ML20111C271
| ML20111C271 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 01/02/1985 |
| From: | GULF STATES UTILITIES CO. |
| To: | |
| Shared Package | |
| ML20111C255 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.3, TASK-TM COP-1050, NUDOCS 8501090232 | |
| Download: ML20111C271 (26) | |
Text
- ,a
's ENCLOSURE 4 RBS Procedure COP-1050
" Post-Accident Estimation of Fuel Core Damage"
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B501090232 850102 i
~PDR-ADOCK 05000458
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, .o RIVER BND STATION i
AFFE0 VAL SEEET
- s. STATION OPERATING PROCEDURES s
NO. cop 1050 TITLE POST ACCIDENT ESTIMATION OF FUEL CORE DAMAGE i l SAFETT IIIATED TES M NO TECE ICAL REVIEW EEQUIRED TRS NO REY. FACES IMIEF. TECE.
- 30. ISSUED REVIN REVIW AFFROVED BT EFFECT
, SIGRATURE/DATE SIGEATURE/DATE SIGE&TURE/DATE DATE O 1 THRU 23
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. (LATER'S)
PROC. NO. COP-1050 TITLE POST ACCIDENT ESTIMATION OF FUEL CORE DAMAGE 0 - I REV. NO.
i The following is a list of " missing" ano/or " tentative inforniation" g
-- (LATER'S) -- in this procedure. The responsible Section Superviser i shall update this procedure as the inforraation becomes available, j (LATER) LOCATION 5ECTION/ m P PAGE BRIEF DESCRIPTION I
'3.2 2 COP-1001 NOT ISSUED 3.3 2 COP-1002 NOT ISSUED !
3.9 2 COP-0425 NOT ISSUED j 3.11 2 COP-1033 NOT ISSUED I,
)
I 3.13 2 COP-103d NOT ISSUE 0 1 4.1.4 3 REACT 0k COOLANT V01.UME 4.1.8 3 ,,, TOTAL F. ASS OF ZIRCONIUM
, ATTACHMENT 3 9 TOTAL PASS OF ZIRCONIUM REACTOR COOLAf;T ff:P 5bFFRE55 ION .
ATTACHMENT 4 ,
10 POOL VOLUME I ATTACHMENT ^ 10 TOTAL ACTIVITY OF I-131 '
VOLUME OF REACT 0?. Gnni E~
ATTACHMENT 5 11 SUPPRESSION POOL ATTACHMENT 5 11 TCTAL ACTIVITY OF CS-137 ;
ATTACHMENT 6 12 TOTAL ACTIVITY OF XE-133 i
ATTACHMENT 7 13 TOTAL ACTIVITY OF KR-85 l l' -
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. POST ACCIDENT ESTIMATION OF FUEL CORE DAMAGE g
TABLE OF CONTENTS I SECTION PAGE NO. g 1.0 PURPOSE 2
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2.0' DISCUSSION 2
3.0 REFERENCES
2 4.0 DEFINITIONS 4 5.0 PRECAUTIONS / LIMITATIONS 5
[
6.0. REQUIRED ECUIPMEhT/NATERIALS 5 7.0 PERFCRMANCE CONTROL 5 .
8.0 PROCEDURE -
5I 8.1 Deterriretire of Fuel Core Damage by Nyaragen Prcduction 5 l
8.2 Determination of Fuel Core Damage by Iodir.e Concer.tretien 6' 8.3 Detennination of Fuel Core Damage by Cs-137 {'
7 8.4 Determinatier of Fuel Core Damage by Xe-133 7 8.5 Deterrir.ation of Fuel Core Damage by Kr-85 8 ;
8.6 Estimation of Fuel Core Damage 6 8.7 Estication of Fuel Core Damace frcr Standardized Graphs 8 -
9.0 ACLEPTAhCE CRITERIA 9
-l ATTACHMENTS !
l ATTACHMENT I - F0WER CORRECTION AND REFERENCE PLAhT COPPECTION. 10
}
ATTAChhENT 2 - DENSITY CORRECTION FACTOR (X) FOR LIQUID 11 l
!/.PFLF TEPPERATURE CHANGES
}
l ATTACFMENT 3 - HYDROGEN DATA CALCULATI0h SHEET 12
' l ATTACl# TENT 4 - IODINE DATA CALCULATION SHEET 13 g l' ATTACHMENT 5 - CESIUM DATA CALCULATION SHEET 14 l
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~ ATTACHMENT 6 - XENON DATA CALCULATION SHEET 15 -j L .' ATTACHMENT 7 - KRYPTON DATA CALCULATION 'SHEET 16.
L- . 2 --
- ' ATTACHMENT 6 ' CCfE C/FAGE' GPAPH 17 l
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-I N/A n/A i. COP-1050 REY. O PAGE 1 0F 23 e i l' !
. POST ACCIDEhT ESTIPATION OF i FUEL CORE DAMAGE g l TABLE OF CONTENT 5 CONT'D i
SECTION PAGE NO.
ATTACHMENT 9 - I-131 CONCENTRATION VS FUEL CORE DAMAGE 18 j ATTACHMENT 10 - Cs-137 CONCENTRATION VS FUEL CORE DAMAGE 19 :
ATTACHMENT 11 - ye-133 CONCENTRATION VS FUEL CORG DAMAGE E0 l
ATTACHMENT 12 - Kr-85 CONCENTRATION VS FUEL CORE DAMACE 21 :
ATTACHMENT 13 - H CONCENTRATION STANDARDIZATION 22 ATTACHMEtiT 14 -'F CONCENTRATION VS "> METAL WATER REACTION 23 I
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l ti I N/A N/A I COP-1050 REV. O PAGE 2 0F 23 !
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-1.0 PURPOSE 1.1 The purpose of this procedure is to previde the method and calcul-ations necessary to estfrate fuel core damage from a major reactor accident.
2.0 DjSCUSSION ?
2.1 This procedure provides a quick first estimate of the extent of fuel I core damage based on the methods outlined in Refererce 3.C. . A I
- direct method, employing specific RSS data is covered in Subsection 8.1 to 8.6. An indirect method, using plant perimeter correction factors which relate RBS to the Reference Plant (GE Standard 3579 1
MW BWR 6/III) is outlined in Subsection 8.7. This last methcc l utilizesstandardizedgraphsofnuclideconcentrationvscoredamage
- supplied with Reference 3.8 (Attachments 9 to 12) to estimate the extent of core damage.
l 2.2 Calculation for the determincation of percent fuel damage shall be I perfomed by the Sampling Team and reviewed by the Chemistry / Core I-Damage Assessment Coordinatcr. I
?.3 Analytical results are cbtained by sampling the Containment cr.d/cr Drywell Atmosphere, Supression Pool and/or Reactor Coolant System:
,i 2.A The Chemistry Core Damage Assessment' Coordinator will inform te Technical Support . Center Emergency Director with the percent fuel dame.ce results as soon as the infcrmation is available and verified. I I
2.5 Measurements of Cs-137 and Kr-85 activities may not be perrib'e until the reector has been. shut down for several weeks to allow the I' decay of the shorter lived isotopes.
f
3.0 REFERENCES
I I.s 3.1 'JSNRC Reg Guide 1.97, Irstrumentation for Light Water Cooled huclear
+
Pcwer Piants to Assess Plant and Environs Cerditions During and _f Following an Accident, 1980
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3.2' COP-1001, Post-Accident Sampling of Primary Coolant (LATEP) 3.3 ,CCP-1002, Post-Accident Sampling of Containment Atmosphere (LATER) 3.4 . RBS FSAR, Section 13.3.5.2, Emergency Planning Assessment Actiert 3.5 RBS FSAR, Volume 1, Chapter 1.1, Introouction and General Descr'r-
_ tion of P1 art g
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3.6 RBS FSAR, Volume.8, Table 4.2-4, Fuel Data 6<
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COP-1050 REY. O PAGE 3 CF 23
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I i I 3.7 US Atomic Energy Comission, Safety Evaluatier cf the River Bend I I Station, Sept. 1974 !
l 3.8 NEDO-22215. 82NED090, Procedures for the Determination cf the Extent !
8 of Core Damage L'r.dcr Accident Conditions.
3.9 COP-0425,- Detennination of the H2 and 02 gas - Gas Chrr.r.rtecraphy .
Method.
3.10 NEDC-30088 Responses to NRC Post-Irnpler:entr.tien Review Criteria fnr Post-Accident Samplir.g System 3.11 COP-1033, Pest-Accident Isotopic analysis for Liquid Activity (LATER). I I
3.12 EIP-2-015, Post-Accident Sampling Operations i l
3.13 COP-1030, Post-Accident Isotopic Analysis for Particulate / I Iodine / Gaseous Activity I 4.0 DEFINITIONS 4.1 River Bend Station General Inforr:r.tfen 4.1.1 MWT - 2894 f' I 4.1.2 Asscirblies - 624 I' i 1 4.1.3 Fuel Reds / Assembly - 62 l 4.1.4 Reactor Coolant Vclurre (LATER)
I 4.1.5 Supression Pool Volume - 126.600 ft3 (3.57 E9 cc) 4.1.6 Certainment Net Free Air. Volume - 1.120E6 ft3 (3.17 E10 cc) j l 4.1.7 Crywell Net Free Air Volume - 2.47E5 ft3 -(t;.SS E9 cc) !
I i I 4.1.8 Containment + Drywell Net Free Air Volume - 1.367 ft3 (;.E7 I E10cc) I 4.1.9 Zirconium - (LATER) Ib. tntal 4.2 .Fydtcgen burn - The initiation of the hyorcgen igniters to decrease I the hydrogen concentration in containment.
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4.3 Evidence of Fydrog'en Burn ~- Evidence 'of hydrocen turn is 'cepicted by T an increase of ccedersation and a drop of pressurc in containrrent.-
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ammme N/A N/A COP-1050 REY. O PAGE 4 0F 23 q ,
. 5.0 PRECAUTIONS /LIMITAT:0fi5 1
5.1 Iodine and Xenon analysis are based upon equilibrium fuel power I isotopic conceritrations. If fuel damage is suspected to have I occurred during times of reduced power or near the tir.:e of signifi- I.
cant power change, the Iodine and Xenon invertcry rrbst be compensat- I, ed accordingly. Refer to Attachmert I to calculate the pcher I correction factor Y.
5.2 Cs-137 and V.r-85 concentrations will be ccrrecteo by multiplying the 1 average capacity factor for the previous 3 years due to their long half Ifves. Refer to Attachment I to calculate the pcher correction factor Z.
5.3 Core damage below 17, is assumed to be a non-accident condition.
5.4 If the isotcpic analysis show the absence of Ruthenium and Tellurium, then assume that fuel melting has not occurred. hcwever, I' the presence of these nuclides does ret _ recessarily confirm fuel melting.
' 5. 5 The determination of percent failed fuel is highly depencent on core temperaturereachgdduringtheaccidentcordition. Core temperatyre l in excess of 1600 F indicate. possible claddirg danage. Temperaturas i' 4
in excess of 40000 F indicate pcssible fuel melting. (
I 5.6 The reactor coolant temperature shall be cerpensated by a Density ;
Cortection Factor. Refer to Attachment 2. g 5.7 The performed estimates are , dcre under the presumption that no I reactor coolant cleanup systems are operated after the acciderit.
6.0 REQUIRED ECL'IPPENT/ MATERIALS N.A.
7.0 _P_ERFORMA_NCE CONTROL !
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, 8.0 PROCEGuRE 8.1 Determination of Fuel Core Damage by Hydrogen Production 8.1.1 Obtain sample analysis by grab sarplirg via the Post Acci-dent Sample Panel if the hydregen igniters have beer l en- f g
. ergized cr. there is evidence of hydrogen burn.- The H C concentrations :are determined per Reference 3.9. 2' v 2 1 l
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i f!/A N/A COP-1050 REY. O PAGE 5 0F 23 1 I I! I I l
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NOTE I Use one Attachment 3 form for Crp;cil !
Volume and one for Containment Volurre.. I 8.1.2 Refer to Attachment 3 fer calculation using this rethod.
8.2 Determination cf Fuel Core Damage by Iodine Concentration NOTES
- 1. During a reactor accident, all of the iodine is assumed to remain in the reactor coolant.
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- 2. If mixing of -reacter ecolant with I water of the suppression pool occurs, the suppression reel may be sampled I and the volumes of the reacter coolant I loop plus the weter volume of the ,
suppression pool have to be used to .
. calculate the total emount of T-131 '
release from the core. - I,
- 3. If the accident is such that oc mixing I with suppression pool water occurs, I
, the reactor coolant locp rtust be I sarr. pled ard only .its water volurre te I used in the calc 61ation, f
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8.2.1 Obtain a liquid sample per Reference 3.2 and perferr the '
analysis per refererce 3.11 to determine the I-131 concen- I tration. I 8.2.% Convert the treasured specific activity to the tctal I-131 cttivity released by using the apprrpriate water volume.
I L.i.3 Lbtain correction factors frcr: Attachment 1 and 2 for X and Y to normalize the data for corr.parison.
8.2.4 Calculate the percent.of core damage by dividing the cor- !
recteo released I-131 activity by the total I-131 activity of the reactor core.
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14 g
C.2.5 .Fefer to Attachment 4 for calculaticr.'using this method. ;.
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O 8.3 Detemination of Fuel Core Damage by Cs-137 I
NOTES
- 1. During a reactor accident, all of the Cesium is . assumed to remain in the I reactor coolant.
- 2. If mixing of reactor ccolant with I water of the suppression pool occurs, the suppression pool may be sampled and the veltmes of the reactor coolant loop plus the water volume of the suppressicn pool have to be used to calculate the total amount of Cesium I ielease from the core.
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- 3. . .I# the accident if such that no mixing I with suppression pool water occurs, the reactor coolant loop must be
, sampled and only its water voluire be used ir the calculation. -
I' 8.3.1 Obtefr c liquid sample per Reference 3.2 and perforin the analysis per Reference 3.11 to determine the Cs-137 concer.-
tratiers. I I
l 8.3.2 Convert the reasured specific activity to a total Cs-137 I activity releasea by using the appropriate water volume, f
l S.3.3 CLtrir cerrection factors from Attachment I and 2 for X and I Z to normalize the data fcr cok 1rison.
l l 8.3.4 Calculate the percent of core damage by ditic' ire the cor-i~
rected released Cs-137 activity by the total available Cs 137-activity of the reactor cere.
8.3.5 Refer to Attachtert E fcr calculation using this methud.
8.4 Determination of Fuel Core Damage by Xe-13?
g j 8.4.1 Obtair. a gaseous sample per Reference 3.3 and perform the analysis per reference 3.13 to determine Xe-133. concer-p . ., tration.
l 8.4.2 Obtain the Power Correcticn Factor (Y) from Attachment 1.
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e 8.4.3 Calculate the percent of core damage by dividi.:g thr: cer-rected released Xe-133 activity by the total available Xe-133 cctivity of the reactor core.
8.4.4 Refer to Attachment 6 for calculation using this method.
8.5 Determination of Fuel Core Damage by Kr-85 8.5.1 Obtain a ' gaseous sample per Reference 3.3 and perform the ;
analysis per Reference 3.13 to determire (Kr-85) concen-tration.
8.5.2 Obtain the Power Correction Factor Z frem Attachment 1. g 8.5.3 Calculate the percent of core damage by dividirg the cor-rected released Kr-85 activity by the total available yr-85 activity in the reactor core.
8.5.4 Refer to Attachment 7 for calculation using this method.
8.6 Estimation of Core Damage 8.6.1 Enter the results from an'alysis of methods 8.1, 8.2, 8.,3, 8.4 and 8.5 on Attachment 8 and calgulate the averggp estimated cladding derage. L 8.6.2 Sign Attachment 8 and include all other attachments ccrplet-ed.
8.6.3 Submit to the Chemistry / Core Damage Assessment Cecrdfrator for verification.
f l 8.6.4 Evaluate the results of Attachment 8 erd retify the Techni- l cel Surport Center - Emergency Director as per Reference 3.12 of the Core Status Estimation.
I.
S.7 Estimation of Fuel Cere Ot.c.sge irun Standardized Graphs 8.7.1 Calculate the corrected released !-!31 ectivity (Item 3. of-Attachment 4) as per Steps 8.2.1 te 8.7.2.
l 8.7.2 Divide by the primary coolant mass and naultiply by the primary coolant mass correction facter (Section 4.1) g i
l- 8.7.3 For.the standardized specific I-131. activity of Step 8.7.2 -l b.-
I- . obtain the correspceding value. for the estimateo fuel . core g
' g l~~
damage'from Attachment 9. . .
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COP-1050 REV. O PAGE 8 0F 23 l I
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8.7.4 Record the result on the appropriately mar keo /.ttachment 8 form.
8.7.5 Calculate the corrected released Cs-137 activity as per Steps 8.3.1 to 8.3.3 (Item 2. of Attachment 5) 8.7.6 Repeat 5tpes 8.7.2 to 8.7.4 for Cs-137 by i sirr Attachment 10 in Step 8.7.3.
6.7.? Calculate the corrected released Xe-133 activity as per Subsection 8.4 (Item 2 on Attachrent 6).
8.7.8 Divide by the containment gas volure and multiply by the contairment gas volume correction facter (Section 4.1),
- 8. 7.9- For the standardi ed ye-133 activity concentraticn cf Step 8.7.8 obtain the corresperdf rg value for the estimated fuel I core danage frcm Attachment 11. I 8.7.10 Record the result en the appropriately marked Attactrent 8 fonn.
8.7.11 Calculate the corrected released Kr-65 activity as per Subsection 8.5 (Item 2 cn Attachment 7)- (, l 8.7.12 Repeat Steps 8.7.8 to 8.7.10 for Kr-85 by usiro Attachment 11 in Step 8.7.9.
8.7.13- Obtain and record on Attachncnt P the value of,the fuel core I damage derived frca the containment H concentration of the OrtphAttachment13usingthemethcobutlinedonAttachment I
, 14.
8.7.14 Calculate the average Estimated Cladding Damage frcr the I four entries for I-131, Cs-137, Xe-133, Kr-85 and H '
2 8.7.15 Proceed as per Steps 8.6.2 to 8.6.4. I i
I 9.0 ACCEPTAhCE CRITERIA N.A. J l l l l- I .
L l "END" I
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I ATTACHMENT - 1 1
FCFER CORRECTION AND REFERENCE .'d. ANT CCRRECTION l
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- 1. Correction factor for I-131 and Xe-133 (Y) I I To correct for core isotopic inventory if fuel damage is suspected, calculate the correction factor for I-131 and Xe-133 by using the folicwing equation: '
Y= 100%
Average thermal power over the last 30 days in percent j
- 2. Correction factor for Cs-137 and Kr-85 (Z) j For calculatiers employing Cs-137 and Kr-85 use the fo11cwing equation for the I g
correction factor Z: l I Z= 100%
Average thermal power over the last 3 years in percent l .
9
} Refer to Reference 3.8 if 'more accurate Power Correction is desirea I' ;
- 3. Reference Plant Parameters (Reference 3.8)
I Rated Thermal Power: 3579 n.'t f'
- of Fuel Bundles: 740
, f Total primary coolant rrass: 3.92E9 g !
(reactor + suppression pool) ;
Total ccr.tzirrent and drywell !
I gas space vclur.e 4.0 E10 cc i
Primary Coolant Mass Cerrection Factor (Fg ) {
4-l I 3 E9 g I p" , RBS Coolant Mass , ,
g Ref. Plant coolan' Mass 3.92E99
- 5. Containment Gas Volure Correction Factor (Fg ) ,
t p . , RDS Containment.& Drywell. Gas Vol. , 3.87 E10 cc ,
0.967 9 Ref. Plant Containment Gas Vol. 4.00E10 cc g l
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ATTACHMENT - 1 PAGE 1 GF 1 CCP-1CEO REY. O PAGE 10 0F 23 l ,
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- ATTAthMEf;T - 2 DEflSITY CORRECTION FACTOR (X) FCR LIOU?O !
SAMPLE TEMPERATURE CHAf;GFC I flormal Reactor Coolant System sample temperature is approximately 90 F. I i
Cetermine the appropriate Reactor Cociant temperature at the tice cf sterling and I select the associated density correction factor X from the table.
Reactor Coolant Sample TemperatureOF g
REAGIOR COOLANT UENSITY CORRECTION TEMPERATURE FACTOR (X) 100 .998 150 .985 200 .968 g
250 .947 l 300 .923 I I
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.. 350 .595 g
j l 400 .864 i-450 .825 500 .788 I l
550 .740 I i
560 .729 l j 570 .716 580 .708
{
, .g 590 .694 600 .681 f
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l 'ATTACMENT - 2 PAGE 10F 1 CCP-1050 REV. O PAGE 11 0F 3
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ATTACPNENT - 3 HYDROGEN DATA CALCULATION SHEET ,
- 1. Volume Considered 3
l[l Containment (Volume = 1.12 E6 ft )
3 l[l Crywell (Volume = 2.47 E5 ft )
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- 2. H, Volume at Accident Conditions I
- a. From H2 Data 3
( %H)(2 ft Vol.)/100% = ft3 g 2
- b. Frcm 02 Depletion I
( %02 Normal) - ( %02 PostAcid) = %0 Depl.
2 3 3
( %02 Depi)x(2)x( ft Vol)/100% =
ft g2 (02)
- c. Total Volume of Lib. H 2 in Vol. (Add a. and b.) ,
'y 3 3 3 I
( ft H)+( 2 t H2 (02))
= ft H 2 Tot.
- 3. h, Volun.e at Sterdard Pressure and Tefnperature (sct) l 6
c T= c (In considered volume) l p, psia (In considered volume) I l
3 U g ( ft g Tot.) (1 + C/273 C) ( psia /14.7 ps'ia) 2
= __ scf H in 2 l[lCor:tair. ment j l[{ Orywel I !
- 4. Total Mass of Zirconium Reacted f
.scf H2Cont. + scf H2Dryw. = scf H2 Tot.
scf H2 /(8.0 scf Hp/lb Zr React.) = lb Zr
. - I- ,
\
I S. _ Percentage of Core Claddire 100% x ( lbZrReact)/((LATER)IbZrincore)
= % of Total Zr Reacted 4
ATTACHMENT - 3 PAGE 10F 1 COP-1050 O PAGE 12 0F 23 l . lREV.
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ATTACPMEt:T - 4 IODINE DATA CALCULATICN SHEET !
- 1. Total Activity of I-131 A0TE i
Detennine the transient involved to account I for the total iodine activity. Make the necessary corrections as related to the specific incident.
t
- a. Measured in the Suppression Pool + Reactor Ccolant I (N/A if suppression pool not used) l
( uC1/ml)(LATERml)(IE-6) =
C1 (SP & RC) f
- b. Metsured in the Reactor Coolant l
( uC1/ml)(LATERml)(IE-6) =
Ci (RC) I.
(N/A f f suppression pool used)
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- 2. Decay Calculation to Time of Reactor Shutdown T g
l Activity e counting time T x e 4 0.693(T - Tg )/ T1/2 =ActivityreleasedatT,f
( C1)
- e + 0.693 ( h)/193.2h = Ci 0 To ,
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- 2. Fcher and Density Correction (I-131 Activity from 2.)(Y)(X) = corrected Activity released at T '
( 01) ( )( ) = Ci (Y = Average Capacity Factor for previous 30 cays) l I
- 4. Percent'of' Core Damage l
'I 100% x ( Ci released /(LATER) Ci available) = %
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i ATTACHMENT - 4 PAGE 10F 1 COP-1050 REY. O PAGE 13 0F 23 I
ATTACHMENT - 5 .
CESIUM DATA CALCULATI0fi SHEET i
- 1. Total Activity of Cs-137
- a. Measured in the Suppression Pool (N/A if suppressier: poci not used)
=
( uCi/ml)(LATERml)(IE-6) Ci (SP & RC)
- b. Measured in the Reactor Coolant (N/A if suppression pool used)
=
( uC1/ml)(LATER ml)(IE-6) Ci (RC) g
- 2. Power and Density Correctice (Activity Cs-137 trom a. or b.) (X)(Z) = Activity released at T c
( Ci ) ( )( ) = Ci released 0 T o j (Z = Average Capacity Factor for Previous 3 Years) ,
i g.
'I 3. Percent of Available Cesium 137 Released I 100% x ( Ci released /)(LATER) Ci available = %
t 1
I D
w ATTACHMENT - 5 PAGE 10F 1 COP-1050 REY. O PAGE 14 CF 22
, _4 - , - - _ _ . - - _ _ _ . ~ , - _ _ r - ,
. 1 I ATTACHMENT - 6 XENON DATA CALCULATION SHEET l i l I
- 1. Total released Xe-133 Activity I a. Measured in the Certair. ment
( uCi/cc)(3.87E10 cc)(IE-6) = C1(Drywellar.dCent) i
! Assumes approximate ecual distribution in drywell and containment
- b. Measurea in the drywell
( uCi/cc)(6.99E9 cc) (IE-6) = C1 (Drywell)
- c. Xenon 133 Pecsured in the containment
( uCi/cc) (3.17 E10 cc) (IE-6) =
Cf (Cont.)
NOTE l
! !!se either 'a. or the sum of b. and - I 1
- c. to fit the specific situaticn.
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- 2. Decay Evaluation to Time of Reactor Shutdown (Activity 9 time T of count) e + 0.693 (T-T,)/TI/2 = Activity rel. O T !
e Ci 0 T}e + 0.693 ( h)/126.5h
(
l - Ci released @ Tg y
l 3. Power Correction !
i (Activity Xe-133 from 2.)(Y) = Ccrrected Activity 0 Tc
{[ !
=
l ( C1) ( ) Ci @ T,
- 4. Percent of Available Xe-133 Released l
l f 100% x ( Ci released)/(LATER) Availetle = , , %
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I I ATTAChl.1EhT - 7 KRYPTON DATA CALCULATION SHEET I I I i i
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- 1. Total Released Kr-85 Activity l
- a. Measured in the Centainment l
I C1 (Lrp ell & Cent) !
( uCi/cc)(3.87E10cc)(IE-6) =
Assumes approximate ec;uti distribution in drywell and ccntainment f
- b. Measureo in the Drywell !
i
( uCi/cc)(6.99E9 cc) (IE-6) = Ci (Drywell)
- c. Measured in the Containment I
( uCi/cc) (3.17E10 cc) (IE-6) =
C1 (Cont)
NOTE: Use either a., b., cr c. or the sum of b. and c. to fit the specific situation of the Kr-05 distribution present. :
(. !
- 2. Power Correcticn (Cf Kr-05 from above) (Z) = Corrected Activity @ T,
( Cf) ( ) = Ci @ Tg l .
(Z = Average Capacity Factor for Previous 3 Years) l i j I
- 3. Percert cf Available Kr-85 Released i
100% x ( , Ci released)/(LATER) Ci Available = %
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n . . . 1.- . .
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l ATTAtletENT - 7 PAGE 10F 1 COP-1050 REY. O PAGE 16 0F 23
. I ATTACHMENT - 8 CORE DAMAGE GRAFH l
i METHOD: CALCULATION (8.1to8.6) l l GRAPHS (8.7) l_i g l
% CORE DAPAGE l
i i i HYOROGEN -
1
'l I-131 '
f.
i \
XE-133 j e
Cs-13, ; ,
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,_ Kr-SS a I 0 20 :;
40 60 80 100 l
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! ESTIP/iTED !
I CLADDING y i DAMAGE O cG JO % 60 80 100
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Performed b'y' '
Signature Date Reviewed by g TSEIstry Core Damage Assessment Ccordination Late v
i ATTAC H NT - 8 PAGE 10F 1 COP-1050 REV. O FACE 17 0F 23
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ATTACFFFFT . t ;-15 L(lt.CEf TRATION VS FUEi. CORE CAMAGE l
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Figure 1. Relattenship .Between 1-131 Concentration in the Primary Cool' ant (Reacter Water + Peel Water) and the Estant of Cort Da age in Reference Plant j 2-4 t
RPC-000.067 !
I I
ATTACHMENT - 9 PAGE 1 0F 1 COP-1050 REV. O PAGE 18 0F 23 ,
l t
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. 7- t 10 Cs-137 CONCEP.Trd.T10h VS FLEL CCPE DAPA0E i ATTACHMEtli - ,
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" .g, , Figure 2.
Relationsh1P Between Cs-137 Ceneentration in the Primary Coelent l
(neseter veter + Peel Water) and the Essent of Care Da age in j Reference Plant 1
23 RPC 000.068 i
ATTAClelENT - 10 PAGE 10F 1 C0p-1050 REV. O PAGE 19 0F 23 ;
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o .,,--
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l .' h-133 CONCENTRATION VS FUEL CCRE D/F. AGE f ATTACFFFtT -
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Tigure 3. Relationship Between Ie-133 Concentration in the Contairement Ces
} prywell + forus Cas) and the Estent of Core Da age in Reference }
. ' F1snt - ' g 26 . l l
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1 RPC 000.069!
I I I
l ATTACHMENT - 11 PAGE 1 0F 1 COP-1050 REY. O PAGE 20 0F 23
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ATTACH.* TENT - 12 ,
Kr-85 CONCENTRATION V5 FLEL CCFE CMME i l
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Figure 4. Relationship Between Er-85 Concentration in the Containment Gee l
(Drwell - Torus Cas) and the Intent of Core Desage in Referen'ce Flant }
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RPC-000.070 l 1
ATTACMENT - 12 PAGE 10F 1 l COP-1050 REY. O PAGE 21 0F 23 .
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I Aill.ChPEPT - 13 ! PgC($NCEt:TRATI0f3 ST/FCtFCIZAll(fi !
i
_l J. Obtain the total sef F s t ?t.ra. reiedsed into the colitcirrant and I I t'r e dry well (Item 3 of Attachinent 3) da pr St.trection 8.1.
- 2. Calculete the i F,:
4
't H scf H E x 100% =. I ) =
l p = Cont. Voi 1.12 Er. ++- a l' t l3. For the resuit6 tit L H value obtain the corresperc'irg stit.o for the 7. Netai-Lt'.rt Ferhtien (%W) for the 74F (Pt au f Feterence I
P1 art -(right set tic.cl axis). l f' .
1
- 4. Calculate the " i"W for RBS: i 748 V 3
% W = (thWref) T (1.36 EG
,; y 0.ES = 5 l
l I.
.15 fltrber of Fuel Bundles = f?d
{
d
- F:E5 Cor.taintnent !?et A.r 'loit r r 2.1? E6 ft !
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, , , , I ATTACHMENT - 13 PAGE I 0F 1 ' (CF-1050 KEV. C I
25 i
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!, fAGE iE 0F
- ',.* I I
H CChCEhTRATION VS % METAL WATER REACTIOf; I I ATTACHMENT - 14 7 I I
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.=.-,..,o., l Flaute A-1. Myd, ogen Concentratten for Mark 1/!! and 111 Containmente se a Function of Metal-Water Resetten I l
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h 9
RPC 000.071 ATTACHMENT - 14 PAGE 1 0F 1 COP-1050 REV. O PAGE 23 0F 23 l ,.