rev 0 to emergency procedure epp-9, 'determination of core ... · nine mile point nuclear...
TRANSCRIPT
NINE MILE POINT NUCLEAR STATION
EMERGENCY PROCEDURES
PROCEDURE NO ~ EPP-9
DETERMINATION OF CORE DAMAGE UNDER ACCIDENT CONDITIONS
DATE AND INITIALS
APPROVALS SIGNATURES REVISION 0 REVISION l REVISION 2
Chemistry & RadiationManagement Superintendent
'Station SuperintenNMPNST. W. Roman
~cY6i~ .
General SuperintendentNuclear GenerationChairman of S.O.R.C.T. J. Perkins
Summar of Pa es
Revision 0 (EffectivePAGE
iiii>1-26
DATE
March 1984
NIAGARA MOHAWK POWER CORPORATION
THIS PROCEDURE NOT TO BEUSED AFTERSUBJECT TO PERIODIC REVIEW ~
,;„ 8403i30391 84030S.'. PDR ADOCK 05000220""
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EPP-9
DETERMINATION OF CORE DAMAGE UNDER ACCIDENT CONDITIONS
TABLE OF CONTENTS
1.0 PURPOSE
2.0 PROCEDURE BASIS
3.0 EQUIPMENT REQUIRED
4.0 RESPONSIBILITIES
5.0 PROCEDURES FOR DETERMINATION OF CORE DAMAGEI
6.0 REFERENCES
TABLES
TABLE TITLE
FLOW CHARTS
Core Inventory of Major Fission Products in a ReferencePlant Operated at 365l MWt for three
years'ission
Product Concentrations in Reactor Water and DrywellGas Space During Reactor Shutdown Under Normal Conditions
Ratios of Isotopes in Core Inventory and Fuel Gap
Best — Estimate Fission Product Release Fractions
Sequence of Analysis for Estimation of Core Damage
EPP-9 -i March 1984
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EPP-9
DETERMINATION OF CORE DAMAGE UNDER ACCIDENT CONDITIONS
TABLE OF CONTENTS
(Continued)
FIGURES
PIGURE TITLE
Relationship Between I-131, Concentration in the PrimaryCoolant (Reactor Water and Pool Water) and the Extent ofCore Damage in Reference Plant
Relationship Between Cs-137 Concentration in the PrimaryCoolant (Reactor Water and Pool Water) and the Extent ofCore Damage in Reference Plant
Relationship Between Xe-133 Concentration in theContainment Gas (Drywell and 'jtorus Gas) and the Extent ofCore Damage in Reference
Plant'elationship
Between Kr-85 Concentration in the ContainmentGas (Drywell and Torus Gas) and the Extent of Core Damagein Reference Plant x
Hydrogen concentration for Mark I/II Containments as aFunction of Metal-Water Reaction
APPENDICES
Appendix
Sample Calculation of Fission Product Inventory Correctic nFactor
EPP-9 -ii March 1984
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EPP-9
DETERMINATION OF CORE DAMAGE UNDER ACCIDENT CONDITIONS
1 ~ 0 PURPOSE
The purpose of this procedure is to determine the degree of reactorcore damage from the measured fission product concentrations ineither the reactor water or containment gas samples taken underaccident conditions. The procedure involves calculations of fissionproduct inventories in the core and the release of inventories intothe reactor water and containment atmosphere under postulated designbasis loss-of-coolant accident condi.tions. The fuel gap fissionproducts are assumed to Qe released upon the rupture of fuelcladding; the majority of fission product inventories in the fuelrods would be released when the fuel is melted at higher temperatures.
After the initial core damage estimate is made, confirmation and
refinement of the analysis can be achieved using the approachoutlined in Flow Chart 1 and Section 6;0 ~ of this procedure. Thisincludes assessment of core damage usidg (a) Containment hydrogenanalysis (b) Containment High Radiation monitors (c) Water levelindications and (d) Ba, Sr, La, Ru analyses.
2.0
2.1
PROCEDURE BASIS
Reference Plant
The estimation of core damage will be 'calculated by comparing themeasured concentrations of= major fission products in gas and/orliquid samples, after appropriate normalization, with reference plantdata from a BWR-6/238 with Mark III containment. Fission productinventories in the primary system were calculated based on postulateddesign basis loss-of-coolant accident conditions after three years(1095 days) of continuous operation at 3651 MWt or 102K of ratedpower by using a computer code developed at Los Alamos and adopted tothe GE computer system. The inventories of major fission products inthe core at the time of reactor shutdown are given in Table 1.
2 ~ 2 Parameters for Reference Plant and NMP-1
The pertinent plant parameters for the reference plant and the NineMile Point —Unit 1 plant are given below:
Rated reactor thermal power
Number of fuel bundles
Reference Plant
3579 MWt
E91P-1
1850 MWt
748 bundles 532 bundles
EPP-9 -1 March 1984
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2.2 (Cont.)
Reference Plant h>IP-1
Reactor water mass
Suppression pool water mass
Total primary coolant mass (Reactorwater plus suppression pool water)
3.67 x 109g
3.92xl09g
2. 16xl09g*
2.38x10 g*
2.46 x 108g 2.17xl08g
Drywell gas volume
Torus/Containment gas volume 32. 5x109cc 3.70x109cc*
7.77xl0 cc 5.10xl0 cc
Total containment and drywell gasspare volume 4.0x10 cc 8.80x10 cc*
*assumes torus downcomer submergence of 3 ft. (570,000 gal total)., Adjust ifnecessary to account for HPCI or Containment Spray'.Raw Water Additions.
I
3.0 EQUIPMENT REQUIRED
3.1 Apparatus
3 ~ 1 ~ 1
3.1. 2
GeLi — 1 and GeLi — 2 Gamma Spectroscopy System
\'ppropriatedilution equipment as specified in S-CAP-60 and Nl-PSP-13
3.2 Reagents
None
4.0 RESPONSIBILITIES
4.1 The Chemistry and Radiation Management Department is responsible forperforming sampling and analysis of reactor water and containmentatmosphere as necessary to support the calculations of Sections 5.0and 6.0 (See Sections 5.2.1 through 5.2.3, 6.1 and 6.4).
4.2 The Reactor Analysis Department is responsible for performing fissionproduct inventory correction factor calculations (See Section 5.6).
4 3 The Technical Support Department is responsible for calculatingcore damage in accordance with the methodology of Section 5.3, and ofSections 5.4, 5.5, 5.7, 5.8 and 5.2.8, or the methods of Section 6 1,6.3 or 6.4, based on the isotopic data and inventory correctionfactors supplied.
EPP-9 -2 March 198 4
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5.0 PROCEDURES FOR DETERMINATION OF CORE DAMAGE
5.1 Description
Gas/water samples taken from the Post Accident Sampling system areanalyzed for major fission product concentrations with the GeLi-1 orGeLi-2 gamma spectrometers. After incorporation of appropriate decay"and normalization correction factors to the isotopic analysis resultsfor I-131, Cs-137, Xe-133 and Kr-85, the extent of fuel or claddingdamage; can be determined by reference to Figures 1 through 4-Measurements of Cs-137 and Kr-85 activities are not very likely untilthe reactor has been shut down for longer than a few weeks and mostof the shorter-lived isotopes have decayed.
If the concentration falls into a range where the release of thefission product from the fuel gap or from molten fuel cannot bedefinitely determined,'additional data may be needed to determine thesource of fission product release. For example, in addition tolonger-lived isotopes, some shorter-lived isotope concentrations maybe measured in the sample. The ratios of isotopes released fromeither the fuel gap or from the molten fuel are significantlydifferent as shown in Table 3, thus the source (fuel or gap) ofrelease may be identified.
(Refer to Section 5.3). Furthermore, some less volatile elements inthe core may also start to release as the fuel starts to melt. Ifthe less volatile fission products such as isotopes of Sr, Ba, La,and Ru are found to have unusually high concentrations in the watersample as compared to baseline reactor water concentrations, sjnedegree of fuel melting may be assumed. The isotopes 2.7h Sr-92(1.384 MeV) and 40h La-140 (1.597 Mev) in a mixture of fissionproducts should be relatively easy to identify and measure from agamma spectrum.
5.2 Estimation Procedure
5.2.1 Obtain the samples from the Post Accident Sampling System inaccordance with Nl-PSP-13, "Sampling and Analysis of Reactor Waterand Containment Air Using the PASS".
5.2.2 Using the GeLi-1 or GeLi-2 Gamma Spectrometer, determine th econcentrations of fission products, namely I-131, Cs-137, Xe-133, andKr-85. ( wi in water, Cgi in gas)
5.2.3 Correct the measured concentrations for sample dilution, pressure anddecay (to the time of reactor shutdown). See steps 5.6.1.4 and5.6.1.5 of Nl-PSP-13.
5.2.4
5.2 5
Correct the measured gaseous activity concentrations for temperatureand pressure difference in the sample vial and the containmentatmosphere per Section 5.4 of this procedure.
I
Calculate the fission product inventory correction factor, Ii, perSection 5.6.
EPP-9 -3 March 1984
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5.2.6 Calculate the plant parameter correction factors ( w and g) perSection 5. 7.
5.2. 7Ref Ref
Calculate the normalized concentrations, wi or gi by using th ecorrection factors per Section 5.5.
5.2.8Ref Ref
Utilize wi or gi to estimate the extent of fuel or claddingdamage from Figures 1-4.
5.3 Identification of Release Source by Isotopic Ratio
5.3.1 Determine the concentrations of the shorter-lived isotopes shown inTable 3 with the GeLi-1 or GeLi-2 Gamma Spectrometers.
5.3.2
5.3. 3
Correct the measured fission products to the time or reactor shutdown.
Calculate isotopic ratios where
Noble Gas Ratio noble as isotope concerltrationXe-1 concentration
Iodine Ratio iodine isoto e concentrationI-131 concentration
5.3.4
5.4
Determine release source by comparing results obtained in Sectio'n5.3.3 to the noble gas and iodine ratios supplied in Table 3.
l'emerature and Pressure Corrections for Gas Sam le Vial
Cgi = Cgi (vial) x P2 Tl
Pl T2where
Cgi (vial)Cgi
P2 T2
= sample vial isotopic concentration= containment isotopic concentration= atmospheric pressure and temperature, respectively
(i.e., 14.7 psia and 298'K)containment pressure and temperature, respectively
5.5Ref Ref
Calculation of Normalized Concentration wi or ~iNOTE: Omit isotope decay correction if already accounted for in
step 5.2.3.
Ref ~itCwi = Cwje x FIi x Fw
or
Ref Kitgi Cgie x Ii x g
EPP-9 -4 March 1984
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5.5 (Cont.)
where
Ref
RefCgi
concentration of isotope i in the reference plant coolant(uci/g)
concentration of isotope i in the reference plantcontainment gas (uCi/cc)
Cgi
Cwi measured concentration of isotope i in the operatingcoolant at time, t (uCi/g) (See Section 5.8)
measured concentration of isotope i in the operatingcontainment gas at time, t (~iCi/cc) (See Sections 5.4 and5.8)
decay correction to the time of reactor shutdown
decay constant of isotope i (day 1)
FIi
Fg
Fw
time between the reactor shutdown and the sample time (day )
inventory correction factor for isotopic i (see Section 5.6)
containment gas volume correctiori factor (see Section 5.7)
primary coolant mass correction factor (see Section 5.7)
5.6 Fission Product Inventor Correction Factor
NOTE: See Appendix A for an example
FIi = Inventory in reference lantInventory in operating plant
)0
XiTj) e
-1095 >i3651 (1-e
where:
P j = steady reactor power operated in period j (1Ãt)*"
duration of operating period j (day)**
time between the end of operating period j and time ofthe last reactor shutdown (day)
**In each period, the variation of steady power should b< limitedto + 20%.
EPP-9 -5 Harch 1984
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5.6 (Cont.)
For a particular short-lived isotope, i, a calculation for only a
period of approximately 6 half-lives of reactor operation time beforereactor shutdown should be accurate enough. It should be pointed outthat the computer calculation of core inventory takes into accountthe fuel burnings, plutonium fission and neutron capture reactions.The correction factor calculated from this equation may not beentirely accurate, but the error is insignificant in comparison tothe uncertainties in the fission product release fractions (Table 4)and other assumptions.
5.7 Plant Parameter Correction Factors
Fw ~
Fg
operating lant coolant mass ( )*reference plant coolant mass (3.92 x 1 g
o crating lant containment as volume (cc)*)reference plant containment gas volume x 10 cc
5.8 Sample Concentration (Cwi or Cgi) Averagin'
If the fission product concentrations are measured separately for thereactor water and suppression pool water or the drywell gas and thetorus gas, the measured concentrations wi of gi would beaveraged from the separate measurements:
Cwi = (conc. in Rx water)x(Rx water mass)+(conc. in ool)x(pool water mass)Reactor water mass + pool water
gi (conc. in drywell)x(drywell gas vol)+(conc. in torus)x(torus as vol)Reactor water mass + poo water
6.0 ASSESSMENT OF CORE DAMAGE USING OTHER SIGNIFICANT PARAMETERS
6 1 Containment Hydrogen Measurement
6.1.1 Determine the % hydrogen in the primary containment by reference toI/11 and 812 H2/02 monitoring systems, or by gas chromatographicanalysis of a containment atmosphere sample obtained from the, PASS
(see IV.A.22 and Nl-PSP-13) ~
6. 1.2 Using the curve in Figure 5, determine the % metal-water reaction forthe reference Plant, % MWref. The reference Plant used here has a
Mark I/II Containment (500, bundles/350,000 ft containment ~volume )and is not the reference plant described in 2.1.
*assumes torus down comer submergence of 3 ft. (570,000 gal total). Adjust ifnecessary to account for HPCI or Containment Spray Raw Water additions.
EPP-9 -6 March 1984
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6. 1.3 Find the % metal-water reaction a t NMP-1, %MW > using th~ equationbelow:
% MW (%MW ref) (0-94) (0-89)
where:
0.94 ~ ratio of number of bundles at reference plant tobundles at NMP-1 (500/532)
0.89 = ratio of NMp-1 containment volume to reference plartcontainment volume (8.80 x 109*/9.90 x 10 )
6 ' High Range Containment Monitors
6.2.1
6.3
6.3.1
See EPP-8. This procedure provides a method for estimating thefraction of the total core inventory available for releas~ (hence>the % core damage) based on readings from 811 and (/12 High RangeDrywell Penetration Monitors in the main control room.
Reactor Water Level IndicationsI
Reactor water level indications can be used to establish if there hasbeen an interruption of adequate core cooling. Significant periods,of core uncovery, as evidenced by reactor vessel water levelreadings, would be an indication of a situation where core damage islikely. Water level measurement may be useful in distinguishingbetween bulk core damage situations caused by loss of adequatecooling to the entire core and localized core damage situationscaused by a flow blockage in some portion of the core.
6.4 Ba, Sr, La, Ru Analyses
6.4.1 Isotopically analyze a sample of reactor water in accordance witlNl-PSP-13.
6.4.2 Determine thq concentration (C„i in p Gi/g — see sectionthose less volatile elements (i.e., Ru-103, Sr-91, Sr-92 s
La-140) which are indicators of core melt from the isotopicprintout, and which can be determined from the isotopic.
5.') ofBa-14 0,
analysis
6.4.3 Calculate the normalized concentration of each isotope, i, inreference plant Ref in accordance with section 5.5 of this
CWi
theprocedure.
6.4.4 Calculate the fraction of each isotope released from the core, FR
(approximately equal to the fraction of core meltdown) using theequation.
RefF i = Cw' ( 3 ~ 92 x 10-)
Ii
*assumes torus down comer submergence of 3 ft. (570,000 gal total). AdJ'ust j.fnecessary to account for HPCI or Containment Spray Raw Water additions.
EPP-9 -7 March 1984
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6.4.4 (Cont.)
where 3.92 x 10 Total primary coolant mass (g), reference plantRefIi total core inventory of isotope i in the reference plant
(see Table 1)
7.0 REFERENCES
7 ~ 1 Lin, Chien C, Procedure for the Determination of Core Damage UnderAccident Conditions, General Electric Co., NEDO 22215, 1982
7.2 Nuclear Services Department, Post Accident Sampling SystemEvaluation, General Electric, 1983.
7.3 Counting Room Instrument Procedure No. V.A.7-N, "Operation aridCalibration of the GeLi-1 and GeLi-2 Gamma Spectroscopy System".
7.4
7.5
Process Survey Procedure Nl-PSP-13, "Sampling and Analysis of ReactorWater and Containment Air Using the PASS"..
/Chemical Analytical Procedure S-CAP-60, "Dilution of Liquid and GasSamples of High Activity"
EPP-9 -8 Harch 1984
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PLOW CHART I
SEQUEttCE OF ANALYSIS FORESTIHATIOt< OF CORE DAHAGE
O
HydrogenAna lys is(Con firm)
O
YesCon ta inmen tRadia tion(Confirm)
O
YesWaterLevel(Confirm)
O
YesNORHAL OP ERAT ION
HittOR CLAD DAHAGE
De termineOptimumSaaq1 ePoint
Core DamageEs t ima teFrom PASS
'ydrogenAnalysis(Con fi rm)
Yes
O
ContainmentRadia tion(Con firm)
Yes
O
WaterLevel(Con firm)
Analysis ForBa, Sr, La, Ru
tiAJOR CL'AD "DAMAGE
FUEL OVERIIEATFUEL HELT
YesDeterm)nationOf FissionProduct Ratios
CLAD DAHAGE
POSSIBLE FUEL OVERIIEATNO CORE HELT
-9 Harch 1984
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Table 1
CORE INVENTORY OF MAJOR FISSION PRODUCTS IN AREFERENCE PLANT OPERATED AT 3651 MMt POR THREE YEARS
Chemical Grou ~leoeo e* Half-Life
Ma)or Gamma Ray EnergyInventory** (Intensity)
106 Ci KeV (y/d)
Noble gases
Halogen s
Alkali Metals
Kr-85mKr-85Kr-87Kr-88Xe-133Xe-135
I-131I-132I.133I-134I-135
Cs-134Cs-137Cs-138
4. 48h10.72y76.3m2.84h5.25d9.llh
8. 04d2. 3h
20.8h52.6m
6.61h
2.06y30.17y32 ~ 2m
24.61.1
47.166.8
202.026.1
96.0140201221189
f
19. 612.1
178.0
151 (0. 753)514(0.0044)403(0.495)196(0.26),1530(0.109)81(0.365)
250(0.899),
364(0.812)668(0.99.773(0.762)530(0.86)847(0.954),884(0.653)
1132(0.225),1260(0.286)
605 (0. 98), 796 (0. 85)662 (0. 85)463(0.307).1436(0.76)
Tellurium Group Te-132 78.2h 138 228(0-88)
Noble Metals ,Mo-99RQ-103
66. 02h39.4d
183155
740(0.128)497(0.89)
Alkaline Earths Sr-91Sr-92aa-140
9. 5h2. 7 j}1
12.8d
115 750 {0. 23), 1024 {0.325)123 1388 (0. 9)173 537(0.254)
Rare Earths
Refractories
Y-92La-140Ce-141Ce-144
Zr-95Zr-97
3. 54}i40.2h32.5d
284.3d
64.0d16. 9}1
124184161129
161166
934 (0. 139)487 (0. 455), 1597 (0. 955)145 (0. 48)134(0.108
724(0.437),757(0.553)743(0.928)
*Only the representative isotopes Lfhich have relatively large inventory andconsidered to be easy to measure are listed here.
*~At the time of reactor shutdown.
EPP-9 -10 }farch 1984
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Table 2
FISSION PRODUCT CONCENTRATIONS IN REACTOR WATER
AND DRYVELL GAS SPACE DURING REACTOR SHUTDOWN UNDER NORMAL CONDITIONS
~Isoto e
I-131
Cs-137 c
Xe-133
Kr-85
U er Limit Nominal
29 0.7
0.3 0.03
Reactor Water, uCi/
4a10
5a4x10
5b10
6b4x10
D ell Gas (uCi/cc)Nominal
Observed experimentally, in an operating BWR-3 with,MK I containment, dataobtained from GE unpublished document, DRF 268-DEV-0009.
bAssuming lOX of the upper limit values.
Release of Cs-137 activity would strongly depend on the core inventory whichis a function of fuel burnup.
EPP-9 -ll March 1984
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Table 3
RATIOS OF ISOTOPES IN CORE INVENTORY AND FUEL GAP
~Ieoto e
Kr-87
Kr-88
Rr-85m
Xe-133
Half-Life
76.3 m
2. 84}1
4.48h
5.25d
Activity Ratio* inCore 'Invento
0.233
0. 33
0. 122
1.0+
hc tivity,Ratio* inFuel Ga
0.0234
0.0495
0.023
1.0*
I-134 „
I-132I-135I-133I-131
52.6 m
2.3 h
6. 6lh20.8 h
8. 04d
2.3
1.46
l. 97
2.091.0*
0. 155
0.127
0.364
0.685
1.0*
noble as isoto e concentration*Ratio for noble gasesXe-133 concentration
Iodine isoto e concentration for iodinesI-131 concentration
EPP-9 -12 Harch 1984
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Table 4
BEST-ESTIMATE FISSION PRODUCT RELEASE FRACTIONS
~Ca Release Meltdovn Release Oxidation Release Va orization Release
Hoble Gases(Xe,Kr)
Lover UpperHominal Limit Limit
0.030 0.010 0.12
Lo~er Upper Lover UpperRominnl Limit Limit Hominal Limit Limit
0.873 0.485 0.970 0.087 0.078 0.097
Lover UpperHominal Limit Limit
I
0.010 0.010 0.010
Nalogens(I,Br)
0.017 0.001 0.20 0.885 0.492 0.983 0.088 0.078 0.098 0.010 0.010 0.010
A3 kn 3.1 Me ta ls(Cs, Rb)
0.050 0.004 0.30 0 ~ 760 0.380 0.855 0.190 0.190 0.190
Tellurium Group(Te,Se,Sb)
0.0001 3xlO 0.04 0. 150 0.05 0.250 0.510 0.340 0.680 0.340 0.340 0.340
'Hob le Metals(Ru, Rh, Pd, Mo,Tc)
0.030 0.01 0.10 0.873 0.776 0.970 0.005 0.001 0.024
Alkaline Earths(Sr,Ba)
lx10 3x10 0.0004 0.100 0.02 0.20 0.009 0.002 0.045
Rare Earths(Y,La,Ce,Hd,Pr,Eu,Pm,Sm,Np,Pu)
Ref'ractories(7.r, Nb)
0.003 0.001 0.01
0.003 0.001 0,01
0.010 0.002 0.050
EPP-9 -13 March 1984
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FUEL MELTDOWN
10
010I-
ZC00O
ILC
C'2
10
ZOICIZ
Z0U
10
UPPER RELEASE LIMIT
BEST ESTIMATE
LOWER RELEASE LIMIT
///
///
/r / ///
CLADDING FAILURE
UPPER RELEASE LIMIT
1.0
//
/
BEST ESTIMATELOWER RELEASE LIMIT
NORMALSHUTDOWNCONCENTRATIONIN R E ACTOR W ATE R
UPPER LIMIT:NOMINAL:
29.0 uci/p0.7 IECI/p
0,10.1 1.0 '10
5 CLADDING FAILURE
E
I
1.0 10 100~ E FUEL MELTD|TLTN~Figure 1. Relationship BetMeen I-131 Concentration in the Primary Coolant
(Reactor t'ater + Pool t"ater) and the Extent of Core Damage inReference Plant
EPP-9 -14 Harch 1984
~ ~ g
. ~
~ Il ~
10
FUEL MELTDOWN
UPPER RELEASE I.IMIT
BEST ESTIMATE
10
U
10Z
0OO
C
C
10
OI-irIZ~IIUZ0Un 10O
LOWER RELEASE LIMIT /rr r
r r rr
IS
CLADDING FAILURE
UPPER RELEASE LIMIT
BEST ESTIMATE
LOWER RELEASE LIMIT
0.1 NORMALSHUTDOWNCONCENTRATIONIN REACTOR WATER
UPPER LIMIT:NOMINAL.
03 vci/pO.LI yci/II
0 2
0.1 1.0 10
% CLADDING FAILURE
1.0 'Ip I00
% FUEL, MELTDOWN
Figure 2. Relationship Between Cs-137 Concentration in the Primary Coolant(Reactor Water + Pool Water) and the Extent of Core Damage inReference Plant
EPP-9 -15 March 1984
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10
8
9 102
'U
ZW
?
Z010
?0IK'?OZ0CJ
1.0n
X
0.1
/
/
/
FUEL MELTDOWN
UPPER RELEASE LIMIT
BEST ESTIMATE
LOWER RELEASE LIMIT
//
/
///
CLADDINGFAILURE
UPPER RELEASE LIMIT
BEST ESTIMATE
LOWER RELEASE LIMIT
NORMALOPERATINGCOHCENTRATIONIN DRYWELL
10 yci/cc4
~Ci/cc5UPPER LIMIT:NOMINAL:
// /// -/'//
/'0
0.1 1.0 10
% CLADDING FAILURE
1A) 10 100
4 FUEL MELTDOWN
Figure 3. Relationship Between Xe-133 Concentration in the Containment Gas(Dryuell + Torus Gas) and the Extent of Core Damage in ReferencePlant
EPP-9 -16 March 1984
yI f
~l~ s li
~ ~
10
F UE L ME LTD OWN
10
UPPER REI.EASE LIMIT
bEST ESTIMATE
LOWER RELEASE LIMIT
1.0n
I
W
Z
IR'
o 10ZZ'
ItIZoR0o 102
hC
/ j
CLADDINGFAILURE
UPPER RELEASE LIMIT
8EST ESTIMATE
LOWER RELEASE LIMIT
10HORMAL OPERATIONCONCENTRATIONfk ORYWELL
UPPER LIMIT:HOMINAL.
4 x 10 yci/cc4 x 10 yci/cc
100.1 1.0 10
X CLADDINGFAILURE
'1.0 10 100
> FUEL MELTDOWN
Figure 4. Relationship Between Kr-85 Concentration in the Containment Gas(Dryuell + Torus Gas) and the Extent of Core Damage in ReferencePlant
EPP-9 -18 Harch 1984
c'u I tl
68
62
48
44
W~v
I
IPCC
Y
8O
38
32
28
74
20
'S
12
00 10 20 40 60 'O 70
'II METAL-WATER REACTION
Figure 5 Hydrogen Concentration for Mark I/II Containnentsas a Function of Metal-Water Reaction
a'pi<( P
Appendix A
SAMPLE CALCULATION OF FISSION PRODUCT INVENTORY CORRECTION FACTOR
Inventor of nuclide i in reference lantFIi Inventory of muclide i in operating
plant'1095
X3651 (1-e ' "i
0-), 'T. -,X' ~
P (1'-e 'e
where
P. steady reactor power, operated in. period j. (MWt)j-1
decay constant of nuclide i (day )iT duration of operating period j (day)
T. time between the end of. operating period j and time of last0j
reactor shutdown (day)
3651 ave. operation po~er (in N't) for the reference plant.
1095 continuous operation time (in day) for the reference plant.
Assuming a reactor has the following power operation history:
Operati'onPeriod Davs Since Startu
Operation TimeT~'. (day)
0Average Power
(mt)
1A
1B
1 -,6061 — 70
.60 . 254 1000
0
2A
'2B
71 - 270
271 - 300
200 44 2000
0
301 - 314 14 0 3000
EPP-9 -19 March 1984
s.( ( II ( fg
C~
i p pter> )hnocn<iix A (Conc'. )
-1~ For I-131 (X 0.0862 da . )
(1Or0862xl095)
I(I-131) 100 1'0.0862x60 -.0.0862x254
2ppp 10.0862x200
-0.0862x44 . . - -0.0862xl4 -0.0862xpe +.3000(1-e )e
365 1.
M + .45'+'.2103 „:
~ For
h rhh
F I(Cs-137)
Cs-137 (X 6;29.x':10 =da )
-5365 1 ( 1
6 29x 1 0 'x)(95)
0-6.29xlp=-- x60 --6 29xlp x254
.1 . -6. 29x 1 0-'„.,x200 '"-6. 29x 1 0 x44
ppp 1-6.29x1 px 1 4-6.29x 1 0xp
)e
243. 163.74 + 24.93 + 2.64
EPP-9 -20 Harch 1984
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4 Worksheet 1
Core Dama e Estimate Based on I-131 Cs-137, Xe-133 and Kr-85 Concentrations
NOTE: Follow Section 5.2 of procedure while completing this worksheet.
J
1) List the radionuclide concentrations (C„or C ) decayed to the time ofgre~ctor shutdown as determined from steps 5.6.1.4 and 5.6.1.5 of
h,l-PSF-13 ~ Attach Sample Analy is Dat- Sheet
Cw(I-131) =
Cw(Cs-137) =
uncorrectedCg(Xe-133)
uCi/ml
~ Ci/ml
~ Ci/cc at 14 ' psia298'K
uncorrectedg(Kr-85) / ~ci/cc at 14 ~ 7 psia,
298'K
2) 'Correct the measured gaseous activities for T, P differences between th e
sample vial and containment atmosphere per section 5.4.
uncorrectedx P (298) =
Cg iT2 1 .7
For Xe-133: x ~298)( 14. 7) u Ci/cc
For Kr-85: x (298)(14. 7) ici/c c
3) Calculate the Fission Product Inventory Correction Factor. for I-131,Cs-137, Xe-133, Kr-85. Use Worksheet lA, 1B, lc or 1D a'ppropriate.See Appendix A for an example.
FI(I-131)
FI(Cs-137) =
FI(Xe-133) =
FI(Kr-85)
4 ) Calculate the Plant Parameter Correction Factors per section 5.7. lfdowncomer submergence equals 3 feet:
Fw 2.38E9/3.92E9 0.61F = 8.80E9/4.00E10 0.22g
EPP-9 -21 Harch 1984
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Worksheet 1(Continued)
5 ) Calculate the Normalized Concentrations of the isotopes as shown below:
Re f Ref.Cwi Qi C„i Qi x FIi x Fw or Fg
For I-131:Ref
Ce(1 131) r r 0.61qCi ml
For Cs-137:Ref
C (Ce-13)) =
y C1 Ill
RefCg(Xe-133)
y01 ml
For Kr-85:Ref
g(Kr-83) y r 0'33
6) Refer to Figures 1-4 to determine the best estimate os the extent of coredamage.
Best Estimates:
I-131:
Cs-137:
Xe-133:
Kr-85:
% clad% fuel% clad% fuel
% cladfuel
% clad% fuel
failuremeltfailur emeltfailur emeltfailuremelt
Ave: % clad failure% fuel melt
7) Submit all data sheets/worksheets for review by a Technical Suppor tDepartment Supervisor.
8) Confirm and refine the core damage estimate by applying the parametersfound in sections 5.3 and 6.1-6.4 of the procedure.
Worksheet Completed by:Worksheet Reviewed by:
EPP-9 -22 Harch 1984
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Morksheet 1A
Fission Product Inventor Correction Factor for I-131 (i 0.0862 day )
FI(I 131) Inventory of I-131 in reference plantInventory of I- in operating plant
F I (I-131) 3651
j p (1 e (0 ~ 0862) Tj) (0 ~ 0862) T j )j e
Operation Period Days Since Startup Operation TimeT (day) T.
Average Powerp: (yNa)
2A.2B
3A3B
FI(I 131)
FI(I-131) 3651
Calculation Performed By:Calculation Reviewed By:
EPP-9, -23 Harch 1984
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Worksheet 18
Fission Product Inventor Correction Factor. for Cs-137() = 6.29E-5 day )
FI(Cs-137) Inventory ofInventory of
FI(Cs-137)
P~ (1-e
Cs-137 in reference lantCs-1 in operating plant
243.2
-(6.29E-5) Tg) -(6.29E-5) T'g )e
Operation Period Days Since Startup Operation Time Average PowerT. (day) T'. (Set)
3 J— —j
2A2B
3A3B
FI(cs-137) =
FI(Cs-137) ~ 243.2
243.2
Calculation Performed By:Calculation Reviewed By:
EPP-9 -24 1farch 1984
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MORKSHEET IC
Fission Product Inventor Correction Factor for Xc-133
FI (Xe 133) Inventor of Xe-133 in reference lan tInventory of Xe-133 in operating plant
FI (Xe-133) 3651
p (1 e (0 ~ 132) Tj) (Oo 132) T jOperation Period Operati.on Time
Days Since Startup Tj (day)Ave. Powe r
T j Pj (rug)
2A2B
3A3B
FI (Xe-133)
FI (Xe-133)
Calculation Performed By =
Calculation Reviewed By =
EPP-9 -25 Harch 1904
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WORKSlfEET lD
Fission Product Inventor Correction Factor for Kr-85
FI (K 85) = Inventor ofInventory of
I (Kr-85)
Kr-85 in reference lan tKr-85 in operating plant
643
p . (1 e -(1.77E-4) Tj) -(1.77E-4) T'jj j e
Operation Period Operation TimeDays Since Startup Tj (day)
Ave. Powe rT j Pj (~)
IA1B
2A2B
3A3B
FI (Kr-85) =
FI (Kr-85)
Calculation Performed By =
Calculation Reviewed By
EPP-9 -26 Harch 1984
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