ML033460390

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Revision 2 to EP-AA-110-301, Core Damage Assessment (Bwr).
ML033460390
Person / Time
Site: Oyster Creek
Issue date: 12/05/2003
From:
Exelon Nuclear
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML033460390 (26)
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Text

EP-AA-110-301 TMI INSTRUCTION.MEMQL qal Date 1a 5a3 Verif: ___B_ Box No. T V T2 k30Q34 C311 C 311 Control Rm Control Room Master Book. OOB M. Mixon JPIC Corporate Spokesperson, JPIC IKON Control Rm Shift Emergencv Director. 00B M. Mixon JPIC Director, JPIC IKON Control Rm Shift Communicator. OOB M. Mixon JPIC Technical Spokesperson, JPIC IKON Control Rm Damage Control Communicator. OOB M. Mixon JPIC Radiation Protection Spokesperson, JPIC Control Rm Operations Communicator. OOB IKON M. Mixon JPIC Administrative Coordinator, JPIC IKON Central File, SOB D. Marshbank PLAIN JPIC Coordinator, JPIC Docu-ment Control Desk,Label S8-:-S-~; IKON

-NRC - .- -1a JPIC Access Controller, JPIC IKON EP Department, NOB-2 R. Brady I OSC Master Book. Rad Field Ops. Svc Bldg Field Monitoring Team 1. Rad Field Ops Svc Bldg T. Berstler 1 T. Berstler OSC Director, Rad Field Ops. Svc Bldg T. Berstler Field Monitoring Team 2. Rad Field Ops Svc Bldg T. Berstler OSC Assistant OSC Director, Rad Field Ons. Svc Bldg

_ T. Berstler TSC Master Book, OSF-l IKON OSC Damage Control Communicator. Rad Field Ops. Svc Bldg TSC Station Emergency Director, OSF-1 ,T.,Berstler, IKON OSC Operations Group Lead, Rad Field Ops. Svc Bldg T. Berstler TSC Director, OSF-I lKON OSC Radiation Protection Group Lead. Rad Field OPs Svc Bldg TSC Logistics Coordinator, OSF-1 T. Berstler IKON OSC Chemistry Group Lead. Rad Field Ops Svc Bldg. T. Berstler TSC Security Coordinator, OSF-I IKON OSC Mech Maint Group Lead, Rad Field Ops. Svc Bldg I TSC State/Local Communicator, OSF-I T. Berstler IKON OSC T&C/Elect Maint Grout) Lead. Rad Field Ops. Svc Bldg T. Berstler TSC Operations Manager, OSF-I IKON OSC Spare Group Leader Binder. Rad Field OPs Svc. Bldg TSC ENS Communicator, OSF-I T. Berstler IKON OSC Shift Dose Assessor. Rad Field Ops Svc Bldg' T. Berstler TSC Operations Communicator, OSF-I IKON RP Shift Dose Assessor, Rad Field OPs Svc Bldg T. Berstler TSC Technical Manager, OSF-1 IKON 1 Training Department, Trng Bldg TSC Technical Communicator, OSF-1 C. Flory IKON NRC Region 1, Label N.McNammara TSC Tech Support Area, OSF-1 IKON NRC Onsite, Svc. Bldg I TSC Maintenance Manager, OSF-1 P. Sauder IKON Simulator Rm Simulator Master Book, Sim Bldg IKON TSC Damage Control Communicator, OSF-I Simulator Rm Shift Emergency Director, Sim Bldg I IKON IKON TSC Radiation Protection Manager, OSF-I IKON Simulator Rm Shift Communicator, Sim Bldg TSC Radiation Controls Coordinator, OSF-1 IKON IKON Simulator Rm Damage Control Communicator, Sim Bldg IKON TSC Radiation Controls Engineer. OSF-1 IKON Simulator Rm Operations Communicator, Sim Bldg.

TSC HPN Communicator OSF-I IKON IKON 'Record Box, SOB + History Pkg. S. Zimmerman JPIC Master Book, JPIC PLAIN IKON I .1.

Please update your file with the attached listed below, destroy the superseded/cancelled document(s). Also, please sign the acknowledgment at the bottom of this memo and return to Debbie Marshbank, Configuration Cntrl., Rm. 135 SOB, TMI.

TC TC/PROC Page Procedure Number Rev Number CLD Change !P Level Copies Memo Only EP-AA-1 10-301 A_ 9 Controlled Copies, Staple, 3 Hole Punch 2 Plain Copies, Staple, 3 Hole Punch Plain Copies, Staple, 3 Hole Punch (TC Dist)

I hereby acknowledge receipt of this memo and have complied with the instructions Signature Date

-7C:7 09117/03

CONTROLLED COPY EP-AA-1 10-301 Exe tbn,,

Nuclear Exekrn. Revision 2 Level 2 - Reference Use 1 of 25

~~~~~~~~~~~~Page CORE DAMAGE ASSESSMENT (WR)

1. PURPOSE 1.1. This procedure provides emergency response personnel with the methodology to estimate the degree of possible core damage at Exelon Nuclear's Boiling Water Reactor (BWR) stations, with the exception of Oyster Creek Generating Station (OCGS). Refer to EP-AA-1 10-302 for methodology to estimate potential core damage for a Pressurized Water Reactor (PWR).

1.2. This Core Damage Assessment process is designed to assist in estimating core damage after an accident with potential clad or core damage conditions, and is intended to provide an acceptable alternative to existing station core damage assessment models and methods utilized by Reactor Engineering to assist in the following:

  • Determining if the fuel barriers are breached to evaluate the appropriate Emergency Action Level (EAL) classification.
  • Providing input on core configuration (coolable or uncoolable) for prioritization of mitigating activities.
  • Determining the potential quantity and isotopic mix of a radiological release to project offsite doses.
  • Predicting the radiation protection actions that should be considered for long term recovery activities.
  • Satisfying inquiries from local and federal government agencies and provide evidence that the utility knows the plant conditions.

1.3. Core damage may be assessed by:

  • Evaluating the drywell radiation levels (and confirmed by evaluating the extent of time the core was uncovered),
  • History of Core Cooling
2. TERMS AND DEFINITIONS 2.1. BWR - Boiling Water Reactor 2.2. Cladding - The outer coating (usually zirconium alloy), which covers the nuclear fuel elements to prevent corrosion of the fuel and the release of fission products into the coolant.

EP-AA-110-301 Revision 2 Page 2 of 25 2.3. Containment Type -

  • Clinton (Mark ll)
  • Dresden (Mark I)
  • LaSalle (Mark II)

. Limerick (Mark II): 764 assemblies Cont. Volume (384,570 ft3) = Suppression Pool (149,380 ft3) + Drywell (235, 190 ft 3)

  • Peach Bottom (Mark ): 764 assemblies Cont. Volume (303,600 ft3) =Suppression Pool (127,800 ft3) + Drywell (175, 800 ft3)
  • Quad Cities (Mark I) 2.4. Core Release Fraction - The fraction of each isotope in the core inventory that is assumed to be released from the core under given core conditions.

2.5. Core Uncovery Time - For BWRs this is the period of time when reactor water level is less than that required for minimum steam cooling, or about >

20% of the core active fuel is uncovered.

2.6. Cladding Failure

1. Also referred to as "Cladding Oxidation", "Gap Release" or "Clad Rupture" in other documents.
2. 100% clad failure refers to the rupture of 100% of the fuel rods in the core. This would result in all fission products contained in the gap space being released to the reactor coolant system.

2.7. Equilibrium - Conditions associated with evaluation of different volumes of liquid or gas that contain concentrations of radioactive materials or hydrogen, when these concentrations are approximately the same. This is normally an extended period of time following accident initiation.

2.8. Fission Products - The nuclei (fission fragments) formed by the fission of heavy elements or by subsequent radioactive decay of the fission fragments.

2.9. Fuel Melt

1. Referred to as "Core Melt," "In-Vessel Melt" or "Over-temperature" damage in reference documents.
2. 100% fuel melt refers to high temperatures in the fuel pellets in 100%

of the fuel rods in the core. This would result in all the fission products contained in the fuel pellet matrix being released to the reactor coolant system.

2.10. Gap - The space inside a reactor fuel rod that exists between the fuel pellet and the fuel rod cladding.

EP-AA-1 10-301 Revision 2 Page 3 of 25 2.11. Gap Release - The release into containment of fission products in the fuel pin gap.

2.12. In-Vessel Core Melt - A condition during a reactor accident in which some of the cladding or reactor fuel melts as a result of overheating the fuel and remains inside the reactor vessel.

2.13. In-Vessel Core Melt Release - A release into containment from the reactor vessel, which assumes the entire core has melted, releasing a representative mixture of radioisotopes.

2.14. Minimum Steam Cooling RPV Water Level (MSCRWL) - The lowest RP water level at which the covered portion of the reactor core will generate sufficient steam to maintain the hottest clad temperature below 15QOoF.

2.15. Minimum Zero-Iniection RPV Water Level (MZIRWL) - The lowest RPV water level at which the covered portion of the reactor core will generate sufficient steam to maintain the hottest clad temperature below 1800OF, assuming no injection into the RPV.

2.16. Shutdown - As defined by station emergency operating procedures.

2.17. Slump - Relocation of molten reactor core during an accident.

2.18. _ Source Term - The amount and isotopic composition of material released or the release rate, used in modeling releases of material to the environment.

2.19. Spiked Coolant - Reactor coolant containing increased concentrations of non-noble isotopes, sometimes seen with rapid shutdown or depressurization of primary system.

2.20. Spiked Coolant Release -The release into containment of 100 times the non-noble gas fission products found in the coolant.

2.21. Subcritical - The reactor condition when the number of neutrons released by the fission is not sufficient to achieve a self-sustaining nuclear chain reaction.

Defined under station emergency operating procedures.

2.22. Suppression Chamber - May also be referred to as Wetwell or Torus. The Large steel pressure vessel containing a large volume of water that acts as a heat sink for the Drywell.

2.23. TID - Total Isotopic Distribution

EP-AA-1 10-301 Revision 2 Page 4 of 25 J 2.24. Vessel Melt-Through

1. Referred to as "Ex-Vessel Melt" or "Melt Release" in reference documents.
2. Core debris is relocated to the primary containment building after the reactor pressure vessel has failed.
3. RESPONSIBILITIES 3.1. The TSC Core/Thermal Hydraulic Engineer shall serve as the Core Damage Assessment Methodology (CDAM) Evaluator.

3.2. The TSC Radiation Controls Engineer shall coordinate radiological and chemistry information with the Core/Thermal Hydraulic Engineer in support of core damage assessment.

3.3. The TSC Technical Manager shall coordinate core damage assessment activities.

4. MAIN BODY 2 4.1. REFER to Attachment 1, BWR CDAM User Guide for instructions on use of the Core Damage Assessment Methodology (CDAM) Software Program.
5. DOCUMENTATION 5.1. A Summary Form and method specific reports are generated by the BWR CDAM Software for use in documenting the results of the assessment.
6. REFERENCES 6.1. NEDO-22214, Procedures for the Determination of the Extent of Core Damage Under Accident Conditions 6.2. NEDC-33045P, Rev 0 (July 2001), Methods of Estimating Core Damage in BWRs 6.3. WCAP-14696 (July 1996) Westinghouse Owners Group Core Damage Assessment Guidance.

6.4. WCAP-14696-A (November 1999), Westinghouse Owners Group Core Damage Assessment Guidance.

6.5. NUREG-1228, "Source Term Estimation During Incident Response to Severe Nuclear Power Accidents" I

EP-AA-1 10-301 Revision 2 Page5 of 25 6.6. Station Commitments  ;

6.6.1. Peach Bottom CM-1 T04511 (Attachment 1, 5.6) 6.6.2. Limerick Bottom CM-2 T04512 (Attachment 1, 5.6)

A.'

7. ATTACHMENTS 7.1. Attachment 1, BWR CDAM User Guide

EP-AA-1 10-301 Revision 2 Page 6 of 25 Attachment 1 BWR CDAM User Guide Page 1 of 20

1. OVERVIEW 1.1. As a windows based application designed in Microsoft Access, BWR CDAM, uses many standard user interfaces. Instructions are not provided in basic computer operations in the windows) environment. The user must be familiar with these to efficiently operate the program.

1.2. It is also assumed user is familiar with basic reactor physics and core damage fundamentals. Emergency Response Organization training will provide an overview of core damage assessment methodologies.

1.3. The program should be used by qualified personnel as a tool to estimate the type and amount of core damage.

2. DETERMINE APPROPRIATE AND AVAILABLE ASSESSMENT METHODS Mid-West Region Stations lREFER to EP-MW-1 I0-1 001 or a listing of appropriate plant parameter points to be used following a LOCA.

2.1. The magnitude and type of event, transport mechanism and time after shutdown will be influencing factors on the method(s) utilized to determine the extent of core damage. Damage estimates can be developed using one or more methods as they become available or applicable.

2.1.1. Indications Of Core Damage

1. The primary indicators of core damage that are available during the early phases of an event:

- Drywell/Containment Radiation Monitor Readings

- Drywell/Containment Hydrogen Readings

2. Auxiliary indicators that are used to confirm and better define the possible type of damage are:

- Reactor Pressure Vessel Level Indication System readings

- Estimation of maximum temperature reached within the core

- Estimated core uncovery time

- Abnormal Source Range Monitor readings

EP-AA-1 10-301 Revision 2 Page 7 of 25 Attachment I BWR CDAM User Guide Page 2 of 20

3. Long Term Indicators (once liquid or gaseous samples can be safely obtained) are:

- Isotopic Ratios

- Presence of high levels of rare isotopes

- Quantity of isotopes present in samples 2.1.2. SELECT the assessment method(s) most appropriate for the existing conditions. Methods available for assisting in the determination of the extent of core damage include the following:

Method Use Comment Containment Early Indication of Core Uncertainties due to variables in release Radiation Monitor Damage of fission products from RCS and effects of containment sprays.

Core Conditions Indication of onset of May not be reliable during later phases of Core Damage core overheating due to changes in core geometry.

RPV Level Indication of Core Indicates possible damage not useful in Uncovery estimating the quantity of damage.

Source Range Indication of Core Loss of water level leads to increase in Monitor Uncovery gamma detection.

Containment Early Indication of Core Significant uncertainties due to variable Hydrogen Monitor Damage Hydrogen generation in core and in release of Hydrogen from RCS and effects of containment sprays.

RCS Samples and Late Indication of Core Very large uncertainties until all systems Containment Sump Damage -Suppression have reached equilibrium. Useful in and Atmosphere Pool Samples provide planning long term recovery.

Samples indication of Rx Vessel Failure

3. START UP THE CDAM PROGRAM 3.1. ACCESS the application by one of the following:

3.1.1. OPEN the BWR CDAM desktop icon on applicable computers.

1. START the BWR CDAM program for the plant that has declared an emergency. Programs are labeled BWR CDAM.
2. SELECT the appropriate icon or run from the 'start bar' and type in the file path and name as follows C:ACDAM\BWR CDAM.MDB

EP-AA-1 10-301 Revision 2 Page 8 of 25 Attachment I BWR CDAM User Guide Page 3 of 20 3.1.2. If the assigned Core Damage Assessment Computer cannot access the application or the CDAM program will not run, then install BWR CDAM on any computer from CDs or Disks located in the TSC or the EOF Library.

1. INSTALL CDAM by copying appropriate file to computer's hard drive.
2. UPDATE the "properties" of the file by deselecting write protection.
4. SELECTION AND PERFORMANCE OF ASSESSMENT 4.1. SELECT the assessment method(s) most appropriate for the existing conditions. Methods available for assisting in the determination of the extent of core damage include the following:
  • Containment Radiation Analysis - (Section 5.2)
  • Core Conditions Analysis (Cooling History) - (Section 5.3)
  • Containment Hydrogen Analysis - (Section 5.4)
  • Nuclide Analyses (Ratios and Abnormal Isotopes) - (Section 5.5)
  • Liquid Samples Analysis - (Section 5.6)
  • Gaseous Samples Analysis - (Section 5.7)

Basic Program Flow Diagram

EP-AA-1 10-301 Revision 2 Page 9 of 25 Attachment 1 BWR CDAM User Guide Page 4 of 20

5. PROGRAM SCREENS AND INPUTS 5.1. When the program is started the following screen appears:

NOTE: The value boxes are empty when the program is originally launched. The examples below may deviate from the CDAM displays during use due to different software versions in use in the Mid-Atlantic and Midwest regions. The display differences do not Mid-Atlantic versi on lists impact the functionality of the program. Where Limerick, Peach Bottom station title differences exist, the titles applicable to (and Oyster Cree !kwhich the Mid-Atlantic stations are contained in "( )."

is currently rI01 applicable'

-, E

-.;-+;f  ? ?e v. 9 g,,<

..cDresden rFLasalle  : .Quad Cties'

  • N,< b>  ?

- v < - ? o: .v 5t a ?,: S >*, u x )< ? b < Xy - .ffi-szs mtnt z.~~~~ . Mbt, Melt Clad w>.W

..... ..... C.

' . ' ? ? ^ > ?:; .. ' t ?- . -. 2 . _ . 4

  • s _ f > ? ¢ ', ? ? y R4 ??,t ^ f 3 i .

'Rad Monitors See c%3 7.X_

g;Af 3A >ie3f <?* - IZ .iRIf

_'X';_ .Cre^onitons._',-. Core Cooling

  • 'COntainmnent:

m- II - I --

. -- .1-I II

,g e Uncovery Tine: I -. 1,

. I ii i SRNCount Rate:- I . -i I

See 5.5 , 4*

Core- Temp:\ II - -, -- - "- I.'. . 1::

. - II .1. . i<4 E.kn~ I e

t .

.i

-:: e .

ottydrcogen H.',.

- Zi- - *

,.,^f

'^=<,

^ 'l 3

fi

~~NuikeAT t Nucide Analmysis a ?eRios .-

I I

Ii 8? e , ?<3"- C. Sf SAbnorma1 Isotopes:.l

_f3G& 'aeri est--- C .,

mls.>l.?k

  • ~~~~~~~~~  ; ...................

S.. 7-v Iiui..

-~7TI See 5.8 I . I. I '

" , - .- I_ -' ,

EP-AA-1 10-301 Revision 2 Page 10 of 25 Attachment I BWR CDAM User Guide Page 5 of 20 CAUTION Selecting an "Affected Station" resets all inputs to default values.

5.2. SELECT the Affected Station before other Assessment Methods."

CAUTION Pressing the "Quit" button exits the program. When the program is closed all data is reset. Program saves no information to disk; printed reports serve as record of core damage assessments.

5.3. Drywell/Containment Radiation Monitor Method 5.3.1. PRESS the "Cont Rad Monitors" button on the Summary Screen to open the following form:

I i.W See 5.3.3

. ;CortSpraysOlf 'f i~ConSjrayisOn. 1TimesiinceS/D hrs ' 12.0 Monitor IR/hrd . Assessment Results Ml-lad C.5.2.0E+03', -- D 'e Estiriate:~4 Preliminary results I ee5..2 l1..OE.03 di (affect of

~~~ Readin~~~~loby h ER/Hr): 1F.70E+051:8l.i1E--L0 input data) are shown'

-vad--g i- used rtthedamae1~Raigf/i 7l+3 81 .D here.

Contanmt I Dang~timat~>,<Z 5 Njot:e: Th Jtight rnbnitored or ' - . '

lejiiWaied reading Withi' 100/ IR 2.21 E.051 811 E+03 t-ootain en i;used fr :th

~dmage'assesment Co~ua~n 1~ ReadigRIr221E0 81E.O A See 53.8 1 I ~~r~~wer~~ph' t~~raphy j !h' e

EP-AA-1 10-301 Revision 2 Page 11 of 25 Attachment I BWR CDAM User Guide Page 6 of 20 NOTE: Program allows entry from 2 high range monitors for Drywell location and 1 for Torus or Containment /

Suppression Chamber, however a reading may be entered from any monitor or measurement taken external to suppression chamber, which accurately indicated containment radiation levels. If two entries are made only the highest is used.

5.3.2. ENTER the highest Drywell radiation monitor reading that occurred in these boxes

1. If Drywell radiation monitor readings are not available, then enter the containment / Suppression Chamber radiation monitor reading.

5.3.3. SELECT Drywell/Containment Spray status:

1. If the Drywell/Containment Spray system was operated for the majority of the time since the estimated time of the onset of core damage then choose "Drywell Spray On."
2. If the Drywell/Containment Spray system was not operated or only operated briefly (e.g., <10% of time since the estimated time of the onset of core damage) then choose "Drywell Spray Off."

5.3.4. ENTER the time after reactor shutdown, which corresponds the time the containment radiation reading was taken. Value must be between 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after shutdown, which corresponds to the time period in which this method is considered effective.

NOTE: Pressing "Reset" button resets values on this form only.

5.3.5. PRESS "Containment Graph" or "Supp Chamber Graph" button to display a screen similar to the following:

(See example display on next page.)

EP-AA-1 10-301 Revision 2 Page 12 of 25 Attachment I BWR CDAM User Guide Page 7 of 20 i,9i<1E48 - .....[th~s graph is appeoprna'e ol' Use ..... a...................... .Mox,,,R d .

r_E=-...

°i - - ::*  :: .=22E02 l w-1E. s

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....... .-. .-.--  : ~-::  :. ........... ...... ...... . . ..... 4......a........ ..-.-...................-...-.

..... S..... 5......

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i,c 14. -; zw- - iut-;x-- Iz, ... '. 2 2___________L___._.____._______.

r .. :P I: nn pa ___O

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.. f 4,1^+3  ;.... ..f:.:-- .,-- .e.:..... .-....... . ...... ot-r

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C -1E l 1.=;;;=;; -; ........................................

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=; ..:........:::.. - .-....  ;...,...

sl42 i ,,t>g--,,,:,: -c---

a ..........

'.1H'^iso<>vj3st................. P.4

...tt;ij ,6z<

.............. . .... .... . . - ~ Se 53.

NOTE: Graph shows high and low containment radiation levels which correspond to 100% Melt or Clad or 1%

Melt or Clad damage. A dot shows the last containment radiation level entered into the program for assessment.

5.3.6. PRESS the "Print" button to print a report of containment radiation method inputs and best estimate of damage.

5.3.7. PRESS the "Done" button to return to the Containment Radiation Monitor Evaluation Screen.

5.3.8. PRESS the "BACK" button to return to the Summaryv Screen.

5.4. Core Conditions Methods NOTE: Each of these four methods is an independent assessment method.

5.4.1. PRESS the "Core Conditions" button on the Summary Screen to open the following form:

(See example form on next page.)

EP-AA-1 10-301 Revision 2 Page 13 of 25 Attachment 1 BWR CDAM User Guide Page 8 of 20 R- a____ 7 N a X F-s( ?e 5.4.2 RPV Water I ela(nclhesl , Core Uncoverime fioursl- ^ Cre Temieraturei (rF)l

.:C4S~w'(;-1F 25 6 6

'Assessment Results 4200 t/ .,zxa Thte cote i jpartiallr uncovewd e3600;-.

but i~ coole by,2teain. lad ltemrer.tuiez ate perted to 3000 /

Jernror below 1500 F Ho cage.

damrag'e i cirpected.

[ -Se e 5.4.3 So urce Range lonjfk Rte~ U-*-- .5 CoreTemrnre 10 SRM oxNrmal. NoF et Assessment Results . - Assessment Results Assessment Results-

.0 to h hour..4nrluicty eletracen I~F. arid 24OOW f.

Thecae hac ernaine d covers . fe No coedAz] . . Veny mpird Zirc.Walfr eation.

.Locail damagre may have, espected. .It9droueil Is felazsed and the fuel occuired do to othei ev'ents. no Dire darreipe irccipected- cladinr Iais.

.oee 5.4.5 I

Puint I

5.4.2. Under Reactor Pressure Vessel (RPV) Water Level ENTER the lowest recorded (or estimated) RPV level (range 0 to -350 inches) and core spray flow at time of lowest reading NOTE: Steps 5.4.3 through 5.4.6 are based on inputs from Reactor Operators, TSC Staff and other engineering personnel (including outside sources such as General Electric personnel).

5.4.3. Under Source Range Monitor REVIEW plant parameter history and if the SRM (Wide Range Monitor at Peach Bottom) had indications of a reading 10 times those expected check "Yes."

5.4.4. PRESS the "Core Levels" button to view information regarding water levels associated with the Station reactor and vessel level indications.

(See example form on next page.)

EP-AA-1 10-301 Revision 2 Page 14 of 25 Attachment 1 BWR CDAM User Guide Page 9 of 20

. , .~, .- .

X=X_Bsot' s evic rat on Top of Active Fue) a Minimum Ste'ri Coolng RxSta Cool RLx We-186

~~""~ VMinimum Zerolnjiection Rxlvai& LeveI:' -201 6 Au~cti~ '-24 Bottom- ofA tiv u -0 FBack, 5.4.5. ENTER the estimated time the reactor core (20% of top of active core) was uncovered without steam (level below the Minimum Steam Cooling Rx Water Level) or spray cooling reactor core.

5.4.6. ENTER the estimated highest temperature reached in the reactor core.

5.4.7. PRESS the "Print" button to print a report of inputs and results of core temperature methods of core damage assessment.

5.4.8. PRESS the "Back" button to return to the Summary Screen.

5.5. Containment Hydrogen Evaluations CAUTION This CDAM assumes no ignitor operation. Ignitor use limits containment hydrogen concentration affecting the reliability of this method.

EP-AA-1 10-301 Revision 2 Page 15 of 25 Attachment 1 BWR CDAM User Guide Page 10 of 20 5.5.1. PRESS the "Cont Hydrogen" button on the Summary Screen to open the following form:

/#s.this e h. grap e> s

.~~~~s 'f  ? . 'X 'e.appropae v';,A oniy-afleia 9 LOhas ' occured

, 4 s, 5Drv ',e~

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',P9 '.,ztMAe Cla ?Axe .~4~t . , ,{.i'-U-exs O.

X See 5.5.2 ~Cnann

. . . > < 8 . X *, 9 >,

  • t 4*+z+.* ***xt *t4t %l -  ? .Z 25, 8x -.  ;-  ;

See 5.5.3veag 3 x~~~~~~* 4 * *s4 * **e b i e Eo ib um ~...

Lt A 4 r 15 . ___ *>**  ; 4Z>>itr O idized§>gtt s<

.*~~4 ****k~ See-P AX q ~~444 *N * * * * -% Me t:¢bs4 3Xp>JS>3§ 0 5~~~~~0 4, i~~See 5.t-.-Y' < Ki

~

below. ~ ~ ~ [~ l 5.5.. ENTER highest Drywell and/or Suppression Chamber hydrogen level measured.

NOTE: Suppression Chamber reading can only be entered if user selects 'Ono" under Equilibrium in step 5.5.3 below.

5.5.3. SELECT the applicable System Equilibrium status based on the following:

1. If Containment and Suppression Chamber monitors read the same or only atmospheres are assumed equalized, then SELECT Yes" for equilibrium.
2. If containment and suppression chamber atmospheres are not in equilibrium or only containment H2 reading is available, then SELECT "No" for equilibrium.

5.5.4. PRESS the Print" button to print a report of inputs and results of core level methods of core damage assessment.

EP-AA-110-301 Revision 2 Page 16 of.25 Attachment I BWR CDAM User Guide Page 11 of 20 5.5.5. PRESS the "Back" button to return to the Summary Screen.

5.6. Nuclide Analysis (CM-1, CM-2) 5.6.1. PRESS the "Nuclide Analysis" button on the Summary Screen to open the following form:

5.6.2. ENTER the time since reactor shutdown (time between shutdown and sample being drawn).

5.6.3. ENTER isotopic sample results in uCi/cc. Sample results are to be decay corrected back to time after shutdown that the sample was drawn.

1. Noble Gases are ratioed to Xe-133
2. Halogens are ratioed to 1-131 5.6.4. If the ratios evaluated above are greater than predicted melt ratio, then melt damage is predicted 5.6.5. If the ratios evaluated above are less than clad ratio, then clad damage is predicted.

EP-AA-1 10-301 Revision 2 Page 17 of 25 Attachment 1 BWR CDAM User Guide Page 12 of 20 5.6.6. If abnormal levels of rare isotopes are present then check "Yes" and check which isotopes are present.

5.6.7. PRESS the Print" button to print a report of inputs and results of core level methods of core damage assessment.

5.6.8. PRESS the "Back" button to return to the Summary Screen.

5.7. Liquid Samples 5.7.1. PRESS the "Liquid Samples" button on the Summary Screen to open the following form:

~~~ _

Sampe TypeLocafion ________ ower'Histoo~y§~-

i 131 (Short lied d .: 14 of Days in Periodl >Avg

. Power- 1) 1095i 0 Reactor> Coot 4? St ___ See 5.7.6 e tand 6Pps ihT Rw 6 *i5R§>t>a l Ai I omao 5-t 1;~1 -8See sBs<5.7.37 See5.7.4~

~ ~ ~ ~ ~ ~~~ee5..

tss>- - t~~~~~tRetorndr'1 E4 V k Il,igh-~* of

-" ~~~

~~~~~ Bad~~~~ see 5.7.17

  • ~~~~~~~~~~~~~~~ 61 U s;+Sd _

a Tihe ftlS5D.hrf: - I 2.20E+01f JlT 77 < . sBe: > Ihi +rS_ S Se 5.7.9.+.,

5.7.4. ENTR Sample Inormation

-~aet be dea corete bac to tie ater shudw that tev52sample 5.2 1.

SELctivt apria isotopsaerslsi ~/c(~/l.Sml eut 5.7. 2. Sample 5.7.4.

Tim NTEeIngrawn EAfe SI (raco shton is theorit timsbtwenshtdwnan are to be decay corrected back to time after shutdown that the sample was drawn

2. Time After S/D (reactor shutdown) is the time between shutdown and sample being drawn.

5.7.5. SELECT the appropriate System Equilibrium status:

EP-AA-1 10-301 Revision 2 Page 18 of 25 Attachment I BWR CDAM User Guide Page 13 of 20

1. If sample was taken from only one location and systems are in equilibrium, then check "yes" for "Systems in Equilibrium," otherwise check "no."

5.7.6. ENTER power history (past to present, i.e. oldest steady state history as record number) of core since last refueling. Shutdown times are entered as the number of days with Ave Power (%) set at 0.

1. For short-lived isotopes, EXTEND Power History at least 30 days.
2. For long-lived isotopes, EXTEND power history at least 100 days, however the power history for the extent of the cycle is preferred.
3. LIMIT variations in steady state power to +/- 20% within each operational period entered.

5.7.7. Once all data has been entered, PRESS the "Calculate" button to display the

% Damage Estimates.

5.7.8. PRESS the "Volumes" button to display the follow screen:

,, See 5.78.2 S~

~ ~ ~ Isfii??s<o

?X@89 ~~upe5- "r Ch b Ligtid [rml i _

Cnainmenit Atmoshr (c 44E0

-- L?8, iSupreinChamber Atmosphercc 3.E eD esden Staiione

Res~Back See 5.7.8 See 5.7.8.4 N
1. Program enters default RCS volume, which the user may change based on RPV Level Readings at time of sample.
2. Program enters default Suppression Chamber volume, which the user may change based on readings at time of sample.
3. Program enters default Containment free air volume which user may change based on conditions at time of sample. Unless there has been significant flooding of drywell this value will not change.

EP-AA-1 10-301 Revision 2 Page 19 of 25 Attachment 1 BWR CDAM User Guide Page 14 of 20

4. Program enters default Suppression Chamber free air volume which user may change based on conditions at time of sample. If there has been a significant increase or decrease in the water level in the Suppression Pool or Torus then the free air volume will change.

NOTE: Pressing the "Reset" button will reset all volumes to default values.

5. PRESS the "Back" button to return to the Liquid or Gaseous screen, which user used to call volume form.

5.7.9. PRESS the "Graph" button to display the following screen:

b §>-s§

<>< (>%PFel i lt 2$ '.sjoi i s k'S - 10t < amp e lnformtion

-17

____________iI_l_ orrected Actlivty d

<1E5' > . . - _-- '£Sample Location ,

Reactor Coolant E IE+i eC in Equbiibium

.X.->System

, fuj3 ,,., w es, >____________ ' 'ti>iYes~ O loO/ . ~A v - \ e>' i > ................... f.. .^ Damae Estimates_,

. 1 E 2 ........... M lt; Clad ltX.

< vnXEr1 ................. .l.Best- l OLi l iZ t..e.'g-

0. l Lrzw il^'-i i' I E1 _ See5.7.9.2_l (See Note on next page.)

EP-AA-1 10-301 Revision 2 Page 20 of 25 Attachment 1 BWR CDAM User Guide Page 15 of 20 NOTE: Graph on previous page shows High, Low, and Best melt curves; High, Low, and Best clad damage curves, and a red line across graph indicating entered corrected sample activity.

1. PRESS the Print" button to print a graph and summary of inputs.
2. PRESS the "Back" button to go back to liquid or gaseous form which called this form.

5.7.10. PRESS the "Back" button to return to the Summary Screen.

5.8. Gaseous Samples 5.8.1. PRESS the "Gas Samples" button on the Summary Screen to open the following form:

_s=Ad b fwAe

.Sample;ype/Locati~n ~ ~owe¢^Histo.

~~ $ hd V~~L Diri Pefl6dl~v oe ~

~~Cont~tosP1~Sup~hamb&AM-OS~ .Bth~ Ifi t D~~~~~~~~~~~~~~~~~19o -Rr0;a

Sample lnformaion ^ Se 5.8.5

.Actviy fpCi'cc' 2. OE IL m;

if,> k[ , x Dord 'ij ~; i I lP o TimeAlter /D hi.OEO

,<Sytemres (p1-N 123E+O2 [l -TTf amage Estimate:

Sy¢t>4<emremp(9.

Sxle espsg f

ZOOE+

illt T -l Highet S

Clad [t t.

VOlumes mple~ 8.70E__01_

Best: L.........J Grap ..

SY!stem inEtaliNukLowest:

_ .Effi* tB S _

mBc l~~~~~~~~~~~~~~~See 5.8.9l 5.8.2. SELECT appropriate isotope.

5.8.3. SELECT and sample location.

1. If samples are available form both locations, then SELECT "Both" option.

EP-AA-1 10-301 Revision 2 Page 21 of 25 Attachment 1 BWR CDAM User Guide Page 16 of 20 5.8.4. ENTER Sample Information:

1. ENTER sample activity for selected isotope in uCi/cc (uCi/ml). Sample results are to be decay corrected back to time after shutdown that the sample was drawn
2. ENTER Time After S/D that sample was taken.
3. ENTER the pressure and temperature of the system sampled
4. ENTER the end pressure and temperature of sample.

5.8.5. ENTER power history (past to present, i.e. oldest steady state history as record number 1) of core since last refueling. Shutdown times are entered as the number of days with Avg Power (%) set at 0.

1. For short-lived isotopes, EXTEND Power History at least 30 days.
2. For long-lived isotopes, EXTEND power history at least 100 days, however the power history for the extent of the cycle is preferred.
3. LIMIT variations in steady state power to +/- 20% within each operational period entered.

5.8.6. Once all data has been entered PRESS the "Calculate" button to display the

% Damage Estimates.

5.8.7. PRESS the "Volumes" button to display the following screen (same as 5.7.8):

See 5.8.7.1 Cnr'mentAtmo phefe (cc] 4__47E__9 - See 5.8.7.

i, '-  :'i: i;OChame phr (c]

'jiAtmo I;33E0

- -Stpe3f ot See 5.8.7.3 79 l S.8t7.5 Not l/

DrSee -Reet Sc e 5.8.7l

1. Program enters default RCS volume, which the user may change based on RPV Level Readings at time of sample.

EP-AA-1 10-301 Revision 2 Page 22 of 25 Attachment 1 BWR CDAM User Guide Page 17 of 20

2. Program enters default Suppression Chamber volume, which the user may change based on readings at time of sample.
3. Program enters default Containment free air volume which user may change based on conditions at time of sample. Unless there has been significant flooding of drywell this value will not change.
4. Program enters default Suppression Chamber free air volume which user may change based on conditions at time of sample. If there has been a significant increase or decrease in the water level in the Suppression Pool or Torus then the free air volume will change.

NOTE: Pressing the "Reset" button will reset all volumes to default values.

5. PRESS the "Back" button to return to the Liquid or Gaseous screen, which user used to call volume form.

5.8.8. PRESS the "Graph" button to display the following screen:

,,~~~~~i M~f-, -721 E2 , . amp.FulMA9..:10t Xc  ? 5_?_ *

,________________ .Sample Locati

____________ - - ~~~Contamnnent oph~

_______,_____ 5 - .- >X 0S ystem in Equilibiu V .

I~ ~ ______ . ________ %Damage Estimates  :

. ES1

,; 1 ee 58. Se 5

. I= Bad'

-A1E-2-. . ..

NO4TE Graph shw Hih Low an Bes met-curves; ¢ 2l-lSee 5.8.8.1 l e5...

NOTE: Graph shows High, Low, and Best melt curves; High, Low, and Best clad damage curves, and a red line across graph indicating entered.

EP-AA-1 10-301 Revision 2 Page 23 of 25 Attachment I BWR CDAM User Guide Page 18 of 20

1. PRESS the "Print" button to print a graph and summary of inputs.
2. PRESS the "Back" button to go back to liquid or gaseous form which called this form.

5.8.9. PRESS the "Back" button to return to the Summary Screen.

6. CORE DAMAGE

SUMMARY

REPORT 6.1. Once the program user enters data for all available assessment methods and the program calculates damage based on inputs, SELECT the "Print" button to print a summary of all methods used.

6.2. The values presented in the Assessment Methods section of the summary report show that they are in percent (%). Containment Hydrogen values are also in percent (but do not show the % symbol)..

(Sample report on next page.)

EP-AA-1 10-301 Revision 2 Page 24 of 25 Attachment I BWR CDAM User Guide Page 19 of 20 CDAM Method: Core Damage Summary Station: Clinton 0 Dresden 0 LaSalle En Quad Cities Assessment Methods: Melt Clad Containment Radiation Monitorse Containment: 2 7 l 7-W Supperssion Chamber: <1%

Core Conditions Core Cooling: Clad Damage Core Unco/ery Time: lNo Core Danage SRM Count Rate: No CoreDanage Core Temp: Clad Failure Containment Hydrogerl <1 l208 Sample Analysis Ratios: Fuel Melt Abnormal Isotopes: 6 of 19 Present RCS: Liquid Samples: l80% l 0%

Chamber: Gas Samples: 23 l 100 l

  • These methods should NOT be used for qualdtattve or quantitative assessment except in be case of a LOCA Ana"'s Estimate:

0 No Core Damage 3 Cladding Failure 0 Fuel Melt Amount:

NRC Core Condition Category: l Degree of Minor Intermediate Major Degradation (c,10°) (10°W -50%i (>5onp%)

No Core Damage 1 1 1 Cladding Failure 2 3 4 F el Overheat 5 6 7 F uel Melt 8 9 10 Generated Dv:

Name: Date: 1 2f0O5102 Time: 8:29 AM Core Damage Summary Exelon BWR CDAMvI.0 J

EP-AA-1 10-301 Revision 2 Page 25 of 25 Attachment I BWR CDAM User Guide Page 20 of 20 6.3. The Individual tasked with assessing core damage shall then ANALYZE the report to determine best estimate of type and amount of damage.

NOTE: The CDAM program does not use the Fuel Overheat Condition Category 6.4. Based on estimated type and amount of damage and following table (table also printed on summary report) ASSIGN NRC Core Condition Category (1-4 or 8 -10).

NRC Core Condition Categories Degree of Minor Intermediate Major Degradation (<10%) (10% to 50%) (>50)

No Core Damage 1 1 1 Cladding Failure 2 3 4 Fuel Overheat 5 6 7 Fuel Melt 8 9 10

7. QUITING, OR EXITING, THE PROGRAM NOTE: When the program is closed all data is reset.

CAUTION Program saves no information to disk; printed reports serve as record of core damage assessments.

7.1 . PRESS the "Quit" button on the Summary Screen exits the program.

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