ML031920268

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Revision 9 to EP-PS-324, Fuels Lead Engineer - Emergency Plan-Position Specific Procedure.
ML031920268
Person / Time
Site: Susquehanna  Talen Energy icon.png
Issue date: 06/24/2003
From:
Susquehanna
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
EP-PS-324, Rev 9
Download: ML031920268 (64)
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Text

Jun. 24, 2003 Page 1 of 1 MANUAL HARD COPY DISTRIBUTION DOCUMENT TRANSMITTAL 2003-29913 TRANSMITTAL INFORMATION:

TO: t .OD 06/24/2003 LOCATION: DOCUMENT CONTROL DESKL: T.

FROM: NUCLEAR RECORDS DOCUMENT CONTROL CENTER (NUCSA-2)

THE FOLLOWING CHANGES HAVE OCCURRED.TO THE HARDCOPY OR ELECTRONIC MANUAL ASSIGNED TO YOU:

324 - 324 - FUELS LEAD ENGINEER - EMERGENCY PLAN- POSITION SPECIFIC PROCEDURE REMOVE MANUAL TABLE OF CONTENTS DATE: 02/26/2003 ADD MANUAL TABLE OF CONTENTS DATE: 06/23/2003 CATEGORY: PROCEDURES TYPE: EP ID: EP-PS-324 REPLACE: REV:9 UPDAq E DISTRIBUTED WITHIN 5 DAYS IN ACCORDANCE WITH DEPAE ALL CHANGES AND ACKNOWLEDGE COMPLETE IN YOUR NIMS . FOR ELECTRONIC MANUAL USERS, ELECTRONICALLY REVII Remove Tab 2 , ACKNOWLEDGE COMPLETE IN YOUR NIMS INBOX.

Deleted 045

Tab 1 EP-PS-324-1 FUEL DAMAGE WORKSHEET General The following information should be kept current at all times after facility activation. This information is used by Dose Calculators to perform dose projections for support of Protective Action Recommendations required within fifteen minutes of a General Emergency classification.

Engineering Support is required to provide an estimate of percent fuel damage to the TSC Dose Calculator and the EOF Dose Assessment Staffer in a timely manner, allowing sufficient time for a dose projection to be performed.

The following information is the best estimate possible within the time and using available data.

If no information is provided to the Dose Calculator, default values will be used to determine dose projections. This may result in more severe conditions prompting a non-conservative protective action recommendation.

1.0 ISOTOPIC DETERMINATION: (choose one)

UNKNOWN MIX (Containment Rad <5R hr)

NORMAL COOLANT LEAK (Containment Rad <5R/hr)

LOCA No Fuel Damage (Iodine Spike, Containment Rad <1 OR/hr) -

LOCA CLAD FAILURE (Containment Rad 1.5E+02 - 5.OE+04 R/hr)

LOCA FUEL MELT (Containment Rad 8.OE+03 - 1.OE+06 R/hr)

FUEL HANDLING ACCIDENT EP-AD-000-454, Revision 3, Page 1 of 2

Tab 1 EP-PS-324-1 2.0 CORE CONDITION: (choose one)

Gap Release (Core uncovered for 15-30 minutes) l In Vessel Severe Damage (Core Uncovered >30 Minutes)

'9 Vessel Melt Through EP-AD-000-454, Revision 3, Page 2 of 2

' ( ~~~~~~~~~( (

Tab 3 EP-PS-324-3 DOSE PROJECTION WORKSHEET GENERAL:

After initial activation, the following information should be kept current at all times. This information is used by the Health Physics Dose Calculator to perform dose projections In support of Protective Action Recommendations required to be made within 15 minutes of a General Emergency classification.

The Engineering Support function will provide the following information to the Dose Calculator in a timely manner to allow sufficient time to perform a dose projection.

The following Information Is the best estimate possible considering the time and information available. If the Information is not provided, the Dose Calculator will use default values which may yield dose projections much more severe than actual conditions prompting a non-conservative Protective Action Recommendation.

1.0 GENERAL INFORMATION RX SHUTDOWN TIME: 1 RELEASE START TIME: [RELEASE STOP TIME:

2.0 TYPE OF RELEASE (select one)

EP-AD-000-459, Revision 2, Page 1 of 4

( ~~~~~~~~~~(

Tab 3 EP-PS-324-3 3.0 DBA ACCIDENT TYPES [ISOTOPIC DETERMINATION (CHOOSE ONE)]

UNKNOWN MIX (Containment Rad <5 R/hr)

NORMAL COOLANT LEAK (Containment Rad <5 R/hr)

LOCA (No Fuel Damage, Iodine Spike, Containment Rad <10 R/hr)

LOCA CLAD FAILURE (Containment Red 1.5E+02 - 5.OE+04 R/hr) l_____%(Give Estimate)

LOCA FUEL MELT (Containment Rad 8.0E+03 - 1.0E+06 R/hr) l_% (Give Estimate)

FUEL HANDLING ACCIDENT NOTE: Quick methods to determine the isotopic mix/type of fuel damage and estima"e percentages are located in HELP tabs entitled Core Damage Estimate I (Primary System Breach Inside Containment) and Core Damage Estimate II (Small or no Primary System Breach Inside Containment).

EP-AD-000-459, Revision 2, Page 2 of 4

( ( (

Tab 3 EP-PS-324-3 4.0 RELEASE TYPES (Select release type and go to release selected) 4.1 Orvwell Release (circle one In each column)

Core Condition (Choose one, provide estimate In /6) Sprays Hold-up Time Treatment Release Rate Gap Release (Core uncovered for 15-30 minutes) On <1 Hour Filtered 100%/o/hr In Vessel Severe Damage (Core uncovered > 30 minutes) Off 2-12 Hours Unfiltered 100%/oday Vessel Melt Through 24 Hours Design (10/o/day) 4.2 Wetwell Release (circle one In each column)

Core Condition Water Hold-up Time Release Rate (Choose one, provide estimate In /') Conditions Treatment Gap Release (Core uncovered for 15-30 minutes) Subcooled <1 Hour Filtered 100%/hr In Vessel Severe Damage (Core uncovered > 30 minutes) Saturated 2-12 Hours Unfiltered 100%/o/day Vessel Melt Through 24 Hours Design (10/oday)

EP-AD-000-459, Revision 2, Page 3 of 4

( (.

Tab 3 EP-PS-324-3 4.3 Secondary Containment Bypass Release (circle one In each column)

Core Condition (Choose one, provide estimate in /6) Treatment Release Rate Gap Release (Core uncovered for 15-30 minutes) Filtered 100%/a/hr In Vessel Severe Damage (Core uncovered > 30 minutes) Unfiltered 100%/o/day Vessel Melt Through Design (1%/day) 4.4 Spent Fuel Pool Release (circle one In each column)

Core Condition Hold-up Release (Choose one) Accident Type Time Treatment Rate Gap Release Zircaloy Fire in One 3 Month Batch <1 Hr Filtered/ 100%/hr (Core uncovered for 15-30 minutes) Sprays on In Vessel Severe Damage Gap Release from One 3 Month Batch 2-12 Hrs. Unfiltered/ 100%/o/day (Core uncovered > 30 minutes) Sprays off Vessel Melt Through Gap Release from 15 Batches EP-AD-000-459, Revision 2, Page 4 of 4

Tab4 EP-PS-324-4 CORE DAMAGE ESTIMATE I (Primary System Breach Inside Containment)

NOTE: It is important to quickly provide a status of the present situation and a prognosis on whether the situation is expected to degrade, improve, or remain the same, (i.e., within 5 to 10 minutes of a change in plant status).

1.0 INDICATORS USED 1.1 Containment Radiation Use Attachment 1, A, B, or C, as applicable, to determine the amount and type of fuel damage using containment radiation monitors. These figures were taken from the US NRC Response Technical Manual, RTM-96. Obtain the containment radiation levels from SPDS or the Control Room indicators.

NOTE (1): Correction for the pre-release background-radiation levels may be required as listed below.

Gap or In-Vessel Melt - The background radiation monitor value is normally low (5 4 R/hr) relative to 1% gap or in-vessel melt release. Consequently, the monitor reading does not require correction for background level in determining the type and amount of fuel damage. If the background radiation monitor reading is > 4 R/hr, the monitor reading should be corrected for the background level in determining the type and amount of fuel damage.

Spiked or Normal Coolant - The radiation monitor value requires correction for the background level. Correct the monitor reading to account for the normal background level in determining the type and amount of fuel damage.

NOTE (2): Containment radiation will go up if there is fuel damage. The increase will depend on the type of fuel damage, and whether or not there was a LOCA, Drywell andlor Wetwell sprays were used, and the amount of blowdown from the Reactor Vessel to the Suppression Pool.

In the case of a LOCA, the fuel damage estimate depends strongly on whether or not containment sprays are being used.

Special care should be taken to confirm the operation of containment sprays.

EP-AD-000-457, Revision 7, Page 1 of 10

Tab 4 EP-PS-324-4 1.2 Containment Hydrogen Use Attachment 2, taken from the US NRC Response Technical Manual RTM-96, to determine the amount and type of fuel damage using Hydrogen Concentration. Obtain the containment Hydrogen levels from SPDS or the Control Room indicators.

NOTE: Containment Hydrogen will increase if there is a LOCA inside the containment and significant fuel damage.

1.3 Coolant Fission Product Concentration vs. Core Damage Coolant sampling will indicate the amount of fuel damage, but in most cases, will take too long for use in dose projections. If PASS sample data becomes available, the Nuclear Fuels Engineer is responsible for assuring a fuel damage calculation based on the measured fission product inventories is performed. The results of this analysis should be compared to previous ealculations using other methods.

1.4 Plant Translent Precipitating Fuel Damage If the core experienced a loss of coolant accident and is not covered within 15 minutes, refer to Attachment 3 taken from the US NRC Response Technical Manual RTM-96. The amount of time the core was uncovered can be determined using SPDS. Using the attached figures will provide an estimate of potential fuel damage. Coolant samples must be taken to accurately assess fuel damage.

The type of transient experienced by the reactor leading tofuel damage can be an indicator of the amount and type of fission products released.

  • If the core experienced an overpower/pressure transient, a gap release may have occurred.
  • If the core experienced a mechanical failure, which could produce flow blockage, there may be localized fuel melt.
  • If the core experienced a mechanical perturbation, such as a seismic event or a large steam line break causing a large delta pressure across the core, a gap release could result.
  • If the Reactor failed to shut down (ATWS) with a subsequent loss of cooling, there may be fuel melt.

EP-AD-O00-457, Revision 7, Page 2 of 10

Tab4 EP-PS-324-4 Containment Radiation Monitor Response Direct Release Path to Drv well I.E+07 (Sprays Off) 1.E+06 100%

-50%

1.E+05 _ - 10% 53% 1DO%

10%

E 1

_ _04 _ -1O% _ -0% _- 0% _

0 1%~~~~10 2_1%

-- 5%10%-_

C- 1.E F03 -_1% __5%,

0 _1% ._

E C

'E I.E402-

°0 .E+01 - E10%

1.12+00 - 0 ~~~~~~~~~~~~~0 _ 0%-100X IWO E01- -ss-50%f 5% s

%0 1%

0 1.E 1_ _-5% 100% *_;1 I.E-02 1% 10 I.E.03  ! 0.

I.E-04 1%

I.E-05-h 24h 1h 24h In 24h 1h 24h In-Vessel Melt Gap Spiked Coolant Normal Coolant Core Damage and Time After Reactor Shutdown Note 1: This figure should be used only when there Is a primary system breach Inside containment and a direct release path to the Drywell.

Note 2: See Attachment 3 to determine If fuel melt occurred (core uncovered or fuel blockage).

ATTACHMENT 1A EP-AD-000-457, Revision 7, Page 3 of 10

Tab 4 EP-PS-324-4 Containment Radiation Monitor Response Direct Release Path to Dry well (Sprays On)

I.E+07 I.E+06 I.E+05 I.E+04 1E .E.03 0

E C

3 I.E+02 C

0 U

2 16I.EeO1 0

I0 0

  • i I.E.OO 40 I.E-02 I .E-03 i .E-04 I.E-05 1h 24h 1h 24h 1h 24h 1h 24h In-Vessel Melt Gap Spiked Coolant Normal Coolant Core Damage and Time After Reactor Shutdown Note 1: This figure should be used only when there Is a primary system breach Inside containment and a direct release path to the Drywell.

Note 2: See Attachment 3 to determine if fuel melt occurred (core uncovered or fuel blockage).

ATTACHMENT 1B EP-AD-000-457, Revision 7, Page 4 of 10

Tab 4 EP-PS-324-4 Containment Radiation Monitor Response Direct Release to Wetwell and Not to Drvwell I.E+06 I.E+05 I.E+04 I.E+03

-1.E+02 C

C)

.E I.E+01

  • g i.E+00 0

1.E-01 I.E-04 I .E-05 1.E-06 1h 24h 1h 24h lh 24h 1h 24h In-Vessel Melt Gap Spiked Coolant Normal Coolant Core Damage and Time After Reactor Shutdown Note 1: This figure should be used only when there Is a primary system breach Inside containment and a direct release path to the Wetwell without a primary release to the Drywell.

Note 2: See Attachment 3 to determine If fuel melt occurred (core uncovered or fuel blockage).

ATTACHMENT IC EP-AD-000-457, Revision 7, Page 5 of 10

Tab 4 EP-PS-324-4 CONTAINMENT HYDROGEN VS CORE DAMAGE

% MoaI-ws Peacdion & Cam Dwmage Staft 60 40 D0 4.,pon~ft Mek Throug 20 t0 <.SW Fadmek 0

0.1 1 10 H2VO% In Containment Saw= NLVMIM2. P.4-3; dunyg =ame NUREG4524. VoL S.;

Da ecuae NMUMEG470O; NUREG=C4Wl41; NUREGICR4567, Tabl 439, p. 71.

A1TACHMENT 2 EP-AD-000-457, Revision 7, Page 6 of 10

Tab 4 EP-PS-324-4 WATER INJECTION REQUIRED TO COOL CORE BY BOILING CAUTION:

I These rates are those required to remove decay heat from a 3000 MW(t) plant by boiling. If there is a break requiring make up or injected water, more water than indicated will be required to both keep the core covered and cooled.

CAUTION:

If the core has been uncovered, the fuel temperature will have increased significantly.

Additional flow will be required to accommodate the heat transfer necessary to return to equilibrium fuel temperature.

NOTE:

These curves are based on a 3000 MW(t) plant operated at a constant power for an infinite period and then shutdown instantaneously. The decay heat power is based on ANS-5.1/N18.6.

Assuming the injected water is at 800 F, these curves are within 5% for pressures between 14 psia to 2500 psia. These curves are within 20% for injected water temperatures up to 2120 F.

ATTACHMENT 3 (Page 1 of 4)

EP-AD-000-457, Revision 7, Page 7 of 10

Tab 4 EP-PS-324-4 WATER INJECTION REQUIRED TO COOL CORE BY BOILING While the op of the actiVe c La unCovered, assume that the fuel il heat fp at 2-2*F/seC. the Szcreased cor temperatty* vii result in fuel pin damage ax shown below.

ROT:

These eim-esare reasonable (factor of pal

2) if the core Ls uncovered vithn -a . -

tow hours of shutdon (Including failore to srm, I f there is sufficient Izijection, core eatup may be stopped or slowed due P o*wmwm*M dofsiwi" C to stean cooling.

Steam cool may not ._

prevent Core damage under accidentdfir codlton. dishtInI-I tW nPbd mgem am

. "M~f*zMMf -sUb..._

i pradum hii phi wo am NOMM g iuce: UURG-090oo, 2/-4524, _UMG-0956 ATTACHMENT 3 (Page 2 of 4)

CAUTION: If the core Is severely damaged, It may not be In a coolable state even If covered again with water.

NOTE: If there is sufficient injection, core heatup may be stopped or slowed due to steam cooling. Steam cooling may not prevent core damage under accident conditions.

EP-AD-000-457, Revision 7, Page 8 of 10

Tab 4 EP-PS-324-4 WATER INJECTION REQUIRED TO COOL CORE BY BOILING wz=RZZ (gin) REQUIRID To RlPLU0 WATER 105 An D0IMM Mg TO VIM HVLaT rCR A 3000 NMt)

KAN (1/2-24 BOURS lrM SHUTDOWN)

Sol lea c InJeco soo soo SA SsD -

100 1lC6.00 3BY 00 DU to D FOR A 3000 (t) 300 SUN (I to 30 DAS amR Smn~TOW 50 . - 30 40 - _ _~~~~~~~~~~~~~A 40 20 , s 0.5 2 3.......... 3 4 1 ...

1205 t0 ATTACHMENT 3 (Page 3 of 4)

EP-AD4-00-457, Revision 7, Page 9 of 10

Tab 4 EP-PS-324-4 WATER INJECTION REQUIRED TO COOL CORE BY BOILING Care dauge vs.mthe tbtreacor are is mcovrd T=me PWR or 20% of BWR aive are is

(_) (OF) (C) Posske Coe d 0 >600 >315

  • Now 0.5 so 0.7S 1800-2400 980-1300
  • Locaf fad m f *l
  • Busmin of d i fi sm pmdnwsin (eadic Zr-H2o seactwish upid H2 gemion)
  • RapAd eIdm gfiilme (pp r PI Im from ft core) 0.5 to IS 2400-4200 130-2300 e RaM reline of vodtile fission

- v- d core s

dAl re ease fr= ore)m

  • Possie rehnat (slup) of molte
  • Posble wolable care 1to 3+ >4200 >2300
  • Mebugh of vessd wilb possile conimmm f*lmce ad rellus of addisiona I kSI-P hes-volatie fission

.Sow=. NURBEG=c4245. NUREG1cR462. NURBG=.46Z9. KuREG=-s37. NuRmG000 NUREG4)9S6. N4UREG-4 10. and NUREG-1465.

ATTACHMENT 3 (Page 4 of 4)

EP-AD-000-457, Revision 7, Page 10 of 10

Tab 5 EP-PS-324-5 CORE DAMAGE ESTIMATE II (Small or no primary system breach inside Containment)

This instruction provides a method of estimating the percentage of fuel that has failed using the Containment Post-Accident Radiation Monitor (CPARM) readings on panel 1C601 (2C601) during an accident. Since the Containment Post-Accident Radiation Monitor readings are readily available, this calculation provides a quick assessment of core damage. This estimate only applies if there is a small or no primary system breach within containment.

1.0 LIMITATIONS OF THE METHOD 1.1 This procedure will only determine qualitatively the amount of fuel damage. The method uses Containment Post-Accident Radiation Monitor Readings to calculate the percentage of failed fuel during an accident where the fission products are released from the fuel rod cladding. The methodology is based on assumptions with large uncertainties that can significariffy affect the results.

1.2 To use this method, the accident scenario up to the time of the Containment Post-Accident Radiation Monitor Reading must be well understood to estimate the fuel temperatures required by this procedure.

1.3 In addition, a Containment Post-Accident Radiation Monitor Reading and the time the reading was obtained must be available.

2.0 RESPONSIBILIES 2.1 The Nuclear Fuels Engineer, Lead Technical Supoort Engineer, or designee collects information and makes estimates and determinations described in this procedure.

3.0 INSTRUCTIONS 3.1 Determine if Cladding Failure, Fuel Overheat, or Fuel Melt has occurred:

3.1.1 Claddina Failure is expected if peak cladding temperature remains less that 22000F, but the Containment Post-Accident Radiation Monitor readings have increased.

3.1.2 Fuel Overheat is expected If peak cladding temperature exceeds 22001F, but the maximum volume-averaged fuel pellet temperature remains less that 45000°F.

3.1.3 Fuel Melt is expected If any volume-averaged fuel pellet temperature exceeds 45000F.

EP-AD-000-270, Revision 6, Page 1 of 3

Tab 5 EP-PS-324-5 3.2 Since the fuel melt temperatures are dependent on the event progression, specific guidelines cannot be given to cover all scenarios. Some judgment will have to be made or specific temperature calculations will have to be performed during the event. However, the following provides guidelines for a few known scenarios.

3.2.1 If a main steamline high radiation trip causes the scram and the core remains covered, usually cladding failure can be assumed and is possibly due to debris fretting, short term DNB, or PCI. However, if channel flow blockage is suspected, overheat or melting may occur.

3.2.2 For loss-of-inventory-after-the-reactor-is-shutdown scenarios, use Attachment 3 to Tab 4 to estimate if Fuel Melt has occurred.

3.3 Determine the Time After Reactor Shutdown that a Containment Post-Accident Radiation Monitor Reading was obtained.

3.4 Determine if the event has resulted in a primary system breach inside primary containment (increase in drywell pressure/temperature and inventory makeup to the vessel is required to maintain level in the vessel). If the total primary system water released to the drywell is equivalent to less than 9,000 gallons or no primary system breach has occurred inside primary containment, use Figure 1.

Otherwise, use Core Damage Estimate I (Tab 4).

Note: The 9,000 gallon value Is about 10% of the fluid volume of the reactor vessel and primary piping (main steam, reactor recirculation, and feedwater).

3.5 Determine Fraction of Fuel Failed (FFF) as follows:

FFF= CPARM Reading x 100 Expected 100% Fuel Failure CPARM Reading EP-AD-000-270, Revision 6, Page 2 of 3

Tab 5 EP-PS-32 4 -5 FIGURE A CONTAINMENT HIGH RANGE RADIAION MONITOR READINGS THAT ARE EXPECTL WITH 100% OF THE FUEL FAILED FOR AN EVENT WITH NO PRIMARY $YSTA BREACH INSIDE CONTNMENT if .................

p....-.

. .................. 4.......................... . .

\ ~1 . -.1 . -)..T ----- T------

,,, ,.at,

\. *. FIJEL MEL'. .

~~~~~~~~~~~~~~~~~~~~~~..... . t_<..

..e..... . .

C-.

a_

0do 0

........... .......... .......... ~~~. ~~~~~~..... .. ......... _ ~~~~..

2 . ... .. .. 4..... ..

6- FUEL OVERHEAT. ............  ;

_O

-0 O10, I

~~~~~~~~~~~.. .. ..X....... . ............. .......... .4 . 4..... ....... _.....

  • S . . .. ... ...... .....

z ----'4.'--''4'--

0 >)S;; @@< <* .--.. .............................................

~..

xW. ..

'CLADDING FAIWRE 10 i -' t 'l 'l 'l~~~~~~~~~~~~~

II i1 15 . *0 Post-Shutdown Time (HRS)

EP-AD-ooo-2 7 0, Revision 6, Page 3 of 3

Tab 6 EP-PS-324-6 EMERGENCY CLASSIFICATION CHECK E 1.0 TIMING OF CLASSIFICATION 0o 1.1 UNUSUAL EVENT An UNUSUAL EVENT shall be declared within 15 minutes of having information necessary to make a declaration.

0 1.2 ALERT An ALERT shall be declared within 15 minutes of havinginiformation necessary to make a declaration.

0 1.3 SITE AREA EMERGENCY A SITE AREA EMERGENCY shall be declared within 15 minutes of having information necessary to make a declaration.

0 1.4 GENERAL EMERGENCY A GENERAL EMERGENCY shall be declared within 15 minutes of having information necessary to make a declaration.

EP-AD-000-200, Revision 19, Page 1 of 38

Tab 6 EP-PS-324"6 CLASSIFICATION OF EMERGENCY CONDITIONS USE OF EMERGENCY CLASSIFICATION MATRIX NOTE: CONFIRM THAT INDICATORS AND/OR ALARMS REFLECT ACTUAL CONDITIONS PRIOR TO TAKING ACTION BASED ON THE INDICATOR OR ALARM.

The matrix is worded in a manner that assumes parameter values indicated are the actual conditions present in the plant.

The matrix is designed to make it possible to precisely classify an abnormal occurrence into the proper emergency classification based on detailed Emergency Action Level (EAL) descriptions.

It is impossible to anticipate every abnormal occurrence. Therefore, before classifying any abnormal occurrence based on the EALs in the matrix, one should verify that the general conditions prevalent in-plant and offsite meet the general class description of the emergency classification. In addition, prior to classification, one should be aware of the ramifications in-plant and particularly offsite of that classification. Special consideration of offsite consequences should be made prior to declaring a GENERAL EMERGENCY.

EP-AD-0O0-200, Revision 19, Page 2 of 38

Tab 6 EP-PS-324-6 CLASS DESCRIPTIONS UNUSUAL EVENT - Events that are occurring or have occurred which indicate a potential degradation of the level of safety of the plant.

No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.

ALERT - Events that are occurring or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.

SITE AREA EMERGENCY - Events that are occurring or have occurred which involve actual or imminent major failures of plant functions needed for protection of the public. Any releases are not expected to exceed EPA Protective Action Guideline exposure levels except inside the emergency planning boundary.

GENERAL EMERGENCY - Events that are occurring or have occurred which involve actual or imminent substantial core degradation or melting with potential for loss of containment integrity. Expectation is that releases will exceed EPA Protective Action Guideline exposure levels beyond the emergency planning boundary.

EP-AD-000-200, Revision 19, Page 3 of 38

Tab 6 EP-PS-324-6 CATEGORY INDEX TO THE MATRIX FOR THE CLASSIFICATION OF EMERGENCY CONDITIONS TABLE OF CONTENTS CATEGORY EVENT PAGE I AIRCRAFT/TRAIN ACTIVITY ............................................. 5 2 CONTROL ROOM EVACUATION ............... .............................. 6 3 FUEL CLADDING DEGRADATION ............................................. 7 4 GENERAL.................................................................................................... 10 5 INJURED/CONTAMINATED PERSONNEL ............................................. 11 6 IN-PLANT HIGH RADIATION ............................................. 12 7 LOSS OF AC POWER ............................................. 13 8 LOSS OF CONTROL ROOM ALARMS AND ANNUNCIATORS ................... 14 9 LOSS OF DC POWER ............................................. 15 10 LOSS OF DECAY HEAT REMOVAL CAPABILITY ....................................... 16 11 LOSS OF REACTIVITY CONTROL ............................... 17 12 LOSS OF REACTOR VESSEL INVENTORY ............................... 19 13 NATURAL PHENOMENA ................................. 21 14 ONSITE FIRE/EXPLOSION ................................ 23 15 RADIOLOGICAL EFFLUENT ................................ 25 16 SECURITY EVENT ............ 29 17 SPENT FUEL RELATED INCIDENT ................................ 31 18 STEAM LINE BREAK ................................. 33 19 TOXIC/FLAMMABLE GASES ................................ 36 20 TECHNICAL SPECIFICATION SAFETY LIMIT ................................ 37 21 DRY FUEL STORAGE ................................. 38 EP-AD-000-200, Revision 19, Page 4 of 38

Tab 6 EP-PS-324-6 1 - AIRCRAFT/TRAIN ACTIVITY UNUSUAL EVENT EAL# 1.1 Aircraft crash or train derailment onsite as indicated by:

Visual observation or notification received by control room operator.

ALERT EAL# 1.2 Aircraft or missile strikes a station structure as indicated by:

Direct observation or notification received by control room operator.

SITE AREA EMERGENCY EAL# 1.3 Severe damage to safe shutdown equipment from aircraft crash or missile impact when not in cold shutdown, determined by:

(A and B and C)

A. Direct observation or notification received by control room operator.

and B. Shift Supervisor evaluation.

and C. Reactor Coolant temperature greater than 200OF as indicated on Panel I C651 (2C651).

GENERALEMERGENCY EAL# 1.4 None.

EP-AD-000-200, Revision 19, Page 5 of 38

Tab 6 EP-PS-324-6 2 - CONTROL ROOM EVACUATION UNUSUAL EVENT EAL# 2.1 None.

ALERT EAL# 2.2 Control Room evacuation as indicated by:

(A and B)

A. Initiation of control room evacuation procedures.

and B. Establishment of control of shutdown systems from local stations.

SITE AREA EMERGENCY EAL# 2.3 Delayed Control Room Evacuation as indicated by:

(A and B)

A. Initiation of control room evacuation procedures.

and B.- Shutdown systems control at local stations not established within 15 minutes.

GENERAL EMERGENCY EAL# 2A None.

EP-AD-000-200, Revision 19, Page 6 of 38

Tab 6 EP-PS-324-6 3 - FUEL CLADDING DEGRADATION UNUSUAL EVENT EAL# 3.1 Core degradation as indicated by:

(A or B)

A. Valid Off-gas Pre-treatment Monitor high radiation alarm annunciation on Panel 1C651 (2C651) or indication on Panel 1C600 (2C600).

or B. Reactor coolant activity, determined by sample analysis greater than or equal to 2 pCi/cc of 1-131 equivalent.

ALERT EAL# 3.2 Severe fuel cladding degradation as indicated by:

(A or B or C or D)

A. Valid Off-gas Pre-treatment monitor High-High radiation alarm annunciation on Panel 1C651 (2C651) or indication on Panel 1C600 (2C600).

or B. Valid Reactor coolant activity greater than 300 LCi/cc of equivalent 1-131, as determined by sample analysis.

or C. Valid Main Steam Line High radiation trip annunciation or indication on Panel 1C651 (2C651).

or D. Valid containment post accident monitor indication on Panel 1C601 (2C601) greater than 200 R/hr. (An 8R/hr correction factor must be added manually to the indication to offset a downscale error if primary containment temperature exceeds 225 degrees Fahrenheit. Reference EC-079-0521.)

(CONTINUED ON NEXT PAGE)

EP-AD-000-200, Revision 19, Page 7 of 38

Tab 6 EP-PS-324-6 3 - FUEL CLADDING DEGRADATION (continued)

SITE AREA EMERGENCY EAL# 3.3 Severely degraded core as indicated by:

(A or B)

A. Reactor coolant activity greater than 1,000 gCi/cc of equivalent 1-131 as determined by sample analysis.

or B. Valid containment post accident monitor indication on Panel 1C601 (2C601) greater than 400 R/hr. (An 8 R/hr correction factor must be added manually to the indication to offset a downscale error if primary containment temperature exceeds 225 degrees Fahrenheit. Reference EC-079-0521.)

(CONTINUED ON NEXT PAGE)

EP-AD-000-200, Revision 19, Page 8 of 38

Tab 6 EP-PS-324-6 3 - FUEL CLADDING DEGRADATION (continued)

GENERAL EMERGENCY EAL# 3.4.a Fuel cladding degradation. Loss of 2 out of 3 fission product barriers (fuel cladding and reactor coolant pressure boundary) with potential loss of the third barrier (primary containment) as indicated by:

(A or B)

A. (1 and 2)

1. Valid containment post accident monitor indication on Panel 1C601 (2C601) greater than 400 R/hr. (An 8 R/hr correction factor must be added manually to the indication to offset a downscale error if primary containment temperature exceeds 225 degrees Fahrenheit. Reference EC-079-0521.)

and

2. (aorborc)
a. Containment pressure greater than 40.4 PSIG, indicated on Panel 1C601 (2C601).

or

b. A visual inspection of the containment indicates a potential for loss of containment (e.g. anchorage or penetration failure, a crack in containment concrete at tendon).

or

c. Other indications of potental or actual loss of primary containment.

or B. (1 and 2)

1. Reactor coolant activity greater than 1,000 gCi/cc of equivalent 1-131 as determined by sample analysis.

and

2. Actual or potential failure of reactor coolant isolation valves to isolate a coolant leak outside containment as determined by valve position indication on Panel 1C601 (2C601) or visual inspection.

OR EAL# 3.4.b Core melt as indicated by:

(A and B)

A. Valid containment post accident monitor indication on Panel 1C601 (2C601) greater than 2000 R/hr. (An 8 R/hr correction factor must be added manually to the indication to offset a downscale error if primary containment temperature exceeds 225 degrees Fahrenheit. Reference EC-079-0521.)

and B. Containment high pressure indication or annunciation on Panel 1C601 (2C601).

EP-AD-000-200, Revision 19, Page 9 of 38

Tab 6 EP-PS-324-6 4-GENERAL UNUSUAL EVENT EAL# 4.1 Plant conditions exist that warrant increased awareness on the part of plant operating staff or state and/or local offsite authorities as indicated by:

Events that are occurring or have occurred which indicate a potential degradation of the level of safety of the plant. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.

ALERT EAL# 4.2 Other plant conditions exist that warrant precautionary activation of PPL, State, County, and local emergency centers as indicated by:

Events that are occurring or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.

SITE AREA EMERGENCY EAL# 4.3 Other plant conditions exist that warrant activation of emergency centers and monitoring teams or a precautionary notification to the public near the site as indicated by:

Events that are occurring or have occurred which involve actual or imminent major failures of plant functions needed for protection of the public. Any releases are not expected to exceed EPA Protective Action Guideline exposure levels except inside the emergency planning boundary.

GENERAL EMERGENCY EAL# 4.4 Other plant conditions exist, from whatever, source, that make release of large amounts of radioactivity in a short time period available as indicated by:

Events that are occurring or have occurred which involve actual or imminent substantial core degradation or melting with potential for loss of containment integrity. Expectation is that releases will exceed EPA Protective Action Guideline exposure levels beyond the emergency planning boundary.

EP-AD-000-200, Revision 19, Page 10 of 38

Tab 6 EP-PS-324-6 5.- INJURED/CONTAMINATED PERSONNEL UNUSUAL EVENT EAL# 5.1 Transportation of externally contaminated injured individual from site to offsite medical facility as deemed appropriate by Shift Supervisor.

ALERT EAL# 5.2 None.

SITE AREA EMERGENCY EAL# 5.3 None.

GENERAL EMERGENCY EAL# 5.4 None.

EP-AD-000-200, Revision 19, Page 11 of 38

Tab 6 EP-PS-324-6 6 - IN-PLANT HIGH RADIATION UNUSUAL EVENT EAL# 6.1 Unanticipated or unplanned concentrations of airborne activity exist in normally accessible areas, which are not due to planned maintenance activities, as indicated by:

Concentrations exceed 500 times the DAC values of 10CFR20 Appendix B, Table I values for a single isotope, or for multiple isotopes where CA C,9

+ +-.>.

CC CN>0 500 DACA DACB DACC DACN ALER EAL# 6.2 Unexpected in-plant high radiation levels or airborne contamination which indicates a severe degradation in the control of radioactive material as indicated by:

Area Radiation Monitor reading 1000 times normal annunciation on Panel 1C601 (2C601) or indication on Panel 1C600 (2C600).

SITE AREA EMERGENCY EAL# 6.3 None.

GENERAL EMERGENCY EAL# 6.4 None.

EP-AD-000-200, Revision 19, Page 12 of 38

Tab 6 EP-PS-324-6 7 - LOSS OF AC POWER UNUSUAL EVENT EAL# 7.1 Loss of offsite power or loss of all onsitd AC power supplies as indicated by:

(A or B)

A. Loss of power to Startup Transformer 10 and 20 annunciation or indication on Panel OC653.

or B. Failure of all diesel generators to start or synchronize to the emergency buses by indication or annunciation on Panel OC653.

ALERT EAL# 7.2 Loss of all offsite power and all onsite AC power supplies as indicated by:

(A and B)

A. Loss of power to Startup Transformer 10 and 20 annunciation or indication on Panel OC653.

and B. Failure of all diesel generators to start or synchronize to the emergency buses by annunciation or indication on Panel OC653.

SITE AREA EMERGENCY EAL# 7.3 Loss of all offsite power and loss of all onsite AC power supplies for greater than 15 minutes as indicated by (A and B and C)

A. Loss of offsite power.

and B. Failure of all diesel generators to startup or synchronize to the emergency buses by indication or annunciation on OC653.

and C. The above conditions exist for greater than 15 minutes.

GENERAL EMERGENCY EAL# 7.4 None.

EP-AD-000-200, Revision 19, Page 13 of 38

Tab 6 EP-PS-3246 8 - LOSS OF CONTROL ROOM ALARMS AND ANNUNCIATORS UNUSUAL EVENT EAL# 8.1 None.

ALERT EAL# 8.2 Loss of all control room annunciators as indicated by:

In the opinion of the Shift Supervisor, all Control Room annunciators and the Plant Process Computer are lost, or insufficient annunciators are available to safely operate the unit(s) without supplemental observation of plant systems.

SITE AREA EMERGE[4CY EAL# 8.3 AJI annunciators lost and plant transient initiated while annunciators are lost as indicated by:

(A and B)

A. In the opinion of the Shift Supervisor, all Control Room annunciators and the Plant Process Computer are lost, or insufficient annunciators are available to safely operate the unit(s) without supplemental observation of plant systems.

and B. (1 or2 or3 or4)

1. Low-Low reactor water level indication on Panel IC651- (2C651) followed by ECCS initiation on Panel 1C601 (2C601).

or

2. Reactor coolant temperature change greater than 100OF per hour indication on recorder TR-1 R006 on Panel 1C007 (2C007) (Reactor Building elevation 683').

or

3. High reactor pressure indication on Panel 1C651 (2C651) and followed by scram indication on Panel 1C651 (2C651).

or

4. Any indication that transient has occurred or is in progress.

GENERAL EMERGENCY EAL# 8.4 None.

EP-AD-000-200, Revision 19, Page 14 of 38

Tab 6 EP-PS-324-6 9 - LOSS OF DC POWER UNUSUAL EVENT EAL# 9.1 None.

ALERT EAL# 9.2 Loss of onsite vital DC power as indicated by:

(A and B)

A. Less than 210 volts on the 250 VDC main distribution Panel buses, 1D652 (2D652) and I D662 (2D662) as indicated by trouble alarms on Panel 1C651 (2C651).

and B. Less than 105 volts on the 125 VDC main distribution buses 1D612 (2D612), 1D622 (2D622), 1D632 (2D632), and 1D642 (2D642) as indicated by trouble alarms on Panel 1C651(2C651).

NOTE: Buses are not tripped on undervoltage condition.

SITE AREA EMERGENCY EAL# 9.3 Loss of all vital onsite DC power sustained for greater than 15 minutes as indicated by:

(A and B and C)

A. Less than 210 volts on the 250 VDC main distribution Panel buses, 1D652 (2D652) and 1D662 (2D662) as indicated by trouble alarms on Panel 1C651(2C651).

and B. Less than 105 volts on the 125 VDC main distribution buses 1D612 (2D612), 1D622 (2D622), 1D632 (2D632), and I D642 (2D642) as indicated by trouble alarms on Panel 1C651 (2C651).

and C. The above condition exists for greater than 15 minutes.

NOTE: Buses are not tripped on undervoltage condition.

GENERAL EMERGENCY EAL# 9.4 None.

EP-AD-000-200, Revision 19, Page 15 of 38

Tab 6 EP-PS-324-6 10 - LOSS OF DECAY HEAT REMOVAL CAPABILITY UNUSUAL EVENT EAL# 10.1 None.

ALERT EAL# 10.2 Inability to remove decay heat while in plant condition 4, inability to maintain the plant in cold shutdown as indicated by:

Inability to maintain reactor coolant temperature less than 200OF with the reactor mode switch in shutdown; exception is when testing per Special Test Exception TS 3.10.1 which allows maximum temperature of 212 0F.

SITE AREA EMERGENCY EAL# 10.3 Inability to remove decay heat while the plant is shutdown as indicated by:

(A and B and C)

A. Reactor Mode switch in shutdown.

and B. Reactor Coolant System temperature greater than 200OF and rising.

and C. Suppression Pool temperature greater than 1201F and rising.

GENERAL EMERGENCY EALU 10.4 Inability to remove decay heat while the plant is shutdown with possible release of large amounts of radioactivity as Indicated by:

(A and B and C)

A. Reactor mode switch in shutdown.

and B. Reactor coolant system temperature greater than 2000F and rising.

and C. Suppression pool temperature greater than 2900 F indicated on the computer output (MAT 12,13,14,15 or 16).

EP-AD-000-200, Revision 19, Page 16 of 38

Tab 6 EP-PS-3246 11 - LOSS OF REACTIVITY CONTROL UNUSUAL EVENT EAL# 11.1 Inadvertent Criticality as indicated by:

Unexpected increasing neutron flux indication on Panel 1C651 (2C651).

ALERT EAL# 11.2 Failure of the Reactor Protection System or the Alternate Rod Insertion System to initiate, and complete a scram that brings the reactor subcritical as indicated by:

(A or B) and (C and D and E)

A. Trip of at least one sub-channel in each trip system (RPS A and RPS B) as indicated by annunciators and trip status lights on Panel 1C651 (2C651).

or B. Trip of both trip systems (ARI A and ARI B) as indicated by annunciators on Panel 1C601 (2C601).

and C. Failure of control rods to insert, confirmed by the full core display indication on Panel I C651 (2C651) or process computer indications.

and D. Failure to bring the reactor subcritical confirmed by neutron count rate on the neutron monitoring indication on Panel 1C651 (2C651).

and E. Reactor power >5% as indicated on Panel 1C651 (2C651).

(CONTINUED ON NEXT PAGE)

EP-AD-000-200, Revision 19, Page 17 of 38

Tab 6 EP-PS-324-6 11 - LOSS OF REACTIVITY CONTROL (continued)

SITE AREA EMERGENCY EAL# 11.3 Loss of functions needed to bring the reactor subcritical and loss of ability to bring the reactor to cold shutdown as indicated by:

(A and B and C and D)

A. Inability to insert sufficient control rods to bring the reactor subcritical as indicated by count rate on the neutron monitoring instrumentation on Panel 1C651 (2C651).

and B. (1 or 2)

Failure of both loops of standby liquid control to inject into the vessel indicated by:

1. Low pump discharge pressure indication on Panel 1C601 (2C601).

or

2. Low flow indication on Panel 1C601 (2C601).

and C. Reactor coolant temperature greater than 200 0F, indicated on Panel 1C651 (2C651).

and D. Reactor power >5% indicated on Panel 1C651 (2C651).

GENERAL EMERGENCY EAL# 111.4 Loss of functions needed to bring the reactor subcritical and transient in progress that makes release of large amounts of radioactivity in a short period possible as indicated by:

(A or B) and (C and D)

A. Trip of at least one sub-channel in each trip system (RPS A and RPS B), indicated by annunciation or trip status lights on Panel 1C651 (2C651).

or B. Trip of both systems (ARI A and ARI B) as indicated by annunciators on Panel 1C601 (2C601).

and C. Loss of SLC system capability to inject, indicated by instrumentation on Panel 1C601 (2C601).

and D. Reactor power greater than 25% of rated, indicated on Panel 1C651 (2C651).

EP-AD-000-200, Revision 19, Page 18 of 38

Tab 6 EP-PS-324-6 12 - LOSS OF REACTOR VESSEL INVENTORY UNUSUAL EVENT EAL# 12.1 Valid initiation of an Emergency Core Cooling System (ECCS) System as indicated by:

(A or B)

A. Initiation of an ECCS System and low, low, low reactor water level (-129) annunciation or indication on Panel 1C651 (2C651).

or B. Initiation of an ECCS System and High Drywell Pressure annunciation or indication on Panel 1C601 (2C601).

ALERT EAL# 12.2 Reactor coolant system leak rate greater than 50 gpm as indicated by:

(A or B)

A. Drywell floor drain sump A or B Hi-Hi alarm on Panel 1C601 (2C601) and 2 or more drywell floor drain pumps continuously running as indicated on Panel 1C601 (2C601).

or B. Other estimates of Reactor coolant system leakage indicating greater than 50 gpm.

SITE AREA EMERGENCY EAL# 12.3 Known loss of coolant accident greater than make-up capacity as indicated by:

Water level below (and failure to return to) top of active fuel for greater than three minutes as indicated on fuel zone level indicator on Panel 1C601 (2C601).

(CONTINUED ON NEXT PAGE)

EP-AD-000-200, Revision 19, Page 19 of 38

Tab 6 EP-PS-324-6 12 - LOSS OF REACTOR VESSEL INVENTORY (continued)

GENERAL EMERGENCY EAL# 12.4.a Loss of coolant accident with possibility of imminent release of large amounts of radioactivity as indicated by:

Water level below (and failure to return to) top of active fuel for greater than 20 minutes as indicated on fuel zone level indicator on Panel 1C601 (2C601).

OR EAL# 12.4.b Loss of Reactor Vessel inventory. Loss of 2 out of 3 fission product barriers (fuel cladding & reactor coolant pressure boundary) with potential loss of the third barrier (primary containment), as indicated by:

(A or B)

A. (1 and 2 and 3)

1. High drywell pressure annunciation or indication on Panel lC601 (2C601).

and

2. (a orb orc)
a. Containment pressure exceeds 40.4 PSIG as indicated on 'Panel 1C601 (2C601).

or

b. A visual inspection of the containment indicates a potential or actual loss of containment (e.g. anchorage or penetration failure).

or

c. Containment isolation valve(s) fail to close as indicated by valve position indication on Panel 1C601 (2C601).

and

3. Reactor Vessel level drops below (and fails to return to) top of active fuel for greater than three minutes as indicated on fuel zone level indicator on Panel 1C601 (2C601).

or B. (1 and 2)

1. Failure of reactor pressure vessel isolation valves to isolate coolant break outside containment as indicated by valve position indication on Panel lC601 (2C601) or visual inspection.

and

2. Reactor vessel level drops below (and fails to return to) top of active fuel for greater than three minutes as indicated on fuel zone level indicator on Panel 1C601 (2C601).

EP-AD-000-200, Revision 19, Page 20 of 38

Tab 6 EP-PS-324"6 13- NATURAL PHENOMENA UNUSUAL EVENT EAL# 13.1 Natural phenomenon occurrence as indicated by:

(A or B or C)

A. Tornado impact on site.

or B. Hurricane impact on site.

or C. Earthquake detected by seismic instrumentation systems on Panel 0C696.

ALERT EAL# 13.2 Natural Phenomenon Occurrence as indicated by:

(A or B or C)

A. Tomado with reported wind velocities greater than 200 mph impacting on site.*

or B. Reported hurricane or sustained winds greater than 70 mph.*

or C. Earthquake at greater than operating basis earthquake (OBE) levels as indicated on Panel 0C696.

  • Telephone numbers for the National Weather Bureau are located in the Emergency Telephone Directory.

(CONTINUED ON NEXT PAGE)

EP-AD-000-200, Revision 19, Page 21 of 38

Tab 6 EP-PS-324-6 13 - NATURAL PHENOMENA (continued)

SITE AREA EMERGENCY EAL# 13.3 Severe natural phenomenon occurrence, with plant not in cold shutdown, as indicated by:

(A and B)

A. Reactor Coolant Temperature greater than 2000F as indicated on Panel 1C651 (2C651).

and B. (1or2or3)

1. Reported hurricane or sustained winds greater than 80 mph.*

or

2. Earthquake with greater than Safe Shutdown Earthquake (SSE) levels as indicated on Panel OC696.

or

3. Tornado with reported wind velocities greater than 220 mph impacting on site.*

GENERAL EMERGENCY EAL# 13.4 None.

  • Telephone numbers for the National Weather Bureau are located in the Emergency Telephone Directory.

EP-AD-000-200, Revision 19, Page 22 of 38

Tab 6 EP-PS-324-6 14 - ONSITE FIRE/EXPLOSION UNUSUAL EVENT EAL# 14.1 Significant fire within the plant as indicated by:

(A and B)

A. Activation of fire brigade by Shift Supervisor.

and B. Duration of fire longer than 15 minutes after time of notification.

OR Explosion inside security protected area, with no significant damage to station facilities, as indicated by:

Visual observation or notification received by control room operator and Shift Supervisor evaluation.

ALERT EAL# 14.2 On-site FirelExplosion as indicated by:

(A or B)

A. Fire lasting more than 15 minutes and fire is in the vicinity of equipment required for safe shutdown of the plant and the fire is damaging or is threatening to damage the equipment due to heat, smoke, flame, or other hazard.

or B. (1 and 2)

Explosion damage to facility affecting plant operation as determined by:

1. Direct observation or notification received by control room operator.

and

2. Shift Supervisor observation.

(CONTINUED ON NEXT PAGE)

EP-AD-00O-200, Revision 19, Page 23 of 38

Tab 6 EP-PS-324-6 14 - ONSITE FIRE/EXPLOSION (continued)

SITE AREA EMERGENCY EAL# 14.3 Damage to safe shutdown equipment due to fire or explosion has occurred when plant is not in cold shutdown, and damage is causing or threatens malfunction of equipment required for safe shutdown of the plant as determined by:

(A and B and C)

A. Direct observation or notification received by control room operator.

and B. Shift Supervisor evaluation.

and C. Reactor Coolant Temperature greater than 2000 F as indicated on Panel 1C651 (2C651).

GENERAL EMERGENCY EAL# 14.4 None.

EP-AD-000-200, Revision 19, Page 24 of 38

Tab 6 EP-PS-324-6 15 - RADIOLOGICAL EFFLUENT UNUSUAL EVENT EAL# 15.1 Any unplanned release of gaseous or liquid radioactivity to the environment that exceeds 2 times the Technical Requirements Manual limits for 60 minutes or longer.

EAL# 15.1 (1 or 2 or 3)

1. Valid Noble Gas vent stack monitor reading(s) that exceeds a total site release rate of 2.OE+6 giCi/min and that is sustained for 60 minutes or longer.

OR

2. Confirmed sample analyses for airborne releases indicates total site release rates at the site boundary with a release duration of 60 minutes or longer resulting in dose rates of:

a) Noble gases >1000 mremlyear whole body, or b) Noble gases >6000 mremfyear skin, or c) 1-131, 1-133, H-3, and particulates with half lives >8 days >3000 mrem/year to any organ (inhalation pathways only).

OR

3. Confirmed sample analyses for liquid releases indicates concentrations with a release duration of 60 minutes or longer in excess of two timbe the Technical Requirements Manual liquid effluent limits.

(CONTINUED ON NEXT PAGE)

EP-AD-000-200, Revision 19, Page 25 of 38

Tab 6 EP-PS-324-6 15 - RADIOLOGICAL EFFLUENT (continued)

ALERT EAL# 15.2 Any unplanned release of gaseous or liquid radioactivity to the environment that exceeds 200 times Technical Requirement Manual limits for 15 minutes or longer.

EAL# 15.2 (1 or 2 or 3)

1. Valid Noble Gas vent stack monitor reading(s) that exceeds a total site release rate of 2E+8 glCi/min and that is sustained for 15 minutes or longer.

OR 2 Confirmed sample analyses for airborne releases indicates total site release rates at the site boundary for 15 minutes or longer resulting in dose rates of:

a) Noble gases >1 .OE+5 mremlyear whole body, or b) Noble gases >6.OE+5 mrem/year skin, or c) 1-131, 1-133, H-3, and particulates with half4ives >8 days >3.0E4-5 mrem/year to any organ (inhalation pathways only).

OR

3. Confirmed sample analyses for liquid releases indicates concentrations in excess of 200 times the Technical Requirements Manual liquid effluent limits for 15 minutes or longer.

(CONTINUED ON NEXT PAGE)

EP-AD-000-200, Revision 19, Page 26 of 38

Tab6 EP-PS-324-6 15 - RADIOLOGICAL EFFLUENT (continued)

SITE AREA EMERGENCY EAL# 15.3 Dose at the Emergency Plan boundary resulting from an actual or imminent release of gaseous radioactivity exceeds 100 mrem whole body TEDE or 500 mrem child thyroid CDE for the actual or projected duration of release.

EAL# 15.3 (1 or 2 or 3 or 4 or 5)

1. Valid Noble Gas vent stack monitor readings(s) that exceeds a total release rate 6.2E8 gtCi/min for greater than 15 minutes and Dose Projections are not available.

Note: If the required dose projection cannot be completed within the 15 minute period, then the declaration must be made based on a valid sustained monitor reading(s).

OR

2. Valid dose assessment using actual meteorology indicates projected doses greater than 100 mrem whole body TEDE or 500 mrem child thyroid CDE at or beyond the EPB.

OR

3. A valid reading sustained for 15 minutes or longer on the RMS perimeter radiation monitoring system greater than 100 mR/hr.

OR

4. Field survey results indicate Emergency Planning boundary dose rates exceeding 100 mR/hr expected to continue for more than one hour.

OR

5. Analyses of field survey samples indicate child thyroid dose commitment at the Emergency Planning Boundary of 500 mrem for one hour of inhalation.

(CONTINUED ON NEXT PAGE)

EP-AD-000-200, Revision 19, Page 27 of 38

Tab 6 EP-PS-324"6 15 - RADIOLOGICAL EFFLUENT (continued)

GENERAL EMERGENCY EAL# 15.4 Dose at the Emergency Planning Boundary resulting from an actual or imminent release of gaseous radioactivity exceeds 1000 mrem whole body TEDE or 5000 mrem child thyroid CDE for the actual or projected duration of the release using actual meteorology.

EAL# 15.4 (1 or2or3or4or5)

1. Valid Noble Gas vent stack monitor readings(s) that exceed a total release rate of 6.2E9 j+/-Cimin for greater that 15 minutes and Dose Projections are not available.

Note: If the required dose projection cannot be completed within the 15 minute period, then the declaration must be made based on a valid sustained monitor reading(s).

OR

2. Valid dose assessment using actual meteorology indicates projected doses greater than 1000 mrem whole body TEDE or 5000 mrem child thyroid CDE at or beyond the EPB.

OR

3. A valid reading sustained for 15 minutes or longer on the RMS perimeter radiation monitoring system greater than 1000 mR/hr.

OR

4. Field survey results indicate Emergency Planning Boundary dose rates exceeding 1000 mR/hr expected to continue for more than one hour.

OR

5. Analyses of field survey samples indicate child thyroid dose commitment at the Emergency Planning Boundary of 5000 mrem for one hour of inhalation.

EP-AD-000-200, Revision 19, Page 28 of 38

Tab 6 EP-PS-324-6 16 - SECURITY EVENT UNUSUAL EVENT EAL# 16.1 Security threat or attempted entry or attempted sabotage as indicated by:

(A or B or C)

A. A report from Security of a security threat, attempted entry, or attempted sabotage of the owner controlled area adjacent to the site.

or B. Any attempted act of sabotage which is deemed legitimate in the judgment of the SHIFT SUPERVISORIEMERGENCY DIRECTOR, and affects plant operation.

or C. A site specific credible security threat notification.

ALERT EAL# 16.2 Ongoing Security Compromise as indicated by:

(A or B)

A. A report from Security that a security compromise is at the site but no penetration of protected areas has occurred.

or B. Any act of sabotage which results in an actual or potential substantial degradation of the level of safety of the plant as judged by the SHIFT SUPERVISOR/EMERGENCY DIRECTOR.

SITE AREA EMERGENCY EAL# 16.3 An ongoing adversary event threatens imminent loss of physical control of plant as indicated by:

(A or B)

A. Report from Security that the security of the plant vital area is threatened by unauthorized (forcible) entry into the protected area.

or B. Any act of sabotage which results in actual or likely major failures of plant functions needed for protection of the public as judged by the SHIFT SUPERVISOR/EMERGENCY DIRECTOR.

(CONTINUED ON NEXT PAGE)

EP-AD-000-200, Revision 19, Page 29 of 38

Tab 6 EP-PS-324-6 16 - SECURITY EVENT (continued)

GENERAL EMERGENCY EAL# 16.4 Loss of physical control of facilities as indicated by:

(A or B)

A. Report from Security that a loss of physical control of plant vital areas has occurred.

or B. Any act of sabotage which results in imminent significant cladding failure or fuel melting with a potential for loss of containment integrity or the potential for release of significant amounts of radioactivity in a short time as judged by the SHIFT SUPERVISOR/EMERGENCY DIRECTOR.

EP-AD-000-200, Revision 19, Page 30 of 38

Tab 6 EP-PS-324-6 17- SPENT FUEL RELATED INCIDENT UNUSUAL EVENT EAL# 17.1 Unanticipated or unplanned concentrations of airborne activity exist in normally accessible areas, which is not due to planned maintenance activities, as indicated by:

Concentrations exceed 500 times the DAC values of 10CFR20 Appendix B, Table I values for a single isotope, or full multiple isotopes where CA. Ca CC CN A D C.DACc CDA Cs Ž'>500 DCDA DAC + ..

ALERT EAL# 17.2 Unexpected in-plant high radiation levels or airborne contamination which indicates a severe fuel handling accident as indicated by:

Refuel floor area radiation monitor reading 1000 times normal annunciation on Panel 1C601 (2C601) or indication on Panel 1C600 (2C600).

(CONTINUED ON NEXT PAGE)

EP-AD-000-200, Revision 19, Page 31 of 38

Tab 6 EP-PS-324-6 17- SPENT FUEL RELATED INCIDENT (continued)

SITE AREA EMERGENCY EAL# 17.3.a Major damage to irradiated fuel with actual or clear potential for significant release of radioactive material to the environment as indicated by:

(A and B)

A. Dropping, bumping, or otherwise rough handling of a new OR irradiated fuel bundle with irradiated fuel in the pool.

and B. (1 or2)

1. Refueling floor area radiation monitor reading 1000 times normal annunciation on Panel I C601 (2C601) or indication on Panel I C600 (2C600).

or

2. Reactor Building vent stack monitoring system high radiation annunciation or indication on Panel 0C630 or 0C677.

OR EAL# 17.3.b Damage to irradiated fuel due to uncontrolled decrease in the fuel pool level to below the level of the fuel as indicated by:

(A and B)

A. (1 or2)

1. Uncovering of irradiated fuel confirmation by verification of significant leakage from spent fuel pool.

or

2. Visual observation of water level below irradiated fuel in the pool.

and B. (1 or2)

1. Refueling floor area radiation monitor annunciation on Panel 1C651 (2C651) or indication on Panel 1C600 (2C600).

or

2. Reactor Building vent stack monitoring system high radiation annunciation or indication on Panel 0C630 or 0C677.

GENERAL EMERGENCY EAL# 17.4 None.

EP-AD-000-200, Revision 19, Page 32 of 38

Tab 6 EP-PS-324-6 18 - STEAM LINE BREAK UNUSUAL EVENT EAL# 18.1 None.

ALERT EAL# 18.2 MSIV malfunction causing leakage as indicated by:

(A and B)

A. Valid MSIV closure signal or indication on Panel 1C601 (2C601).

and B. (1 or2)

1. Valid Main Steam Line flow indication on Panel lC652 (2C652).

or

2. Valid Main Steam Line radiation indication on Panel I C600 (2C600).

(CONTINUED ON NEXT PAGE)

EP-AD-000-200, Revision 19, Page 33 of 38

Tab 6 EP-PS-324-6 18 - STEAM LINE BREAK (continued)

SITE AREA EMERGENCY EAL# 18.3 Steam line break occurs outside of containment without isolation as indicated by:

(A or B or C or D)

A. (1 and 2)

1. Failure of both MSIVs in the line with the leak to close as indicated by position indication on Panel 1C601 (2C601).

and

2. (a orb)
a. High MSL flow annunciation on Panel 1C601 (2C601) or indication on Panel 1C652 (2C652).

or

b. Other indication of main steam leakage outside containment.

or B. (1 and 2)

1. Failure of RCIC steam isolation valves HV-F008 and HV-F007 to close as indicated on Panel 1C601 (2C601).

and

2. (aorborcordoreorf)
a. RCIC steamline pipe routing area high temperature annunciation on Panel 1C601 (2C601), or indication on Panel 1C614 (2C614).

or

b. RCIC equipment area high temperature annunciation on Panel 1C601 (2C601) or indication on Panel 1C614 (2C614).

or

c. RCIC steamline high flow annunciation on Panel 1C601 (2C601).

or

d. RCIC steamline tunnel ventilation high delta temperature annunciation on Panel 1C601 (2C601).

or

e. RCIC turbine exhaust diaphragm high pressure annunciation on Panel 1C601 (2C601).

or

f. Other indication of steam leakage from the RCIC system.

(CONTINUED ON NEXT PAGE)

EP-AD-000-200, Revision 19, Page 34 of 38

Tab 6 EP-PS-324-6 18 - STEAM LINE BREAK (continued)

SITE AREA EMERGENCY (continued) or C. (1 and 2)

1. Failure of HPCI steam isolation valves HV-F002 and HV-F003 to close as indicated by position indicator on Panel 1C601 (2C601).

and

2. (a orb orc ord ore orf)
a. HPCI steamline pipe routing area high temperature annunciation on Panel 1C601 (2C601), or indication on Panel 1C614 (2C614).

or

b. HPCI equipment area high temperature annunciation on Panel 1C601 (2C601) or indication on Panel 1C614 (2C614).

or

c. HPCI steamline high flow annunciation on Panel 1C601 (2C601).

or

d. HPCI steamline tunnel ventilation high delta temperature annunciation on Panel 1C601 (2C601).

or

e. HPCI turbine exhaust diaphragm high pressure annunciation on Panel 1C601 (2C601).

or

f. Other indication of steam leakage from the HPCI system.

or D. Any other un-isolatable steam line breaks.

GENERAL EMERGENCY EAL# 18.4 None.

EP-AD-000-200, Revision 19, Page 35 of 38

Tab 6 EP-PS-324-6 19 - TOXIC/FLAMMABLE GASES UNUSUAL EVENT EAL# 19.1 Nearby or onsite release of potentially harmful quantifies of toxic or flammable material as indicated by:

Visual observation or notification received by the control room operator.

ALERT EAL# 19.2 Entry of toxic or flammable gases into the facility, with subsequent habitability problem as indicated by:

Visual observation, direct measurement, or notification received by the control room operator.

SITE AREA EMERGENCY EAL# 19.3 Toxic or flammable gases enter vital areas, restricting access and restricted access constitutes a safety problem, as determined by:

(A and B)

A. Shift Supervisors evaluation.

and B. Visual observation, direct measurement, or notification received by control room operator.

GENERAL EMERGENCY EAU# 19.4 None.

EP-AD-000-200, Revision 19, Page 36 of 38

Tab 6 EP-PS-324-6 20 - TECHNICAL SPECIFICATION SAFETY LIMIT UNUSUAL EVENT EAL#: 0.1 Abnormal occurrences which result in operator complying with any of the Technical Specification SAFETY LIMIT ACTION statements indicated by:

(A orB orC or D)

A. Exceeding THERMAL POWER, low pressure or low flow safety limit 2.1.1.1.

or B. Exceeding THERMAL POWER, high pressure and high flow safety limit 2.1.1.2.

or C. Exceeding REACTOR VESSEL WATER LEVEL safety limit 2.1.1.3.

or D. Exceeding REACTOR COOLANT SYSTEM PRESSURE safety limit 2.1.2.

ALERT EAL# 20.2 None.

SITE AREA EMERGENCY EAL# 20.3 None.

GENERAL EMERGENCY EAL# 20.4 None.

EP-AD-00O-200, Revision 19, Page 37 of 38

Tab 6 EP-PS-324-6 21 - DRY FUEL STORAGE UNUSUAL EVENT EAL# 21.1.a. Situations are occurring or have occurred during the transport of the irradiated spent fuel to the onsite storage facility, which jeopardize the integrity of the spent fuel or its container as indicated by:

(A or B)

A. Radiological readings exceed 2 Rlhour at the external surface of any transfer cask or horizontal storage module.

or B. Radiological readings exceed 1 Rlhour one foot away from the external surface of any transfer cask or horizontal storage module.

OR EAL# 21.1.b. Situations are occurring or have occurred at the irradiated spent fuel storage facility, which jeopardize the integrity of the dry cask storage system as indicated by:

(A or B)

A. Radiological readings exceed 2 Rlhour at the external surface of any transfer cask or horizontal storage module.

or B. Radiological readings exceed 1 Rlhour one foot away from the external surface of any transfer cask or horizontal storage module.

ALERT EAL# 21.2 None.

SITE AREA EMERGENCY EAL# 21.3 None.

GENERAL EMERGENCY EAL# 21.4 None EP-AD-000-200, Revision 19, Page 38 of 38

Tab 7 EP-PS-324-7 PAR AIRBORNE RELEASES MONITOR CONDITIONS:

- PLANT STATUSIPROCEDURES

- ONSITE RAD CONDITIONS

- STATUS OF RELEASE

- DOSE PROJECTIONS PA-4 NOTES:

1. PA-# CAN BE USED TO REFER TO PROCEDURE STEPS FOR MORE DETAILED INFORMATION ON THE ACTION TO BE TAKEN.
2. DOSE PROJECTIONS DO NOT INCLUDE DOSE ALREADY RECEIVED.
3. TEDE - WHOLE BODY (TEDE) IS THE SUM OF EFFECTIVE DOSE EQUIVALENT RESULTING FROM EXPOSURE TO EXTERNAL SOURCES. THE COMMITTED EFFECTIVE DOSE EQUIVALENT (CEDE) FROM ALL SIGNIFICANT INHALATION PATHWAYS AND THE DOSE DUE TO GROUND DEPOSmON.
4. CDE - COMMITTED DOSE EQUIVALENT TO THE CHILD THYROID.

EP-AD-00-126, Revision 10, Page 1 of 6

Tab 7 EP-PS-324-7 PAR LIQUID RELEASES ENTRY:

INDICATIONS OF A POTENTIAL UQUID RELEASE

-UNISOLABLE RADWASTE TANK RELEASE

-LEAK TO COOLING TOWER BASIN

-LEAK TO SPRAY POND PL-1 EC Values I RADIONUCLIDE (PiCIhI)s CO-60 3E4 Sr-91 2E' MO-99 2E46 Te-132 9E' 1-131 IE_ __ _

1-133 7E-4 1-134 4E-4 1-135 3E-6 7

Cs-I134 IE-Cs-136 GE4 Cs-137 1E-6 I Ba-I139 2E-4 I Ea-140 SE-$

Ba-1411 3E' Np-239 2E45 I

I I

I I

I NOTES I

1. PLY CAN BE USED TO REFER TO PROCEDURE STEPS FOR MORE DETAILED INFORMATION ON THE ACTION TO BE TAKEN.
2. CALLS TO DANVILLE ARE COURTESY INFORMATION CALLS ONLY.

PROTECTIVE ACTION RECOMMENDATION CALLS MUST BE MADE BY DEPIBRP.

Yes

-v RAD PERSONNEL NOTIFY DEPIBRP FOR DOWNSTREAM USERS TO DIVERT WATER SUPPLY &ESTIMATED TIME OF ARRIVAL OF RELEASE AT DANVILLE PL-9 EP-AD-000-1 26, Revision 10, Page 2 of 6

Tab 7 EP-PS-324-7 PUBLIC PROTECTIVE ACTION RECOMMENDATION GUIDE AIRBORNE RELEASES Ol PA-I MONITOR CONDITIONS FOR PAR APPLICATION The following conditions should be continuously evaluated to determine if a PAR should be implemented or changed:

0 Plant status and prognosis for changes in conditions 0 Onsite radiological conditions 0 Status of actual or potential radioactive releases 0 Offsite dose projections or actual offsite radiological conditions 0 Escalation in Emergency Classification (i.e., General)

(Go to PA-2)

PA-2 HAS A GENERAL EMERGENCY BEEN DECLARED?

o YES- If a GENERAL EMERGENCY has been declared, a PAR must be made within 15 minutes of the emergency declaration. The PAR requirement is found in NUREG-0654. (Go lo PA-3)

O NO- If a GENERAL EMERGENCY has not been declared, continue to monitor plant status, parameter trends, and prognosis for termination or escalation of the event. (Go to PA-1)

PA-3 IS THERE A VALID DOSE PROJECTION INDICATING DOSES OF > I REM TEDE OR 2 5 REM CDE CHILD THYROID AT A DISTANCE OF > 2 MILES?

o YES - If the projected doses at 2 miles are > 1 REM TEDE or 2 5 REM CDE child thyroid, then full evacuation (0-10 miles) is recommended.

(Go to PA-5) o NOIUNKNOWN - (Go to PA-4) o PA-4 RECOMMEND EVACUATION 0-2 MILES; SHELTER 2-10 MILES Limited Evacuation (0-2 miles) and sheltering is appropriate for events that are significant enough to cause a General Emergency classification and dose projections are low, unknown, or below full evacuation guidelines.

O PA-5 EVACUATE 0-10 MILES Full evacuation of members of the general public is recommended at this point based on the emergency classification and dose projections.

EP-AD-000-126, Revision 10, Page 3 of 6

Tab 7 EP-PS-324-7 LIQUID E PL-1 ENTRY This section is entered when there are indications of a potential unplanned radioactive liquid release.

Indications of potential unplanned releases include:

  • an unisolable radwaste tank release

O PL-2 CHEMISTRY/ENVIRONMENTAL SAMPLING-DIRECTOR (ESD)

- TAKES AND ANALYZES SAMPLE (Go to PL-3)

PL-3 IS THERE AN UNPLANNED RELEASE TO THE RIVER?

o . YES - An unplanned release to the river has occurred when event-related radioactive materials are released to the river that are not controlled by the release methodologies described in the ODCM and applicable Chemistry procedures.

(Go to PL-4) o NO- If there is no unplanned release to the river, then no notifications are required and monitoring should continue.

o PL-4 RAD PERSONNEL NOTIFY DEPIBRP THAT A RELEASE HAS OCCURRED Depending on which facility is activated, the notification to BRP will be made by the RPC (TSC), Dose Assessment Supervisor, or Radiological Liaison at the EOF.

DO NOT MAKE ANY PROTECTIVE ACTION RECOMMENDATIONS AT THIS TIME.

(Go to PL-5)

EP-AD-000-1 26, Revision 10, Page 4 of 6

Tab 7 EP-PS-324-7 LIQUID (CONT'D)

PL-5 IS RELEASE 2 TECHNICAL REQUIREMENTS. LIMITS (AT THE RELEASE POINT)?

0 YES- Releases are at or greater than Technical Requirements limits when Chemistry determines that the limits are exceeded based on methodologies described in the ODCM and applicable Chemistry procedures.

(Go to PL-6) o NO- If the release is < Technical Requirements limits, then no further notifications are required and monitoring should continue.

o PL-6 RAD PERSONNEL NOTIFY DANVILLE THAT A RELEASE HAS OCCURRED Depending on which facility is activated, the notification to Danville will be made by the RPC (TSC), Dose Assessment Supervisor, 6r Radiological Liaison at the EOF.

DO NOT MAKE ANY PROTECTIVE ACTION RECOMMENDATIONS AT THIS TIME.

(Go to PL-7)

OI PL-7 CHEM/FTD EVALUATES RELEASE VERSUS PAGs The results of the sample analysis are compared to the PAGs for radionuclides in drinking water. The analysis calculates the expected concentration at Danville, taking into account the dilution afforded by the river.

PL-8 DOES RELEASE EXCEED PAGs (AT DANVILLE)?

O YES - If a single isotope exceeds its effluent concentration (EC) value or the sum of EC fractions exceeds 0.85, then a protective action recommendation should be made for downstream water users (e.g.,

Danville) to DIVERT DRINKING WATER supply to a backup supply or terminate user intake until the release has passed.

(Go to PL-9) 0 NO- If the PAGs are not exceeded, monitoring should continue and the State should be notified that no PAR for the liquid release is required.

(Go to PL-10)

EP-AD-000-1 26, Revision 10, Page 5 of 6

Tab 7 EP-PS-324-7 LIQUID (CONT'D) 0 PL-9 RAD PERSONNEL NOTIFY DEP/BRP OF PAR Depending on which facility is activated, the PAR notification to DEP/BRP will be made by the RPC (TSC), Dose Assessment Supervisor, or Radiological Liaison at the EOF. The PAR FORM shall be used to document the PAR.

DO NOT COMMUNICATE THE PROTECTIVE ACTION RECOMMENDATION TO DANVILLE. THE DEP/BRP IS RESPONSIBLE FOR THIS COMMUNICATION AND ANY COMMUNICATION TO OTHER DRINKING WATER SUPPLIERS OR WATER USERS.

0 PL-10 RAD PERSONNEL NOTIFY DEP/BRP No PAR is required. Depending on which facility is activated, the RPC (TSC), Dose Assessment Supervisor, or Radiological Liaison at the EOF shall notify DEPIBRP that no PAR is required.

EP-AD-000-126, Revision 10, Page 6 of 6

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