ML20092J928
| ML20092J928 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 04/23/1984 |
| From: | PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | |
| References | |
| OL-A-035, OL-A-35, NUDOCS 8406270311 | |
| Download: ML20092J928 (25) | |
Text
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EXHIBIT No. ' 8{
BocMtT MUMeta ' Staff
~
Pit 00. si uTIL FAc..d},h -gML Applicant Intervencr EP-C-326 Rev.0
..r 3/21/84
' gg jg identified Received ' Rciected Page 1 of 4 Date- #/ FV MJMilmd -
[ Reporter: --w
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PHILADELPHIA ELECTRIC COMPANY T PEACH BOTTOM 2 AND 3/ LIMERICK 1 .
2N
-EMERGENCY PLAN IMPLEMENTING PROCEDURE EP-C-326 PROCEDURES FOR ESTIMATING CORE DAMAGE Es 95 g
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DU_ RING Addf655T d656fTf655~ g g w$g, g -< q
. pm _L 5 1.0 P U,R,P O S E ,
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,C. o M t This procedure provides the method for acquiriiig oniGiite radiological data and the use of such data for estimating the extent of core damage during accident conditions.
2.0 .ES PONS,IBILITIES 2.1 The Core Physics Coordinator is responsible for assuring that this procedure is implemented as required, and the results of the core' damage estimates are reported to the Emergency Support Officer. 2.2 The Core Physics Coordinator is responsible for assigning personnel to perform thip procedure. 3.0 A,,P,P,,E_N,DI CES
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3.1 EP-C 326-1 Radiological Data -
~3.2 EP-C 326-2 Hydrogen Concentration Data 3.3 EP-C 326-3 Containment Radiation Monitor Data *-
L 3.4 EP-C 326-4 Metal Water Reaction Figures 3.5 EP-C 326-5 Percent of Fuel Inventory. Airborne in containment vs. approximate Source and Damage Estimate o 3.5.1 EP-C 326-5(A) PBAPS Containment Radiation"
. Monitor Curves 3.5.2 EP-C 326-5(B) LGS Containment Radiation Monitor Curves 3.6 EP-C 326-6 Methods for Estimating I-131 concentration in coolant and Xe-133 concentration'in containment atmosphere N0$$$3 PDR Q
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EP-C-326 Rnv.0 I 3/21/84 ; Page 2.of 4 : MJMalmd l 3.6.1. EP-C 326'-6(A) Manual procedure for determining I-131 concentration in coolant 3.6.2 EP-C 326-6(B): Manual procedure for determining ! Xe-133 concentration in containment atmosphere i 3.6.3 EP-C 326-6(C) Corporate Computer Method for ! determining I-133 concentration in-coolant , 3.6.4 ~ EP-C-326-6(D)' Corporate Computer Method for : Determinign Xe-133 concentration in containment ! atmosphere
" 3.7 EP-C .326-7 XE-133 Concentration vs. Fuel Damage f i
- 3.8 EP-C 326-8-I-131 Concentration vs. Fuel Damage l
4.0 P REQUISITES . I e 4.l' Onsite personnel shall have prepared and transmitted the necessary data for conduct of this procedure. 4.2 Station Daily Thermal / Electric Output, A-1 Form Log, f shall be updated to current core conditions. ' I
. i
. 5.0 . SPECIAL EQUIPMENT ----- t None ! t i 6.0 SYMPTOMS t None. e. 7.0 ACTION LEVEL
.. An Alert, Site or General Emergency has been declared at i
Limerick 1 or Peach Bottom 2 or 3. 6 t i l y -e, ,-,~-..--,--..w,---%,_.,- . w w w.r,- w e ew -w.,,,,m,,,,.c.,_,,,,-
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EP-C-326 Rev.0 ! 3/21/84 i Page 3 of 4 l MJMalmd ; i 8.0 PRECAUTION ! None ! 9.0 PROCE DURE 9.1 Actions t 9.1.1 .Obtain the radiological sample data listed in Appendix EP-c 326-1 from the Health Physics and ! Chemistry Coordinator at telephone number - at the Peach Bottor Emergency Operations e t.a6. Facility or at telephone number . at the pI F Limerick this form.Emergencyf,j,,acility
,. , , 6 ,,, and log the data on l
. 9.1.2 Obtain the hydrogen and oxygen concentration ;
! data listed on Appendix EP-C-326-2,from thu ! Health Physics and Chemistry Coordinator and : log data on this form. 1
?
9.1.3 Perform the manual or computer procedure given l
- in Appendix EP-C 326-6 to estimate the extent i of core damage based on the radiological sample ;
-data obtained. O !,
9.1.4 Determine Percent Metal Water:Ronction. 9.1.4.1.:Using the data from 9.1.2 and f Appendices EP-C-326-2 and EP-C-326-4, " ' il -.- . determine the 9 MW reaction and status , 3 of containment atmosphere relative to. [ potential for hydrogen or oxygen burn, i L 9.1.5 Using Appendix EP-326-3, estimate the extent of ,, ! " core damago based on containment radiation ! monitors using curves provided in EP-C 326-5. j i 9.1.6 Return the completed EP-C 326-1, EP-C 326-2,
~
l and EP-C 326-3 to the Core Physics Coordinator. !
- i. ; r 9.1.? The Core Physics Coordinator shall review the !
completed Forms from Step 9.1.6 and inform the , Emergency Support Officer of the extent of any l l core ^ damage. The Core Physics Coordinator ! i l ! l l
. c. 'i
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. EP-C-326 Rav. O !'
. 3/21/84 [
Page 4 of 4 MJMilmd l t
- shall retain completed forms for future
> reference during the emergency. l l
10.0 ' REFE RENCES 10.1 ! Procedures for Determination of the Extent of Core Damage Under Accident Conditions, NEDO 22215,. August 1982 (G.E. Generic Procedure). - l 10.2 LG8 FSAR 11.5.5.4.5 Determination of Extent of Core ; Damage. ! 10.3 BLP-29640 Containment Radiation Monitors Post-LOCADose i Rates vs. TimeCurves. { 10.4 G.E. Document RPE OL CCLO1 dated November 1981, Procedure for Determination of teh Extent of Core ; Damage Under Accident Conditions. l t
'10.5 -Peach Gottom 2, 3 A-1 Poem Log i (DailyThermal/ Electrical Output).
10.6 ' Limerick, ( Daily Thermal / Electrical Output) . 10.7 Corporate Emergency Procedure EP-Cy226 Core Physics ; Coordinator ! 10.8 Bechtel~ Power Corporation letter BLP21558, dated April i 8, 1980, R. H. Elias to E. C. Kistner " Emergency j , Planning Acceptance Criteria: Modification Request gg t 503." i 10.9 Bechtel Power Corporation letter BL28640, " Containment
~ Radiation Monitors Post-LOCA Dose Rates vs Time r Curves." i
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- 3/21/84 Paga 1 of 1 APPENDIX EP-C-326-1 MJMilmd RAD 101DGICAL DATA s
PB2 Pn3 Inst
- 1. Reactor Shutdown Date/ Time /
- 2. Reactor Water Measured Fission Product Concentration, micro-curies /qm I-131 Sample Date/ Time /
- 1. Suppression Pool Water Measured Fission Product Concentration, micro-curies /gm I-31 Sample Date/ Time /
- 4. Torus Gas Measured Fission Product Concentration, micro-curioc/cc Xe-133 Sample Date/Timo /
Vial Temoerature, F Vial P ressuro, psia - - ~ ~ Torus Temperature, F Torus Pressure, psia 9 Drywell Gas Measured Fission y Product Concentration, micro-curies /cc Xe-133 Sample Date/ Time / Vial Temperature, F i,
- Vial Pressure, osia '
Drywell Temperature, F Drywell Pressure, psia
- 6. Core Damage Estimate Based on I-131
- 7. Core Damage Estimate Based on Xe-133
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3/21/84 Page 1 of 1 MJM:lmd APPENDIX EP-C-326-2 PB2 , PB3 LGS 1 HYDROGEN CONCENTRATION DATA
- 1. Drywell H2 Measurement ,
4 of Volume
- 2. Torus H2' Measurement 4 of Volume
'3. Average Containment H2 "
% of Volume (Average of 1 and 9 or if only one value is given, assume it to be the average for containment).
- 4. Drywell 02 Measurement 4 of Volume
- 5. Torus 02 Measurement 4 of Volume
- 6. MWref a (metal-water reaction for reference plant using Item 3 and Figure FM-19-1),
it
- 7. 4 MW - Plant Specific
- O.568 (MWref)
- 8. Containment Atmosphero Status (Combustible or Non-Flammahle
-use Items 3 and 4 and -
Fiqure FM-19-2).
- t MW = MWref L500)(V) For Limerick LN)(350,000) N=764 assemblies V= Torus + Drywell Volur
=149, 380 +3235,190
=384,570 ft For Peach Bottom 1,3 N = 764 assemblies i V = Torus plus Drywell Volume
= 127,900 + 175,800 = 303,600 ft3
- k. .
EU-u s a u - > . 1 3/21/84 APPEND'IX EP-C-326-3 Paga 1 of 1
?!JM:lmd PB2 PB 3 _ LCS1 CONTAINMENT RADIATION MONITOR DATA
- 1. Drywell High Range Reading, R/hr OR
- 2. Drywell Low Range Reading / R/hr . _ _ .
- 3. Torus Low Range Recding, R/hr
- 4. Date/ Time of Reading /
- 5. Date/ Time of Reactor Shutdown /
- 6. Time Following Shut:10wn (4-5), hr 7 t Fuel Inventory Released (from Appendix EP-C-326-SA or D'
- 8. Core Damage Estimate from Appendix EP-C-326-5 ti ll O
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- Y21/84 Page 1 of 5 MJMilmd APPENDIX EP-C-326-5 PERCENT OF FUEL INVENTORY AIRBORNE IN T!!E CONTAINMENT VS. APPROXIMATE SOURCE AND DAMADE ESTIMATE t Fuel
- Curve Inventory Approximate Sourco and Damage Estimato No. Rel eas ed 1 -100. 100% TID 100% fuel damage, potential core melt.
50, 50% TID noble gasos, TMI courca 2 10. lot TID, 100% NRC gap activity, total clad failure, partial core uncovered.
- 3. 3% TID, 100% WASH-1400 gap activity, major clad failure.
3 1. It TID, 10% NRC gap, Max. 10% clad failure. 4 .1 .1% TID, it NRC gap, it clad failure, locci heating of 5-10 fuq} assemblies. 5- .01 .01% TID, .1% NRC gap, clad failure of 3/4 fuel olomont (36 rods). 6 1.0H3 .01% NRC gap, clad failuro of a fow rods,. 3.ON4 100% coolant roloose with spiking. 7 5.ON6 100% coo 3 ant inventory roloase.
.l.ON6 Upper range of normal airborno noble gas activity in containment.
*100% Fuel Inventory = 100% Noble Gases + 25% Iodines t 'll particulaE TID = Total Icotopic Distribution O
e 11/21/83
, I' 9. M;91 Tors fiES.:C.95E C SVES g .
3 an j O' 10 13* 10 - 1 44 Q o s se. o o ; :
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02<' Da E s'-- . - - -- . . _ . . - . . . so' Primary Ccatainment High itapSe Monitors e , l e ~ I - 1 i N . . . E. t,. i Curve Index i fk -. [ . :- _ . . - _ - . . . - - . 10 1. 1C01 Fuel Irxentory (1905 TID 14644 i a< :: i ; : 22. 10E Fuel Inventory u-
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. 4. 1% Fuel Inventory e '
3 - 011 H ~_ 5. Fuel Inventory 2
- - 6. 001% Fuel Inventory at-.L l t -
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# EP-C-326-6 3 /21/84-Re(
Page 1 of 5) MJM:Imd APPENDIX EP-C-326-6 ' c i A. Procedure for determining fuel damage estimate based on , measured I-131 concentrations' . I t
- 1. Calculate Py (I-131) '
l
= 3651 i P1 x .4531 + P2 x .2T70 + P3 x .135"5 + P4 x .0741 Cont'd be14 i
t
+P5 x .0405 + P6 x .02i2 + P7 x .0121 l
Where P1, P2' P3'~P4' F5' E6' #7 are the average reactor power I output (MWT) during week 1, 2, 3, 4, 5, 6 and 7 respectively l prior-to shutdown. Those values are obtained from the A-1 , log sheets maintained for both Peach Bottom 2 and 3. ' i Plant Parameter Correction Factor
- 2. i Rr P B Pw = 0.956 '
Far linerick Fw = 1.C01 u
- 3. Calculato the time (t) between reactor shutdown and the I-131 sample time (in days) utilizing information from form EP-C-326-1 ,g ,
I' o 0862 t
- 4. Calculate Ch*f(I-131) =C w (I-131)e xFI (I-131)xFw Where C.(I-131) is the measured I-131 concer.tration fr.om' ;
Form EP-C-326-1
- a. If a fuol failure estimate is to be based on reactor .
- water-sample, utilize the I-131 concentration. measured 1 in reactor water as Cw (I-131) 1 If A fuel failure estimate is to be based on suppression b.
~
pool water, utilize the I-131 concentration measured
- in suppression pool water as Cy (I-131) l L
l , l f
/
t.c-u ._, y 1/21/84 Rev. Page 2 of 9 MJM:lmd
- c. If I-131 concentration measurements are available in both reactor water and suppression pool water, determine an average concentration by weighting by the mass of reactor water and mass of suppression pool water.
For Peach Bott e 2,3 Cw (I-131) = Cw in ax Water x 2.67x100 + Cw in SP water x 3.4a x
- For Limerick 1: 3.747 x 10 9
= w " "x Water x 2.93x108 + ~ ( in' SI' water x 3.63x109
( (1-131) 3.923x109
- 5. Using Cw Ref. (I-131) and Appendix 7, obtain the fuel damage estimate.
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EP-C-J2u v-3/2178 4-aev Page 3 of f MJM:lmd D. Procedure for Determining .uel Damage Estimate Based on bkssured Xe-133 Concentrations
- 1. Calculate FI (Xe-133) 3651
.P y X 249 + P3 x .0975 + P4 x .0402 x .613 + P2 P P P4 are the average reactor power output Whore (MWt) dur P1,ing2' t hd<first, second, third, and fourth week respectively prior to shutdown. These values.are obtained from the A-1 log sheets maintained for both Poach Dottom 2,3.
- 2. For Peach Bottom 2,3 For Limerick 1 Pg = 0.214 Fg'= 0.272
- 3. Calculate the time (t) between reactor uhutdown and thu Xe-133 sample time (in days) utilizing information from Form EPH:-326-1.
- 4. Calculate Cg"*# (Xe-133) = Cg (Xe-133) e0 .132 t x Fy (Xe-133) xF Where Cg(Xe-133) is the measured,Xe-133 concentration from Form EPH:-326-1.
Si
- a. If the fuel-failure estimate is to be based on the -
drywoll gas sample, utilizo the Xe-133 concentration
' measured in the drywell corrected for pressure and temperature differences between the drywell and .
sample vial.(see FORM epi >32&-1) . gg
- Pdw x Tvial Cg (Xe-133) = Cg (vial) ic Pvial x Tdw
[
- b. If the fuel failure estimate is to be based on the !
torus gas sample, utilize the Xe-133 concentration ( measured in the torus corrected for pressure and ; temperature differences between the torus and sample ! vial (see vonx EPH}-326-1). .. P,nr x Tvial : Cg (Xe-122) = Cg (vial) ' x Pvial x Tt ur I
s.. . ..
~
3 /21/84 -Rev . 0 ,
'.acgo 4 of 9 MJM:lmd
- c. If Xe-133 concentration measurements.are available in both drywell and torus, determine an average ,
concentration by weighting by the volume of the i drywell and torus For Ibach Ibtton 2,3: C9 (Xe-13 3 ) = Ug(Xu-133, stop a x 4 . tJ U + Cg(Xu-133)stop b.s.~~
~
86 , For Lincrick -1: . Cg(Xe-133) = Q(Xe-133, ste a x 6.66 + Cg(Xe-133) step 6 x 4.23 , 10.89 l
- 5. Using Q lter. (1-131) aint Appendix 8, obtain tin fuel dancye,e estunate.
n Y t W g F 4 I
- P e
[ EP-C-326-6 ( C ) Rev . 0 ; 3/21/84 ! Page 5 of 9 l MJM:lmd Procedure for determing,fue'l damage estimate based on measured I- t 171 concentrations. l l
- 1. Logon to the PECO Corporate Computer System using normal logon procedures.
Using standard processes, enter into the SPF 2 format ! I (Note SPF 2 is used for all editing of datasets information)
- 2. The necessary dataset and the corresponding member can !
be found in dataset 'EP. DIAMOND. DATA 2' member ' DAMAGE I'. (Note: .DATASET *EP. DIAMOND. DATA 2'is controlled ~ L by H..J. Diamond, Alternate Core Physics Coordinator; l all Datasets and Job Control information are located under his computer identification number.) : l i ! 3. After entering into the proper dataset member, the l necessary data concerning the Post Accident Sainpling
- l. System (PASS) must be entered into the proper format I within the member (Note: A sample member and the o l format is shown in Appendix 6, Figure 1.) I t
i ! 4. The Description of formating is as follows:
4.1 First-line
Type in "PBAPSI' of the procedure ,
- p. is to be run for Peach Bottom reactor or
~
" Limerick" if the procedure is to calculate .
fuel damage at-a Limerick unit. [ i
, 4.2 Second line: Type in " UNIT X" where 'X' corresponds to the unit number of the reactor for which this procedure is to be used.
L 4.3 Third line: Type in the word " IODINE" which I will specify the iodine analysis to be run. , 4.4 Fourth line: Input average power of each of the i seven (7) weeks prior to shutdown, P1, P2, P3, !
- l. - P4,-P5, P6, P7; where P1 is the first week I
i _
~
prior to shutdown and is input first, P2 is input second, etc. The values are obtained from the log sheet maintained for both Peach - Bottom Units 2 and 3 and Limerick Unit 1. l (Notes Keep the decimal points on line 4 in i the name locations as shown in Appendix 6,- Figure 1, so the program will run correctly.) ,
: EP-C-326-6(C),acy.o 5 3/21/84 ;
Page 6 of 9 ; MJMilmd i 4.5 Fifth lin'es Type in the phrases " TIME", "RX I- ['
-CONC.", and " POOL I-CONC." which corresponds to
-the time difference between the point in time !
the vial-sample was taken and the time the j reactor was shutdown, the reactor water Iodine- l 131 concentration measurements, and-the ! suppression pool water Iodine-131 concentration ! measurements respectively. (Note: Remember to keep decimal points in the same locations as in : L Appendix 6, Figure 1.) [ I l L ! 4.6 Sixth line: Input the corresponding values for !
" TIME" "Rx I-CONC.", and " Pool I-Conc." as i found on form EP-C-326-1.- Note: If only one !
I-131 measurement is taken, leave the other ! blank. Also, remember to keep decimal points and exponentials in the same locatione as in' , i f -Appendix 6, Figure 1. l
- 5. Enter, using the SUVZ option, into dataset j "T423HJD. JOB.CNTL" member " DAMAGE".whichis the needed i
- job control 'o execute the PASS analysis program.
Change.(on line number FT05F001) the member name ! enclosed in parenthoses to "DAMAGEI". ' Submit the l program by typing "SUB" at the command input located i at the top of the job control member. ; I i i i
- 6. The Output of the iodine program shall show the !
E following datar the inputted post shutdown sample i time (Section 4.4), the inputted vial iodine 1
~
l concentration (Section 4.4), the calculated FI i (Inventory Correction Factor), the calculated CWREF (Iodine-131 concentration with respect to the il l =. . reference value), the estimated % of cladding failure. and the estimated % of. fuel meltdown. 5 i ! T 6.1 NOTE: The Program shall estimate the % [ cladding failed and-the % of fuel meltdown that has occurred using the data found on figure 1 & ~ i 2, respectively, of Appendix 6. The computer i output shall be assumed correct if reactor.is l lLn the shutdown mode. If the reactor is not fully shutdown (as is the case of an ATWS), the . l program will yield a conservative estimation of , L _ _ the relative fuel cladding damage within the ; reactor.. ! ? l i , i L i t ;
,p. -- -- -,
wr --,n, .... ,y ..,,,wm,-,,-,.--,, ,-,-,-,,,-,-,,,,,,-,,-,,,,,,,,,,,,,----,,e,,,--.
EP-C-326-(D) gey,y 3/21/94 7 L Page 7 of 9 MJM:lmd : Procedure for determing fuel damage estimate based on measured i
~~
Re!TTI Eoncentratf6n. ; t Logon to the PECO Corporate Computer System using normal j 1. logon procedures. i Using standard processes enter into the SPF2 format j (NOTE: SPF2 is used for all editing of dataset l information.) [ f 1 The necessary dataset and the corresponding member can ! 2.
- be found in dataset 'EP. DIAMOND. DATA 2' member ' DAMAGE l .XE'. ,
- 3. After entering into the proper dataset member, the t necessary data concerning the Post Accident Sampling l l System (PASS) must be entered into the proper format '
within the member (NOTE: A sample member and format is i shown in Appendix 6, Figure 2. g 4. The description of formatting is as follows: l 4 .1 - First line: Type in "PBAPS" of the procedure is to be run for Peach Bottom reactor or " Limerick" if ; the procedure is to' calculate fuel damage at a Limerick unit. 4 ! 4.2 Second line: ' Type in " Unit X" where 'X' , i corresponds to the unit number of the reactor for
. which this procedure is to be used.
It , j 4.3 Third line: Type in the word " Xenon" which will : l specify the xenon analysis to be run. ( t I 4.4 Fourth line: Input average power of each of the i four (4) weeks prior to shutdown, P1, P2, P3, P4. ,, t The first week prior to shutdown is input first, P2 is input second, etc. The values are obtained from the A-1 log-sheet maintained for both Peach Bottom Units 2 and 3 and Limerick Unit 1. (Notes Keep the decimal points on line 4 in the same locations l
- as shown in Appendix 6, Figure 1, so the program ; will'run correctly.) :
4.5 Fifth line: Type in the phrases " TIME", which i corresp6nds to the time difference between when the
- i f
. . t
. _ _ . . - - - - ~ . . . , , . , _ - . - - _ - - - - - - -._. ~ ,-- --.
= - -
c EP-C-32u-(v; nev., '
'1 : 3 Pagaf21/84 of 9 :
MJM:lmd i vial sample was taken and the reactor was , shutdown;"D/W'XE,-CONC." which. corresponds to the Xenon-133 concentration 1 measured in the torus; PDW" which corresponds to the pressure of the drywell; ,
"TDW" which is the temperature of the drywell; '
"PV", which is the temperature of the sample vial:
"PT", which is-the pressure of the torus; and "TT", i which is the-temperature of the torus.
4.6 Sixth line: Input the corresponding values for
" TIME" "D/W.XE-CONC.", " TOR XE-CONC.", "PDW", '
"TDW", "PV", "TV", "PT", .and "TT" as found on EPC 326-1. Note: If only one Xe~-133 measurement is i taken, leave the:other blank. Also, remember to i
, keep decimal points and exponentials in the same locations as in Appendix 6, Figure 1. l
- 5. Enter, using the SPF 2 option, into dataset ;
'T423HID. JOB.CNTL", member " Damage" which is the needed I job control to execute the PASS analysis Program. !
Change (on 1ino number PTO'3 F001 ) the memimr.iname ( wl i eli
.is enclosed in parenthesis) to "DAMAGEXE". Submit the {
program by typing "sub" at the command input located at , the top of the job control.
- 6. The output of the xenon program shall show the following i data: the; inputted post shutdown sample time (Section 4.4), the inputted xenon vial concentration (Section ?
14.4), the.drywell xenon sample concentration, drywell pr)ssure and drywell temperature, if inputted (Section 4.4), the torus zonon sample' concentration, torus t
. pressure,.and. torus temperature, if. inputted (Section !
'4.4), the inputted vial pressure and vial temporarure t
(Section 4.4). Also printed will be the calculated it corrected concentration of xenon taking into account the difference in pressure and temperature between the vial and the drywell and/or torus, the calculated FI !
.(Inventory Correction Factors), the calculated'CgREF ;
i (Xenon-133 concentration with respect to the reference ! lL .value) the estimated % of cladding failure, and the
- estimated % of fuel meltdown. (See Note in Part A, i l Section 6.1 of Appendix 6). j I -
h u r I I L : t i f F
. - ?
~~
. EP-C-326-6 Rev.O 3/ ?.1/ 84 Page 9 of ;
MJM:Imd i
+
i l I, ( l 2003NE 3393- 3393- 3893-3393. 3393 3393. 3 3 9 3.~ TINC 3* CONC. ' 0.07 3.000E+03
)
i
?
. i I
APPENO3) 6 FIGURE & I
?
i I I MENON I 3393 - 3393. 3393 3380 . . . 13MC MC. VIAL PDw VIAL P-VIAL TDW 0.01 3.000E+0C 10 300. 39.7 195. , l,
- 4 i
f APPENO3M 6 FIGURE I l ' t O . i
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W-C-320 i 3 /21/84 Rev . 0 ) Page 1 of 1 l; Xe-133 Concentration vs Fuel Damage MJM:lmd i Relationship between Xe-133 Concentration in the Containment ( } Torus Gas) and the Extent of Core Damage in Reference Plant ; e i . " r l Fuel Meltdown ! _ _ Upper Release Limit x -sF -- t t Dest Estimate -
-YY --"*
I
. .. . . 1 i i . ..
. .i Lower Release Limit '/ !'
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' ' \. Upper Release Limit .. . . 4
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' Lower Release Limit __
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if if: : i' i .! ! ; isi!: .; + i siiii' l 0 E F 1 I I 1 4 :
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f ,. .,, ii i. .1 , n ... 3 i t l ,
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d ' l ;'-- f ?!ormal Operation Conc. in Drywell -
.'. i . . . ., .o ! ' 4. Upper Limit
- 10'4 uCi/cc -
f ' Nominal : 10-5 uC1/cc
- " ~ ' ~ !
g, 5. --
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w - t t 1.0 10 100 !
~i t 3 Cladding failure l.0 50 100 k 5 Twel Meltdown i !
t
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o uu.. .;
- 3/21/84 nevJ I-131 Concen'. ration vs Fuel Damage Page 1 of 11 2 :lmd j Relationship between I-131 Concentration in the Primary Coolant ,
Wate.r
.+ Po,ol . . . . ..
Water) and the, . i iExtent ii ofe Core i..i.i Damage '.in Reference t
. ' i Fuel Meltdown ....I w.'
r--- :. , ,l l l .. upper Release Limit .g , ., .-...- , Best Estimate - - Lower Release Limit., q. 'Yy'
.: + = . , ..i-. s .4 i t u . I u 1. s . j, l gui M$dt N.: 14-!E %:.di-i-i .
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. 1 1 1 Sest Estimate
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j
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x . . . .m .en .s .e 1- i - ,_ ..- . . . . <
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- Nornal Shutdown Conc. in Reactor Water -
,, 3- . ,
'q' --
.. m ,< . .. . . , , a .... - i
..r -
, ,e ; :, Upper Limit : 29.0 uCi/g - . . . i
" w ,- .n. . . . . . w. . .. . . . .yom;na3 : 0.7 uCi/g .. 3 ;
- c. e. ,,
.qp w - < 1 .
rc2.., .
-~ n':TP 4iU Dn O .4
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1M 'I l 4
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t
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- - .1 . :2 ' ' '
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01 0.:1 1.0 10 100 1 5 Cla(ding Failure 4 . 10 190 i 1.0 .; Fuel Meltdown - i
- 1 i,
3
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