rev 0 to emergency procedure epp-9, 'determination of core ... · nine mile point nuclear...

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


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

EPP-9


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


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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s

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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1I

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

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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,A

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

I ~ 4 s I

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


. ~

~ 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


~~

I<

a I) ~

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


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


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


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


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I

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Y

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


~ I

<i y, ~

y

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:


0 ~ ~ ~ i

%a

a

))( yg

)a

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

(~ I ~ g

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

('pa~~ )

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 =


4~' ~, g ia

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


tyr.

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1 ~

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