ML20117P143
| ML20117P143 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 05/13/1985 |
| From: | GULF STATES UTILITIES CO. |
| To: | |
| Shared Package | |
| ML20117P105 | List: |
| References | |
| TASK-2.B.3, TASK-TM COP-1050, NUDOCS 8505200420 | |
| Download: ML20117P143 (25) | |
Text
1 s 's, I RIVER BEND STATION 7 APPROVAL SHEET STATION OPERATING PROCF.DURES l NO. em -1050
- TITLE POST ACCIDENT ESIIMATION OF FUEL ;
CORE DAMAGE SAFETY RELATED YES NO[
TECHNICAL REVIEW REQClRED YES NO]
REV. PAGES INDEP. TECH.
NO. ISSUED REVIEW REVIEW APPROVED BY EFFECT SIGNATURE /DATE SIGNATURE /DATE SIGNATURE /DATE DATE O 1 THRU 23 q, '
r ,
y V 10/18/84 LATER'S J V M 2.7/ yff*/
PAGE v 1 1 Thru 17 LATER'S /fdn gr ///[83 //J f 01/15/85 PAGE 2 1 Thru 22 h/M /J/*f-/ W C~-/J -#[
5-n -es On i va C , ; L '/
CHEM i
8505200420 850513 PDR ADOCK 05000458 E PDR ,
SOM-OO3
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- 9, \,
RIVER BEND STATION PROCEDURE (CHANGE) REVIEW FORM PROC. NO. COP-1050 TITLE POST ACCIDENT ESTIMATION OF FUEL CORE DAMAGE REV. NO. 2 REASON FOR PROCEDURE (CHANGE): Incorporation of Coments YES NO l
}4l IS THIS A NEW PROCEDURE l l_ lXl DOES THIS REVISION CHANGE THE SCOPE OF THE PROCEDURE
-l l lXl DOES THIS REVISION CHANGE THE INTENT OF THE PROCEDURE
@ l 3' S THIS A REVISION TO A SAFETY RELATED PROCEDURE PREPARED BY E.Dreyer /o$-427- 6 RESPON. SUPV. [1.J[/
PROCEDURE (CHANGE) CLASSIFICATION YES NO lXl l l SAFETY RELATED/QA APPLICABLE lXl l l TECHNICAL REVIEW f EQ I D CLASSIFIED BY W DATE ~/ i TECH STAFF SIGNATUREp CROSS DISCIPLINE REVIEWS REQUIRED lXlALARA l l Operations l l Maintenance lXlTechStaff lXl Chemistry
.l lISICOORDINATOR lXlQA l lRADPRO l l Security l lAPM-S l lAPM-MAINT l lRadwaste lXlAPM-0/C/R l lOTHER TECHNICAL REVIEW 4M i DATE 7/ f SOM-001 m
RIVER BEND STATION NUCLEAR SAFETY EVALUATION APPLICABILITY CHECK LIST / SAFETY EVALUATION
, 10CFR 50.59 I
I l (1) PART A CHECK LIST APPLICABLE T0: 6ab/ofu Revision 2__ TCN l l (2) SAFETY EVALUATION APPLICALBIITY CHECKLIST - PART-A l 1 -
l l
This procedure, procedure changs or' modification to which this evaluation]
l is applicable represents:
l l (2.1) YES NO / A change to the plant as described in the FSAR? l l (2.2) YES NO / A change to procedures as described in the FSAR?l l (2.3) YES NO / A test or experiment not described in the FSAR? l
-1 (2.4) YES NO V A change to the Technical-Specification or l l Operating License?
1 l 1
I l
If the answer to question 2.1, 2.2, 2.3 or 2.4 is "YES", complete ITE!! l l (3). If the answer to all of the above is "N0" omit ITE!! (3) (8) l l and (9) i l I
I l
l(3) SAFETY EVALUATION - PART B l' (3.1) YES NO Will the probability of an accident previously l
l
.l evaluated in the FSAR be-increased? l l -(3.2) YES NO Will the consequences of an accident previously l l
evaluated in the FSAR be increased? l l (3.3) YES N0 ?!ay the possibility of an accident which is l I different than any already evaluated in the FSAR l l be created" l (3.4) YES NO Will the probability of a malfunction of l l 1 equipment important to safety previously l l
evaluated in the FSAR be increased. l l (3.5) YES NO Will the consequences of a malfunction of l l
equipment important to safety different that l l
any already evaluated in the FSAR be increased? l l (3.6) YES NO
!!ay the possiblity of ~a malfunction of equipment l l
l important to safety different that any already l evaluated the FSAR be created? l l (3.7) YES NO Will the margin of safety as defined in the i
l l
basis to any Technical Specification be reduced? l
[
}
l l If the answer to any of the above questions is "YES", an unreviewed safetyl l question may be involved. If so, NRC review /spproval prior.to implemen- l l tation is required (confer with licensing). Explain the basis for each l l answer (yes or no) on the SAFETY EVALUATION CONTINUATION SHEET.
l(4) ret 1 ARKS: (Attach additional pages if necessary) l l .
I e N m. 'l l(5) PREPARED BY: (MU DATE /3 WAR { l (6) SUPV. REVIEWED c DATE J-/,f -[>3 l l
(7) TECH STAFF REVIEW (F G. 1): M DATE 6[O/Ab i NA'!E l
l(8) FRC APPROVAL: DATE l l NA!!E l(9) ~ NRB REVIEW: l DATE S0t!-019a l Rev.-0
! POST ACCIDENT ESTIMATION OF !
l FUEL CORE DAMAGE !
I !
l !
1 i l
- TABLE OF CONTENTS
, l u
l SECTION PAGE N0.
1.0 PURPOSE / SCOPE / APPLICABILITY 3 ;
l
2.0 REFERENCES
3- ,
, 3.0 DEFINITIONS 4 !
.f4.0 REQUIRED EQUIPMENT / CHEMICALS 4
-l, 5.0 PRECAUTIONS 4 ,!
! 6.0 LIMITATIONS 4 l
7.0 PREREQUISITES / INITIAL CONDITIONS 5 l 8.0 PROCEDURE / INSTRUCTIONS / REQUIREMENTS 5 i li
! 8.1 Sequence of Analysis 5 l 8.2 Selection of Sample Point l 5 ;
8.3 Determination of the Core Damage from Fission Products 7 -j Concentration i
! 8.4 Determination of the Cladding Damage from Hydrogen 7 !
l Concentration in the Containment Atmosphere 1 8.5 Determination of Core Damage from Contaimment Radiation 8
-I Level ;
e i
9.0 RESTORATION 9 s
l i
I 10.0 DATA REQUIRMENTS 9 f !
l 1 l 11.0 ACCEPTAhCE CRITERIA 9 ,
l '
l I ;
l l
l l
v m
l
! l ll i !
l- N/A lI N/A !! COP-1050 lI REV. 2 PAGE 1 0F 22 t il I
lI
- e --- -
w --
- wtv---m -- e- *- y
e i
!- POST ACCIDENT ESTIMATION OF
.l FUEL CORE DAMAGE
. l l
L l
u l
-l l l TABLE OF CONTENTS l SECTION PAGE N0.
ATTACHMENTS i
l ATTACHMENT 1 - DECAY' CORRECTED FISSION PRODUCT CONCENTRATIONS 10 !
l ' ATTACHMENT 2 - FISSION PRODUCTS INVENTORY CORRECTION FACTORS 11 l l ATTACHMENT 3 - NORMALIZED CONCENTRATIONS OF THE FISSION PRODUCTS 13 l i
ATTACHMENT 4 - CORE DAMAGE ESTIMATES 15 l ATTACHMENT 5 - SEQUENCE OF ANALYSIS. 16 l ATTACHMENT 6 - I-131 CONCENTRATION VS FUEL CORE DAMAGE 17 l l ATTACHMENT 7 - Cs-137 CONCENTRATION VS FUEL CORE DAMAGE 18 l l ATTACHMENT Xe-133 CONCENTRATION VS FUEL CORE DAMAGE 19 l l ATTACHMENT 9 - Kr-85 CONCENTRATION VS FUEL CORE DAMAGE 20 l l ATTACHMENT 10 - CONTAINMENT H, VS CLADDING DAMAGE 21 l ATTACHMENT.11 - CONTAINMENT AIRBORNE ACTIVITY FROM RADIATION LEVEL 22 l
i .l l
l l
s ll l
l i !
. i l l i '
' l :
, l
, l
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l- _N/A l N/A ll COP-1050 l REV. 2 l PAGE 2 0F 22 ll l ! ll- l l l
l l l l 1.0 PURPOSE / SCOPE / APPLICABILITY li 1.1 The purpose of this procedure is to provide a preliminary estimate ,I of the reactor core damage from the measured fission product concen- l tration; under accident conditions, j i
l 1.2 The procedure involves calculation of fission product inventories of l the primary system under postulated loss of coolant accident j conditions. j i
1.3 A BWR-6 with a Mark III containment is used as a reference plant. ,!
Application of reference plant core damage graphs to the RBS reactor j
, is possible by applying pertinent correction factors discussed in j j the procedure. l 1 4 l 1.4 The fuel gap fission products are assumed to be released upon the l rupture of fuel cladding. j i
1.5 The majority of fission product inventories in the fuel rods would l l be released when the fuel is melted at high temperatures. j i i l 1.6 The fission product concentrations are obtained by sampling the l l primary coolant and/or the containment atmosphere via post accident i j samplingsystem(PASS). q i
l 1.7 The Chemistry / Core Damage Assessment Coordinator will perform the
- core damage assessment estimate and report the results to the l Technical Support Center Emergency Director. l L
2.0 REFERENCES
2.1 NED0-22215 82NED090, Procedure for the Determination of the Extent of Core Damage, Under Accident Conditions j 2.2 USNRC Reg Guide 1.97, Instrumentation for Light Water Cooled Nuclear [
- Power Plants to Assess Plant and Environs Conditions During and u
Following an Accident, 1980 ,
i 2.3 NED0-30088, Responses to NRC Post-Implementation Review Criteria for l Post-Accident Sampling System j }
l 2.4 RBS FSAR, Section 13.3.5.2, Emergency Planning Assessment Actions 2.5 RBS FSAR, Volume 1, Chapter 1.1, Introduction and General Descrip-tion of Plant j 2.6 RBS FSAR, Volume 8, Table 4.2-4, Fuel. Data and Table 4.3-1 Reactor j j Core Dimensions j i
i u i
! e
! i
. ii . I i ll l l
, N/A ', N/A ',' COP-1050 ', REV. 2 ,
PAGE 3 0F 22 I I 15 1 1 I
F j j
. i l 2.7 US Atomic Energy Commission, Safety Evaluation of the River Bend Station, Sept. 1974 l l ;
I
- I l 2.8 EIP-2-015, Post-Accident Sampling Operations l
1 2.9 '
CORDAM. BAS, Chemistry Computer Routine l 2.10 COP-0425, Determination of the H and 0 Gas - Gas Chromatography 2 2 ,
l Method u 1 1
li 2.11 COP-1001, Post-Accident Sampling of Primary Coolant j 1
i 2.12 C0P-1002, Post-Accident Sampling of Containment Atmosphere ) ;
2.13 C0P-1030, Post-Accident Isotopic Analysis for Particulate, Iodine and Gaseous Activity u 2.14 COP-1033, Post-Accident Isotopic Analysis for Liquid Activity
, 3.0 DEFINITIONS e l 1 l 3.1 Metal-Water Reaction - The reaction between Zircaloy from the fuel i
- cladding and the water under accident conditions is called Met- l
-l al-Water reaction. % Metal-Water reaction represents % cladding ;
j failure. j i
i li 3.2 PAS - Post-Accident Sample li l' 4.0 REQUIRED EQUIPMENT / CHEMICALS !
j N/A ;
l l
I l
5.0 PRECAUTIONS '
Ir N/A ! I 6.0 LIMITATIONS '
l 1 6.1 The performed estimates are done under the presumption that no j reactor coolant cleanup systems are operated after the accident.
[
6.2 Measurement of Cs-137 and Kr-85 activities may not be possible until j
, the reactor has been shut down for several weeks to allow the decay j
-l of shorter lived isotopes. l l l l 6.3 If isotopes of Sr, Ba, La and Ru are found in significant concen-
" l trations some degree of fuel melting may be inferred. However, the j extent of fuel melting cannot be determined based on the concen- l l g trations of these nuclides because of the lack of baseline data, j l
l i
, ! i i il l l l i
N/A l 1
N/A ll 18 C0P-1050 lI REV. 2 li PAGE 4 0F 22 ll ;
i l
. . _ - - - ~.. . . . . _ . _ _ _ _ _ - . _ . . _ _ _ _ _ _
I i.
o g
. i -
l l
6.4 Containment hydrogen concentration measurements via PASS or contain-j ment gas analyzers provides a measure of the extent of metal-water l l
l reaction which is used as an estimate of cladding damage. l I
6.5 Core damage below 1 % is assumed to be a non-accident condition in j view of cladding failure, j l
7.0 PREREQUISITES / INITIAL __ CONDITIONS 7.1 Post-Accider.t sampling and analysis completed per References 2.10, thru 2.14. !'
8.0 PROCEDURE / INSTRUCTIONS / REQUIREMENTS s
l i
8.1 Sequence of Analysis j l
l 8.1.1 Attaciiment 5 shows as an example the sequence of steps to l
- assess the extent of damage to the- core, after its pos
- ibility has been established. Indicators for core damage l
, or its possibility are increased radiation levels (i.e. l l Off-Gas Pre-Treatment, Main Steam Line, Drywell Post l Accident Area Monitor) or decreased water levels (i.e., j Reactor water level less than 160") l l
8.1.2 The sequence of analysis from Attachment 5 may be changed by l the Chemistry / Core Damage Assessment Coordinator according l to the availability of data and/or information about the l extent and course of the accident which caused the core l ll damage. (Subsections of this procedure may therefore be l performed out of sequence.) l l
1 8.2 Selection of Sample Point NOTE h Depending on the actual extent of the l l accident, the Chemistry / Core Damage j Assessment Coordinator might require the l analysis of additional samples taken from l the Drywell and/or Suppression Pool, under l consideration of the res volumes l (Item 10.1, Section 10) pective to correct the individual fission product concentration l l l
J results of the reactor coolant or the l l
containment atmosphere. l l
l
, l l l l
l
. .. . , l 1 i ll 1 l l i
N/A l N/A ll ii COP-1050 l REY. 2 l PAGE 5 0F 22 l i
i
!- l l !
l 8.2.1 Refer to the following table for selecting the recommended l sample location for gas samples depending on the event type causing the core damage: l i
,'
- Event Type Sample Location !
n li Non-Breaks (e.g., MSIV Closure) Containment Atmosphere j l n
I Small Breaks Drywell (before depress.)
I Containment Atmosphere i
,! (afterdepress.) !l u
1 l Large Breaks (liquid or steam) Drywell ,
j in Containment Large Breaks outside containment Containment Atmosphere I
! 8.2.2 For liquid sampling per Reference 2.2, the optimum sample point for all events is the jet pumps as long as there is ll j sufficient reactor pressure to provide a sample location. If n
there is not sufficient reactor pressure to allow a sample i l to be taken from the jet pumps, then the sample should be
- taken from the sample point in the RHR system.
i NOTE il Refer to Attachment 3 for explanation of j symbols. i l
L
! 8.2.3 If separate measurements of the activity concentrate.. are
! made in the containment and drywell atmosphere, the li resultant average concentration is calculated for RBS by using. u ll C =C Drywell x 0.174 + CCont. x 0.826 i
U 8.2.4 Fission product concentration correction for temperature and pressure difference in sample vial (T l y
y' P) y and in ;
containment (Tc, P c ) are : ;
PT cv u
C9j(Cont) =C x 99(vial) PT yc 1
h 8.2.5 If separate measurements of the specific activity are made for the suppression pool and reactor water, the resultant
- l average concentration is calculated for RBS by using
C,j =C Rx x 0,053 +C supp x 0.847 e !
i I I II I i !
! N/A I N/A !! COP-1050 ! REV. 2 I PAGE 6 0F 22 !
i e is e i i
+
i .
8.3 Determination of Core Damage from Fission Products Concentration 1
8.3.1 Obtain primary coolant and/or containment atmosphere sample ,
l
, .from the post-accident sampling system per Reference 2.11 ;
j and/or Reference 2.12 and record sample time on Attachment j
! I-l 8.3.2 Obtain fission products concentration in the samples per Reference 2.13. Record concentrations of I-131, Cs-137, Xe-133 and Kr-85 on Attachment 1. !
, NOTE
, L e
f Calculations in Steps 8.3.3 to 8.3.6 are l performed per computer by using Reference j
, 2.9. j L
8.3.3 Record the shutdown time on Attachment 1 and correct
, measured concentrations for decay to the time of reactor j shutdown.
l 8.3.4 Calculate the fission products inventory correction factors as per Attachment 2. l l j i
i li 8.3.5 By using the correction factors determined in Sections 10.2 thru 10.4, calculate the normalized concentrations of the l j
fission products as per Attachment 3. j i
( 8.3.6 Use Figures 1 through 4 (Attachments 6 through 9) to esti- l j mate the extent of fuel or cladding damage. Record the j j results of the estimated damage on the Attachment 4. l i i
! 8.4 Determination for Cladding Damage from Hydcogen Concentration in the l Containment Atmosphere 8.4.1 Determine hydrogen concentration in the containment atmo-sphere post accident sample as % H 7 (dry basis) per Refer-ence 2.10. The Containment Hydroggn monitor reading may be used as an alternative.
8.4.2 Using a curve for Mark III Containment in Fig. 5 (Attachment !
10), determine the cladding damage (% metal-water reaction) l for the reference plant. l l
l I
l
\
l l
l . .. , l l l ll '
l N/A l N/A llei C0P-1050 ;
i REY. 2 ,
i PAGE 7 0F 22 li w
e 4/
t 8.4.3 The % metal water reaction (% MW) for RBS is assumed to be the same as the % metal water reaction for the reference
. plant (% MW Ref.) within the accuracy of estimate, since:
%W = % W ref
- N
- 1.36 E6
% MW ref x 1.05
=
where = N =
RBS number of fuel bundles = 624 V =
RBS Containment Air Volume = 1.443 E6 ft3 8.4.4 Record the results of estiniated damage from Step 8.4.2 on Attachment 4.
8.5 Octermination of Core Damage from the Containment Radiation Level NOTE This method can only be used if substantial core damage and release of fission products into the containment has occured.
8.5.1 Obtain reading of drywell radiation monitor, (R) in R/hr.
8.5.2 Determine elapsed time from plant shutdown to the containment radiation mon; tor reading (t) in hours.
8.5.3 Using Attachment 11, determine the fuel inventory release for the reference plant (I)ref in %.
8.5.4 Determine the inventory release to the containment (I) in %
using the following formula:
(I) = (I)ref
~
) (237 450) (6/D) x 6.09 E4
=(I)ref P where P = reactor power level, MW th V = total containment free volume, 1.443 E6'ft3 0 = distance of detector from reactor biological shield wall, 19.6 ft 8.5.5 Record the result of Step 8.5.4 on Attachment 4.
N/A N/A COP-1050 REV. 2 PAGE 8 0F 22
I, 1
l !
! 8.5.6 Retain all pertinent data, calculations, etc. for permanent !
l , retention, j
\
i 9.0 RESTORATION N/A l
10.0 DATA REQUIREMENTS l
10.1 River Bend Station General Information l 10.1.1 MWT - 2894 j i
i li 10.1.2 Assemblies - 624 li l 10.1.3 Fuel Rods / Assembly - 62 i
i l
li 10.1.4 Water Rods / Assembly - 2 l\\
l 10.1.5 Reactor Coolant Mass 1.99 E8 gm 10.1.6 Supression Pool Volume - 126,600 ft3 (3.57 E9 ml) ll 10.1.7 Total Volume Reactor Cool. + Supr. Pool - 3.77 E9 ml (i 10.1.8 Containment Net Free Ai' Volume - 1.192E6 ft3 (3.38 E10 ml) li
, 10.1.9 Drywell Net Free Air Volume - 2.51E5 ft3(7.11E9ml) li i
,! 10.1.10 Cont. + Drywell Net Free Air Vol. - 1.443E6 ft3 (4.09 E10ml) li
\
10.2 Inventory Correction Factor (Fyj) ,!
p , Inventory of Nuclide i in the Reference Reactor Core li
- 11 Inventory of Nuclide i in the RBS Reactor Core 10.3 Containment Gas Volume Correction Factor (F ) ,
i 9 i .
I = RBS containment gas volume, (4.09 E10 cc)
I Reference plant Containment Gas Volume (4.0 E10 cc) = 1.02 F
j g j i i le l'0. 4 Primary Coolant Mass Correction Factor (Fg) li l p , Operation Plant Coolant Mass (3.77 E9 gm) = 0.96 l w Reference Plant Coolant Mass (3.92 E9 gm) l i i l 11.0 ACCEPTANCE CRITERIA N/A I
]
"END" l !l
! I li ; i !
! N/A ! N/A I! C0P-1050 ! REV. 2 ! PAGE 9 0F 22 !
! ! ll l l l
f o ATTACHMENT - 1 DECAY CORRECTED FISSION PRODUCT CONCENTRATIONS Fission Prcduct Concentrations Shutdown Time, T : hrs , Date o
Sample Time, T: hrs , Date s
Decay Time, T =(Ts -T g ): hrs + 24 = day
- C 4 = _ Fission Product Concentration of the Post Accident Sample.
%, = Decay Constant _of the nuclide, day-1
- C iT = Decay Corrected Fission Product Concentration.
T cit. = Ce i g
FISSION PRODUCT Cg 2 9 T ,2 iT C iT I - 131 8.62 E-2 Cs - 137 6.29 E-5 Xe - 133 1.32 E-1 Kr - 85 1.77 E-4 concentrations expressed in uCi/gm for liquids and uC1/ml for gases.
I Chemistry Core Damage Assessment Coordinator Date Time ATTACHMENT - 1 PAGE 10F 1 COP-1050 REV. 2 PAGE 10 0F 22
r-ATTACHMENT - 2 FISSION PRODUCTS INVENTORY CORRECTION FACTORS Inventory Correction Factors (Fy$)
Equation 1 p Inventory in reference plant of Isotope i If , Inventory ir operating plant of Isotope i
-109529 3651 (1 - e )
(P) (1 - e- ) e- }
where:
=
P 3
Steady reactor power operated in period j, MWt T
3
=
Duration of operating period j, days
=
Tj Time between the end of operating period j and time of the final reactor shutdown, days hj = Decay correction factor of the isotope, days-1 3651 =
Avg. Operation Power (in MWt) for the reference plant 1095 =
Continous Operation time (in Days) for the reference plant Using Equation 1, compute F yj for the fission products using 2 4values provided in the following table. A sample calculation is shown on Page 2 of Attachment 2. Attach Calculation Sheets.
FISSION PRODUCT 21 INVENTORY CORRECTION FACTOR, F y9 I-131 8.62 E-2 Cs-137 6.29 E-5 Xe-133 1.32 E-1 Kr-85 1.77 E-4 ATTACHMENT - 2 PAGE 10F 2 COP-1050 REV. 2 PAGE 11 0F 22
ATTACHMENT - 2 FISSION PRODUCTS INVENTORY CORRECTION FACTORS g
Sample Calculation of Fission Product Inventory Correction Factors, F gg Assuming a reactor has the following power operation history:
Operation Operation Time A
- Period T (day) Tj Days Since Startup_ 3 1A 1 - 60 60 254 (= 314 - 60) 1000 18 61 - 70 --- ---
0 2A 71 - 270 200 44 (= 314 - 270) 2000 28 271 - 300 --- ---
0 3 301 - 314 14 0 (= 314 - 314) 3000
- For I-131 ( 2 = 0.0862 day-1) 3651(1-e-0.0862x1095)
I p(I-131), 1000(1-e -0.0862x60),-0.0862x254 + 2000(1-e-0.0862x200) e -0.0862x44 + 3000(1-e-0.0862x14)e-0.0862x0
= 3651
" I'7 0 + 45 + 2103
- For Cs-137 ( % = 6.29 x 10-5 day -1) p , 3651(1-e-6.29x10-5x 1095) s-137) 4 1000(1-e-6.29x10 x 60),-6.29x10-5 x 254
+ 2000(1-e-6.29x10-5x200),-6.29x10-5 x 44 243.16
- 7.77 x0 3.74+24.93+2.64
+3000(1-e-6.29x10-5x14)c-6.29x10-5 Chemistry Core Damage Assessment Coordinator Date Time
- ATTACHMENT - 2 PAGE 2 0F 2 COP-1050 REV. 2 PAGE 12 0F 22 L..._
l ATTACHMENT - 3 NORMALIZED CONCENTRATIONS OF THE FISSION PRODUCTS
- 1. Normalized Concentration of Primary Coolant Fission Products, CRf
= C it xF gg x F, (A)
= C it xF ig x 0.96 1.1 Normalized Concentration of I-131 C
it
=
uCi/gm (From Attach. 1)
F gg = (From Attach. 2)
Substituting in equation (A) for I-131, CRef= x x 0.96
= uCi/gm 1.2 Normalized Concentration of Cs-137 C
it
= uCi/gm (From Attach 1)
Fy9 = (From Attach. 2)
. Substituting in equation (A) for Cs-137 C{I= x x 0.96
= uCi/gm Ref
- 2. Normalized Concentration of Containment Atmosphere Fission Products, Cgj
= Cit
- Ili
- I g (B)
=C it xF yj x 1.02 I 2.1 Normalized Concentration of Xe-133
=
C it uC1/ml (From Attach. 1)
=
F 79 (From Attach. 2)
SubstitutingforXe-133,inequation(B).
CRef= x x 1.02 g
=
uCi/ml ATTACHMENT - 3 PAGE 10F 2 C0P-1050 REV. 2 PAGE 13 0F 22
ATTACHMENT - 3 NORMALIZED CONCENTRATIONS OF THE FISSION PRODUCTS 2.2 Normalized Concentration of Kr-85
=
C it uCi/ml (Frai Attach. 1)
F gg =
(From Attach. 2)
Substituting for Kr-85, in equation (B).
Ref C,=
g x x 1.02
=
uCi/ml Record the calculated normalized concentrations in the box below and use the information to compute core damage using Figures 1 through 4, (Attachments 5 through 8).
I-131 uCi/gm Cs-137 uCi/gm Xe-133 uCi/ml Kr-85 uCi/ml Chemistry Core Damage Assessmenc Coordinator Date Time ATTACHMENT - 3 PAGE 2 0F 2 COP-1050 REV. 2 PAGE 14 0F 22 1
1
r m
ATTACHMENT - 4 CORE DAMAGE ESTIMATES
- 1. Core Damage Estimates based in the Fission Products Concentrations in the post accident samples
. Record results obtained based on concentrations reported in Attachment
'3.and using Figures 1 through 4 in the following table.
i FISSION PRODUCT % CLADDING FAILURE % FUEL MELTDOWN I-131
.Cs-137 Xe-13 Kr '
-2. Cladding failure estimate based on the Hydrogen Concentration in the containment Record results obtained per Step 8.4.2:
% MW reaction
(% Cladding Failure) 3._ Core damage estimate based on radiation level in the drywell caused by airborne fuel invertory of fission products per Step 8.5.5
% Core Damage
(% Inventory airborne) l Chemistry Core Damage Assessment Coordinator Date Time
- ATTACHMENT - 4 PAGE 10F 1 COP-1050 REV. 2 PAGE 15 0F 22
- .= ... .__ . _ _ . . . . , .
e ATTACHMENT - 5 SEQUENCE OF ANALYSIS Hydrogen Containment Analysis M RedIation M Water Level Y'S :
IIDIMAL OPEllATIOR MIIIOR CLA0 GRPIAE (Confim) (ConfIm) (Confim) b E E E Setemine - Com Osange Optisun Estiaste*
Sample From PA55 -
Point ,,
~
E E E Hydrogen Containment
- Analysis M Radiation M Wa ter Level ~ M Analysts For Ba. Sr. La, Ilu (Conftm) (Confim) (Conff m)
MMOR CLA0 04,ME Deteminetten FEL OVEletEAT FUEL E LT g Of Fission Product Rattes E
.CuOanimat -
POS$18LE FWL OVElWEAT NO CORE ELT f
(
l I
RPC-000.216 ATTACHMENT - 5 PAGE 10F 1 COP-1050 REV. 2 PAGE 16 0F 22
. I
. l I
ATTACHMENT - 6 I-131 CONCENTRATION VS FUEL CORE DAMAGE 10E PUSL400LTOOWN
. UPPER RELEASE Laess?
SEST ESTIMATE /
/
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[ ! BEST ESTIMATE
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f f NORMAL SMU TDOWN y CONCENTMATION j IN mE ACTom WATER
/ UPPER LIMIT E0 mC./s
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a, . .a....i . . . .....I . . . .....i . .......I . . . .....
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'g % CLADOING P AILumE r',
1.0 10 100
'C
, % FUE L MELTOOWN r'i Fig. 1 I-131 Concentration in the Primary Coolant vs the Extent of Core Damage in the Reference Plant RPC-000.161 ATTACHMENT - 6 PAGE 10F 1 COP-1050 REV. 2 PAGE 17 0F 22
ATTACHMENT - 7 ; Cs-137 CONCENTRATION VS FUEL CORE DAMAGE g
g.
PUSL MELT 0cuses upper RELEAa8 uusT SGST ESTIMATE /
/
LourER RELEASE LausT
'Y
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'; s puSL uGLToowN j Fig. 2 Cs-137 Concentration in the Primary Coolant vs the Extent of Core Damage in the Reference Plant
, RPC 000.162 ATTACHMENT - 7 PAGE 10F 1 COP-1050 REV.2 PAGE 18 0F 22
w .w._, -
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ATTACHMENT - 8 Xe-133 CONCENTRATION VS FUEL CORE DAMAGE 1 1h ,
PUGL MELT 90usN
, MR RELEAM LL*fe?
oest astiMATE /
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l Fig. 3 Xe-133 Concentration in the Containment vs the Extent of Core E FUEL MSLTDOWN rl Damage in the Reference Plant RPC-000.163 ATTACHMENT - 8 PAGE 10F 1 COP-1050 REV. 2 PAGE 19 0F 22
. : i ATTACHMENT - 9 g Kr-85 CONCENTRATION VS FUEL CORE DAMAGE is' ,
{ PUE4 tsELTDOUUIL8 upptR RELEAct Laem?
eEst EstiesATE LouvEn mELEAst uns 7 p se - /
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/ leOMMAL QPEm ATION f CONCENTmATION
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o.i s.o io 500 f % CLAOCING F AlLumE 30
$.0 100
': s puGL asELToown i
Fig. 4 Kr-85 Concentration in the Containment vs the Extent of Core Damage in the Reference Plant RPC 000.$64 ATTACHMENT - 9 PAGE 10F 1 COP-1050 REV. 2 PAGE 20 0F 22 l
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, ATTACHMENT - 10 CONTAINMENT H2 VS CLADDING DAMAGE I
as MARKN CONTAINMENT " "
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( % Cladding Failure)
R PC-000.165 ATTACHMENT - 10 PAGE 10F 1 COP-1050 REV. 2 PAGE 21 0F 22
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