ML030920544

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Emergency Plan Implementing Procedures Revisions
ML030920544
Person / Time
Site: Browns Ferry  Tennessee Valley Authority icon.png
Issue date: 03/20/2003
From: Abney T
Tennessee Valley Authority
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML030920544 (52)
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Text

Tennessee Valley Authority, Post Office Box 2000, Decatur, Alabama 35609-2000 March 20, 2003 10 CFR Part 50, App E U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop OWFN, P1-35 Washington, D.C. 20555-0001 Gentlemen:

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.

Specifically, portions of EPIP-1 were revised, namely, Table of Contents, Revision 34; Section II-1.0, Revision 31; Section II-3.0, Revision 30; Section III-1.0, Revision 31; and Section III-3.0, Revision 30. The revisions have an effective date of March 10, 2003.

The enclosed information is being sent by certified mail.

The signed receipt signifies that you have received this information. If you have any questions, please telephone me at (256) 729-2636.

4 . Abney D nager of Licensin

~ and Industry Aff irs Ao 5 Pr-ted -,n rec-c-eyd paper

U.S. Nuclear Regulatory Commission Page 2 March 20, 2003 cc (Enclosure):

NRC Resident Inspector (Enclosure provided by Browns Ferry Nuclear Plant BFN Document Control Unit) 10833 Shaw Road Athens, Alabama 35611-6970 Mr. Stephen J. Cahill, Branch Chief (2 Enclosures)

U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street S.W., Suite 23T85 Atlanta, Georgia 30303-8931 Mr. Kahtan N. Jabbour, Senior Project Manager (w/o Enclosure)

U.S. Nuclear Regulatory Commission One White Flint, North (MS 08G9)

Office of Nuclear Reactor Regulation 11555 Rockville Pike Rockville, Maryland 20852-2738

ENCLOSURE TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, AND 3 EMERGENCY PLAN IMPLEMENTING PROCEDURE (EPIP) REVISIONS EPIP 1 TOC, SECTIONS II-1.0, II-3.0, III-1.0, AND III-3.0 SEE ATTACHED

GENERAL REVISIONS FILING INSTRUCTIONS FILE DOCUMENTS AS FOLLOWS:

PAGES TO BE REMOVED PAGES TO BE INSERTED TOC, Revision 33 TOC, Revision 34 Section II-1.0, Revision 30 Section II-1.0, Revision 31 Section II-3.0, Revision 29 Section II-3.0, Revision 30 Section III-1.0, Revision 30 Section III-1.0, Revision 31 Section III-3.0, Revision 29 Section III-3.0, Revision 30

TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURE EPIP-I EMERGENCY CLASSIFICATION PROCEDURE REVISION 34 PREPARED BY. T W. CORNELIUS PHONE 2038 RESPONSIBLE ORGANIZATION EMERGENCY PREPAREDNESS APPROVED BY: JEFF LEWIS DATE 03/05/2003 EFFECTIVE DATE 03/10/2003 LEVEL OF USE: REFERENCE USE QUALITY-RELATED

REVISION LOG Procedure Number: EPIP-1 Revision Number 34 Pages Affected 1,14,15,17,30,84.86,89,118,120 Description of Change.

IC - 42 EPIP 1, rev. 31 revision is being conducted to change the Site Boundary Radiation Reading from a beta-gamma value to gamma only value. This change does not involve the numerical value. This revision is in compliance with the REP and doesn't affect the BFN EP standard emergency classification and action level scheme This revision is being conducted to ensure consistency NN ith NUMARC/NESP-007. Reg Guide 1 101, and NEI 99-01 (Rev. 4).

IC - 43 EPIP 1, rev. 32 is being conducted to modify infornation that support EAL 1. I-GI,

1. I-G2, and 1.2-G The revision incorporates changes resulting from modifications to calculations that support Minimum Alternate RPV Flooding Pressures (MARFP) and Minimum Steam Cooling Reactor Water Level (MSCRWL) Revisions to these calculations wvere conducted in support of the EOI Program Manual Revision 21 (U3C1 1).

IC-44 EPIP 1. rev. 33 is being issued to modify EAL 6 7-U. The revision incorporates changes resulting from the letter from NEI to NRC (to Mr Bruce A. Boger) dated December 18, 2001 requesting confirmation for EAL basis change to include response to a Site -Specific Credible Threat. Tils teas developed in response to NRC's October 6. 2001 Safeguards Advisor. This is additional information and does not change existing criteria in the EAL Basis.

IC 45 EPIP 1, rev. 34 is being conducted to modify infornation that support EAL 1.1-GI, 1.I-G2, and 3 1-S. The revision incorporates changes resulting from modifications to calculations that support Minimum Alternate RPV Flooding Pressures (MARFP).

Minimum Steam Cooling Reactor Water Level (MSCRWL) and Maximum Safe Operating Area Temperature Limits. Revisions to these calculations *wereconducted in support of the EOI Program Manual Re, ision 22 (U2C 13) 2

EPIP-1 EMERGENCY CLASSIFICATION PROCEDURE TABLE OF CONTENTS PAGE NUMBER REVISION TABLE OF CONTENTS ........ ..................... . ..... 1 34 SECTION I INTRODUCTION CLASSIFICATION INSTRUCTIONS ...... 3 29 GLOSSARY... ........ . .5 29 EVENT CLASSIFICATION INDEX . ........ 11 29 SECTION 11 EVENT CLASSIFICATION MATRIX 1.0 REACTOR ......................... ..... ............ 13 31 2 0 PRIMARY CONTAINMENT .... ............. .21 28 3 0 SECONDARY CONTAINMENT... . ........... .. .. 29 30 4 0 RADIOACTIVITY RELEASES ... ...... ........ 33 30 5 0 LOSS OF POWER . . 39 29 6 0 HAZARDS . .. . 45 30 7 0 NATURAL EVENTS ..... .

. .................... ...... . 616. 28 8 0 EMERGENCY DIRECTOR JUDGEMENT ....... . ............ 67 29 SECTION III BASIS 1 0 REACTOR ... . . .. .... 75 31 2 0 PRIMARY CONTAINMENT .............. .... . ..... 97 28 3 0 SECONDARY CONTAINMENT ....................... . ..... 116 30 4 0 RADIOACTIVITY RELEASE ............................. . ...... 126 30 5.0 LOSS OF POWER ................. ......................... ................ 139 29 6.0 HAZARDS ................. .......... ................... ... 5. 55 30 7.0 NATURAL EVENTS ................ ..... ..... ............... 183 28 8 0 EMERGENCY DIRECTOR JUDGEMENT ....... .. .. 190 29 PAGE 1 OF 207 REVISION 34l

EPIP-1 EMERGENCY CLASSIFICATION PROCEDURE THIS PAGE INTENTIONALLY BLANK

<-I PAGE 2 OF 207 REVISION 34

EMERGENCY EPIP-1 CLASSIFICATION SECTION II DP(pTnnllRpT EVENT CLASSIFICATION MATRIX 1.0 REACTOR REACTOR 10 1.0 REACTOR PAGE 13 OF 207 REVISION 31

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

I I-UI/1.1-Al Applicable when the Reactor Head is removed and the Reactor Cavity is flooded I I-i5 Applicable in Mode 5 when the Reactor Head is installed.

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

  • All control rods except one are inserted to or beyond position 00
  • Determined by reactor engineering CURVES/TABLES:

TABLE I.1 - C2.

vMINIMUM ALTERNATE RPVFLOODING PRESS NUMBER OF OPEN MSRVs MARFP (PSIG) 6 or More 190 5 230 4 290 REVISION 31 PAGE 141 OF 207 1.0 REACTOR I

EMERGENCY EPIP-1 CLASSIFICATION SECTION II PROCEDURE EVENT CLASSIFICATION MATRIX 1.0 REACTOR i3DI BT DESCRIPTION DESCRIPTION

& 9*-

1.1-Ul 1.1-U2 C

Uncontrolled water level decrease in Reactor Cavity Uncontrolled water level decrease in Spent Fuel Pool Ca4 with irradiated fuel assemblies expected to remain with irradiated fuel assemblies expected to remain covered by water covered by water C

3 OPERATING CONDITION OPERATING CONDITION

- Mode 5 -All 1.1-Al IN -A2 Uncontrolled water level decrease in Reactor Cavity Uncontrolled water level decrease in Spent Fuel expected to result in irradiated Fuel assemblies being Storage Pool expected to result in irradiated fuel uncovered assemblies being uncovered OPERATING CONDITION OPERATING CONDITION.

- Mode 5 - All 1.1-SI 1.1-52 Reactor water level CANNOT be maintained above Reactor water level CANNOT be determined M

-162IN (TAF)

OPERATING CONDITION OPERATING CONDITION

- All - Mode I - Mode 3

-Mode 2 1.1-G2 C Reactor water level CANNOT be determined Reactor ,vater level CANNOT be restored and AND maintained above -185 IN EITHER of the following conditions exists:

  • The reactor wfl issnain 9bxrtical w/o bonunder all caridions and less than 4 N1SRVs can be opten orReactorpreure CANNOT be restored and mairtained at least 65 PSI above Sufrfssion Chamber
  • It has NOT been determined that the reactor will rem.Lin subentical wlo boron under all conditions and unable to retoe umdmsurtunMARFP in Table I l-G2 OPERATING CONDITION OPERATING CONDmON

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

- Mode 2 - Mode 2 1.0 REACTOR PAGE 15 OF 207 REVISION 31 l

EPIP-1 EMERGENCY SECTION II CLASSIFICATION 1.0 REACTOR EVENT CLASSIFICATION MATRLX PROCEDURE' NOTES:

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

CURVES/TABLES:

CURVE 1.2-G HEAT CAPACITY TEMP LIMIT 260

..I "... I

...... .I.

. ,. ,. .x _."' f,. ..........................................

ACTION REQUIREDIF CURVE FR EX I.IN R S 31 U-1 21 0 24RPV l ...

.R.

... .~l:

ABOVE P

Press.90 Prss 300E 16 OF 2 g g m 1................0 2504EZ , v. _ Z .iy .SAFE WE5iRS.

200 RPV Press. 500e . eel,,:,::  : ":02liS CLC190 X =:::::s;:eSg 160 *e.<- Re V Pe + i-ss. ............... ... f 15 . s .....

' - l  ; s AFl e ' _- . .

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 IF ABOVE CURVE FOR EXISTING RX REVISION 31 PAGE 16 OF 207 1.0 REACTOR I

EMERGENCY EPIP-1 CLASSIFICATI ON SECTION 11 PROCEDUIfRlE EVENT CLASSIFICATION MATRIX 1.0 REACTOR DESCRIPTION DESCRIPTION 1.3-U C

Reactor coolant activity exceeds 26 pCi/gm dose C,'

equivalent 1-1 31 (Techmcal Specification Limit) as determined by chemistry sample.

cn OPERATING CONDmON

- All 1.2-A 1.3-A Failure of automatic scram functions to bring the Reactor coolant activity exceeds 300 pCi/gm dose Reactor subcntical equivalent Iodine-I 31 as determined by chemistry AND sample Manual scram or ARI was successful OPERATING CONDITION OPERATING CONDITION

- Mode I -Mode I -Mode 3

- Mode 2 -Mode 2 1.2-S ?lc Failure of automatic scram, manual scram, and ARI to bnng the Reactor subcntical M OPERATING CONDITION

- Mode I 1.2-G 0 Failure of automatic scram, manual scram, and ARI. Z Reactor power >3%

AND EITHER of the following conditions exists

  • Suppression Pool temp exceeds HCTL Refer to Curve I 2-G
  • Reactor water level CANNOT be restored and maintained at or above -185 IN Z

OPERATING CONDITION

- Mode l

- Mode 2 1.0 REACTOR PAGE 17 OF 207 REVISION 31 l

EPIP-1 EMERGENCY SECTION II CLASSIFICATION 1.0 REACTOR EVENT CLASSIFICATION MATRIX PROCEDURE CURVES/TABLES:

CURVE 1.5-S HEAT CAPACITY TEMP LIMIT 260 250- l i, SAFE WHEN RX PRESS =

24  : , - R.i$ -. $ $ g~g : > .$' I S BE LOW 6 5 PSI G .

240 $:g.;.^:

  • .$..$' . .B ..s B

......... S B 0$. . ...... . ..

X ;NI--V: ~~~~. ;.::......' .. . : ..:'::

sX*S;:.X>4....;:

210 PVPess..00.

200 RPV Press. 50O.

170o - RPV Press. 700 .....

co180 -. P Press. 900^a*# EBtR 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 IF ABOVE CURVE FOR EXISTING RX REVISION 31 PAGE 18 OF 207 1.0 REACTOR

EMERGENCY EPIP-I CLASSIFICATION SECTION II I f)lu Tb IT F.VFNT Cl ASSIFICATION MATRIX 1.0 REACTOR r1t U* A J 'A l A A *

  • J* A_

A,,_

___p =I DESCRIPTION DESCRIEPTION 1.4-U Z Valid MAIN STEAM LINE RADIATlON HIGH-HIGH alarm, RA-90-135C OR Valid OG PRETREATMENT RADIATION HIGH alarm, RA-90-1 57A OPERATING CONDITION

-Mode -Mode 3

- Mode 2 1.5A Reactor moderator temperature CANNOT be maintained below 212' F whenever Technical Specifications require Mode 4 conditions or during operations in Mode 5 OPERATING CONDITION

-Mode4

- Mode 5 1.5-S ,)

Suppression Pool temperature, level and RPV pressure CANNOT be maintained in the safe area of Curve 15-S OPERATING CONDITION

-Mode l -Mode3

- Mode 2 0

1.0 REACTOR PAGE 19 OF 207 REVISION 31 l

EPIP-1 EMERGENCY SECTION II CLASSIFICATION 1.0 REACTOR EVENT CLASSIFICATION MATRIX PROCEDURE THIS PAGE INTENTIONALLY BLANK REVISION 31 PAGE 20 OF 207 1.0 REACTOR

EPIP-1 EMERGENCY 3.0 SECONDARY CLASSIFICATION SECTION II CONTAINMENT lTr'tTAT f'1 AccTtCAT1ON MATRIX PROCEDURE JV IJ-L A ---- -

SEC ONDARY C ONTAINMENT

--- 3. 0 REVISION 30 3.0 SECONDARY PAGE 29 OF 207 CONTAINMENT

Y, EPIP-1 EMERGENCY SECTION II CLASSIFICATION 3.0 SECONDARY PROCEDURE r'fThT' A TWMfl.NrT FIVINT rLASSIFICATION MATRIX

%-V a ll ki1 s1X1A1 -: VN LSIIAINMT RCDR NOTES:

CURVES/TABLES:

TABLE 3.1 MAXIMUM SAFE OPERATING AREA TEMPERATURE LIMITS APPLICABLE PANEL 9-21 AREA TEMPERATURE ELEMENTS MALX SAFE OPERATING VALUE OF (UNLESS OTHERWISE NOTED) UNIT 2 UNIT 3 74-95A 150 155 RHR A'C PUNIP ROOM 74-95B 210 215 RHR B,'D PUMP ROOM 73-55A 270 270 HPCI TURBINE AREA 71-41A 190 190 RCIC TURBINE XREA (XA-55-3E-29) PANEL 9-3 TI-75-69B 150 150 CSPUNIP ROOM HIGH HUMIDITY OR TENIP HIGHI 71-41B. 41C. 41D 200 250 RCIC STEAM SUPPLY AREA 73-55B.55C. 55D 240 240 HPCI STEAM-ISUPPLY AREA 74-9511 240 240 RHR AC PUMP SUP'LY AREA 74-95G 240 240 RHR B'D PUNIP SUPPLY AREA (XA-55-3D-24) PANEL 9-3 TIS-1-60A 315 315 MAIN STEAM LINE LEAK DETECTION HIGH 74-95E 170 175 RIIR VALVE ROOM (XA-55-5B-32133) PANEL 9-5 170 175 RWCU ISOL LOGIC CHANNEL A'BTENIP HIGH 69-835A.B. C. D AUX INST ROOM _

69-29F 220 220 RWVCU OUTBD ISOL VLV AREA 69-29G 220 220 RWCU HXAREA 69-29H 220 220 RWCU HX EXH DUCT 69-29D 215 215 RWCU RECIRC PUNIP A AREA 69-29E 215 215 RWCU RECIRC PUNIP B AREA 74-95C -195 200 RHR A'C IHXROOMI RHR BRDILX ROOM 74-95D 195 200 74-95F 155 155 FPC HX AREA TABLE 3.2 MAXIMUM SAFE OPERATING AREA RADIATION LIMITS AREA RAD MONITOR MLAX SAFE VALUE MIR/IIR 90-25A 1000 RHR WEST ROOM 90-28A 1000 RHR EAST ROOM 90-24A 1000 HPCI ROOM 90-26A 1000 CS/RCIC ROOM 90-27A 1000 CORE SPRAY ROOM 90-29A 1000 SUPPR POOL AREA 90-20A 1000 CRD-HCU WEST AREA 90-2 1A 1000 CRD-IICU EAST AREA 90-23A 1000 TIP DRIVE AREA 90-13A 1000 NORTH RWCU SYSTENI .REA 90-14A 1000 SOUTH RWCU SYSTEM AREA 90-9A 1000 RWCU SYSTEM AREA 90-4A 1000 MG SET AREA 90-IA 1000 FUEL POOL AREA 90-2A 1000 SERVICE FLR AREA 90-IA 1000 NEW FUEL STORAGE DRY\\ ELL RADIATION UNIT 2 2-RE-90-272A > 345 RiHR 2-RE-90-273A > 164 R,'HR REACTOR COOLANT ACTIVITY > 300 ptCIgm DOSE EQUILAVENT IODINE-131 PA( ;E 30 OF 207 3.0 SECONDARY REVISION 30 CONTAINMENT

EMERGENCY EPIP-1 SECTION 11 3.0 SECONDARY CLASSIFICATION CONTAINMENT PROCEDURE EVENT CLASSIFICATION MATRIX

)

nr~rRTPTwfN DESCRIPTION JJ ]:J \, J .

3.2-A An ot the to owming high radiation alarms on Panel 9-3

  • RA-90-lA, Fuel Pool Floor Area
  • RA-90-250A, Reactor, Turbine, Refuel Exhaust
  • RA-90-142A, Reactor Zone Exhaust
  • RA-90-1 40A, Refueling Zone Exhaust AND Confirmation by Refuel Floor personnel that irradiated fuel damage may have occurred OPERATING CONDmON

-All Cr I 3.1-S PTl tr iI An unisolable Primary System leak is discharging into ty I An unisolable Pnmar% SN stem leak is discharging into 4 Secondar) Containment Secondary Containment tT .4 AND AND :X Any area radiation level at or above the Maximum 17 Ann area temperature exceeds the Maximum Safe -1 Safe Operating area Radiation limit listed in Table 3 2 t2 Operating Temperature limit listed in Table 3 1 0

(I OPERATING CONDITION 0 I OPERATING CONDmON - Mode 3

-Mode 3 - Mode I

-Mode I - Mode 2 _ -

-Mode 2 .

3.2-G \N~t I

0 An unisolable Primary System leak is discharging into An unisolable Pnmar%Sy stem leak is discharging into I

Secondary Containment Secondary Containment AND C,-4 AND An) area radiation level at or above the Maximum t4 AnN' area temperature exceeds the Maximum Safe 1 Safe Operating area Radiation limit listed in Table 3 2 ;0 Operating Temperature limit listed in Table 3 1 AND ik AND I Any indication of potential or significant fuel failure Any indication of potential or significant fuel failure exists Refer to Table 3 1-G/3 2-G exists Refer to Table 3 l-G/3 2-G El 4

OPERATING CONDmON 1 OPERATING CONDITION - Mode 3

-Mode 3 -ModeI 10

-Mode I I 2

- Mode

-Mode 2 An:

REVISION iO

.0 SECONDARY PAGE 31 OF 207 CONTAINMENT

EMERGENCY EPIP-1 SECTION II 3.0 SECONDARY CLASSIFICATION CONTAINMENT VxTvvMT NT rAr&qmATIN MATRIX FRKUUEDUK LT -as -^^..-.. _

THIS PAGE INTENTIONALLY BLANK PAGE 32 OF 207 3.0 SECONDARY REVISION 30 CONTAINMENT

EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 1.0 REACTOR TECHNICAL BASIS 1.0 REACTOR REACTOR 10 1.0 REACTOR PAGE 75 OF 207 REVISION 31 l

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 REVI1SION 31

EMERGENCY EPIP-1 CLASSIFICATION SECTION m PROCEDURE TECHNICAL BASIS 1.0 REACTOR TECHNICAL BASIS 1.0 REACTOR LU Hri pm U A ri mmgi UNUSUAL EVENT (CONTINUED)

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

Technical Specifications 3 5 2 NOTES NOTE I I-UI/I 1-Al Applicable when the Reactor Head is removed and the Reactor Cavity is flooded 1.0 REACTOR PAGE 77 OF 207 REVISION 31 l

EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 1.0 REACTOR iYaui maIi flU PhYA Ii - m ihin UNUSUAL EV7ENT 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. I -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 31 l

EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL 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 31 l

EMERGENCY EPIEP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 1.0 REACTOR WDI_

ALERT (CONTINUED)

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

Technical Specifications 3.5 2 NOTES NOTE l.1-UI/1.1-AI Applicable when the Reactor Head is removed and the Reactor Cavity is flooded.

1.0 REACTOR PAGE 80 OF 207 REVISION 31 l

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, (NUMARC-AA2 example-4) 1.0 REACTOR PAGE 81 OF 207 REVISION 31 l

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

- E0I 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 31

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

A SITE AREA EMERGENCYS 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, (NUMARC-FS)

EOI Program Manual Section VI-J 1.0 REACTOR PAGE 83 OF 207 REVISION 31 l

EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 1.0 REACTOR GENERAL EMERGENCY Reactor water level CANNOT be restored and maintained above -185 IN.

OPERATING - Mode l CONDITION - Mode 2

- Mode 3 BASIS If Reactor water level cannot be restored and maintained above the 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 EQIs 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 31

EMERGENCY EPIP-1 CLASSIFICATION SECTION mI PROCEDURE TECHNICAL BASIS 1.0 REACTOR 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 MlARFP in Table 1.1-G2.

OPERATING - Mode I 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 (MIRFP) 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 31

EMERGENCY EPIP-1 CLASSIFICATION SECTION m PROCEDURE TECHNICAL BASIS 1.0 REACTOR v.VNA - m - invm m - - --

GENERAL EMERGENCY (CONTINUED)

CURVIES/TAB LES TABLE 1.1 - G2 MINIMUM ALTERNATE RPV FLOODING PRESS NUMBER OF OPEN MSRVs MARFP (PSIG) 6 or More 190 5 230 4 290 NOTES NOTE I I-G2 The reactor will remain subcritical under all conditions w/o boron when:

  • All control rods except one are inserted to or beyond position 00
  • Determined by reactor engineering 1.0 REACTOR PAGE 86 OF 207 REVISION 31

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

ALERT Failure of automatic scram functions to bring the Reactor subcritical AND Manual scram or Alternate Rod Insertion (ARI) was successful.

OPERATING - Mode I 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 1I.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 31

EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 1.0 REACTOR UPIk'ErWUIEhIPJ.

SITE AREA EMERGENCY Failure of automatic scram, manual scram, and ARI to bring the Reactor subcritical.

OPERATING - Mode I 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 31

EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 1.0 REACTOR TECHNICAL BASIS 1.0 REACTOR EIFIRIJr4 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 -185 IN.

OPERATING - Mode I CONDITION - Mode 2 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 S

1.0 REACTOR PAGE 89 OF 207 REVISION 31

EMERGENCY EPIP-1 CLASSIFICATION SECTION m PROCEDURETjl TECHNICAL, BASIS 1.0 REACTOR GENERAL EMERGENCY (CONTINUED)

CURV'ES/TABLES CURVE 1.2-G HEAT CAPACITY TEMP LIMIT 250 . .. == SAFE WHEN RX PRESS

........ ...... IS BELOW 65 PSIG 240 - RPV Press 65 . _

230 - .- - - ---

~7220

_RPVPress 300 -- - - - - - -

2 _ RPV Press 500 - - - - -

a- 190 RPV Press 700 ' -

RPV Press 900 SA RPV Press. 1135 SAFE 150 - - - , m 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 IF ABOVE CURVE FOR EXISTING RX 1.0 REACTOR PAGE 90 OF 207 REVISION 31 l

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

-_I]1I __ :IN W.I I UNUSUAL EVENT Reactor coolant activity exceeds 26 pCi/gm dose equivalent 1-131 (Technical Specification limit) as determined by chemistry sample.

OPERATING All CONDITION 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 31 l

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

- 4 W1U (6] Ki&SI6] WihI WAIU Ub'i I i'm ALERT Reactor coolant activity exceeds 300 pCi/gm dose equivalent Iodine-131 as determined by Chemistry sample.

OPERATING - Mode 1 CONDITION - Mode 2

- Mode 3 BASIS Reactor coolant activity samples exceeding 3000 pCi/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, (NUMARC-FA)

- RIMS L36 921201 806 1.0 REACTOR PAGE 92 OF 207 REVISION 31

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

k'i R) WaIffiyK',I I 3 I U[@].W 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 I 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 pCi/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 31 l

EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 1.0 REACTOR ALERT Reactor moderator temperature CANNOT be maintained below 2120 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, (NJMARC-SA3) 1.0 REACTOR PAGE 94 OF 207 REVISION 31

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

_~ . ~

_=_.VE 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 RENWSON 31

EMERGENCY EPIP-1 CLASSIFICATION SECTION III PROCEDURE TECHNICAL BASIS 1.0 REACTOR SITE AREA EMERGENCY (CONTINUED)

CURVES/TABLES CURVE 1.5-S HEAT CAPACITY TEMP LIMIT 260 - _ - - - - - _ __

250 _- __ - SAFE WHEN RX PRESS

. - _IS BELOW 65 PSIG 240 RPV Press 5 -

230 -- -

- 220 - c.=,-- .-

RPV Press. 300 _ - - - - -

a. 200 - RPV Press 500-- - --

190 - - RPV~es 0 RPV Press 900 - -

170 - RPV Press 1135 - - - - -

SAFE 115 12 12.5 13 13 5 14 14.5 15 15.5 16 165 17 17.5 18 185 19 SUPPR PL LVL (FT)

ACTION REQUIRED IFABOVE CURVE FOR EXISTING RX 1.0 REACTOR PAGE 96 OF 207 REVISION 31

EMERGENCY EPIP-1 CLASSIFICATION SECTION III 3.0 SECONDARY PROCEDURE TECHNICAL BASIS CONTAINMENT SECONDARY C ONTAINMENT 3.0 3.0 SECONDARY PAGE 116 OF 207 REVISION 30 l CONTAINMENT

EMERGENCY EPIP-1 CLASSIFICATION SECTION m 3.0 SECONDARY PROCEDURE TECHNICAL BASIS CONTAINMENT EII (kOhl 37.1 I'E&ShM W'1 klii I kIII

  • Ubh' I MDI W'I III 4 DRBU 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 I 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 3.0 SECONDARY PAGE 117 OF 207 REVISION 30 l CONTAINMENT

EMERGENCY EPIP-1 CLASSIFICATION SECTION III 3.0 SECONDARY PROCEDURE TECHNICAL BASIS CONTAINMENT E~1I~S]I ~l~IkS~UWII~k' 1 U~ILM M !~I ~ ~~UIi bN b' SITE AREA EMERGENCY (CONTINUED)

CURVES/TABLES TABLE 3.1 MAXIMUM SAFE OPERATING AREA TEMPERATURE LIMITS APPLICABLE PANEL 9-21 AREA TEMPERATURE ELEMENTS '%IAX SAFE OPERATINGV ALUE OF (UNLESS OTHERWNISE NOTED) UNIT 2 UNIT 3 RHR A'C PUMP ROOMI 74-95A 150 155 RHR B'D PUMP ROOM 74-95B 210 215 HPCI TURBINE AREA 73-55A 270 270 RCIC TURBINE AREA 7141A 190 190 CS PUNIP ROOM HIGH HUMIDITY OR TEMIP (XA-55-3E-29) PANEL 9-3 TI-75-69B 150 150 HIGH RCIC STEMI SUPPLY AREA 7141B. 41C. 41D 200 250 IIPCI STEAM SUPPLY AREA 73-55B. 55C 55D 240 240 RHR A'C PUNIP SUPPLY AREA 74-9511 240 240 RHR B'D PUMIP SUPPLY AREA 74-95G 240 240 MAIN STEAM LINE LEAK DETECTION HIGI1 (XN-55-3D-24) PANEL 9-3 TIS-1-60A 315 315 RHR VALVE ROOM 74-95E 170 175 RNNCU ISOL LOGIC CH ANNEL A'B TEMIP HIGH (XA-55-5B-32'33) PANEL 9-5 170 175 69-835A. B. C. D AUX INST ROOM_

RNNCU OlTBD ISOL \'LV AREA 69-29F 220 220 RWCU IIX AREA 69-290 220 220 RWVCU IIX EXII DUCT 69-2911 220 220 RNNCU RECIRC PUNIP A AREA 69-29D 215 215 RXNCU RECIRC PUMP B AREA 69-29E 215 215 RHR A'C HX ROOM 74-95C 195 200 RHR B'D ILX ROOM 74-95D 195 200 FPC HX AREA 74-95F 155 155 3.0 SECONDARY PAGE 118 OF 207 REVISION 30 CONTAINMENT

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

. _ M W0 IO1Sh IY I I'StMW F U I~MUHILAUU~ L u~ iN5~

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 I 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 REVISION 30 l CONTAINMENT

EMERGENCY EPIP-1 CLASSIFICATION SECTION III 3.0 SECONDARY PROCEDURE TECHNICAL BASIS CONTAINMENT EJ KUIS]hI IIKI I'LSIOJM WmI IIi I mhM U UOIk III bi LI III 4 0U ECU 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 TABLE 3.1 MAYIIIUM SAFE OPERATING AREA TEMPERATURE LIMITS APPLICABLE PANEL 9-21 0

AREA TEMPERATURE ELEMENTS 'LIXX SAFE OPERATING VALUE F (UNLESS OTIIERN ISE NOTED) UNIT 2 UNIT 3 RIIR A'C PUMP ROOM 74-95A 150 155 RHR B D PUMIP ROOM 74-95B 210 215 HPCI TURBINE AREA 73-55A 270 270 RCIC TURBINE AREA 71-41A 190 190 CS PUMP ROOM HIGH IIUMIDITY OR TEIP (XA-55-3E-29) PANEL 9-3 TI-75-69B 150 150 HIGH RCIC STEAM SUPPLY AREA 71-41B. 41C. 41D 200 250 HPCI STEAM SUPPLY AREA 73-55B. 55C. 55D 240 240 RHR A'C PUMIP SUPPLY ARE-A 74-9511 240 240 RHR B'D PUMP SUPPLY AREA 74-95G 240 240 MXIMNSTEAM LINE LEAK DETECTION HIGH (XA-55-3D-24) PANEL9-3 TIS-1-60A 315 315 RHR VALVE ROOM 74-95E 170 175 RWCU ISOL LOGIC CHA.NNEL AB TEMIP HIGH (XA-55-5B-32'33) PANEL 9-5 170 175 69-835A. B. C. D AUX INST ROOM RNNCU OUTBD ISOL VLV AREA 69-29F 220 220 RWVCU HX AREA 69-29G 220 220 RWCU HX EXH DUCT 69-29H 220 220 RWVCU RECIRC PUMP A AREA 69-29D 215 215 RWCU RECIRC PUMP B AREA 69-29E 215 215 RHR A'C HIX ROOM 74-95C 195 200 RHR B'D HIX ROOM 74-95D 195 200 FPC HIX AREA 74-95F 155 155 I TABLE 3.1-G/3.2-G INDICATIONS OF POTENTIAL OR SIGNIFICANT FUEL FAILURE is ith RCS Barrier Intact DRYNN ELL RADIATION UNIT 2 DRIA ELL RADIATION UNIT 3 2-RE-90-272A > 345 R'HR 3-RE-90-272A l> 106 R/JIR 2-RE-90-273A > 164 R'HR 3-RE-90-273A I > 164 R/HIR REACTOR COOLANT ACTIVITY > 300 pLCIgm DOSE REACTOR COOLANT ACTIVITY > 300 ItCI'g DOSE EQUILAVENT IODINE-13 1 EQUILAVENT IODINE-13 1 3.0 SECONDARY PAGE 120 OF 207 REVISION 30 CONTAINMENT

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

'j kOLS]I IJm I'L&6]LI W.'I ILM I IIM U tLI I] VI U[S]WU ALERT Any of the following high radiation alarms on Panel 9-3:

  • RA-90-lA, 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, (NUMARC-AA2-example-1) 3.0 SECONDARY PAGE 121 OF 207 REVISION 30 l CONTAINMENT

EMERGENCY EPIP-1 CLASSIFICATION SECTION III 3.0 SECONDARY DRFlCnrVl1jRF. TECHNICAL BASIS CONTAINMENT IiiY.I I'LOiSJLI WmI UM I i U Iii ru uL@JWU 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 I 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)

- EGI Program Manual, Section V-E 3.0 SECONDARY PAGE 122 OF 207 REVISION 30 l CONTAINMENT

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

oiI&1l ~ I OWNU11I L OUTIW?1MF2 U~n I SITE AREA EMERGENCY (CONTINUED))

CURVES/TABLES TABLE 3.2 It A %'II% IrTTA CA IC nD' AT1N.VP A UP RAPlAlATiON LIMITS AREA RAD MONITOR NLAX SAFE VALUE NMR11Rl RIIR WEST ROOM 90-25A 1000 RHR EAST ROOM 90-28A 1000 HPCI ROOM 90-24-A 1000 CSIRCIC ROOM 90-26 A 1000 CORE SPRAY ROOMI 90-27A 1000 SUPPR POOL AREA 90-29A 1000 CRD-IICU WEST ARE-A 90-20A 1000 CRD-HCU EAST AREA 90-21A 1000 TIP DRIVE AREA 90-23A 1000 NORTH RIVCU SYSTEM AREA 90-13 A 1000 SOUTHI RWVCU SYSTEM AREA 90-14 A 1000 RWCU SYSTEM AREA 90-9A 1000 NIG SET AREA 90-4A 1000 FUEL POOL AREA 90-IA 1000 SERVICE FLR AREA 90-2A 1000 NEW FUEL STORAGE 90-1 x 1000 3.0 SECONDARY PAGE 123 OF 207 REVISION 30 CONTAINMENT

EMERGENCY EPIP-1 CLASSIFICATION SECTION III 3.0 SECONDARY PROnCrEDRE TECHNICAL BASIS CONTAINMENT II I U t4l _91 ] S 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-N0090-93 0050 R2 (Unit 2/3) 3.0 SECONDARY PAGE 124 OF 207 REVISION 30 CONTAINMENT

EMERGENCY EPIEP-l CLASSIFICATION SECTION III 3.0 SECONDARY PROCEDURE TECHNICAL BASIS CONTAINMENT M

1OLS]I 17.1 I'LOLS]I W.1 IIk' I mM U I] VI H[Oh I'll GENERAL EMERGENCY (CONTINUED)

CURVES/TABLES TABLE 3.2 MAXIMUMT1 SAFE OPERATING AREA RADIATION LIMITS AREA RMD MONITOR NIMAXSAFE VALUE NRIIIR RIIR UVEST ROOM 90-25A 1000 RHR EAST ROOM 90-28A 1000 IlPCI ROOM 90-24A 1000 CS'RCIC ROOM 90-26 A 1000 CORE SPRAY ROOM 90-27A 1000 SUPPR POOL AREA 90-29A 1000 CRD-HCU WEST ARE A 90-20 A 1000 CRD-HCU EAST AREA 90-21 A 1000 TIP DRIVE AREA 90-23A 1000 NORTH RU'CU SYSTE\M AREA 90-13A 1000 SOUTH RU'CU SYSTENI AREA 90-14 A, 1000 RU'CU SYSTEM ARE A 90-9 A. 1000 MIG SET AREA 90-4 . 1000 FUEL POOL AREA 90-IA 1000 SERVICE FLR AREA 90-2A 1000 NEW FUEL STORAGE 90-IA 1000 TABLE 3.1-G/3.2-G INDICATIONS OF POTENTIAL OR SIGNIFICANT FUEL FAILURE with RCS Barrier Intact DRVYA ELL RADIATION UNIT 2 DRYWELL RADIATION UNIT 3 2-RE-90-272A > 345 RIIIR 3-RE-90-272A > 106 RIIR 2-RE-90-273A > 164 RIl{R 3-RE-90-273A I > 164 R'IIR REACTOR COOLANT ACTIVITY > 300 jiCl/gm DOSE REACTOR COOLANT ACTIVITY > 300 [tCIgni DOSE EQUILAVENT IODINE-131 EQUILAVENT IODINE-131 3.0 SECONDARY PAGE 125 OF 207 REVISION 30 CONTAINMENT

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