PROCEDURE TECHNICAL BASIS 6.0 HAZARDS UW1Li UUK WA11 ALERT Release of flammable gases within any building or structure listed in Table 6.5/6.6 in concentrations at or above 25% of the Lower Explosive Limit (LEL) for any three readings obtained in a 10 ft. triangular area as indicated by Fire Protection of Industrial Safety personnel using appropriate monitoring instrumentation.

OPERATING - All CONDITION BASIS This event classification is based on flammable gases that have entered a plant structure with potential to affect safe operation of the plant. This event classification applies to buildings and areas contiguous to plant vital areas or other significant buildings or areas (i.e., intake and Standby Gas Treatment buildings).

The intent of this event classification is not to include buildings (i.e., warehouses and administration buildings) or other areas that are not contiguous or immediately adjacent to plant vital areas. It is appropriate that increased monitoring be performed to ascertain whether consequential damage has occurred.

Escalation to higher emergency class is based on system malfunction, fission product barrier degradation, radioactivity release, or Emergency Director Judgement event classification.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-HA3 EXAMPLE 1&2) 6.0 HAZARDS PAGE 175 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION Ill PROCEDURE TECHNICAL BASIS 6.0 HAZARDS

- -- m UWL1 ELI L IIUDKWXI DL ALERT (CONTINUED)

CURVES/TABLES DIESEL GENERATOR BUILDINGS TURBINE BUILDING INTAKE PUMPING STATION RADWASTE BUILDING CABLE TUNNEL (INTAKE TO TURBINE BUILDING)

STANDBY GAS TREATMENT BUILDING 6.0 HAZARDS PAGE 176 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION HI PROCEDURE TECHNICAL BASIS 6.0 HAZARDS UNUSUAL EVENT ANY of the following conditions exists:

• Bomb device discovered within the plant protected area but NOT within a vital area.

• Attempted or imminent attempt by a hostile force to penetrate the plant protected area barrier.

• Civil disturbance ongoing on the owner controlled property outside the protected area that threatens to interrupt plant operations.

• Hostage/Extortion situation that-threatens to interrupt plant operations.

OPERATING - All CONDITION BASIS This event classification represents a potential degradation in the level of safety of the plant due to potential damage to permanent plant structures, intrusion or attempted intrusion of the protected area or disturbance that may affect plant operations; therefore, the Unusual Event classification is appropriate.

For the purposes of this event classification the following definitions apply:

Civil Disturbance - A group of 20 or more persons violently protesting station operations or activities at the site; Extortion - An attempt to cause an action at the station by threat or force; Hostage - Person(s) held as leverage against the station to ensure that demands will be met by the station.

Escalation to Alert is based on bomb device being discovered within a plant vital area or actual intrusion into the plant protected area by a hostile force.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-HU4 EXAMPLE 1 & 2)

- Browns Ferry Physical Security/Contingency Plan 6.0 HAARDS PAGE 177 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION LU PROCEDURE TECIINICAL BASIS 6.0 HAZARDS

________________ - I. A ALERT Bomb device discovered within ANY plant vital area OR Actual intrusion into the plant protected area by a hostile force.

OPERATING - All CONDITION BASIS The first part of this event classification represents a potential substantial degradation in the level of safety of the plant due to the threat to vital safety systems. Actual discovery of the device is required to meet this classification.

Any detonation of an explosive device could result in declaration under Fire/Explosion event classifications.

The second part of this event classification represents a substantial degradation in the level of safety of the plant due to the extreme measures that must be taken by an intruder to enter the protected area. Hostile intent must be suspected to meet this classification. A civil disturbance which penetrates the protected area boundary can be considered a hostile force.

Escalation is by actual intrusion into a vital area by a hostile force.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-HA4 EXAMPLE 1 & 2)

- Browns Ferry Physical Security/Contingency Plan 6.0 HAZARDS PAGE 178 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION I][

PROCEDURE TECHNICAL BASIS 6.0 HAZARDS I

I SITE AREA EMERGENCY Intrusion into ANY plant vital area by a hostile force.

OPERATING - All CONDITION BASIS This event represents a substantial degradation in the level of safety of the plant due to the extreme measures that must be taken by an intruder to enter a vital area.

Hostile intent must be suspected to meet this classification. This event represents a possible threat to the public and an escalated threat to plant safety above that contained in the Alert event classification in that a hostile force has progressed from the protected area to the vital area.

Escalation is by actions occurring involving a hostile force which results in a loss of physical control of equipment or functions required to reach and maintain safe shutdown or remove decay heat from any unit.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-HSI EXAMPLE 1) 6.0 HAZARDS PAGE 179 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION mI1 PROCEDURE TECHNICAL BASIS 6.0 HAZARDS NKUII'4M' GENERAL EMERGENCY Intrusion by a hostile force into Control Rooms, backup control areas, or plant vital areas which results in a loss of physical control of equipment or functions required to reach and maintain safe shutdown or remove decay heat from any unit.

OPERATING - All CONDITION BASIS This event represents an extreme threat to the safety of the plant and possible substantial threat to the public. This event encompasses conditions under which a hostile force has taken physical control of vital areas required to reach and maintain safe shutdown.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-HG1 EXAMPLE 1 & 2) 6.0 HAZARDS PAGE 180 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION mII PROCEDURE TECHNICAL BASIS 6.0 HAZARDS

____ M UNUSUAL EVENT Vehicle crash (for example; aircraft or barge) into plant structures or systems within the protected area boundary.

OPERATING - All CONDITION BASIS This event classification is intended to address such items as plane, helicopter, or barge crash that may potentially damage plant structures containing functions and systems required for safe shutdown of the plant.

The crash should be of sufficient impact to cause structural damage to a plant system or structure. Visual observation of structural damage is sufficient to trigger this event classification.

Escalation to higher emergency class is based on crash into any plant area affecting equipment required for safe shutdown or by system malfunction event classifications.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-HUI EXAMPLE 4) 6.0 HAZARDS PAGE 181 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION mI -

PROCEDURE TECHNICAL BASIS . 6.0 HAZARDS I

IMII[UESEUIi11:

ALERT Vehicle crash (for example; aircraft or barge) into ANY Plant vital area.

OPERATING - All CONDITION BASIS This event classification is intended to address such items as plane, helicopter, or barge crash into a plant vital area. In cases where structural damage has occurred it may be assumed that the area and associated equipment has been subjected to forces beyond design limits and thus damage may be assumed to have occurred to plant safety systems.

It should not be interpreted that a lengthy damage assessment is necessary prior to classification. Declaration of Alert with subsequent manning of support personnel will provide adequate personnel to make a detailed assessment. The Site Emergency Director should also consider any security aspects of the crash.

Escalation to higher emergency classification is based on system malfunction, fission product barrier degradation, radioactivity release, or Emergency Director Judgement event classification.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-HAl EXAMPLE 5) 6.0 HAZARDS PAVTV 1R) A' 1n7

EMERGENCY EPIP-1 CLASSIFICATION SECTION m 7.0 NATURAL PROCEDURE TECHNICAL BASIS EVENITS4 10NT NATURAL EVENTS 7.0 7.0 NATURAL PAGE 183 OF 207 EVENTS REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION ))I 7.0 NATURAL PROCEDURE TECHNIrAT RASIS lVxrl~nrQ

, - AU V e,1l A .0 WAIUU[@ A UNUSUAL EVENT Valid annunciation in Unit One control room, Panel 1-XA-55-4B, Window 29, START OF STRONG MOTION ACCELEROGRAPH AND Assessment by Unit One and Two Control Room personnel that an earthquake has occurred.

OPERATIN - All CONDITIOP BASIS The purpose of this event classification is to recognize an earthquake of low intensity that should not affect the ability of safety systems to function. Window 29 on Panel I-XA-55-4B alarms at > .Olg Triaxial acceleration to alert the operator of seismic activity.

The assessment by Control Room personnel includes a determination considering the following:

• Apparent ground motion

• Report by other plant personnel

• Verification from the National Earthquake Center (refer to 0-AOI-100-5, Earthquake).

Escalation to Alert is based on the occurrence of an earthquake greater in magnitude than that analyzed in the Final Safety Analysis Report (FSAR) as design basis earthquake [.lOg Horizontal Operating Basis Earthquake (OBE)] or system malfunction event classification.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC-HU1 EXAMPLE 1)

- Browns Ferry Nuclear Plant, FSAR, Volume 1, Section 2.5.5.1

- O-AOI-100-5, Earthquake 7.0 NATURAL PAGE 184 OF 207 EVENTS REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III 7.0 NATURAL lPROCEDURE TECIINICAL BASIS EVENTS IWitIMU[S

• A ALERT Any of the following annunciations in Unit One Control Room, Panel 1-XA-55-4B

• Window 22, SEISMIC TRIGGER A

• Window 23, SEISMIC TRIGGER B

• Window 30, SEISMIC TRIGGER C AND Assessment by Unit One and Two Control Room personnel that an earthquake has occurred.

OPERATING - All CONDITION BASIS The purpose of this event classification is to recognize an earthquake of greater magnitude (lOg) than the Operating Basis Earthquake (OBE) described in the Final Safety Analysis Report (FSAR). Seismic events of this magnitude can cause damage to plant safety systems or functions. This equipment may be assumed to have been damaged when subjected to forces beyond design limits. Actuation of the emergency centers ensures additional personnel are available to make a subsequent, more thorough, assessment of damage to structures or equipment.

The assessment by Control Room personnel includes a determination considering the following:

• Apparent ground motion

• Report by other plant personnel

• Verification from the National Earthquake Center (refer to 0-AOI-100-5, Earthquake).

Escalation to higher emergency classification is based on system malfunction, fission product barrier degradation, radioactivity release, or Emergency Director Judgement.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC HAI EXAMPLE 1)

- Browns Ferry Nuclear Plant, FSAR Volume 1, Section 2.5.5.1 and Volume 7, Section 12.2

- 0-AOI- 100-5, Earthquake 7.0 NATURAL PAGE 185 OF 207 EVENTS REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION m 7.0 NATURAL PROCEDURE TECHNICAL BASIS EVENTSl EWNTF UNUSUAL EVENT Report by plant personnel of Tornado striking within the protected area boundary.

OPERATINM - All CONDITION1 BASIS This event classification is intended to recognize that a tornado touching down within the protected area boundary may have potentially damaged plant structures containing functions or systems required for safe shutdown of the plant.

Escalation to higher emergency class is based on confirmation of damage to plant structures containing functions or systems required for safe shutdown or system malfunction event classifications.

REFERENCES Reg Guide 1.101 Rev. 3, (NUMARC HUI EXAMPLE 2) 7.0 NATURAL PAGE 186 OF 207 EVENTS REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III 7.0 NATURAL PROCEDURE BASIS TECHNrAl,TECWI VIr..TRo RAA CAT. 0 KS] IFi1 3101111 [u I VA LI pinn ALERT Tornado striking plant vital area OR Onsite wind speed above 90 MPH as indicated using the meteorological data screen of the Integrated Computer System (ICS).

OPERATING - All CONDITION BASIS This event classification addresses wind loads that may have exceeded the design basis wind loads for plant structures containing functions or systems required for safe shutdown.

Environmental data is detected by instrumentation at the meteorological tower, transmitted to the plant, and displayed by the Integrated Computer System. The 90 MPH value was chosen because it is below the 100 MPH design basis value and within the upper range of the instrument scale.

The Alert classification is appropriate because manning of emergency centers will ensure adequate personnel to perform thorough damage assessment of structures and equipment.

Escalation to higher emergency classification is based on system malfunction, fission product barrier degradation, radioactivity release, or Emergency Director Judgement event classifications.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC HAl EXAMPLE 2)

Browns Ferry Nuclear Plant, FSAR Volume 7, Section 12.2 and Volume 1, Section 2.3 7.0 NATURAL PAGE 187 OF 207 EVENTS REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION mii 7.0 NATURAL PROCEDURE TECENIVAL BASISR lixrVWNT.

Z~ V ~11 A U1610] I ____

UNUSUAL EVENT Wheeler Lake level greater than elevation 565 FT.

AND Water entering permanent plant structures due to flooding.

OPERATINM - All CONDITIOIN BASIS This event classification is intended to address situations that are precursors to more serious events. Water entering any permanent plant structure can cause equipment malfunctions or failures which lead to degraded plant conditions or potential degradation in the level of safety of the plant. Elevation 565 FT.

corresponds to the elevation of the intake pumping station deck and access passages into most permanent plant structures.

Escalation to Alert is based on lake level greater than elevation 565 FT. and breech or failure of any water tight structure or affecting equipment required for safe shutdown.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC HUIl EXAMPLE 7)

- Browns Ferry Nuclear Plant, FSAR, Volume 7, Section 12.2 I

7.0 NATURAL PAGE 188 OF 207 EVENTS REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION m11 7.0 NATURAL PROCEDURE TFCYNr.Al. RA.QT. lvilrWNTe T1WHTCAT RA~~ V J111i%

P~~TT A0 A

TW -

ALERT Wheeler Lake level greater than elevation 565 FT.

AND Either of the following conditions exists:

• Breech or failure of any watertight structure causing flooding of the structure.

• Affecting equipment required for safe shutdown.

OPERATIN( - All CONDITIOI BASIS This event classification is intended to address flooding that may affect equipment or components required for safe shutdown of the plant. Water entering watertight structures will lead to equipment failures if the structure cannot be sealed and places the plant outside the design analysis for flooding representing a potential degradation in the level of safety of the plant. Actual indication of equipment being affected which is required for safe shutdown is an indication of actual degradation in the level of safety of the plant.

Escalation to higher event classification is based on system malfunction, fission product barrier degradation, radioactivity release, or Emergency Director Judgement event classifications.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARCHAI EXAMPLE 7)

Browns Ferry Nuclear Plant, FSAR, Volume 7, Section 12.2 7.0 NATURAL PAGE 189 OF 207 EVENTS REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION m PROCEDURE TECHrNICAL. RACT.4 Oan MODF~

TA A 0 *  !CM'r EMERGENCY DIRECTOR JUD GEMENT 8.0 8.0 OTHER PAGE 190 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 8.0 OTHER EkUiIhILVAEIMkUIU[WIU[OJF:NkI UNUSUAL EVENT Inability to reach required shutdown condition within Technical Specification Limiting Conditions for Operation (LCO) limits.

OPERATING - Mode 1 CONDITION - Mode 2

- Mode 3 BASIS Most Technical Specification LCOs require the plant to be brought to a required operating condition when Technical Specification required configuration cannot be restored within a specified time. Depending on the circumstances, this may or may not be an emergency or precursor to a more serious condition. .The initiation of plant shutdown required by plant Technical Specifications requires a report under 10 CFR 50.72(b) non-emergency events. The plant is within its safety envelope when being shutdown within the allowable action statement time in the Technical specifications. An immediate notification of Unusual event is required when the plant is not brought to the required operating mode within the allowable action statement time in the Technical Specifications. Declaration of an Unusual event is based on the time at which the LCO-specified action statement time period elapses under Technical Specifications and is not related to how long a condition may have existed. Other required Technical Specification shutdowns that involve precursors to more serious events are addressed by other system malfunction, hazards, or fission product barrier degradation event classifications.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC SU2)

- Technical Specifications 8.0 OTHER PAGE 191 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 8.0 OTHER

- -0 k*I&'r*] Et(,1L MI IL'I ru nre1':au .

UNUSUAL EVENT Unplanned loss of onsite communications listed in Table 8.2-U that defeats the Plant Operations Staffs ability to perform routine operations OR Unplanned loss of ALL offsite communications listed in Table 8.2-U.

OPERATING - All CONDITION BASIS The purpose of this event classification is to recognize a loss of communications capability that either defeats the Plant Operations Staffs ability to perform routine tasks necessary for plant operations or the ability to communicate with offsite authorities. Table 8.2-U contains all credible means of routine and emergency onsite and offsite communications.

Offsite communications loss encompasses all means of communications with offsite authorities or support organizations. This part of the event classification is intended to exists only when extraordinary means are being utilized to make communication possible (i.e., relaying of messages or sending individuals to offsite locations with messages).

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC SU6 )

8.0 OTHER PAGE 192 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION m PROCEDURE TECHNICAL BASIS 8.0 OTHER UNUSUAL EVENT (CONTINUED)

CURVES/TABLES uInlML C'UMMUI ICATI1 OFFSITE COMMUNICATIOI PLANT PHONF SYSTEM NOD. 1 BELL LINES TWO WAY RADIO (CH Fl, F2, ,4. and F5) DIGITAL MICROWAVE SOUND POWER PHONES _ NRC (F7S-2000) i CELLULAR PHONES (IF AVAILABLE) 8.0 OTHER PAGE 193 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION mII PROCEDURE TECHNICAL BASIS 8.0 OTHER UNUSUAL EVENT Unplanned loss of most or all safety system annunciators or indicators which causes a significant loss of plant assessment capability for greater than 15 minutes AND Compensatory non-alarming safety system indications are available (SPDS, ICS)

AND In the opinion of the Shift Manager, increased surveillance is required to safely operate the plant.

OPERATING - Mode I CONDITION - Mode 2

- Mode 3 BASIS This event classification is intended to recognize the difficulty associated with monitoring changing plant conditions without the use of a major portion of the annunciation or indication equipment.

Quantification of most is arbitrary and should be based on the Site Emergency Director's Judgement that there is increased risk that a degraded plant condition could go undetected. If in the opinion of the Site Emergency Director additional personnel are required to provide increased monitoring of system or plant operation to safely operate the unit(s) then this event classification should be considered and subsequent declaration of event should be based upon whether a degradation in the level of safety of the plant is represented by the extent of the loss in assessment capability. This judgement should take into consideration those annunciators identified in Alarm Response Procedures (ARPs), Abnormal Operating Instructions (AOIs), Emergency Operating Instructions (EOIs),

Emergency Preparedness Implementing Procedures (EPIPs), and the operators ability to recognize entry conditions based on symptoms that require or support procedure execution.

8.0 OTHER PAGE 194 OF 207 1k REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION m PROCEDURE TECHNICAL BASIS 8.0 OTHER L A UNUSUAL EVENT (CONTINUED)

BASIS Individual system annunciators and/or indicators may impact the function of system or component operability. System and component operability are addressed by Technical Specifications. Inoperability of multiple safety systems beyond Technical Specification compliance is addressed at 8.1-U.

Unplanned excludes loss of annunciators due to scheduled maintenance or testing activities. Compensatory non-alarming safety system indications include the Integrated Computer System (ICS) and SPDS.

15 minutes was chosen as a threshold to exclude transient or momentary power losses.

Escalation to Alert is based on a transient or significant power change required or in progress or loss of compensatory non-alarming indications.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUTMARC SU3) 8.0 OTHER PAGE 195 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION Im1 PROCEDURE TECHNICAL BASIS 8.0 OTHER M.

k@IISJ iYi IL'fJLlI UhU EUi WA IlIUM IUMNj ALERT Unplanned loss of most or all safety system annunciators or indicators which causes a significant loss of plant assessment capability for greater than 15 minutes AND In the opinion of the Shift Manager, increased surveillance is required to safely operate the plant AND EITHER of the following conditions exists:

• Compensatory non-alarming safety system indications are NOT available (SPDS, ICS).

• A significant transient is in progress.

OPERATING - Mode I CONDITION - Mode 2

- Mode 3 BASIS This event classification is intended to recognize the difficulty associated with monitoring changing plant conditions without the use of a major portion of the annunciation and indication equipment during a transient.

Quantification of most is arbitrary and should be based on the Site Emergency Director's Judgement that there is increased risk that a degraded plant condition could go undetected. If in the opinion of the Emergency Director, additional personnel are required to provide increased monitoring of systems or plant operation to safely operate the units and either compensatory monitoring indications are not available or a significant transient is in progress then this event classification should be considered. Subsequent declaration of event should be based upon whether a substantial degradation in the level of safety of the plant has occurred due the extent of the loss of assessment capability.

8.0 OTHER PAGE 196 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION mI PROCEDURE TECHNICAL BASIS 8.0 OTHER ALERT (CONTINUED)

BASIS This judgement should take into consideration those annunciators identified in Alarm Response Procedures (ARPs), Abnormal Operating Instructions (AOIs),

Emergency Operating Instructions (EOIs), Emergency Preparedness Implementing Procedures (EPIPs), and the operators ability to recognize entry conditions based on symptoms that require or support procedure execution.

Loss of individual system annunciators and/or indicators may impact the function of system or component operability. System and component operability are addressed by Technical Specifications. Inoperability of multiple safety systems beyond Technical Specification compliance is addressed by separate event classification at 8.1-U.

Unplanned excludes loss of annunciators due to scheduled maintenance or testing activities. Compensatory non-alarming safety system indications include the Integrated Computer System (ICS) and SPDS.

15 minutes was chosen as a threshold to exclude transient or momentary power losses.

Escalation to Site Area Emergency is based on inability of the operating crew to monitor a significant transient in progress due to loss of annunciators and compensatory non-alarming indications during the transients.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC SA4)

NOTES NOTE 8.3 Significant transient includes response to automatic or normally initiated functions such as scrams or runbacks involving greater than 25% core thermal power change, Emergency Core Cooling System (ECCS) injections, or thermal power oscillations of 10% or greater.

8.0 OTHER PAGE 197 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECr1TCAT. RARERZ ,D n mon

-- ~O.U IJMflL SITE AREA EMERGENCY Loss of most or all annunciators associated with safety systems AND Compensatory non-alarming safety system indications are NOT available (SPDS, ICS)

AND Indications needed to monitor safety functions are NOT available (Refer to Table 8.3-S)

AND A significant transient is.in progress OPERATING - Mode 1 CONDITION - Mode 2

- Mode 3 BASIS This event classification is intended to recognize the inability of the Control Room operators to monitor the plant response to a transient. A Site Area Emergency is considered to exist if the Control Room staff cannot monitor safety functions needed for protection of the public.

Annunciators that apply to this event classification are those that are required in support of Alarm Response Procedures (ARPs), Abnormal Operating Instructions (AOIs), Emergency Operating Instructions (EOIs), and Emergency Preparedness Implementing Procedures (EPIPs).

Compensatory non-alarming safety system indications include the Integrated Computer System (ICS) and SPDS .

8.0 OTHER PAGE 198 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION PROCEDURE SECTION mII TECHNICAL BASIS 8.0 OTHER M

K] I M .I I01 SITE AREA EMERGENCY (CONTINUED)

Table 8.3-S includes indications necessary to shutdown the reactor, maintain the core cooled and in a coolable geometry, maintain RCS intact, maintain containment integrity.

REFERENCES Reg Guide 1.101 Rev. 3, (NUMARC SS6)

CURVES/TABLES r] n REACTOR LEVEL SUBCRITICALLITY DRYWELL TEMPERATURE DRYWELL PRESSURE SUPPRESSION CHAMBER PRESSURE SUPPRESSION CHAMBER TEMPERATURE SUPPRESSION POOL LEVEL NOTES NOTE 8.3 Significant transient includes response to automatic or normally initiated functions such as scrams or runbacks involving greater than 25% core thermal power change, Emergency Core Cooling System (ECCS) injections, or thermal power oscillations of 10% or greater.

8.0 OTHER PAGE 199 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECHICAL BASIS 8.0 OTHER UNUSUAL EVENT Other events are in process or have occurred which indicate a potential degradation in the level of safety of the plant. No radioactive releases are expected which require offsite response. Refer to Table 8.4-U for examples.

OR Any Loss or Potential Loss of Containment OPERATING - All CONDITION BASIS This event classification is intended to address conditions not explicitly addressed elsewhere that, in the judgement of the Site Emergency Director, warrant Unusual Event classification. Examples are provided in Table 8.4-U but classification is not restricted to only those events listed in the table. BFN EAL's were developed primarily utilizing the symptom based grouping methodology. This approached is consistent with the BFN EOI methodology. It is important to note here that the consideration of fission product barriers has been incorporated within this symptom based approached. Barrier-based EAL's refer to the level of challenge to principal barriers used to assure containment of radioactive material. For radioactive materials that are contained within the reactor core, these barriers are:

fuel cladding, reactor coolant system pressure boundary, and containment.

The level of challenge to these barriers encompasses the extent of damage (loss or potential loss) and the number of barriers currently under challenge.

Site Emergency Directors should be continuously aware of all challenges to these barriers and the number of barriers loss or potentially loss. Also Site Emergency Directors should consider that when the loss or potential loss thresholds is imminent (i.e., I to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />) use judgement and classify as if the thresholds are exceeded.

The threshold for fission product barrier loss is considered to be consistent with the following:

Primary Containment barrier - Refer to 2.5-U.

REFERENCES - Reg Guide 1.101 Rev.' 3, (NUMARC HU5, FU) 8.0 OTHER PAGE 200 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION HI PROCEDURE TECNICAL. BA.QT.P. 0 A ^'1"XMD O.U NJL~aVRLE UNUSUAL EVENT (CONTINUED)

CURVES/TABLES EXAMPLE I PLANT TRANSIENT RESPONSE UNEXPECTED i Ad 1(b UNANALYZED SAFETY SYSTEM SONFIGURATION A THREATRI QING SAFE SHUTDOWN INADEQUATE PERSONNEL TO ACHIEVE OR MAINTATN. !AP14T RTRTlAUTKT DEGRADED PLANT CONDITIONS BEYOND LICENSE BASIS THREATENING SAFE OPERATION OR

.SAFE SHUTDOWN EMERGENCY PROCEDURES NOT ADEQUATE TO MAINTAIN SAFE OPERATION OR ACHIEVE SAFE SHUTDOWN NOTES Note 8.4-U Table 8.4-U contains only example events that may justify unusual event classification. This event classification is intended to address unanticipated conditions not explicitly addressed elsewhere, but that warrant declaration of an emergency because conditions exist which the Site Emergency Director believes to fall under the Unusual event classification. Additionally this EAL should be considered in making emergency classifications regarding challenges to fission product barriers not specifically address elsewhere in the EAL matrix.

8.0 OTHER PAGE 201 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 8.0 OTHER ALERT Other events are in process or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant. Radioactive releases are expected to be within a small fraction of the EPA Guidelines.

OR Any Loss or Any Potential Loss of Either Fuel Cladding or RCS Pressure Boundary OPERATING - All CONDITION BASIS This event classification is intended to address conditions not explicitly addressed elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Site Emergency Director to fall under the Alert classification. Examples are provided in Table 8.4-A but classification is not restricted to only those events listed in the table. BFN EAL's were developed primarily utilizing the symptom based grouping methodology. This approached is consistent with the BFN EOI methodology. It is important to note here that the consideration of fission product barriers has been incorporated within this symptom based approached. Barrier-based EAL's refer to the level of challenge to principal barriers used to assure containment of radioactive material. For radioactive materials that are contained within the reactor core, these barriers are:

fuel cladding, reactor coolant system pressure boundary, and containment. The level of challenge to these barriers encompasses the extent of damage (loss or potential loss) and the number of barriers currently under challenge. Site Emergency Directors should be continuously aware of all challenges to these barriers and the number of barriers loss or potentially loss. Also Site Emergency Directors should consider that when the loss or potential loss thresholds is imminent (i.e., I to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />) use judgement and classify as if the thresholds are exceeded.

8.0 OTHER PAGE 202 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECrYMVAT. RAQIR Qa MOD~.

OM %l I -

ALERT (CONTINUED)

The threshold for fission product barrier loss is considered to be consistent with the following:

Fuel clad - Approximately 5% cladding failure. A Reactor coolant sample that yields a results of 300 XCi/gm Iodine- 13 1 equivalent is indicative of this amount of cladding failure (Refer to 1.3-A).

RCS barrier - Reactor coolant leakage of at least 50 GPM from the primary system (Refer to 2.4-A).

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC HA6, FA)

NOTES Note 8.4-A This event classification is intended to address unanticipated conditions not explicitly addressed elsewhere, but that warrant declaration of an emergency because conditions exist which the Site Emergency Director believes to fall under the Alert classification. Additionally this EAL should be considered in making emergency classifications regarding challenges to fission product barriers not specifically address elsewhere in the EAL matrix.

8.0 OTHER PAGE 203 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION mII PROCEDURE TECHNICAL BASIS 8.0 OTHER inIEEguM! I.

SITE AREA EMERGENCY Other events are in process or have occurred which involve actual or likely major failures of plant functions needed for protection of the public. Releases are not expected to result in exposure levels that exceed EPA Guidelines except near the site boundary.

OR Any Loss or Potential Loss of Both Fuel Cladding and RCS Pressure Boundary OR Potential Loss of Either Fuel Cladding or RCS Pressure Boundary and Loss of any additional barrier OPERATING - All CONDITION BASIS This event classification is intended to address unanticipated conditions not explicitly addressed elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Site Emergency Director to warrant Site Area Emergency classification. BFN EAL's were developed primarily utilizing the symptom based grouping methodology. This approached is consistent with the BFN EOI methodology. It is important to note here that the consideration of fission product barriers has been incorporated within this symptom based approached. Barrier-based EAL's refer to the level of challenge to principal barriers used to assure containment of radioactive material.

For radioactive materials that are contained within the reactor core, these barriers are:

fuel cladding, reactor coolant system pressure boundary, and containment.

The level of challenge to these barriers encompasses the extent of damage (loss or potential loss) and the number of barriers currently under challenge.

Site Emergency Directors should be continuously aware of all challenges to these barriers and the number of barriers loss or potentially loss. Also Site Emergency Directors should consider that when the loss or potential loss thresholds is imminent (i.e., I to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />) use judgement and classify as if the thresholds are exceeded.

8.0 OTHER PAGE 204 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION I1I PROCEDURE TECHNCAL. RAsRs so d'%nr1TV' TECHNIC~.

RA ~M IV f~I nzW SITE AREA EMERGENCY (CONTINUED)

Loss or potential loss of any two fission product barriers must be considered along with inability to monitor fission product barriers during extreme conditions.

The threshold for fission product barrier loss is considered to be consistent with the following:

Fuel clad - Approximately 5% cladding failure. A Reactor coolant sample that yields a results of 300 gCi/gm Iodine-13 1 equivalent is indicative of this amount of cladding failure (Refer to 1.3-A).

RCS barrier - Reactor coolant leakage of at least 50 GPM from the primary system (Refer to 2.4-A).

Primary Containment barrier - Refer to 2.5-U.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC HS3, FS)

NOTES Note 8.4-S This event classification is intended to address unanticipated conditions not explicitly addressed elsewhere, but that warrant declaration of an emergency because conditions exist which the Site Emergency Director believes to fall under the Site Area Emergency classification.

Additionally this EAL should be considered in making emergency classifications regarding challenges to fission product barriers not specifically address elsewhere in the EAL matrix.

8.0 OTHER PAGE 205 OF 207 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION 111 PROCEDURE TECHNICAL BASIS 8.0 OTHER GENERAL EMERGENCY Other events are in process or have occurred which involve actual or imminent substantial core degradation or melting with potential for loss of containment integrity. Releases are expected to exceed EPA Guidelines for exposure levels offsite beyond the site boundary.

OR Loss of Any Two Barriers and Potential Loss of Third Barrier OPERATING - All CONDITION BASIS This event classification is intended to address unanticipated conditions not explicitly addressed elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Site Emergency Director to fall under the General Emergency classification. BFN EAL's were developed primarily utilizing the symptom based grouping methodology. This approached is consistent with the BFN EOI methodology. It is important to note here that the consideration of fission. product barriers has been incorporated within this symptom based approached. Barrier-based EAL's refer to the level of challenge to principal barriers used to assure containment of radioactive material.

For radioactive materials that are contained within the reactor core, these barriers are:

fuel cladding, reactor coolant system pressure boundary, and containment.

The level of challenge to these barriers encompasses the extent of damage (loss or potential loss) and the number of barriers currently under challenge.

Site Emergency Directors should be continuously aware of all challenges to these barriers and the number of barriers loss or potentially loss. Also Site Emergency Directors should consider that when the loss or potential loss thresholds is imminent (i.e., I to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />) use judgement and classify as if the thresholds are exceeded.

8.0 OTHER PAGE 206 OF 207 1 REVISION 28

EMERGENCY EPIP-1 CLASSIFICATION SECTION m PROCEDURE TEC1WVMAIr.1RAqirc 5.U UI nIiK I '4 S

GENERAL EMERGENCY (CONTINUED)

Loss or potential loss of all fission product barriers must be considered along with inability to monitor fission product barriers during extreme conditions.

The threshold for fission product barrier loss is considered to be consistent with the following:

Fuel clad - Approximately 5% cladding failure. A Reactor coolant sample that yields a results of 300&iCi/gm Iodine- 13 1 equivalent is indicative of this amount of cladding failure (Refer to 1.3-A).

RCS barrier - Reactor coolant leakage of at least 50 GPM from the primary system (Refer to 2.4-A).

Primary Containment barrier - Refer to 2.5-U.

REFERENCES - Reg Guide 1.101 Rev. 3, (NUMARC HG2, FG)

NOTES Note 8.4-G This event classification is intended to address unanticipated conditions not explicitly addressed elsewhere, but that warrant declaration of an emergency because conditions exist which the Site Emergency Director believes to fall under the General Emergency classification. Additionally this EAL should be considered in making emergency classifications regarding challenges to fission product barriers not specifically address elsewhere in the EAL matrix.

8.0 OTHER PAGE 207 OF 207 REVISION 28

TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURE EPIP-13 RADIOCHEMICAL LABORATORY PROCEDURE REVISION 7 PREPARED BY: TONY FELTMAN PHONE: 3666 RESPONSIBLE ORGANIZATION: EMERGENCY PREPAREDNESS APPROVED BY: GILBERT V. LITTLE DATE: 11/03/99 EFFECTIVE DATE: 11/04/99 f LEVEL OF USE: REFERENCE USE VALIDATION DATE: NOT REQUIRED QUALITY-RELATED

REVISION LOG Procedure Number: EPIP-13 Revision Number: 7 Pages Affected: 1-3 Pagination Pages: NONE Description of Change:

NIC-07 - OTHER - Change references only.

IC (4/99) Revision updates position titles, and enables the Lead Radiological Laboratory Analyst (LRLA) for the conduct of the procedure.

RADIOCHEMICAL BROWNS FERRY LABORATORY PROCEDURE EPIP-13 NUCLEAR PLANT RADIOCHEMICAL LABORATORY PROCEDURE 1.0 PURPOSE This procedure provides guidance to Chemistry personnel during a Radiological Emergency.

1.0 SCOPE This procedure outlines the actions to be followed by Radiochemical Laboratory Analysts (RLAs) and other chemistry personnel during a radiological emergency.

This procedure describes those Radiochemical Laboratory actions required during an emergency involving radiochemical problems.

NOTE: Unit 1, Unit Operator will initiate EPIP-1 3 by calling the Radiochemical Laboratory Shift Supervisor/Lead Radiochemical Laboratory Analyst (LRLA).

3.0 INSTRUCTIONS 3.1 Notification of Unusual Event 3.1.1 No offsite radiochemical problems are postulated during a NOTIFICATION OF UNUSUAL EVENT. This situation should not have any major impact on the Radiochemical Laboratory.

3.1.2 Although the lab will not automatically be called, should assistance be needed, RLAs will follow standard practices and procedures during any response work.

3.2 Alert 3.2.1 Card into PREAS in the RADCON lab.

INITIALS TIME 3.2.2 RLAs report to the Radiochemical Lab Shift Supervisor/ LRLA. INITIALS TIME 3.2.3 Prepare to implement 2/3-TI-331 Post Accident Sampling Procedure and CI-900 INITIALS TIME Analysis Procedures.

3.2.4 Verify proper operation of laboratory assigned survey instruments. INITIALS TIME PAGE 1 OF 3 REVISION 7

RADIOCHEMICAL BROWNS FERRY LABORATORY PROCEDURE EPIP- 13 NUCLEAR PLANT 3.0 INSTRUCTIONS (CONTINUED) 3.3 Site Area Emergencv 3.3.1 Card into PREAS in the RADCON lab. If not INITIALS TIME previously conducted.

3.2.2 RLAs report to the Radiochemical Lab Shift INITIALS TIME Supervisor/ LRLA.

3.2.3 Prepare to implement 2/3-TI-331 Post INITIALS TIME Accident Sampling Procedure and CI-900 Analysis Procedures.

3.2.4 Verify proper operation of laboratory INITIALS TIME assigned survey instruments.

3.4 General Emergency 3.4.1 Card into PREAS in the RADCON lab. If not INITIALS TIME previously conducted.

3.4.2 RLAs report to the Radiochemical Lab Shift INITIALS TIME Supervisor/ LRLA.

3.4.3 Prepare to implement 2/3-TI-331 Post INITIALS TIME Accident Sampling Procedure and CI-900 Analysis Procedures.

3.4.4 Verify proper operation of laboratory INITIALS TIME assigned survey instruments.

PAGE 2 OF 3 REVISION 7

RADIOCHEMICAL BROWNS FERRY LABORATORY PROCEDURE EPIP-13 NUCLEAR PLANT 3.0 INSTRUCTIONS (CONTINUED) 3.5 Site Evacuation 3.5.1 RLAs proceed to lab, if habitable, report to Radiochemical Shift Supervisor/LRLA. If INITIALS TIME uninhabitable, report to a location determined by RADCON.

3.5.2 Inform Chemistry OSC Manager of new location. INITIALS TIME 4.0 ATTACHMENTS None LAST PAGE PAGE 3 OF 3 REVISION 7

TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURE EPIP-14 RADIOLOGICAL CONTROL PROCEDURES REVISION 14 PREPARED BY: TONY FELTMAN PHONE: 3666 RESPONSIBLE ORGANIZATION: EMERGENCY PREPAREDNESS APPROVED BY: GILBERT V. LITTLE DATE: 11/03/99 EFFECTIVE DATE: 11/04/99 LEVEL OF USE: REFERENCE USE VALIDATION DATE: NOT REQUIRED QUALITY-RELATED

REVISION LOG Procedure Number: EPIP-14 Revision Number: 14 Pages Affected: 6,11,15,18 Description of Change:

IC THIS REVISION IS BEING CONDUCTED TO; APPLY HUMAN FACTORS TO THE DOSE CALCULATION FOR TEDE, REMOVE THE RADIATION SAFETY AND CONTROL UNIT RESPONSIBILITIES, DUE TO THE REDUCTION OF THE UNIT, THE INVENTORY REQUIREMENT WILL ROLE INTO EPIP-17,AND CHANGE THE WEATHER SERVICE CONTACT TELEPHONE NUMBER.

IC THIS REVISION IS BEING CONDUCTED TO; HUMAN FACTOR DOSE PROJECTION PROCESS, AND UPDATE THE NATIONAL WEATHER SERVICE TELEPHONE NUMBER.

RADIOLOGICAL fTTh1A BROWNS FERRY CONTROL PROCEDURES EPrIr NUCLEAR PLANT 1.0 PURPOSE The purpose of this procedure is to describe actions and responsibilities of Radiological Controls (RADCON) personnel during a radiological emergency at Browns Ferry.

2.0 SCOPE EPIP-14 will be initiated when RADCON Shift Supervisor or designee receives indications that the Emergency Plan has been activated or information regarding processes contained within this procedure are required.

EPIP-14 contains instructions for RADCON during the implementation of the Emergency Plan event classifications. The procedure additionally contains instructions for RADCON during Site Assembly and Evacuation, RADCON Lab Habitability, Issuance of Potassium Iodide, the use of the Health Physics Network, the Alternate Personnel Decontamination Facility and methods for Projecting Total Effective Dose Equivalent (TEDE) from airborne radioactivity releases.

The method for projecting TEDE from airborne radioactivity releases may be requested by operations to support emergency classification and/or protective action recommendations. The use of this method should only be utilized in the absence of more sophisticated dose models, when the Central Emergency Control Center (CECC) is not activated.

3.0 INSTRUCTIONS 3.1 Notification of Unusual Event 3.1.1 No offsite radiological problems are postulated during a Notification of Unusual Event. (NOUE). This situation should not have any major impact on RADCON.

3.1.2 Although RADCON will not automatically be called, should assistance be needed, RADCON will follow standard practices and procedures during response activities.

PAGE 1 OF 17 REVISION 14

RADIOLOGICAL T n 1A BROWNS FERRY CONTROL PROCEDURES EPrIr NUCLEAR PLANT 3.0 INSTRUCTIONS (CONTINUED) 3.2 Alert 3.2.1 When a Site Assembly is conducted, (see Section 3.4) all RADCON personnel will report to their assigned assembly areas.

3.2.2 If radiological conditions warrant, RADCON personnel will periodically take radiation, airborne (particulate and iodine), and contamination surveys of all assembly areas inside the protected area. This is to include the Control Rooms, Technical Support Center (TSC), Operations Support Center (OSC), RADCON Lab, Chemistry Lab, OSC Staging Area, Plant Assembly Room, Maintenance Building and Engineering Building.

3.2.3 A RADCON technician will accompany any personnel dispatched into areas of potential radiological hazard.

3.2.4 RADCON personnel will assist in the development of recovery plans as deemed necessary by the recovery organization. Recommendations will be made to keep exposure as low as reasonable achievable and to recommend and approve any clean up activities.

3.3 Site Area Emergency or General Emergency 3.3.1 RADCON technicians report to the lab as directed by their Shift Supervisor or designee. A site evacuation will be conducted at the SAE or GE classification, if not already completed (see Section 3.5).

3.3.2 RADCON personnel will periodically take radiation, contamination and airborne surveys as necessary to ensure no radiological hazards exist in occupied Emergency Response Facilities, (TSC, OSC, Staging Area, RADCON Lab, Chemistry Lab, Control Rooms, or other Operations areas).

3.3.3 A RADCON technician will accompany any personnel dispatched into areas of potential radiological hazard.

3.3.4 Equipment listed in CECC-EPIP-9, Attachment J, Section 1.0 may need to be transported to the environmental monitoring van. Nuclear Security (NS) will allow equipment to be removed from the protected area.

PAGE 2 OF 17 REVISION 14

RADIOLOGICAL T BROWNS FERRY CONTROL PROCEDURES EPIP-1 NUCLEAR PLANT 3.0 INSTRUCTIONS (CONTINUED) 3.3 Site Area Emergency or General Emergency (continued) 3.3.5 Initial offsite environmental assessment will be conducted per CECC-EPIP-9.

3.3.6 Dispatch RADCON technician to the site access control point established by NS personnel. Survey vehicles and personnel leaving the site using RM-1 4 friskers (or equivalent) and smear techniques, if radiological conditions warrant.

3.4 Site Assembly and Evacuation 3.4.1 RADCON technicians proceed to the RADCON Lab and read your badge into the accountability reader. If uninhabitable, see Section 3.5.

3.4.2 Sign the Accountability Roster.

3.4.3 If any plant personnel are unaccounted for, NS will form search teams, each having at least one RADCON technician as a part of the team.

3.4.4 RADCON will survey personnel and vehicles leaving the site at the NS access control point, if radiological conditions warrant. Contaminated individuals will be evacuated to the Power Service Shop No. 4 Locker Room at Muscle Shoals Reservation, as directed by the OSC.

3.4.5 Should conditions exist that RADCON cannot survey all people and vehicles leaving the site, RADCON will set up a monitoring station as directed by the SED.

3.5 Radiological Control Lab Habitability 3.5.1 [NRC/C] When conditions within the Radcon Lab become uninhabitable the RADCON technicians will proceed to mechanical equipment room, control bay, Unit 3, elevation 617. [NRC/C 81-19-17]

3.5.2 [NRC/C] Report location to the RADCON Manager in the TSC.

[NRC/C 81-19-17]

PAGE 3 OF 17 REVISION 14

RADIOLOGICAL BROWNS FERRY CONTROL PROCEDURES EPIP-14 TfTD1A NUCLEAR PLANT 3.6 Issuina Potassium Iodide (KI) 3.6.1 If the TSC RADCON Manager has reason to believe that a person's projected cumulative dose to the thyroid from inhalation of radioactive iodine might exceed 10 rems (see Attachment A), the exposed person should be started immediately on a dose regimen of KI. This decision shall be immediately communicated to the SED.

3.6.1.1 If the TSC is not staffed or the RADCON Manager position has not been filled, then the senior onsite RADCON Supervisor has the authority to issue KI utilizing the bases describe in step 3.6.1.

3.6.1.2 The initial dose of KI should be not delayed since thyroid blockage requires 30 to 60 minutes. Anyone authorized to initiate KI shall be familiar with the Food and Drug Administration approve package insert and be sure that each recipient is similarly informed.

3.6.1.3 Prior to issuing KI to an individual, the person should be asked if he/she is allergic to iodine. If the person indicates a possible sensitivity to iodine they should not be issued KI.

3.6.2 KI is stored in the plant RADCON supply cage and the REP Van instrument kits.

3.6.3 RADCON normally will not dispense a bottle of KI to TVA Personnel involved in activities to support a radiological emergency. RADCON will however dispense a single individual dose of KI to team members dispatched from the OSC.

3.6.4 Follow the dosage outlined on the package insert (Attachment B). A copy of the Food and Drug Administration approved package insert shall accompany each bottle of KI issued. If KI is distributed in individual doses, verbal instructions of the significant information on the package insert by a knowledgeable individual is sufficient.

3.6.5 Complete the KI Issue Report (Attachment C) or an RWP time sheet as appropriate for issuance of Ki. An RWP time sheet may be used for this documentation instead of completing the Attachment C. If the RWP time sheet is used to document distribution of the KI, note the time of KI distribution on the back of the time sheet.

PAGE 4 OF 17 REVISION 14

RADIOLOGICAL rr81r BROWNS FERRY CONTROLPROCEDURES EPIP-1 NUCLEAR PLANT 3.7 Use of the NRC Health Physics Network (HPN) 3.7.1 The HPN contact with the NRC will be made by the RADCON group.

3.8 Browns Ferry Alternate Personnel Decontamination Facility 3.8.1 The BFN alternate personnel decontamination facility is located at the Power Service Shop No. 4 Locker Room on the Muscle Shoals Reservation. It will be activated when the BFN personnel decontamination facility is inaccessible or incapable of handling the number of contaminated personnel.

3.8.2 When the decision is made to transport contaminated personnel to the alternate decontamination facility, BFN RADCON shall make notifications to the CECC, and the Power Service Shops.

The notification to the CECC shall include all available information at that time. Interface with state and local authorities (i.e., transportation route considerations) will be made available via the CECC.

The notification to the Power Service Shops shall include a request that the Shop 4 sewer lift station sump be emptied, followed by tagging out the power supply to the two pumps. (The sump and control panel are located adjacent to the North East corner of the Gas and Diesel Building, approximately 500 feet east of Shop 4). In the event a volume of effluent in excess of 1800 gallons is anticipated, additional containment capabilities will need to be arranged. The primary point of contact is the Supervisor, Maintenance Group, with the back-up being the Mechanical Supervisor. Notification phone numbers are listed in the Radiological Emergency Notification Directory (REND).

PAGE 5 OF 17 REVISION 14

RADIOLOGICAL a- FbTT 1 BROWNS FERRY CONTROL PROCEDURES EPrIr-14' NUCLEAR PLANT 3.0 INSTRUCTIONS (CONTINUED) 3.8 Browns Ferry Alternate Personnel Decontamination Facility (continued) 3.8.3 Browns Ferry RADCON is responsible for the following:

• Providing appropriate personnel and equipment to operate the alternate decontamination facility.

• Calculating the amount of radioactive material in the decontamination effluent. Effluent releases will be in accordance with Standard Program and Process (SPP) - 5.1.

• Documenting appropriate records on all contaminated personnel.

• Ensuring the alternate decontamination facility is secured following decontamination activities and assisting in recovery efforts.

3.9 Method for Proiectina Total Effective Dose Equivalent (TEDE) from Airborne Radioactivity Releases The method for Projecting Total Effective Dose Equivalent (TEDE) from Airborne Radioactivity Releases is by manual method through the application of matrix tables.

3.9.1 Manual Method for Projecting Total Effective Dose Equivalent (TEDE) from Airborne Radioactivity Releases 3.9.1.1 The Radcon Shift Supervisor/designee or the TSC Radcon representative is responsible for completing Attachment D of this procedure when releases involves a stack release or Attachment E when the release involves a building or ground level release.

3.9.1.2 This method for projecting the TEDE from airborne radioactivity releases should only be utilized in the absence of more sophisticated dose models.

3.9.1.3 This method may be requested by the Shift Manager prior to any emergency classification declaration. Results of this method may be utilized to classify emergency conditions, make protective action recommendations or by TSC personnel conducting evaluations of current plant conditions.

PAGE 6 OF 17 REVISION 14

RADIOLOGICAL BROWNS FERRY CONTROL PROCEDURES EPIP-14 NUCLEAR PLANT 3.0 INSTRUCTIONS (CONTINUED) 3.9 Method for Projecting Total Effective Dose Equivalent (TEDE) from Airborne Radioactivity Releases 3.9.1.4 When requested the appropriate attachment of this procedure should be completed immediately and the results reported to the Shift Manager or SED.

4.0 ATTACHMENTS Attachment A - Occupational Dose From Inhalation of lodine-1 31 Attachment B - Patient Package Insert Attachment C - Potassium Iodide Issue Report Attachment D - Manual Method for Projecting Total Effective Dose Equivalent (TEDE) from Stack Airborne Radioactivity Releases Attachment E - Manual Method for Projecting Total Effective Dose Equivalent (TEDE) from Building or Ground Level Airborne Radioactivity Releases Attachment F - Projected TEDE Assessment Survey Form PAGE 7 OF 17 REVISION 14

RADIOLOGICAL BROWNS FERRY CONTROL PROCEDURES EPIP-14 NUCLEAR PLANT ATTACHMENT A (Page 1 of 1)

OCCUPATIONAL DOSE FROM INHALATION OF IODINE-131 1.OOE-03 1.OOE-04 U) 1.OOE-05 0

to 1..

n 0 1.OOE-06 L.

0 0

(a 1.OOE-07 1.OOE-08 1.OOE-09 0.1 1 10 100 Exposure Time (hours)

PAGE 8 OF 17 REVISION 14

RADIOLOGICAL BROWNS FERRY CONTROL PROCEDURES EPIP-14 NUCLEAR PLANT ATTACHMENT B (Page 1 of 1)

PATIENT PACKAGE INSERT THYRO-BLOCK If you take potassium iodide, it will fill up your thyroid gland.

Tablets This reduces the change that harmful radioactive iodine will (POTASSIUM IODIDE TABLETS< USP) enter the thyroid gland.

(pronounced poe-TASS-e-um EYE-oh-dyed)

(abbreviated: KI) WHO SHOULD NOT TAKE POTASSIUM IODIDE The only people who should not take potassium iodide are TAKE POTASSIUM IODIDE ONLY WHEN PUBLIC HEALTH people who know they are allergic to iodide. You may take OFFICIALS TELL YOU IN A RADIATION EMERGENCY, potassium iodide even if you are taking medicines for athyroid RADIOACTIVE IODINE COULD BE RELEASED INTO THE problem (for example, a thyroid hormone or antithyroid drug).

AIR. POTASSIUM IODIDE (A FORM OF IODINE) CAN Pregnant and nursing women and babies and children may HELP PROTECT YOU also take this drug.

IF YOU ARE TOLD TO TAKE THIS MEDICINE, TAKE IT HOW AND WHEN TO TAKE POTASSIUM IODIDE ONE TIME EVERY 25 HOURS. DO NOT TAKE IT MORE Potassium Iodide should be taken as soon as possible after OFTEN. MORE WILL NOT HELP YOU AND MAY public health officials tell you. you should take one dose every INCREASE THE RISK OF SIDE EFFECTS. DO NOT TAKE 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. More will not help you because the thyroid can "hold" THIS DRUG IF YOU KNOW YOU ARE ALLERGIC IODIDE. only limited amounts of iodine. Larger doses will increase the (SEE SIDE EFFECTS BELOW.) risk of side effects. You will probably be told not to take the drug for more than 10 days.

INDICATIONS THYROID BLOCKING IN A RADIATION EMERGENCY SIDE EFFECTS ONLY. Usually, side effects of potassium iodide happen when people DIRECTIONS FOR USE take higher doses for a long time. You should be careful not to Use only as directed be State or local public health authorities take more than the recommended dose or take it for longer than in the event of a radiation emergency. you are told. Side effects are unlikely because of the low dose Dose and the short time you will be taking the drug.

Tablets: ADULTS AND CHILDREN 1 YEAR OF AGE OR OLDER: One (1)tablet once a Possible side effects include skin rashes, swelling of the day. Crush for small children. BABIES salivary glands, and "iodism" (metallic taste, burning mouth and UNDER 1 YEAR OF AGE: One half (1/2) throat, sore teeth and gums, symptoms of a head cold, and tablet once a day. Crush first. sometimes stomach upset and diarrhea).

Take tablets 10 days unless directed otherwise by State or local public health authorities. A few people have an allergic reaction with more serious symptoms. These could be fever and joint pains, or swelling of Store at controlled room temperature between 150 and 300C parts of the face and body and at times severe shortness of (590 to 860F). Keep container tightly closed and protect from breath requiring immediate medical attention.

light.

WARNING Taking iodide may rarely cause overactivity of the thyroid gland, Potassium iodide should not be used by people allergic to underactivity of the thyroid gland, or enlargement of the thyroid iodide. Keep out of the reach of children. In case of overdose gland (goiter).

or allergic reaction, contact a physician or the public health authority. WHAT TO DO IF SIDE EFFECTS OCCUR DESCRIPTION If the side effects are severe or if you have an allergic reaction, Each THYRO-BLOCK TABLET contains 130 mg of stop taking potassium iodide. Then, if possible, call a doctor or potassium iodide. Other ingredients: magnesium stearate, public health authority for instructions.

microcrystalline cellulose, silica gel, sodium thiosulfate.

HOW POTASSIUM IODIDE WORKS HOW SUPPLIED Certain forms of iodine help your thyroid glands work right. THYRO-BLOCK TABLETS (Potassium Iodide Tablets, USP)

Most people get the iodine they need from foods, iodized salt or bottles of 14tablets (NDC 0037-0472-20). Each white, round, fish. The thyroid can "store" or hold only a certain amount of scored tablet contains 130 mg potassium iodide.

iodine.

WALLACE LABORATORIES In a radiation emergency, radioactive iodine may be released in Division of the air. This material may be breathed or swallowed. it may CARTER-WALLACE, INC.

enter the thyroid gland and damage it. The damage would Cranbury, New Jersey 08512 probably not show itself for years. Children are most likely to have thyroid damage.

PAGE 9 OF 17 REVISION 14

RADIOLOGICAL BROWNS FERRY CONTROL PROCEDURES EPIP-14 NUCLEAR PLANT ATTACHMENT C (Page 1 of 1)

POTASSIUM IODIDE ISSUE REPORT NAME SSN Time of Time of Package Issue Agent Exposure Initial KI Insert

__ Dose Issued 2.

3.

4.

5.

6.

7.

8.

9.

10.

11.

12.

13.

14.

15.

16.

17.

18.

19.

20.

21.

22.

23.

24.

25.

26.

27.

28.

29.

30.

PAGE 10 OF 17 REVISION 14q

RADIOLOGICAL -rnrn BROWNS FERRY CONTROL PROCEDURES ErPIr-14 NUCLEAR PLANT ATTACHMENT D (Page 1 of 3)

[NRC/C]

MANUAL METHOD FOR PROJECTING TOTAL EFFECTIVE DOSE EQUIVALENT (TEDE) FROM STACK AIRBORNE RADIOACTIVITY RELEASES CAUTION: USE THIS ATTACHMENT FOR STACK RELEASES ONLY

1. OBTAIN APPROPRIATE METEOROLOGICAL DATA AND ENTER BELOW:

NOTE: MET DATA CAN BE OBTAINED FROM THE SPDS TERMINALS UNDER SCREEN NAME "METDATA" OR BY ACCESSING THE "MET DATA TERMINAL" LOCATED IN THE TECHNICAL SUPPORT CENTER NOTE: OBTAIN DATA FOR STACK RELEASES AT THE 91 METER INSTR. READINGS.

NOTE: IF ALL MET DATA COLLECTION METHODS ARE UNAVAILABLE CONTACT THE NATIONAL WEATHER SERVICE BY DIALING 9-1-205-621-5650. THE NATIONAL WEATHER SERVICE WILL PROVIDE WIND SPEED AND WIND DIRECTION. THE DEFAULT VALUE FOR STABILITY CLASS WITH NO MET DATA AVAILABLE IS "D" FOR STACK RELEASES.

• STABILITY CLASS:

NOTE: STABILITY CLASS MAY BE ELECTRONICALLY DISPLAYED AS 1,2,3.... THIS CORRESPONDS TO A,BC....

• WIND SPEED: METERS/SECOND To CONVERT MULTIPLY BY To OBTAIN MILES/H 0.45 METERS/SEC METERS/SEC 2.2 MILES/H KNOTS 0.5 METERS/SEC

• WIND DIRECTION

• PLUME DIRECTION To OBTAIN PLUME DIRECTION, ADD 1800 TO WIND DIRECTION IF < 1800 OR SUBTRACT 1800IF WIND DIRECTION IS > 1800.

2. OBTAIN NOBLE GAS RELEASE RATE . MCI/SECOND.

To CONVERT MULTIPLY BY To OBTAIN Cl 1.0X10lCI6 NOTE: This data can be obtained from the SPDS terminals under the screen name "STRW". If SPDS is unavailable, notify operators personnel to gain the information utilizing control room instrumentation, or if unavailable notify the Shift Manager that TI-67, "Determination of Stack and Hardened Wetwell Vent Release Rates" (backup method), procedure must be performed by Radcon and Chemistry personnel.

PAGE 11 OF 17 REVISION 14

RADIOLOGICAL BROWNS FERRY CONTROL PROCEDURES EPIP-14 NUCLEAR PLANT ATTACHMENT D (Page 2 of 3)

3. LOCATE THE "TEDE FACTOR" TABLE THAT CORRESPONDS TO THE APPLICABLE STABILITY CLASS LOCATED ON PAGE 3 OF THIS ATTACHMENT.

4. LOCATE THE COLUMN WITHIN THE TABLE THAT CORRESPONDS TO THE APPLICABLE WIND SPEED.

NOTE: IF WIND SPEED FALLS BETWEEN THE TWO COLUMN VARIABLES CHOOSE THE LOWER VALUE, THIS IS THE MORE CONSERVATIVE VALUE.

5. RECORD THE APPICABLE TEDE FACTORS IN THE CORRESPONDING CALCULATIONS FIELDS.

NOTE: FOR TEDE EXPOSURES REGARDING TYPE II RELEASES, MULTIPLY THE DOSE CALCULATED IN STEP 7 BY 3.2.

NOTE: TYPE I1 RELEASE EXPOSURES SHOULD BE UTILIZED WHENEVER FUEL MELT OR FUEL OVER-TEMPERATURE IS SUSPECTED.

6. RECORD THE RELEASE RATE, OBTAINED IN STEP 2, INTO THE CALCULATION FIELDS.

7. COMPUTE THE TEDE VALUES IN REM/H.

8. COMPLETE ATTACHMENT F, PROJECTED TEDE ASSESSMENT SURVEY FORM, AND FORWARD TO THE SHIFT MANAGER OR THE SITE EMERGENCY DIRECTOR. ENTER TEDE VALUES IN REM/HR ON ARROWS FOR APPLICABLE MILEAGE RINGS. SHOW PLUME DIRECTION ON SURVEY MAP BY THE USE OF AN ARROW.

TEDE DOSE ASSESSMENT CALCULATIONS:

0.62 - 1.99 MILES

( TEDE FACTOR) X ( _LCI/S) = _ TEDE REM/H 2.00 - 4.99 MILES

( TEDE FACTOR) X ( _CI/S) = _ TEDE REM/H 5.00 - 10.00 MILES

( TEDE FACTOR) X ( _CI/S) = _ TEDE REM/H

[NRC/C Inspection Report 81/19]

PAGE 12 OF 17 REVISION 14

RADIOLOGICAL BROWNS FERRY

'--- CONTROL PROCEDURES EPIP-14 NUCLEAR PLANT ATTACHMENT D (Page 3 of 3)

STACK TEDE FACTOR (Rem/h per ,uCi/s) TABLES Stability A Wind Speed miles 1 mn/s_ 2 m/s 3 m/s 4 mIs Sm/s 6 m/s 7 m/s 8 rn/s 9 m/s 10im/s 0.62-1.99 2.9E-l 0 1.5E-10 1.2E-10 9.1 E-11 6.1 E-11 5.5E-1 1 4.9E-1 1 4.2E-1 1 3.6E-1 1 3.0E-1 1 2.00-4.99 1.3E-10_ 7.2E-1 1 5.8E-1 1 4.4E-1 1 3.1 E-1 1 2.8E-1 1 .

2.5SE-1 1 2.2E-1 1 1.9E-11 1.6E-11 5.00-10.00 6.3E-1 1_ 3.4E-1 I 2.9E-1 1 2.3E-1 1 1.8E- 1I 1.6E-11 1.4E-11 1.3E-11_ 1.1 E-11 9.2E-12 Stability B Wind Speed i_

miles 1 mn/sl 2 m/s 3m/s 4m/s 5 m/s 63m/s 7 m/s 8 m/s 9 m/s 10 m/s 0.62-1.99 4.2E-1 0_ 2.4E-10C 1.9E-1C 1.4E-10 9.6E-11 8.6E-1 11 5.7E-11 4.7E-11 2.00-4.99 5.00-10.00 1.7E-10 8.2E-1 1_

8.4E-1 1 4.6E-1 1 6.8E-1 1 3.7E-1 1 5.1 E-1 1 2.8E-1 I 3.5E-1 1 2.OE-11 1

_ :_ 2.1 E-11 1.2E-11 1.8E-11 1.OE-11 Stability C Wind Speed miles 1 mn/s 2 m/s 3 m/s 4 m/s 5 m/s 6 ins 7 in/sl 8 in/sl 9 m/s 10 m/s I 0.62-1.99 2.00-4.99 2.5E-1 0_

2.2E-1 0_

1.6E-10I 1.4E-101 1.3E-1C 1.2E-10 1.OE-10 8.9E-11 7.OE-11 6.2E-11 6.3E-1 1 5.6E-1 1 S.6E-11 l 5.0E-1 1l 4.9E-1 1l 4.4E-1 1l t 4.2E-1 1 3.8E-11l 3.5E-11 3.2E-11 5.00-10.00 1.OE-10_ 5.8E-1 1l 4.8E-11 3.7E-1 1 2.7E-1 1 2.4E-1 1 2.2E-1 1 1.9E-11 1.6E-11 1.4E-11 Stability D Wind Speed 1 m/s 2 m/s 3 m/s 4 m/s Sm/s 6mis T l 7 m/s 8m/s T- 9 m/s r10 m/s I E

4.4E-1 1 7.4E-1 11 8.8E-1 1 3.5E-1 l 6.1 E-1 1 7.3E-111 2.7E-11 4.8E-1 11 5.7E-1 1 1.8E-1 3.6E-1 1 4.2E-1I 1.6E-111 3.2E-11 l 3.8E-11 1.3E-11 _ 1.1 E-11I 2.6E-11 2.3E-11I 3.0E-117 _ 772.6E-1 1 9,2E-12 2,0E-11I 2.2E-1 1 Stability E Wind Speed miles I 1 m/s _ 2 mIs 3 mIs 4 mIs 5 mIs 6 i/s 7 rn/sj 8 in/s 9 m/s 10 m/s 0.62-1.99 8.5E-1 1 4.4E-1 1 3.5E-1 1 2.7E-1 I 1.8E-1 1 1.6E-11 1.4E-11j 1.3E-11_ 1.1E-11 9.1E-12 3.1 E-1 I 2.00-4.99 5.00-10.00 6.3E-1 1 1.8E-10 3.8E-1 1 1.OE-1 C 8.4E-11 2.4E-1 1 6.5E-1 1 1.7E-1 1 4.7E-11 1.SE-11 4.2E-l11

_3.8E-11I :_

1.4E-11j 1.2E-11 3.4E-l11

_ 1.1E-11 2.9E-11 9.1 E-12 2.5E-11 Stability F Wind Speed miles 1 m/s_ 2 m/s l 3 m/s 4 m/s l 5m/s 6 m/s 7 mi/s 8 rn/s 9 m/s 10 m/s 0.62-1.99 8.4E-l11 4.4E-1 1 3.5E-1 1 2.7E-1 I 1.8E-1 1 1.6E-11 1.4E-1 1 1.3E-11 1.1 E-11 9.OE-12 2.00-4.99 6.7E-l11 3.8E-1 1 3.1 E-1 1 2.3E-1 I 1.6E-1 _ 1.4E-11 :_ :_ 1.1 E-11 9.7E-12 8.1 E-12 5.00-10.00 9.9E-l11 6.1 E-11 5.1 E-11 4.0E-11 2.9E-1i1 2.7E-l11 2.1E-11 1.9E-1 1 1.6E-1 1 Stability G Wind Speed 1 mn/s 2 m/s 3 m/s 4 m/s miles 5 m/s 6mi/sl 7 in/s 8 in/sl 9 m/s 10 m/s 0.62-1.99 8.SE-i11 4.4E-1 i 3.5E-1 l 2.7E-1 1 1,8E-1 1 1.6E-11i 1.4E-11i i.3E-1i1 1.1 E-11 9.OE-12 2.00499 6.7E-i 3.8E-1ll 3.1E-11 l 2.3E-11 1.6E-11 l Eli 1.4E-11 1.3E-11i 9.7E-12 8.1 E-12 5.00-10.00 7.2E-i11 4.5E-11i 3.6E-11i 2.8E-1 1 1.9E-1 1 1.7E-11i 1.5E-il1 1.4E-1i 1.2E-1 1 1.OE-1 1 PAGE 13 OF 17 REVISION 14

RADIOLOGICAL rTTfllA BROWNS FERRY CONTROL PROCEDURES EPIP-1 NUCLEAR PLANT ATTACHMENT E (Page 1 of 3)

[NRC/C]

MANUAL METHOD FOR PROJECTING TOTAL EFFECTIVE DOSE EQUIVALENT (TEDE) FROM GROUND LEVEL OR BUILDING AIRBORNE RADIOACTIVITY RELEASES CAUTION: USE THIS ATTACHMENT FOR GROUND OR BUILDING RELEASES ONLY

1. OBTAIN APPROPRIATE METEOROLOGICAL DATA AND ENTER BELOW:

NOTE: MET DATA CAN BE OBTAINED FROM THE SPDS TERMINALS UNDER SCREEN NAME "METDATA' OR BY ACCESSING THE "MET DATA TERMINAL" LOCATED INTHE TECHNICAL SUPPORT CENTER.

NOTE: IF ALL MET DATA COLLECTION METHODS ARE UNAVAILABLE CONTACT THE NATIONAL WEATHER SERVICE BY DIALING 9-1-205-621-5650. THE NATIONAL WEATHER SERVICE WILL PROVIDE WIND SPEED AND WIND DIRECTION. THE DEFAULT VALUE FOR STABILITY CLASS WITH NO MET DATA AVAILABLE IS "E" FOR GROUND OR BUILDING RELEASES.

NOTE: OBTAIN DATA FOR GROUND OR BUILDING RELEASES AT THE 46 METER INSTR.

READINGS.

• STABILITY CLASS:

NOTE: STABILITY CLASS MAY BE ELECTRONICALLY DISPLAYED AS 1,2,3.... THIS CORRESPONDS TO A,B,C....

• WIND SPEED: METERS/SECOND To CONVERT MULTIPLY BY TO OBTAIN MILES/H 0.45 METERSISEC METERS/SEC 2.2 MILES/H KNOTS 0.5 METERS/SEC

• WIND DIRECTION

• PLUME DIRECTION To OBTAIN PLUME DIRECTION, ADD 1800 TO WIND DIRECTION IF< 180 OR SUBTRACT 1800 1F WIND DIRECTION IS > 1800.

2. OBTAIN NOBLE GAS RELEASE RATE __ CI/SECOND.

NOTE: Notify operators personnel to obtain release rate information utilizing control room instrumentation.

To CONVERT MULTIPLY BY To OBTAIN Ci 1.0X10 6 ItCI PAGE 14 OF 17 REVISION 14

RADIOLOGICAL BROWNS FERRY

-_ CONTROL PROCEDURES EPIP-14 NUCLEAR PLANT ATTACHMENT E (Page 2 of 3)

3. LOCATE THE "TEDE FACTOR" TABLE THAT CORRESPONDS TO THE APPLICABLE STABILITY CLASS LOCATED ON PAGE 3 OF THIS ATTACHMENT.

4. LOCATE THE COLUMN WITHIN THE TABLE THAT CORRESPONDS TO THE APPLICABLE WIND SPEED.

NOTE: IF WIND SPEED FALLS BETWEEN THE TWO COLUMN VARIABLES CHOOSE THE LOWER VALUE, THIS IS THE MORE CONSERVATIVE VALUE.

5. RECORD THE APPICABLE TEDE FACTORS IN THE CORRESPONDING CALCULATION FIELDS NOTE: FOR TEDE EXPOSURES REGARDING TYPE II RELEASES, MULTIPLY THE DOSE RATES CALCULATED IN STEP 7 BY 3.2.

NOTE: TYPE II RELEASE EXPOSURES SHOULD BE UTILIZED WHENEVER FUEL MELT OR FUEL OVER-TEMPERATURE IS SUSPECTED.

6. RECORD THE RELEASE RATE, OBTAINED IN STEP 2, INTO THE CALCULATION FIELDS.

7. COMPUTE THE TEDE VALUES IN REM/H.

8. COMPLETE ATTACHMENT F, PROJECTED TEDE ASSESSMENT SURVEY FORM, AND FORWARD TO THE SHIFT MANAGER OR THE SITE EMERGENCY DIRECTOR. ENTER TEDE VALUES IN REM/H ON ARROWS FOR APPLICABLE MILEAGE RINGS. SHOW PLUME DIRECTION ON SURVEY MAP BY THE USE OF AN ARROW.

TEDE DOSE ASSESSMENT CALCULATIONS:

0.62 - 1.99 MILES

( TEDE FACTOR) X ( _ _CI/S) = _ TEDE REM/H 2.00 - 4.99 MILES

( TEDE FACTOR) X ( _ _ CI/S) = _ TEDE REM/H 5.00 - 10.00 MILES

( TEDE FACTOR) X ( ECI/S) = _ TEDE REM/H

[NRC/C Inspection Report 81/19]

PAGE 15 OF 17 REVISION 14

RADIOLOGICAL BROWNS FERRY CONTROL PROCEDURES EPIP-14 NUCLEAR PLANT ATTACHMENT E (Page 3 of 3)

GROUND or BUILDING TEDE FACTOR (Remlh per p.Cils) TABLES Stability A Wind Speed miles 1 m/s 2 MIs 3 mIs 4 m/s 5ms 6ms 7 mIs 8 mIs 9 mIs 10 mi/s 0.62-1.99 5.1 E-10 2.5E-10 2.0E-10 1.5E-1 1.OE-10 9.1 E-11 8.1 E-11 7.1 E-11 6.1 E-11 5.1 E-11 2.00-4.99 1.7E-1 0 8.7E-1 1 7.OE-1 1 5.2E-1 1 3.5E-1 1 3.1 E-1 1 2.8E-1 1 2.4E-1 1 2.1 E-11 1.7E-11 5.00-10.00 6.3E-11 3.4E-11 2.8E-11 2.2E-11 1.6E-11 1.4E-11 1.3E-11 1.1 E-11 9.5E-12 7.9E-12 Stability B Wind Speed miles 1 m/s 2 mIs 3 mIs 4 m/s 5 mIs 6 mIs 7 mIs 8 m/s 9 mIs 10 mi/s 0.62-1.99 2.3E-09 1.2E-09 9.4E-1 0 7.1 E-1 0 4.7E-1 0 4.3E-1 0 3.8E-10 3.3E-1 0 2.8E-1 0 2.3E-1 0 2.00-4.99 2.3E-10 1.2E-10 9.2E-11 6.8E-11 4.SE-11 4.1 E-11 3.6E-11 3.2E-11 2.7E-11 2.3E-11 5.00-10.00 8.2E-11 4.6E-11 3.7E-11 2.9E-11 2.1 E-11 1.9E-11 1.7E-11 1.5E-11 1.2E-11 1.OE-11 Stability C Wind Speed miles 1 m/s 2 mIs 3 m/s 4 mIs 5 mIs 6 mIs 7 m/s S/s l9m/s 10 m/s n

0.62-1.99 6.8E-09 3.5E-09 2.8E-09 2.1 E-09 1.4E-09 1.2E-09 1.1E-09 9.6E-10 8.2E-10 6.8E-10l 2.004.99 9.3E-1 0 4.5E-1 0 3.7E-1 0 2.8E-1 0 1.9E-10 1.7E-1 0 1.5E-1 0 1.3E-10l 1.10E-10 9.3E-1 1 5.00-10.00 1.6E-1 0 9.1 E-1 1 7.SE-1 1 5.8E-1 1 4.2E-1 1 3.7E-1 1 3.3E-11 2.9E-11 2.SE-1 1 2.1 E-1 1 Stability D Wind Speed miles 1 mI/s 2 m/s 3 m/s 4m/s 5 m/s 6 m/s 7 m/s 8 m/s 9 m/s 10 m/s 0.62-1.99 2.0E-08 1.OE-08 8.OE-09 6.OE-09l 4.1 E-09l 3.6E-09l 3.2E-09l 2.8E-09l 2.4E-09 2.OE-09 2.00-4.99 3.3E-09 1.7E-09 1.4E-09 1.OE-09j 6.8E-101 6.2E-101 5.5E-1 0 4.8E-101 4.1E-10 3.SE-10 5.00-10.00 6.9E-10 3.9E-10 3.2E-10 2.SE-10l 1.8E-10l 1.6E-10l 1.4E-10l 1.2E-10l 1.0E-10 8.7E-11 Stability E Wind Speed l miles 1 mi/s 2 m/s 3 m/s 4m/s 5 m/s 6 m/s l7m/s l m/s 9m/s /s l0.62-1.99 3.SE-08 1.7E-08 1.4E-08 1.0E-08 7.OE-09 6.3E-09l 5.6E-09 4.9E-09 4.2E-09l 3.5E-09 l2.00-4.99 6.6E-09 3.3E-09 2.7E-09 2.OE-09 1.3E-09 1.2E-09l 1.1 E-09 9.3E-10 7.9E-10l 6.6E-10 l.00-10.00 1.SE-09 8.2E-10 6.7E-10 5.3E-10 3.8E-10 3.4E-10 3.1 E-10 2.7E-10 2.3E-10j 1.9E-10 Stability F Wind Speed miles 1 rm/s ] 2m/s [ 3m/s 4m/s m/s 6m/sI 7rm/s Sm/s 9m/s 10mi/s 0.62-1.99 6.6E-08l 3.3E-08I 2.7E-081 2.OE-08 1.3E-08 1.2E-08 1.1 E-08 9.3E-09 8.OE-09 6.6E-09 2.00-4.99 1.5E-08l 7.6E-09l 6.1 E-09l 4.6E-09 3.1 E-09 2.8E-09 2.5E-09 2.2E-09 1.8E-09 1.5E-09 5.00-10.00 3.8E-09l 2.1E-09l 1.7E-091 1.3E-09 9.6E-10 8.6E-10 7.7E-10 6.7E-10 5.7E-10 4.7E-10 Stability G Wind Speed miles 1 mr/s 2m/s 3m/s 4m/s l Sm/s 6m/s 7m/s l Sm/s l9msl 10ms 0.62-1.99 1 .5E-07 7.2E-08 S.7E-08 4.3E-08 2.8E-08 2.5E-08 2.3E-08l 2.0E-08 1 0l 1.4E-08l 2.00-4.99 3.7E-08 1.6E-08l 1.3E-08 9.5E-09j 6.3E-09 5.6E-09 S.OE-09l 4.4E-09 3.8E-09l E-O9 5.00-10.00 9.4E-09 5.OE-09 4.1 E-09 3.2E-09l 2.3E-09 2.0E-09 1.8E-09l 1.SE-09 1.3E-09 1.1E-9 PAGE 16 OF 17 REVISION 14

RADIOLOGICAL BROWNS FERRY CONTROL PROCEDURES EPIP-14 NUCLEAR PLANT ATTACHMENT F (Page 1 of 1)

PROJECTED TEDE ASSESSMENT SURVEY FORM EZ STACK RELEASE EGROUND / BUILDING RELEASE TIME OF ASSESSMENT RELEASE RATE __C_/S WIND SPEED Ml I/H STABILITY CLASS WIND DIRECTION PLUME DIRECTION PREPARED BY DATE_ _ _

00 1800 DRAWING NOT To SCALE LAST PAGE PAGE 17 OF 17 REVISION 14