Text

Core Damage Assessment Methodology
	ML20128B200
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Site:	Maine Yankee
Issue date:	05/20/1985
From:	; Maine Yankee
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ML20128B188	List: ML20128B185; ML20128B200;
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	PROC-850520, NUDOCS 8505240441
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{{#Wiki_filter:r-6 0 i s ' s 6 l MAINE YANKEE STATION CORE DAMAGE ASSESSMENT METHODOLOGY l i G S e n 9 6 I l t +,( 9 a e 8 s h 6 5 0 I r L 0 e 9 0 4 e s (o

,' 5 TABLE OF CONTENTS ER11 '(

1.0 INTRODUCTION

1 i 4 2.0 TECHNICAL BASIS,FOR CORE DAMAGE ASSESSMENT....................... 2.1 Estimated Levels of Core Damage............................ 4 2.2 Core Inventories................'........................... 6 2.3 Characteristic Fission Products............................ 10 2.3.1 Fissiou Product Release from Clad Failure......... 12 2.3.2 Fission Product Release from Fuel Overheat........ 12 2.3.3 Fission Product Release from Fuel Melt............ 13 2.4 Principle Radionuclide Used to Characterize Core Damage..................................................... 13 3.0 SAMPLING SYSTEMS................................................. 15 3.1 Containment. Air Sample System.............................. 15 ) 3.2 Reactor Coolant sample System.............................. 16 3.3 Miscellaneous Samples...................................... 17 4.0 DETERMINATION OF MASS AND VOLUME ASSOCIATED WITH SAMPLES.......... 18 g. 4.1 Containment Air Vo1uma..................................... 18 18 ( 4'. 2 ' Liquid Sample Mass...,..................................... ~ 5.0 AUKILIARY INDICATIONS.................'........................... 20 5.1 Containment Hydrogen Concentration......................... 20 5.2 . Core Exit Temperature...................................... 20 5.3 Containment Radiation Monitors............................. 24 4 5.3.1 Rasis for Predicting Radiation Monito Readings.......................................... 25 5.3.2 Interpretation of Predicted Radiation Monitor Readings.......................................... 27 5.4 Primary Inventory Trend System (PITS)...................... 27 6.0 CORE DAMAGE ASSESSMENT APPR0ACH.................................. 31 39 References....................................................... Appendix A - ORIGEN Inventory P'*4ng Operating Cycle............. 41 Appendix B - ORICEN Inventory After Shutdown..................... 46 ( . t

^ LIST OF FIGURES - Fiaure No. Eagg k '21 Power Correlation Factors Versus Effective Full Power Days for Lon5 Half Life Nuclides...................... 9 5.1 Containment Hydrogen'.-Concentration Versus Traction of Core' Zirconium Reacted................................... 22 5.2 Core Exit Thermocouple Numbers and Locations................ 23' Cantainment Accident Area Radiation Monitor Response ~ 5.3 Versus Time of Release for Different Release Percentales....- 26 29 5.4 Maine Yankee PIT System.................................... 6.1 Core Dama5e Severity Classification Usin5 Release 38 Fractions of Various Elements............................... G y.* ,.t j' 's e 4 e 9 e ( -lii-W

.;a ' ~ g LIST OF TABLES 4 Pate Table No._ ( 2.1 Progressive Material Interactions and Damage Expected 5 in Fuel Rods During Core Melt Accidents..................... 7 2.2 ORIGEN2 Input Parameters for Maine Yankee Station........... 2.3 Fission Products and' Actinides in Decreasing Order 11 of Volatility............................................... 14 7 2.4 Principle Radionuclides..................................... d 19 j 4.1 Components o,f Primary Mass Inventory Estimation............. 21 5.1 Avail'able Zirconium in Core Region.......................... 30 5.2 Primary Inventory Trend System.............................. 33 6.1 Primary Coolant Evaluation.................................. .u -rm 34 6.2 Containment Sump Evaluation................................. 35 6.3 Containment Air Evaluation.................................. 6.4 Fraction of Fission Products Released From the Core......... 36 q 37 7 f 6. 5_ Summary of Core Damage Estimates............................ = .t.- .( =E A = n 1 5 2 i .= 1= "-ll 5 _'a -iv-i 2 S_ d ] a

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1.0 INTRODUCTION

This report is prepared in response to an NRC request to provide a method to assess core damage following a severe accident at Maine Yankee. For the purposes of this report, core damage.ls to be inferred from the fraction of fission products released from the fuel. Based on the physical and' chemical properties of the various fission products, estimates of the severity. - of the release can.be made. Methods are given to characterize a release as-resulting from Clad Damage, Fuel Overheat, or Fuel Melt. Measurements of. fission products in the reactor coolant and containment stacsphere are the main sources of information for this assessment. Auxiliary indicators that will provide supporting'dat; are core exit thermocouples, primary inventory trend system, containment hydrogen concentration, and the in-containment high . ange radiation monitors. This Core Damage Assessment methodology is provided to allow early estimates of fission products released from the fuel and thus available for a . potential further release'to the environment. No effort is made.to estimate the physical condition of core structural membe'rs in this methodology. ([ Determinat' ion of.the physical condition. of the core is considered to be a [. recovery period function and will require data and analysis beyond the scope Furth'er,'the concept of " fraction failed fuel" is rejec'ted in h-of this report. this" methodology. Only after.the core has been examined can such a phrase be l-applied and even then the term " fraction failed fuel" would have to be 4 carefully defined. In' this methodology, samples shall be evaluated at the time the sample-I' is'taken. Any decay corrections required shall only be to the time of A severe accident must be viewed as a transient event taking place, f sampling. perhaps, over a period of several hours or even days. Actions taken to L mitigate the accident or in' advertent occurrences may cause transfer of fluids l. from one part of the system to another. For example, fission products confinedtothePrimaryCoolantSystemmayatsomelatertimetkansportto5.he Reactor. Building floor due to leakage. This could occur rapidly or over an extended period of time. For these reasons this Core Damage Assessment is based on evaluation of samples at the time of sampling, but not decay l ( . corrected back to some other time such as reactor shutdown. l I,

'. :i ,5 Primary dependence of.this Core Damage Assessment Methodology is on ' primary coolant and containment atmosphere samples, with supporting data The overall avail'able from auxiliary indicators discussed in Section 5.0.

approach is to first obtain the concentration (uci/gm or uCi/cm ) for each This concentration is then multiplied by the mass or volume of the sample.

, sampled' medium to obtain-an estimate of the number of curies of the various fission products that_are in that medium. This is then divided by the core inventory of:that fission product at the time of sampling to obtain, for each fission produc't, the. frac' tion of the core inventory in the sampled medium. ' For each fission product, measured, the fractions of the core inventory in all sampled media are~added together to obtain the total release fraction. ~ These release _ fractions of some specific radionuclides provide the primary If samples of different media' input to estimate the severity of core damage. were taken at different times, other auxiliary indicators and some engineering judgement should be use'd in estimating the total release fractions of the core inventories and the severity of core damage. ' Application'of this methodology is dependent upon the ability to obtain pi:l( and' analyze the required' samples and to ac.curately predict the mass or volume Post-accident sampling systems have been that the sample represents. Analytical installed and should. assure.the availability of samples. t much. . techniques for determining the nuclides contained in the sample are no different than those employed in normal operation and can be verified by Determination of the mass associated with measurements at other laboratories. In any accident scenario a basic liquid samples, however, may not be easy. question is "Where is the water"? This question must_be addressed before this core damage assessment methodology can,be applied. The sequence of events during the course of the accident will determine the answer to this question. Section 4.0 discusses the mass inventory and provides design values for the Primary System and associated tanks. The technical basis for Core Damage Assessment is given in Section 2.0 This section discusses the fission products expected to be of this report. released under different levels of over temperature conditions in the fuel. The expected core inventory of the fission product indicators is also Based on this information, the sampling program described in I. discussed. >

Section 3 0, and the auxiliary indicators discussed in Section 5.0, the methods described in Section 6.0 can be used to estimate the severity of the release and the category of the accident. ~ { 4 0 0 S e e 4 9 _ (- - t.. 6 l e o e 3- ~ e,m ,,v.---r- + - - - - - e-,.-- ,-r.,----,we,,-----m ...m,- v

'2.0 TECHNICAL BASIS FOR CORE' DAMAGE ASSESSMENT A complete evaluation for a Core Damage Assessment would reqdire ~ 3 This could include ' samples from every volume containing radioactive material. domineralizers, filters, tanks, drains, all building air as well as sumps and , primary'and secondary _ coolant systems. If this were available, the number of curies of each.nuclide in each system could be added to obtain the total '~ number'of curies of every nuclide released. One could then estimate the seveiity of.the release b'ased on the fractions of the various radionuclides escaping the fuel. This information could.then be further refined by including data from the radiation monitor, hydrogen analyzers, analysis of the core thermal-hydraulic conditions, operator actions, and visual inspections to develop a complete accident scenario. Such a task could take months and is not the purpose of this evaluation. The basis for this Core Damage Assessment is.to obtain, as quickly as possible, an evaluation of as many samples as are available from locations containing fission products, and determine the number of curies released from the fuel. The released fraction.of the inventory of each measured nuclide, k;. ( ~ faEtoring in information available from auxiliary indicators, then' allows one to'make a judgementlof the severity of the release. 2.1 Estimated Levels of Core-Damase The categories' chosen for description of core damage are four progressive levels: No Significant Damage Clad Failure, Fuel Overheat, and Fuel Melt. It is important to recognize that a severe accident may result in each of these categories existing in some part of the core. Fuel melt may occur in parts of the core, fuel overheat in others, and clad f ailure or no damage in the remainder. This methodology, describes an approach that can lead to estimates of the maximum category and, perhaps to estimate the fraction in lower categories. Table 2.1 (adapted from Reference 1) shows a correlation of expected fuel damage with temperature. I l l _4

,.c.' TCBLE 2.1 "Proeressive Material Interactions and Damage Expected in a Fuel Rods Durine Core Melt Accidents (Adapted from Reference 1) . (--. o.

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0 1. Ballooning of Zircaloy cladding ( 1300 F) 0 2. Rupture of Zircaloy cladding, (1380-1950 F)- 0 3. Oxidation of metal components and hydrogen generation <(1470-1650 F) 4. Embrittlement of fuel rod by oxidation 0 5. Reaction-between solid 002 and. solid metallic (2550-3450 F) s Zircaloy 0 6. Solid metallic Zircaloy-ZrC2 eutectic melting -( 3450 F) 0 7.. Melting of remaining metallic Zircaloy ( 3600 F) 8. Dissolution and liquefaction of U02 in the 0 Zircaloy-ZrC2 eutectic ( 3450 F) '9. -Possible breach of ZrO2 shell as a result of volume expansion accompanying liquefaction 10. Flow <fown or candling of liquified. fuel and Zircaloy j r ( 4890 F)' 0 11. Melting of remaining solid'ZrC2 0 ~ ( 5100 F) .12. Melting of remaining 002 .~ O 9 6 9 e e t 6 i

,s 'From Table 2.1 one readily~ observes that the severity of fuel damage increases as the fuel temperature increases. Preventing fuel rods from overheating is one of the most important safety functions in a nuclear power . \\ plant. Even though the core exit thermocouples provide some indication about the fuel temperature,;the direct measurement of fuel rod temperature is not ,availabl'e in -current nuclear power reactors. Therefore, one has to rely on other measurements to estimate the severity of core damage. It has~long been recognized that fuel damage always results in i releasing radionuclides from the damaged fuel rod to the reactor coolant, thus -increasing the radionuclide concentrations.in the coolant. The greater the fuel damage, the greater'the radioactivity release would be. The Rogovin Report (Reference 3), provides a rationale for selecting specific radionuclides to estimate.the severity of core damage. Based on the rationale of estimating core damage in Rogovin's report, this methodology uses the . release fractions of certain key radionuclides as a primary indicator for estimating the severity of core damage. 2.2 Core Inventories The confidence level of a Core Damage Assessment depends on the-validity of the sample data..t'e ability to establish the mass or volume h associated with the sample and the accuracy of the assumed fission product core inventories. The first step in a Core Damage Assessment program would be to determine, as accurately as possible, the fission product inventories for the affected cycle using ORIGEN-II (Reference 2). In the interim, representative inventories are-provided in Appendix" A and Appendix B. s The Maine Yankee core is modeled as three regions - one region having three irradiation cycles, one region having two irradiation cycles, and one region having only one irradiation cycle (Reference 18). However, because of the differences in specific power, each region consists of two slightly different batches. The input parameters for ORIGEN-II, used in modeling core The inventory during a typical cycle of irradiation, are shown in Table 2.2. source inventories for selected fission products and actinides during the irradiation cycle are shown in Appendix A. This appendix shows the buildup of t ~ selected nuclides as a function of irradiation time and is the sum of the ~. - _ - __ __ _ _ _. _ __ _ _ _., _ - _ _ _ _. _

T, peke 2 2 ,m ORIGE~'2 Input Fu,1 P rametcrs fer Maine Yankee ~ Specific Power (PW/MTU) and Burnup (PWd/MTU) by Cycp

Fu?1 Type Total Mass (MTU)

Enrichment Cycle 2-Cycle 3 Cycle 15 Cycle 16 CJcle 17 3 30.86 F--O O.3896 2.9% 19.04- .36.2 (10189) (12675) (12961) 4 36.33 30.97 29.66 U-4 10.636 3.5% } (15427) (12985). (12457) { 42.14 31.20 23.85 U-8 16.325 3.5% (17894) (13079) (10017) 36.33 30.9/ l V-4 10.636 3.5% i (15232) (13009) 42.14 31.20 -V-8 16.325 3.5% (17668) (13104) W-4 10.636 3.5% 36.39 (15261) 42.15 W-8 16.325 3.5% (17701) Cycle Exposure Time (Days) 535 350 424.6 419.2 420 Notes 1. Burnup in each cycle is indicated within parenthesis. 2. Rafueling time between cycles is assumed to be.42 days. 3. Time between Cycle 3 and Cycle 15 is assumed to be 17 years. 0 .A

three' regions. Appendix B reflects the radioactive decay after shutdown and shows the total core inventories expected at selected times af ter shutdown. t The inventories given in the appendices are based on full power irradiation during a, typical cycle. If the reactor has not operated at 100% , power pr or to s uh tdown, a power correction factor should be considered. This i correction factor depends on the decay constant of the nuclide. For purposes of this methodology, the correction factor discussed here is only used as a quick estimate of_the' fission product inventories using the inventories given in Appendix A. -(1) Radionuclides With Malf-Lives Less Than 1 Day (See Table 2.4) Power Correction Factor = 2630 MWt (2) Radionuclides With Half-Lives Between 1 Day and 1 Year (See Table 2.4) Power Correction Factor = 2630 MWt (3) Radionuclides With'Nalf-Lives Greater Than 1 Year (See Table 2.4) ^ For a few nuclides of interest that have half-life around one year or longer,'a power correction factor, which accounts for the approximately linear buildup of activity as a function of effective full power days, is shown in Figure 2.1. The effective full power days, or EFPD, are calculated from the beginning of the current cycle up to the time of reactor shutdown as follows: N EFpD = )(' t 2 0 g i=1 where: Power. level in NWt during time interval i Pi = (s 1

I ( o O. y o. Cs-137. Sr-90, Kr-85 so. 2 r o M O Cs-134 l-- p 0 - a.s J / l LL Z O V w.., o i.x w 5 u o P y I. E f O o j- .U cv l o-f f o o, 40.00 50 00 T.00 10 00 2'0,00x 10 ' 3'O. 00 EFFECTIVE FULL POWER DAYS l \\ ? FIGURE 2.1: POWER CORRECTION FACTORS I VERSUS EFFECTIVE FULL POWER DAYS FOR LONG HALT-LIFE FISSION PRODUCTS ( l _g.

..a t =1 Time in days for the interval i g Number of time intervals having'different power" level .[ N ' = .t Due to the production characteristics of Cs-134, its power correction ~ .f actor curve _in Figure 2.1 is differeni from other long half-life fission products, such' as Cs-137 Sr-90, and.Kr-85. -l The power cofraction. factor of a radionuclide determined from one of the above equations.is then multiplied by its inventory at the time of interest after reactor. shutdown using the inventories shown in Appendix B. Some interpolation may be required if the time of interest falls between the -decay time steps in Appendix B. 2.3 Characteristic Fission Products Both the identification and quantity of the fission products measured can provide information about the severity of the release. _Due to thei'r chemical properties and volatility, some nuclides will be released at lower temperaEures than others'. Table 2.3 (adapted from Reference 3) shows the ~ . fission product and actinide radionuclides in decreasing order of volatility. For a Clad Failure release one would expect to see the noble gases l' Based on the, l Krypton and, Xenon with small amounts of Iodine and Cesium. l experience at TMI-2,' Fuel Overheat will only add substantial quantities of ' Cesium and Iodine. Strontium and Barium may be in the range of a few percent and, there may be traces of Mo, Ru, Ag, Ce, Sb, Ir, Cd, Rb, Nb, Y, Pr, Am, Pa, Cm, and Eu (References 4, 5, 6, 7, 8, 9, 10, 14). A Fuel Melt release would release larger fractions of the nuclides I: released in a Fuel Overheat. The ability to measure the characteristic nuclides will depend on the inventory of the nuclide, its half-life, and whether it has gamma rays that will not be obscured by other nuclides. This will be discussed further in (. Section 2.4. TABLE 2.3 Fission Products and Actinides in Decreasina Order of Volatility f \\'. I. Noble Cases Ke Kr II. Halogens I ~ Br III. Alkali Metals Cs J Rb IV. Tellurim Te V. Alkaline Earths Sr Ba V'. Noble ~ Metals I Ru P,h Pd Mo Tc i VII. Rare Earths Y Np ~ Pu VIII. ReEractory Oxides l. 2r l. Nb l 9 l l l I l _II_

q .2.3.1 Fission Product Release from Clad Failure { Durins normal operation, some of the saseous and volatile fission l products would mi5 rate from the fuel pellet and reside in the tap between fuel pellet'and claddins. A. release resultin5 from clad failure alone will release .only fission' products in the' Sap between the fuel pellet and the claddins. - The ability to' identify such a release depends primarily on the absence of fission products indicative of fuel overheat. A clad failure would release primarily noble Esses with smaller amounts of cesium and iodine. the. quantity of fission Bases in the fuel-clad Sap, or In estimatin5 the so-called "Say inventory", the diffusion model of ANS 5.4 (Reference 11) has been used by several studies. However, the averste sap inventory stated in these studies varies si5nificantly. For example, the.averase released fraction of the total core to the Sap for Ze-133 varies from'8.9 x 10 to 8.7 x 10 Thus, the uncertainty in arrivins at the Sap inventory is ~ rather larse. A larse uncertainty is also shown in the small amount of data available from experimental measurements (Reference 12). Because of this, no attempt is made to analytically predict the amount of fission products in th'e ~ -( fue'l-clad' Esp.- The approach for classifyins core damase as resultin5 from . clad failures is to reco5nize a condition where there was a substantial release of noble Esses only accompanied by small cesium and iodine releases with almost no other fission products released. ~ 2.3.2 Fission product Release from Fuel Overheat I The Fuel Overheat.catesory is defined as a release resultin5 from l temperatures hi5h enou5h te cause clad failures but insufficient to melt the l f uel'. The release mechanism for fuel overheat is characterized by Ersin Brains (Reference 13). boundary release and diffusion release from the UO2 This release mechanism b* sins at approximately 1350 C (2460*F), and is driven by the vaporization of noble Bases, iodines, and cesiums previously l accumulated at the train boundaries. Approximately equal amounts of noble l-Eas, iodine, and cesiums are expected to be released from the train boundaries. For hi5h burnup fuel rods, the release fraction of these elements could be as hish'as 60 or '70% of the total fuel rod inventory. In addition, there may be a smaller amounts of strontium and barium, less than 1% of the -..

total fuel rod inventory, and traces of those nuclides listed in section 2.3. Fuel overheat will probably be accompanied by releases of hydrogen from

~

'2r-H reaction of the clad material. This will be discussed further in 2 ~ section 5.1. 2.3.3 Fission Product Release from Fuel Melt A Fuel Melt will result in large releases of the noble gases, cesiums, iodines', and larger fractions of the non-volatile fission products then would be expected from a fuel overheat. Althoush THI-II would be classified as a Fuel Overheat accident, temperatures were apparently very close to a' fuel melting condition. Fuel melt would be indicated by larger fractions released of the Alkaline Earths, Noble Metals, Rare Earths, and Refractory Oxides. Releases of Strontiums and Bariums in excess of a few percent of the inventory may be a Sood indication of fuel melting. In addition, there will probably be large quantities of hydrogen produced by the zirconium water reaction. 2.4 Principle Radionuclide Used to Characterize Core Damate The ability to detect a.~radionuclide-will depend on the shutdown f d inventory, the half life, the time of sampling, the branching ratio of 'the characteristic gamma rays, the energy of the Sammas, the efficiency of the detector, and the absence of other stron8 Emmans at nearly the same energy. ? 'hese factors were; evaluated to develop the list of nuclides in Table 2.4 T showins the nuclides indicative of the different levels of core' damage.. The choi'ce of characteristic Samma rays for the nuclides shown in Table 2.4 was based on the followins: 1. The nuclide inventory present three hours after shutdown, and 2. A fuel melt condition with equal release of all fission products. Under this condition each of the nuclides, except the cesiums, should be detectable. In a fuel melt condition with large releases of the Noble Metals, Rare Earths, and Refractory Oxides, all of the, cesium nuclides may be ( obscured. - ~ ~ - - -

g-

~

TABLE 2.4 Principle Radionuclides p*' (- Level of-

Physical *

(kev) Core Damage Nuclide Half Life-Characteristic Gamma Energy Clad Failure S SM Kr~, 4.48 h 151.18, 451.00 87 Kr 76.3 m 1175.40, 1337.96, 2011.88 f, '.88 Kr 2.84 h 1518.39, 1529.77, 2195.84 133 Xe 5.25 d 81.00, 160.60 135.Me. 9.09 h 249.79 Fuel Overheat' 131N~ 8.04 d 364.48 132 I 2.30 h 667.69, 630.22, 954.55 133 I 20.8 h 529.87 13'4 I 52.'6 m 884.09 135 I. 6.61 h-1260.41, 1131.51 i 134 Cs 2.06 Y 604.66, 795.76 136 Cs 13.1 d 340.57 137-Cs 30.1 Y 661.64 140 Ba 12.8 d 537.38, 304.82 40.2 h 1596.18, 487.02, 815.74 140 La 91 Sr 9.5 h 1024.30 92 Sr 2.71 h 1383.94 Fuel Melt 91 Y 58.5 d 555.37 93 Y 10.1 h 266.9 95 Zr 64.0 d 756.72 k/ ~~' 95 Nb 35.1 d 765.78 97 Nb 72.1 M. 657.92 l 103 Ru 39.3 d 497.08 143 Ce 33.0 h 293.26 239 Np 2.35 d 277.60, 209.75

Note that some nuclides.may not' decay as their physical half lives'because of precursor effects. The effective half life can be determined from the l'.

inventories given in Appendix 'B. i 1 t P s 4852L-SEN j i

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TABLE 2.4 Principle Radionuclides x Level of Physical

Core Damate Nuclide Half Life Characteristic Canna Enerty (kev)

Clad Failure 8'5M Kr 4.48 h 151.18, 451.00 87 Kr 76.3 m 1175.40, 1337.96, 2011.8'8 88 Kr 2.84 h 1518.39, 1529.77, 2195.84 133 Ze 5.25 d 81.00, 160.60 135 Ze 9.09 h 249.79

Fuel overheat-131 I 8.04 d 364.48 132 I 2.30 h 667.69, 630.22, 954.55-133 I 20.8 h 529.87 134 I 52.6 m 884.09 135 I 6.61 h 1260.41, 1131.51 134 Cs 2.06 Y 604.66, 795.76 136 Cs 13.1 d 340.57 137 Cs 30.1 Y 661.64 140 Ba 12.8 d 537.38, 304.82 140 La 40.2 h 1596.18, 487.02, 815.74

~ 91 Sr 9.5 Y 1024.30 92 Sr 2.71 h 1383.94 i Fuel Melt 91 Y 58.5 d 555.37 93 Y 10.1 h 266.9 .k -95 Ir 64.0-d '756.72 l 95 Nb 35.1 d 765.78 97 Nb 72.1 M 657.92 l 103 Ru 39.3 d. 497.08 143 Ce -33.0 h 293.26 i 239 Np 2.35 d 277. 60, 209.75 L

Note that some nuclides may not decay as their physical half lives because of precursor effects..The effective half life can be determined from the inventories giv'en in Appendix B.

s e

8 g. 3.0 SAMPLING SYSTEMS Post-accident sampling systems provide the most important information , /' 'infassessing the status of tha core under accident ~ conditions. The radionuclide concentrations of the containment atmosphere and reactor coolant l ith th'e mass and volume of the sampling media, allow one to ~ , samples, a ong w estimate tha" total curies released for each measured fission product. By ratioing the total releLise curies of each measured fission product to its core The chemical inventory, one can obtain the release fraction of that nuclide. properties of measured fission products and their release fractions provide the primary input to the estimat$on of the severity of core damage. 3.1 Containment Atmosphere' Sample System The containment atmosphere sampling system consists of two parallel, in-line hydrogen analyzers.each with a grab sample capability on its upstream Both in-line hydrogen analyzers have a ranSe of up to 20% hydrogen feed line. Each grab sampling device consists of a stainless steel cone.entration. sampling cylinder. Y }, 4 n Containment atmosphere samples can be taken from the. dome area. s ~ (Elevation 164'-11-1/2") and.the containment purge exhaust duct. area l-(Elevation 23'-6"). The sample intakes from the dome and the exhaust duct area come to a common line before feeding into the containment air sampling' Sampling from these p'oints should provide a representative samp'le of' system'. l* the containment atmosphere. In determining the specific activity in uCi/cm of the containment air, adjustment should be considered for decay time between sampling and ~ In addition, if the analysis, and any dilution used during the analysis. containment air sample is depressurized, as it probably will be, some of the In that case, a sample will be lost to the atmosphere before counting. 3 temperature and pressure corr'ection must be made to ensure a value (uCi/cm ) that is representative of the containment air. l t l x i

3. 2 ~ Reactor Coolant Sanelina System There area six sampling points for the reactor coolant system; the. cold leg of each of the three primary coolant loops, a,

b, the discharges of both HPSI pumps'(P-14A and P-14B), and

c. - downstream of the RHR heat exchanger The cold les sampling points provide information for the primary coolant system. However, for a cold les sample to be representative of the primary coolant system, the primary coolant must be circulating throughout the

. system. During the recirculating mode after a loss of coolant accident, liquid is recirculating from the containment sump through spray pumps, RHR heat exchangers, HPSI pumps, and into the reactor vessel to remove the decay heat. Therefore,, samples taken from the discharge of HPSI pumps provide information ^ '[ .on the containment sump. A sampling Point of post-accident RHR System is connected to the downstream of the RHR heat exchangers of the LPSI System. Taking suction from Loop 2 hot leg, the LPSI pumps provide forced recirculation and mixing of This cooling water.to the' reactor vessel through each safety injection loop. sampling Point would be utilized only if the RER System is deemed ussble in a post-accident' condition. To determine the specific activity of the sampling medium, the result of the samma spectral analysis should be adjusted for decay time between sampling and analysis, and any dilution procedures used in the analysis. By multiplying the specific activity (uci/sm) of the sampling medium for an individual nuclide by the mass of the sampling medium, one could obtain the total activity in that medium for.that nuclide. e 0 4, -e

3.0 Niscellaneous Samoles There are a number of miscellaneous locations from which samples could 1 -be drawn if radiation levels permit. Thesteamgenehatorcontinuousblowdownsystemprovidessamplepoints for obtaining grab samples from the steam generator secondary coolant. Under 1 accident conditions which involve a steam generator tube rupture, sampling the secondary coolant ~ could provide additional information in estimating primary coolant inventory and, assessing the radioactivity released. Grab samples' from several points in the primary auxiliary systems, outside containment, may be-taken. These samples may be analyzed to determine if radioactivity has been transferred from the primary coolant system or containment-sump to other systems, and to factor them into estimating the total released fraction from the core. M 69. 4 l 1 t l i I l L ( - - -...

A DETrawTWATION OF MASS AND VOLUME ASSOCIATED WITH SAMPLES 4.0

( in order to determine the number of curies in a samp el d medium, it is necessary to know the mass or volume of that medium. This section discusses ( methods for determination of these values for the containment air volume, the primary coolant mass, and the accumulated mass in the sump. 4.1 contaimment Air Volume For samples from containment it l's assumed that containment air is well mixed and that no stratification or pocketing exists. Thus, the volume associated with a containment air sample will be the free volume of the 10 3 containment building. This has been determined to be 5.27 x 10 cm (1.86 x 10 ft ) (Reference 15). 4.2 Liauid Samole Mass The total mass of the sampling medium is necessary to determine the L total' activity.as stated in Section 3.2. The mass in the Primary System, for b(~- the.purpos~e of activity estimation prior to recirculation, will be based on Consideration the total system volume'less voided portions of the system. will b* s'iven to injected mass based on information available at the time. During the recirculation mode, reactor coolant samples will include the containment floor. mass. The total mass for such samples may be estimated by sununing the initial primary coolant mass and the mass injected into the containment from the' Safety Injection Tanks (SITS), the Refueling Wate I Storage Tank (RWST), and Spray Chemical Addition Tank (SCAT), as applicable,, I to the specific accident scenario in addition to the mass contained in the recirculation piping. L The methodology described above neglects water vapor in the containment l. and assumes the boundary between the Primary and Secondary Systems is intact. The Table 4.1 provides data for calculatias the water mass. instrumentat' ion listed in the table have readouts available in the Main Control l Room. Data from this table and the accident-sp,ecific sequence of i. events will be used in estimating water mass. 6,

^ TABLE ~ d. Components of Liquid Sample Mass Estimation -(References 16,~17) I. 4 Approximate. Appro'ximate Initial ~ Initial Total Volume Water Volume Water Density In nt Range Relation Between Component (ft ) (ft ) (Ibm /ft ) (ft)' Level and Volume Primary System 10,995 10,268# 42.71' 3 Prossurizer 1,526.5 800 42.71 LT-101X 146-321" 50 ft /ft vascal R; circulation 194 Common Piping R2 circulation 528.8 Train 0 Pipe R: circulation 525.5 Train B Pipe 3 113 f t /f t/ tank SI Tank 3,500 1,497 62 LT-311 0-26.2 ft (Accumulator) (1 of 3) (11,200 gal.) LT-321 0-26.2 ft-(Based on (1 of 3) LT-331 0-26.2 ft Tank I.D.) 3 RW3T 50,210 40,104 62 LT-303K O-330,000 gal. 1257 ft /ft (300,000 gal.) LT-304K O-330,000 gal. Minimum 3 62 LT-3201 0-37.5 ft 63.03 ft /ft Spray Chemical 2,400 2,059 NaOH Addition Tank (Ref: NY (15,400 gal.) Drawing Minimum ' Neglected 11550-FV-24A) EV:por space in pressurizer subtracted from total system volume. ~ Note: Actual initial tank volumes should be used when avallable.

5.0 AUIILIARY INDICATIONS ~ As has been discussed in previous sections, the Ccre Damage Assessment Methodology is primarily dependent on the post-accident sampling program. There are, however, a number of auxiliary indicators that can provide This section addresses important information to aid in the decision process. three of these: hydrogen generation, core exit thermocouples, and the containment redir.cion monitors. 5.1 containment Hydromen concentration Incore temperatures in excess of 800'c (1500 F) can result in This reaction hydrogen production by the reaction of Zirconium and steam. will produce about 17.3 cubic feet of hydrogen per each kilogram of Zirconium Table 5.1 shows the Zirconium available in the core (Reference 18). reacted. If hydrogen is produced,'it should behave in a manner similar to the Based noble gises and thus be found in any location that the noble gases are. on the assumption that hydrogen reaches the containment volume' and is well mixed, a correlation can be mad's of the amount of hydrogen measured in the { ( containment air with.the amount of Zirconium reacted. This correlation is sh'own in Figure 5.1. Based on the measurement of hydrogen in the containment, Figure 5.1 can be used to estimate the fraction of the Zirconium reacted and thus the l l fractionofthecore[ involved.. \\ 5.2 core Exit Temperature i Core exit thermocouples can provide an early indication of abnormal core conditions that could lead to fission product release from the fuel. In i addition, the core exit temperatures can aid in the overall estimation of core damas.. I e N.

TABLFul d Available Zirconium in Core Rezion C1 doing 19569 ks Poison Rods. 691 kg Guide Tubes 2151 kg 1309 km Spacers Total' 23720 ks 9 8 ps a, e 4 4 9 i l e 9 -

/_ r-8 .NERSURED HYDROGEN CONCENTRATION ~ 5 g ' VS d PER-CENT OF ZIRCONIUM REACTED z O wsm 8 5 a zW u E 8 o u, nr S = a: 4-a v2 Ewr g $.oo sb.co sb.co A.oo eb.co sb.oo sb.oo 7b.co eb.oo sb.co abo.co .PER-CENT ZIRCONIUM RERCTED FICURE 5.1 MEASURPD HYDROCEN CONCENTRATION VERSUS FERCENT OF ZIRCONIUM REACTED

.s A na A12 AIC AE CALLED 1 837 584-5th 5:1 Eit jtt E7 36 15 MO8IN-2l 3 Cl4 Ci?. Ci6 C:t ' :".1 C:s. CE C7 C6 01 C4-l5 4 Dil Dit 087 Di6 D!S Dil Dii D9 DY 06 DS D4 D3 6 7 EIC L!t Est EIT E16 E tt El3 Ell E9 E7 E6 ES E4 E3 E2 8 9 10

20 FS F it-F l? -

F i6 - flS F13 Fis F9 F7 F4, 14 F4 F3. F2 11 12' 620 6:e sit .si? 66 6:s sia 6:i os at 66 s$ 64 43 s2 w2: 13 14 14 16 17 li sto ano sie sit sie sis s3 sin ss. s7 se as s4 s3 s2 l 18 19 f' Kti L20 Lit Let L17 LI6 Lit Ll3 Lii L9 L7-L6 LS L4' L) L2 "liiF" 20 21. ,22 23 24 25 26 27 ~T . is mis wi7 wie was w3 l ws: we w7 w6 ws w4 wa ur w2o w U W 28 ! 25 30 nao Rio nos ni7 nos mis al2 nel as n7 a6 as R4 m3 n2 ),,, x n sto ses sie s7 ss6 sit sia sie se s1 56 s o, s4 sa s2 33 34 T20 715 716 767 1:6 71L 13 7:i Tl 77 ~6 ik 74 .i2

2 35 3C.

37 vis vis vi7 vi6 vis vil vi s vs v7 v6 vs. v4 -v3 38 39 40 Wit WI7 wi6 wt$ WIS Wil W9 W7 W6 WS we f 41 42-XIT X16 Ell El3 Ell K9 E7 36 IS 43 44 . LOOP N O, I ved n2 Tio vs COLD LIG 45 FIGURE 5.2 CORE EXIT THERMOCOUPLE NUMBERS AND LOCATIONS r I ,-,~,,,--,e -w, w

Maine Yankee has 45 core exit thermocoup'les iT/cs) located at the top .of the assemblies shown in Figure 5.2. Measurements from these T/Cs are , displayed in the. Control Room. The displayed range is 0 to 2300 F with an accuracy of 1 4 F or 1 0.75Y,whichever is greater. 4 Core exit temperatures can indicate the onset and extent of core voiding. In using tho' core exit temperatures consideration should also be given to the subcooling margin' indicator. If the coolant'is in a saturated or superheated state, clad damage is a 31od possibility. Temperatures in the vicinity of saturation indicate the possibility for voiding and clad heatup I while a significant amount of superheat is a strong indication of extensive voiding. Just as the increase in core exit temperatures can indicate core voiding, a decrease in temperature can provide information to estimate the l time the corecis recovered. Caution must be used in interpreting core exit thermocouple data, the location of the T/Cs and the systein parameters of the primary coolant system must be considered. For example T/Cs near the periphery of the core may receive reflux cooling from the drainage of hot leg water and therefore not be . indicative of the clad temperatures at those' locations. The status of the i reactor coolant pumps must also be considered. Tripping the pumps allows liquid and vapor to stratify in the core making temperature determinations difficult. In general, core exit temperatures can perhaps provide the earliest i indication of potent'ial core damage Quantitative interpretation of the T/C data, however', 'is a difficult task dependent on many parameters. The use of core exit temperatures is therefore considered an auxiliary indicator to be used to verify core damage estimates made from fission product measurements. 5.3 Containment Radiation Monitors The in containment radiation monitors provide early and' direct indication of release of radionuclides from the fuel matrix through the reactor coolant into the containment. If there is a release of radioactivity into the containment following an accident, the radiation monitors can indicate the magnitude of the radioactivity release. j e

7, '

There are two independent channels of high rang

Post-accident radiation monitors inside the containment, namely, 6113A and 6113B. Thess radiation monitors are Model RD-23 from General Atomic, and have an' operating

(. y range from 1R/hr up to 10 R/hr. 5.3.1 Basis for Predictina Radiation Monitor Readinas r In using high range radiation monitors to estimate the magnitude of fission produ'ets.re' leased, one should bear in mind that, the estimated dose . rates of these radiation monitors depend on assumptions used in the estimation. These assumptions include the fraction of inventory released to the containment and the fission product behavior inside the containment. Because of the uncertainty and the incomplete understanding of the physiochemical behavior of the fission products and the thermal-hydraulic conditions inside the containment, it is difficult to estimate the radiation monitor reading under accident conditions, especially when one tries to take, into account of irapor and aerosols in the containment atmosphere. In this methodology, the analysis of correlating the radiation monitor readings'with different released fractions is based on the following. assumptions: 1. Only the noble gases are considered. This approach is based on the experience from TMI-II and the conclusion of the recent IDCOR report, that the dominant isotopes in most of the severe accident cases are' the noble gases (Page 2 of IDCOR 11.3, Reference 21). 2. Radioactive gases released from the fuel are all released into the containment, and uniformly mixed with the total free air volume. Some transient states may exist at certain times after the accident, such that this uniformly mixed assumption is not valid. However, continuous observation of the radiation monitor reading enables one to ascertain whether or not the condition is stabilized. 3. Containment integrity is not challenged after the accident. Therefore, leakage of the containment atmosphere is insignificant ( within the time period of concern (30 days after reactor shutdown).

l ( 1E+06 ~100% NOBLE GAS INVENTORIES 'N' 1E+05 -s x n. A D \\ ',. O 1E+04 1 c N 'N W N O 1% NOBLE GAS INVENTORIES h g 1E+03 s A w x s H \\ ( c w 1E+02 \\ E -{f 'n, y ' j .Q \\ 'O N O g s 1E+01 ? l 's s 0.001% NOBLE GAS INVENTORIES -- N 1E+00 ~* ......2 ....e .........e .........e. ... i l 1 E--01 1E+00 .1EA01 1E+02 '1E+03 TIME (HOURS) FIGURE 5.3: CONTAINMENT ACCIDENT AREA RADIATION MONITOR RESPONSE VERSUS TIME OF RELEASE FOR DIFFERENT RELEASE PERCENTAGES

,.~ = [ The computer code QADMOD-G (Reference 22) is used to estimate the dose. rate,of the radiation monitors using the geometry of the radiation monitors inside the containment. Figure 5.3 shows, as a function of time after reactor (' shutdown, the predicted dose rates for 100%, 1%, and 0.001% of noble gases in ~ f the core inventory released into the containment. 5.3.2 Interpietation of predicted Radiation Monitor Readings In general.. noble gas release fractions below 0.001% are indicative of no fuel damage or minor ~ clad damage. Values above 0.001% of noble gases j t k released are indicative of some fraction of clad damage, and the 1% curve indicates a large fraction of the fuel having clad damage. The values between [ 1% and 100% are in the fuel overheat and fuel melt regions. t k If there were significant amounts of airborne fissiori products other L than noble gases present in the containment air, the radiation monitor D readings would be different from those curves in Figure 5.3. Based on the b current understanding of fission product behavior during LWR accidents, the released fraction of noble gases would likely be nearly 100% when the above ( coEitionexists..For example, if there were 100% noble gases and 25% iodines [ of the core inventory uniformly mixed with the containment atmosphere, k.he radiation monitor readings at 1 hour after reactor shutdown would be 2.9 x t 10 R/hr instead of 1.6 x 10 R/hr shown in Figure 5.3. It should be p emphasized that the use of these predicted dose rate curves'for radiation o l monitors is only for a confimative and indicative purpose, and as a backup to To assess core damage, one shculd rely more on the ' direct measurements. h sampling systems of the reactor coolant and containment atmosphere. 5.4~ Primary Inventory Trend System (PITS) q ~ The PIT system consists of three pressure difference (dp) measurements w which indicate trends in reactor vessel and primary system inventories. The 1 I PITS is designed for use after the reactor coolant pumps are tripped and pressure distribution in the primary system is due to gravitation rather than h forced convection and frictional pressure drops. Then voiding in the primary ~ system and particularly in the reactor vessel can be inferred from the dp n [ measurements. B E

The three dp measurements of the PIT system are shown in Figure 5-4. Their measurement locations are given in Table 5.2 (Reference 19). The first {" dp measurement, 3001 measures pressure ~ drop from the bottom of the e, ore to the ~ top of the pressurizer. This indicates trends in the total primary system inventory. The second dp measurement, 3002 measures the pressure drop from the bottom of the core to the top of the reactor vessel. Vessel inventory is indicated by this measurement. The third and final dp measurement, 3003 measures the pressure drop from the hot les elevation to the top of the reactor vessel. Upper vessel inventory is indicated by this measurement..

The PIT System has limita,tions for its use in core damage assessment.

Ideally, vessel liquid inventory could be inferred from PIT System measurements. The PIT System has two limitations. First, the system is intended to be used only after the pumps are tripped and coolant is fairly static. This limitation is inconsequential for small breaks in which pumps. The second are-expected to be tripped long before core uncovery occurs. limitation of the PIT System is that conditions in the reference line are uncertain. Depending on location, breaks can heat the liquid in the lines connecting dp transmitters with the Primary System. Small amounts of heating .would not make much difference. However, large amounts could lead to boiling [ f within a reference line'.. This could also cause misleading dp indications. Therefore, the PIT System alone would not be used to determine core uncovery. Rather, its measurements would be used in conjunction with other indications of core voiding.and uncovery, I, I l l 9 l L

q / ' \\.. - - - - - - - - - - ~ ~ - - - - - - ~ EL.~ 60.0 ft. 4 ~ PRESSURIZER l EL. 26.214 ft._ _,_ _ _ _ _.._ IL. 2 2. 9 2 7 f t. _ _.- ) r EL. 13. 2 4 0 f t. - - - - - ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ )., .L ,......... 3* e core l _g_ DP 3003 ' ' ~ ~ " ~ ' " " EL. - 4.15 6 f t. I \\ [ I EL. - 14. 003 f t - - DP 3002 DP 3001 f-b - - - - ~. - - -----~~---- EL.- 17.167 ft [ l TIGURE 5.4 i MAINE YANKEE PIT SYSTEM .g A t i r

-TABLE 5~.2 . Primary Inventory Trend System Pros p re Tao Locations Delta'-P Z hi h Z low 5 Measurement (ft) (ft) 3001 60".0 ~ -4 3002? 22.927 -4 2 3003 22.927 11'.833 e e y e G ls l i e i ?

i ~ ;.

,s 6.0 ' CORE DAMAGE ASSESSMENT APPROACH

~ This section discusses mechanisms to estimate'the severity of core / \\ damage following an accident. Four cate5 cries have been chosen to describe the level of damase.. These categories are: 9 I. No'Significant Fuel Damage F II. Cla'd Failure .III. Fuel.0verhe'st. IV. Fuel Melt 1 Based on discussions in the previous sections of this-report and on the experience gained at TMI-II, Figure 6.1 has been constructed to provide an estimate of the cate5ory of fuel damage based on the fraction of various i fission products released. {( It should be noted that the boundaries between levels are not well ~~ kn'own and should not.be considered as definite. The final estimate of the level of7damase must be based'on an.overall evaluation of all available data and not on a sin 5 e parameter determined from Fi5ure 6.1. To use Figure 6.1 1 it is firs.t necessary to account for all fission products released from the Their location will obviously be determined by the event sequence, core. L l l however it is difficult to' envision a severe event that would not result in I fission product releases to the primary coolant and containment atmosphere. t In addition, there may be primary coolant discharged to the sump and perhaps transferred to tanks. Tables 6.1 through 6.3 are provided to aid in d the core. . estimating the fraction of the fission products that have escape Similar tables should be created for other locations if required. As discussed previously, the fraction of the fission product inventory is.to be L determined at the time of. sampling so the core inventory should be obtained from Appendix B for the time the sample was taken. Due to anomalies in the analysis of samples or inaccuracies in the inventories calculated by the ORICEN program,'different nuclides of the same element may show some . / T. differences in the fraction released. An average value should be determined ^

~. After the fraction of for.each. element based on a best en51neering judgement. the inventory contained in each location has been detemined, Table 6.4 can' be

~

used to obtain the total fraction released from the core for the various ( The total fraction as determined from Table 6.4 is used in elements' detected. Table 6.5 is provided to Figure 6.1 to determine the category of damage. sununarize the. estimates of core damaSe detemined from fission product Note release, hydrogen Beneration, and the in-containment monitor response. that it may.be difficult to distin5uish between contiguous cat *5ories and that a desi5 nation I-II, II-III, or III-IV is acceptable in Table 6.5. Because of a lack of data,, Figure 6.1 does not include the noble metals, rare earths, or the refractory oxides. Appreciable fractions of these The accident at elements should be an indication that fuel melt has occurred. TKI-II apparently resulted in temperatures that were very close to meltin5 the fuel, however, the release of the noble metals, rare earths, and refractory oxides was ne51151ble. Once. Table 6.5 has been completed, the best estimate of core damage is available from the information summarized in that table. p s '4 W l' -we.

PR!!'.ARY C00LAET NUCLIDE CONCENTRATION ACTIVITY CDRE INVEWiDR1 FRACil0N Or CD5,E AVERAEE .10C116?.1 (Cl) (Cl) AT is ..(HES) gg g$g KR'87 ~ KR KR.98 . ~. IE 133 .q. IE 135 IE. ~_ I 131 I 132 1 I 19 I 134 I!!5 CS134 CS CS 136 CS 137 BA 140 LA 140 SR 90 $s. sn 92 Y. 93 IR 95 NB 97f. - RU 103 RM 107 t CE 141 CE CE 143 EU !$6' N' 236 NP 239 ~ ACilVlit(Cl) CON *(UCl/6f.) I 10-s(C1/UCI) 1 RA!5(69 COREINVEETGEY : OfilAINEL FR0r. AFFEhDl! E, ADhS!ED !F N!!Esit,P.) FRACTION OF CORE = ACilVlif / INVEWTDP.i G TABLE 6.1 ( PRIMARY COOLANT EVALUATION 33

REACTOR 3U11.913 CONT 41 MENT FLOOR (8WRl3 RECIRCULATID ROBE) INUCLIDE, CONCENTRATION ACTIVITY CORE IN ENTORY FRACTION OF CORE AVERAGE . luc 1/OR) =(C11 (Cl) AT is. _ tMR$1 . E 855 R 87 E 88 N-IE 133

u. lu IE _

I 131 1132 '! 133 I_ 1 134 I 135 CS 134 ~

C5 136 CS _

C$ 137 BA 140 LA 140 IR.90 IR SR IR 95 .NB 97R RU103 RH 107 CE I41 bI N IO3 EU156 NP 238 NP 239 i - ( l-r= I L l AtilV111(C11 s, CONC (UCl/60 I 10-6(CI/UCl) I RA55 TEM CORElhVENitAf : OliTAINEL FRDM APPEGl! t, ADJUSTED IF NECE!SARY FRACT10N OF CDAE = ACTIVlif / INVENTDP.Y TABLE 6.2 t-REACTOR BUILDING CONTAINMENT FLOOR e P

r-

~ ~ 4 CMTAINRENT AIR NUCLIDE CONCENTRATION ACTIVITY CORE IWENTORY FRACT10N OF CORE AVERASE (UCIICH3) (Cl) (Cl) AT is (MS) KR ISR KR 87, KR N KR __ IE 133 IE 135 IE... p 4 m f a 9 ACTIVITY (Cl) = CONC (UC1/CR31 ! 10-6tt!/UCI) ! 5.27t+10tCR3) COREINVENTORY : DITAINED FROR APPENDil 3, ADJUSTED IF NECESSARY FRACT10N OF CORE = AClivlTY / INVENTORY e TABLE 6.3 CONTAINKENT AIR EVALUATION _

-y = ai j - 1 1 ' FRACTION OF FISSION PRODUCTS RELEASED FROM THE CORE i NuCLIDE-CONTAINMENT MA' N SUMP MISC. OTHERS RELEASE TO TOTAL-I AIR = COOLANT TANKS ENVIRONMENT it - KR. XE ' CS ~SA/LA ..H... Y IR ' NS RH EU - i. j i- ..._e_. .....s m.. -.......t ..M.S. ..W. .....H M..... 1 4 I t. 's 1 1 I' r TABLE 6.4 FRACTION OF FISSION PRODUCTS RELEASED FROM CORE g

li t. o.

SUMMARY

OF CORE DAMA6E ESTIMATES INDICATOR CATE60RV FISSION PRODUCT RELEASES : . NOBLE 6ASES

10 DINES

.CESIUMS-NOBLE METALS (RU,RH,MO) RARE EARTHS (Y,NP,EU,CE) ~ REFRACTORY DIIM S (ZR,NB). s HYDROGEN RELEASE MONITOR RESPONSE CATA60 RIES : ! NO DAMA6E II CLAD DAMA6E III FUEL OVERHEAT IV FUEL MELT 4 TABLE 6.5 f

SUMMARY

OF CORE DAMAGE ESTIMATES... .^

4 ~ l Kr,Xe I Cs Sr,Ba l 100% 10% FUEL MELT l gg I FUEL OVERHEAT FA 10-3% 1 l N0'SIGNIFICANI FUEL DAMAGE 10-4% 10-5g FIGURE 6.1 CORE DAMAGE SEVERITY CLASSIFICATION USING RELEASE FRACTIONS OF VARIOUS ELEMENTS

REFERENCES Schsal, R. R., D. Cubicciotti D. Hausknect, H. Morewitz, R. Ritzman, and 1., R. C. Vogel, Estkastion of Fission Product and Core Material -Characteristics, IDCOR Technical Report 11.1, 11.4, and 11.5, Electric ( Power Research Institute, Palo Alto, California, October 1983. 2. CCC-371, orIsiW-II: Isotocic Generation and Devletion Code - Matrix E-enantial Method, Oak Rid e National Laboratory Oak Ridge Tennessee, 5 1982. A Report to the 3. - Rogovin, M., G. T. Frampton, Jr., Three Mlle Island: C-_ -4ssioners and to the Public, undated.. Pelletier, C. A'. C. D' Thomas, Jr., R. L. Ritraan, and F. Tooper, 4. Iodine-131 h havior Durina the THI-2 Accident, NSAC-30. EPRI Nuclear Safety Anal'ysis Center Report, November 1981. Daniel, J. A., T.,L. McVey, E. A. Schlomer, D. G. Keffer, 5. Characterization of Cont==4nants in THI-2 Systems EPRI NP-3694, Electric Power Research Institute, Palo Alto, California, September 1984. - Rain, G. M., and G. O. Hayner, Initial ixa=4 nation of the Surf ace Laver 6. of a 9-Inch Leadscrew Section Ma=aved From Three Mile Island 2, EPRI-NP-3407, Electric Power Research Institute, Palo Alto, California, January 1984. Cox, T. E., J. T. Moran, and C. V. McIsace, Reactor Buildina Basement -7. R=dionuclides Distribution Studies GFND-INT-011, Volume II, EG6G Idaho, ""' Incorporated, Idaho Falls, Idaho, October 1982. Nitschke..R. L., Analysis Data en Samoles from the TKI-2 Reactor Coolant ~ 8. System and Reactor Coolant Bleed Tank, GEND-INF-021 EG&G Idaho, Incorporated, Idaho Falls Idaho, May 1982. e Cameron, D. S., E. D. Barefoot C. D.- Thomas, Jr., J..E.

Cline, 9.

Measurement of contamination on Conta4===nt Coolers C. D. and I and surface Contamination on a Desianation Floor Area on Elevation 305'. TMI-2 Reactor Buildina, SAI-139-81-06-RV, Science-Applications Incorporated, Rockville, Earyland, November 1981. Analysis of TMI-2 Paint Chio Samples, Science Applications Incorporated, 10. Rockville, Maryland, August 1981. Method for calculating the Fractional Release of ~ ANSI /ANS 5.4-1982, 11. Volatile Fission Products From oxide Fuel, November 1982. NUREG/CR-2507. BackEround and Derivation of ANS 5.4 Standard Fission 12. Product Release Model, January 1982. NUREG-0772. Technical Bases for Estimatins Fission Product Behavior 13. During LWR Accidents, USNRC, June 1981. Woodruff, O. J., Three Mile Islend Unit 2 Assistance, Letter to File. 14. Westin5 ouse Electric Corporation, May 1979. h Ia ..!+ 15. Modpdon, A. D., Yankee Post-LOCA Radiation Scurce Tems, YARC-1252, Revision 1 Yankee Atomic Electric Company, Framin5 am, Massachusetts, h September 1982. Kulp, D. A., Maine Yankee Regulatory Guide 1.97 Evaluation, YAEC Memo to 16. R. P. Shone, MYP 83-448, October 3, 1983. - 17. MYC-639, NY CDA, YAEC Calculation January 7, 1985. 18. Solan, G. M., Maine Yankee Fuel Parameters for ORIGEN2 Calculation. YAEC Memo to C. D. Thomas /M. N. Jow, RP-84-258, August.20, 1984. 19. Maine. Yankee Drawing MY-D-00-009. Revision 0. 20. Maine Yankee Drawing 11550-FV-10, Revision 5. 21. IDCOR Technical Report 11.3', Fission Product Transport in Degraded Core, December 1983. 22. CCC-396, QADMOD-G: Point Kemel Gamma-Ray Shielding Code, Oak Ridge National Laboratory, December 1979. S N 6 4 s e e 4 (.

7-F APPENDII A selected Fission Product and Actinide Inventories from ORIGEN During an Operating cycle of the Maine Yankee station p 3 k. 4 8 e f 4 e 6 4 8 e i 5 Q / 1 Appendix A contain's the results of th'a Reference ORIGEN-II calculation for Maine Yankee Station. The values presented represent the total core _ (. activity in curies during an irradiation cycle. Activities are given for operating times of 30 days, 100 days, and 420 days. 9 O e b o x 4 4 0 4 9 e D at 0 0 4 4 s * ~ ' " - '

-C'1N YANKEE PDWER STA110N - ORIGEN IWENT ORIES @R1 eel DURING 1RACIA71DN I,. . CHAREE-30 D - .100t '420 0 NUCLI DE ----------------------- ~33 AS 77 '2.520E-03 2.149E+05 2.109E+05 1.903E+05 34.SE B3 9. 3.895E+ % 3.796E+h6 3.345E+06 35 BR 82 4.567E-04 1.634E+05 1.905E+05 3.05BE+05 35 BRi83 0. 1.021E+07 9.904E+% 8.581E+%

35 BR 84 C.

1,842E+07 1.773E+07 1.492E+07 36 KR 83K 0. 1.022E+07 9.909E+% 8.593E+% 36 KR 85 3.890E+05 4.129E+05 4.673E+05 6.858E+05' 36 KR BSE 0. 2.273E+07 2.185E+071.824E47 -36 KR 87 0. 4.483E+07 4.284E+07 3.5%E+07 - 36 KR 88 - 0. 6.326E+07 6.M3E47 4.934E+07 37 R8 86.1.33BE+04 4.404E44 6.605E+04.1.133E+05- - 37 R8 80 0, 6.400E+07 6.119E+07 5.011E+07-37 R8 - 89 0, 8.283E+07 7.907E+07 6.431E+07 38 SR 89 3.042E+07 4.842E+07 6.856E+0. 6.844E 47 7 38 SR~ 90 3.051E+M 3.241E+06 3.669E+% 5.400E+06 38 SR 91-C. 1.040E+0E 9.980E407 8.307E 47 - - 1.094E+08 1.056E+08 8.982E+07 - 38 SR 92 0,. 6.033E+07 5.791E47 4.821E*07 39 Y 91E 0, - 39 Y 91 4.217E+07 6.039E+07.8.372E+07 B.816E+07 39 Y 92 0.- 1.098E 48 1. 4 0E+0! 9.021E+07 , 39 Y 93 0. 1.224E408 1.190E+0E 1.040E+02 .39 Y 94 0. 1.204E+0B 1.176E40B 1.046E+0B 40IR 95 '5.658E 47 7.654E+07 1.035E40E 1.184E+05 [/ 40 IR. 97 9,514E-11 1.234E+06').223E+0B 1.148E 48 . A.'41 N8 95 7.828E+07 7.37,0E407 E.943E+07 1.199E+08 41_N8.95R 4.200E+05 5.196E+05 7.550E45 8.396E45 41 NB 96 1.413E-0B.1.102E+05 1.334E+05 1.910E+05-

41 NB 97 ' 1.025E-101.242E4E 1.232E+0E 1.15BE45 41 N8 975 9.012E-Il 1.170E+0E 1.159E+0E 1.089E4 8' 41 WP 98M 0.

9.275E+05 9.917E+05 1.130E+06 42 RE 49 2.330E+03 1.305E+08 1.300E+% 1.247E+05, 4} R0 101 0. 1.171E+0B 1.176E+0B 1.158E 4B < 43 1C 99M 2.245E+0! 1.14!E+08 1.13BE+08 1.092E+08 43 TC 101 0. 1.172E+0B 1.176E+0E 1.15BE45 43 TC 104, 6. 6.982E+07 7.403E+07 B.600E+07 '44 RU 103 '3.519E+07 5.867E+07 8.496E+07 1.051E+08 44 RU 105 0. 5.224E+07 5.676E+07 7.024E+07 45 RH 103M 3.172E+07 5.267E47 7.655E+07 9.463E407 45 RH 105 1.328E-01 4.881E+07 5.294E+07 6.511E+07 45 RH 105R 0, 1.463E 47 1.589E+07 1.967E+07 4:i RH 106 1.746E47 2.080E+07 2.357E47 3.730E47 - 45 RH I MM 0. 1.150E+% 1.277E+% 1.698E+% . 43 RW 107 C. 2.535E+07 2.873E+07 3.983E+07 46 PD 109 0. 1.210E+07 1.389E+07 2.044E+07 46 PD 111 C. 3.101E+% 3.419E+06 4.528E+% 46 PD !!!R 0. 4.076E+04 4.655E44 6.667E 44 46 PD 112 1.276E-091.813E+% 1.943E+% 2.406E*06 / / ~43-

r .? C1N YANKEE PDWER STATION ORIBEN ! WENT ORIES (CURIES) DURIN8 IRRADIATIDE - . CHAR 8E. 30 D

100 D 420 D

f. NucunE---------------------------------------

, 47 O 1995 1.996E-C2 1.209E+07 1.388E+07 2.043E407 47 0110n B.875E+04 9.602E+041.148E45 2.373E+05 47 0 111 5.941E+04 2.891E+ M 3.404E 46 4.561E+06 47 O 1115 0. 3.089E46 3.406E+# 4.517E+06 47 0 112 1.511E-09 1.818E+ % 1.949E+ M 2.412E+06 O CD 115,1.817E+001.072E+M 1.112E+M 1.246E46-48 CD 1155 4.2 ME+0.4 6.561E+04 9.334E+04 1.202E+05

iG CD 117 0, 6.377E+05 6.554E+05 7.074E+05

.O CD 1175 0.- 3.456E+05 3.556E+05 3.844E+05 49 IN'!!55 4.973E+00 1.075E+ M 1.115E+06 1.248E 4 6 49 1N'117 0.' 5.884E+05 6.051E+05 6.535E+05 49 IN 1175 0.- 7.451E+05 7.660E+05 8.271E+05 49 IN 118 0. 9.830E+051.000E461.081E46 50 SN 123 1.367E405 1.523E+05 1.821E+05 2.490E+05 -50 SN 123n 0. 9.598E405 9.845E+05 1.051E+06 ' 50 SN 125 3.152E+04 6.937E+05 8.172E45 9.147E45

50 SN 127 0.

4.00!E+06 4.178E4 6 4.628E+06 '50 SN 126 0. 1.058E+07 !.081E+07 !.!!7E+07 ~ _51' SE 122 1.328E46 5.734E+04 6.779E+041.171E+05 51 58 124 2.345E44 2.828E44 3.771E+04 7.445E44 i 51 SB 125 5.206E+05 5.452E+05 6.178E+05 9.55BE+05 I 51 $8'126 4.437E+03 4.256E+04 5.512E+04 7.119E+04 .$1SB127 2.704E+03 6.160E 46 6.55BE+0e 7.310!+04 51 S8 128 - 0. ._.7.666E+05 8.340E+05 9.899E+05 y (V 31 5B 1285 0.' 1.129E+07 1.15?E+07 1.212E 47 * [. 51 58 129. 0; ,2.005E 47 2.06BE+07 2.190E+07 ,$1'Sli 130 C. 6.333E+06 6.59EE+06 7.126E+06 51 SS 131 0. 5.852E47 5.856E+07 5.703E41 52 TE 127 5.083E+05 5.859E+06 6.300E46 7.230E+06 51 TE 127n 5.163E 45 5.493E+05 6.712E+05 9.530E+05 s 52 TE 129~ 6.371E+05 1.903E+07 2.017E+07 2.157E+07 .G2 TE 1295 9.787E+05 1.070E+06 2.752E+06 3.244E+06 i $1 TE 131 1.242E-04 6.083E+07 6.121E+07 6.052E+07 52 TE 1315.5.516!-04 9.375E+06 9.590E 4 6 9.860E+0L -52 TE 132 9.386E+03 9.805E+07 9.898E+07 9.697E+07 52 TE 133 0. 8.561E+07 8.513E+07 8.162E+07 .G2 TE 1335 0.- _5.748E+07 5.647E+07 5.133E+07 02 TE 134 0. 1.315E+08 1.284E+0B 1.157E+0B 53 1 129 1.159E+00 1.224E+00 1.38BE+00 2.194E+00 53 I 130 'O. 1.105E+06 1.289E+06 2.208E+06 53 ! 131 1.377E+% 6.257E+07 6.852E+07 6.832E+07 53 1 132 9.670E+03 9.923E+071.002E+08 9.849E+07 53 1 133 2.731E-07 1.463E+08 1.458E+0B 1.395E+08 $3.1134 0. 1.624E+0B 1.612E+08 1.531E40E-53 1 135 0. 1.364E+08 1.35BE+0B 1.303E+08 i I / r' L

~ RAIN.YANrEE POWER STAT!DN DR18Ek lh7ENT DE!ES (CURIES) DURINE IRRADIATION _h CHAF$E. 30 D 100 D 420D NUCLlDE - - - - - - - - - - - - - - - - - - - - - - - - - - - - - '54 IE 131M 1.194E+05 4.910E+05 7.585E+05 7.601E+05 54 IE 133 4.936E+051.3ME+0B 1.414E+061.396E+08 54 IE 133R 8.760E+00 4.185E+ % 4.461E+0h 4.353E+ % 54 IE 115 0. 3.624E+07 3.582E+07 3.309E+07 54 IE 135M 0. 2.695E+07 2.728E+07 2.731E+07 54 IE 138 0. 1.268E+08 1.248E+0B 1.153E+08 55 CS 134 4.261E+M 4.532E+% 5.246E+M 9.769E+06 55 CS 134R 0. 1.565Ei % 1.B30E+06 3.00BE+06 ~ 55 CS 136 1.803E+05 1.289E+% 1.809E+06 2.932E+06 55 CS 137 3.991E+% 4.23BE+% 4.016E+06 7.426E+06 55 CS 138 0. 1.387E+0E 1.370E+06 1.277E+0B' 56 M 135R 4.965E-06 2.039E+03 2.512E+03 5.901E+03 -56 E 139 C. 1.345E+0B !.331E+0B 1.247E+0B 56 M 140 9.279E+06 1.059E+0B 1.282E+0B 1.201E+0B 56 M 141 0. 1.224E+0E 1.212E408 1.133E+08 57 LA 140 1.06EE+071.107E+0E 1.324E+0B 1.234E+0B 57 LA 141 0. 1.229E+0B 1.217E+0B 1.13BE+0B 57 LA 142 0. 1.205E+0E 1.159E+02,1,102E+ % 56 CE 141 3.562E+07 7.651E+071.120!+0i 1.160E+0i 58 CE 143 5.141E-02 1.!BIE400 1.160!+0B 1.060E+0B 50 CE 144 5.240E+07 5.626Eic7 6.405E+07 E.395E+07 58 CE 146 0. 6.214E+07 6.140E+07 5.731E407 b. 59 Pf.142 3.639E-10 2.413E+%.2.766E+06 4.74EE+% 19 PR 143 1 134E+07 E.93tt+07.1.121E+ M 1.05;E40E a '( 59 PR 144 5.240E+07 5.675E+07 6.468E+07 6.45BE+07 ~ 59 PR 144R 6.20BE+05 6.761E+05 7.696E405 1.00BE+ % 59 PR 145 0. (.957E+077.833E+077.206E+07' 59 PR 146 0. s.2'?E407 6.162E+07 5.754E+07 60 ND 147 2.42EE+06 4.152E+07 4.614E+07 4.542E+07 ~ 60 ND 149 0. 2.'.tt2E+07 2.599E+07 2.599! $7 ~61 PM 147 6.953E+0e 6.', BEE +06 7.901E+061.00BE+07 - 61 Pn 145 9.475E+041.06ti+071.253E4071.759E+07 61 PR 14BM 7.179E+051.334E+M 1.574E+06 2.120E+06 61 Pn 149 4.259E+0! 2.750E+07 2.82EE+07 2.963E407 61 PM 151. 1.937E-04 1.200E407 1.237E+07 1.332E+07 61 Pr. 154 0. 2.692E+06 2.B67E+% 3.4%E+06 62 SR 151 1.585E+04 1.760E+04 2.042E+04 2.496E+04 62 SM 153 5.19BE+00 1.610E+07 1.602E+07 2.66BE+07 63 EU 154 2.826E+05 3.026E+05 3.540E+05 6.721E+05 -63 EU !$5 1.860E+05 2.010E+05 2.362E405 4.248E+05 63 EU 156 7.653E+05 3.BE3E406 5.679E406 1.!!0E+07 63 EU 156 7.653E+05 3.883E+06 5.679E+061.110E+07 64 6D 159 1.163E-11 2.311E+05 2.630E405 3.907E+05 92 U 239 0. 1.241E+09 1.261E409 1.346E+09 ~ 93 NP 238 1.105E+01 B.539E+06 1.00Bi+07 1.BB7E+07 93 NP 239 4.345E+03 1.238E+09 1.260E+09 1.346E+09 93 NP 240,0, 1.085E+% 1.123E+061.287E+06 94 PU 241 '4.142E+% 4.34BE+06 '4.995E+06 8.253E+06 94 PU 243 6.7.97E-07 5.89BE+% 7.209E+061.625E+07' 93 AM 244 0. 6.937E404 6.966E+04 2.!24E+05 96 CR 242 7.86EE+05 9.465E+05 1.302E+06 2.793E+06 DUTPUT UNIT

8 PAGE 126 I

4* I APPElrDIX B Selected Fission Product and Actinide Inventories from ORIGEN Following Shutdo b from a Full Power Operation Cycle of the Maine Yankee Station e .( f d 4 e a 0 e e h e p 9 6 c

,t-t s Appendix B contains the results of the reference ORIGEN-II calculation. for-klaine Yankee Station. The values presented represent the total core activity in curies fo110 win 5 shutdown from a full power, 428 day operatint [ cycle. Activities are liven for two time periods, 0.5 hour - 24 hours and 2 days - 180 days. 9 e e 4 9 I,, o' .y 9 D 0 E 6 4 0 0 6 9 .' ; [* CINE YAWKEE POWER STA110N ORIGEN INVENTOR 1E$ (CUR!E51 AFTER SHUTDOWN FROM A TYPICAL FULL POWER OPERAT1NS CYCLE SHU720WN 0.5HR .1.0HR 1.5HR 2.0HR 3.0HR 5.0HR 6.0HR 10.0HR 12.0HR 12.0 4 24.0% p ..ULIDE '.31 O 77 1.903E451.893E451.BB2E+051.871E451.h0E+051.537E451.792E+051.72 34 SE $3 3.345E+H 1.3ME+06 5.301E45 2.104E+05 B.349E+041.315E44 3.261E421. B2 3.050D05 3.032E+05 3.002E45 2.973E+05.2.944E+05 2.887E+05 2.775E+05 2.61 35 BR 35 BR 83 ' E.500006 7.757E46 6.826E+H 5.953E46 5.166E+06 3.174D06 2.171E+06 9 1.492D07 6.633D06 4.491E+M 2.335DM !.214E+06 3.294E45 2.401D04 4.748E+02 3.471E412.536!40 35 BR 84s 36 O 83K 5.59X46 8.523E+M 8.307E+N 7.972E+H 7.552E+M 6.574E+M 4.500E+M 2.30 36 KR 85' 6.tS8E+05 6.958E+05 6.059E+05 6.85BE45 6.85tE+05 6.859E+05 6.tS9E+05

36 KR 85M 1.824E471.709E+071.502E471.464E+071.355E+071.161E+07 8.51tE+% 5

% KR 67 3.5ME+07 2.69tD07 2.055E+071.564E+071.191E+07 6.907E46 2.322E+H 4.52 36 KR '98 A.934E47 4.372i+07 3.B&9E+07 3.425E+07 3.031E+07 2.374D071.457E 1.133E+05 1.132E+05 1.131E+05 1.130E+05 1.129E45 1.12BE45 1.124E+05 1. 37 RE h 37 0 98 5.011E+07 4.725E+07 4:272E47 3.009D07 3.30'0E47 2.651E+071.627D07 7.8 6.431D07 2.043D07 5.207E461.326E+06 3.375E45 2.198D04 9.194E+0! 2.504E-021 37 0 99 6.H4D07 6.844E47 6.842E47 6.840E47 6.B3BE+07 6.834D07 6.826E47 6.815E+ 38 SR 89 38 SR 90 5.400E46 5.400D06 5.400E46 5.400E46 5.400E+% 5.400E46 5.400E+M 5.400 8.306E47 8.023147 7.735D07 7.458907 7.191E47 6.685E+07 5.777E+07 4.642E 38 SR 91 '30 SR 92 6.982E47 7.906!47 6.957E47 6.122907 5.387E47 4.171E47 2.501E471.161 39 Y 91K 4.921E47 4.791!47 4.713D07 4.606!47 4.4!2!47 4.210E47 3.664D07 2.949E4 6.t!6E47'6.t!6E47 E.B16E47 8.E:5D07 8.815!+07 f.814!47 8.t!2D07 f.807E4 39 Y 91 92 9.021D07 8.966E47 6.821D07 8.607E47 f.339E47 7.699907 6.244E47 4.217E47 39 Y 93 1.040!481.0!6D06 9.827E47 9.496!47 9.175!47 6.567E47 7.468E47 6.078E47 5.299E47 4.619E47, 39.Y 1.046f48 3.752E471.263E47 4.252E461.431E461.622E+0! 2.084E43 3.033D00 39 9 '94 4Q IR 95 1.164E4B 1.163D061.1E3E+0B 1.1E!D081.163D0E 1.!B2E421.181E4 , - so jg 97 1.14tE4,1.125D061.102DM 1.079D061.057E461.015 DOE 9.351E+07 8.26 Bid 7 7.617D07 7 6 ,, N 4 O ' 95 1.199i481.199DM 1.199D061.199D0E l.199 DOE 1.199i+061.199E9E'!.1 410 95R 8.390E45 8.397E+05 8.397E45 B.396E45 0.396E405 8.395E45 8.393E45 41 NE 96,1.910D051.682E4$ 1.054E451.227E451.600E+051.748E+051.647Dc51.507D051.420E451.3 41 NB 97 1.150E461.151E4B 1.1401481.126E4t*1.!!0E+% 1.075D06 9.900E47 8.t$ 97M 1.089E4E 1.065E4B 1.044D021.022D061.002D06 9.614E47 f.957i+07 7.83 41 Ni 41 NB tif.1.130D06 7.546D05 5.039E45 3.365E45 2.247E451.007E451.9939041.767E43 3 !!5902 6.990 1.247D061.240D001.234D061.227E+061.221D061.200Edi 1.183D021.146!+0! 1.123E4E,1.099 42 M3 99 42 r.0101' !.156E42 2.820E47 6.601E+061.640E46 3.955E45 2.300E44 7.779141 43 7C 99?.1.092E461.0919001.091D061.090E4B 1.089E+081.%6E461.07BD06 1.156D06 6.722!4'7 2.530E47 B.201D06 2:464E+M 1.963D05 9.805!42 2.iB6 43 TC 101 'I.600E+07 2.992007 9.545D06 3.0450% 9.714E45 9.885E441.024D031.079E 43TC104 44 RU 103 1.05tt+081.050D081.050!481.050E481.049D021.04tidt 1.047E461.0 44 RU 105 7.024D07 6.69ED07 6.210E47 5.744E+07 5.313!47 4.545E47 3.326E47 45 RH 103f. 9.463E+07 9.462E+07 9.461E47 9.459E+07 9.456D07 9.450E407 9,437D07 9.417D07 43 RH 105 6.511E47 6.515E+07 6.515E47 6.509D07 6.500E47 6.469E47 6.371E47 6.158D07 !.99 45 RH 10$M 1.967D071.881D071.744D071.613E+071.492E471.276E47 9.!39 45 RH 1% 3.730E+07 3.380D07 3.380E47 3.36CC7 3.360E47 3.379E+07 3.37 43 RH 1%M 1.698D061.45tE+M 1.239E461.059D06 9.044D05 6.600E45 3.51 45 RH 107 3.983E+071.919D07 7.388D06 2.834E461.087E461.599E45 3.46 2.044D07 2.005D071.954D071.904D071.956E471.763E471.590D071.363 4.528E+M 1.800D% 7.589D05 3.185E451.459E45 4.964D04 2.626E441 46 FD 109 46 PD ll!M 6.667E44 6.266D04 5.963D04 5.524D04 5.167i44 4.573E44 3.554 46 PD 111 46 PD 112 2.4HE46 2.365Du 2.324E46 2.285D06 2.245E+06 2.169D06 2.025 I t I~ + 4 ^- -m

v 3 i :,- - I RAINE 94m:EE P0NER STATION l DRIEW INVENT 0 RIES (CURIES) AFTER $NUTDDNN FROK A TYPICAL FULL PONER DPERATINE CYCLE [

$WTDONN 0.5HR 11.0H6 1.5HR 2.0NR 3.0HR

!.0HR I 0HR 10.0HR 12.0HR lt.M

24. M NutL I E --.-.--~ ~. -.~ - -.~....-.~.--...--. -

, 47 0 199R 2.04M47 L 60$E47 1. 955E+07 1.905E+07 1.157E47 1.763E+07 1.591E47 1.363E47 1.230E+07 4743 !!0R 2.37M+05 2.37M+05 2.373E45 2.373E+05 2.372E45 2.372E45 2.37M45 2. 47 0 !!! 4.561E+M 4.559E+M 4.552E+4 4.545E+M 4.536E+M 4.519E+M 4.485E+06 4.43 47 O l'!!R 4.514E+M 1.999E+M B.094E45 3.48E451.67tE+05 6.5HE+M 3.752E 47 0 !!2 2.412E+M 2.409E+M 2.40X+06 2.392E46 2.379E+M 2.344E+M 2.254E+M 2.tt0E+M i O CD 115 1.24E+M !.242E+061.235E+061.227E+M 1.220E+M 1.204E+M 1.17X+# !.129 O C8 !!5R 1.202E+05').202E451.201E+051.201E451.200E451.200E451.190E451.19E4 O CI !!7 7.074E45 6.225E+05 5.448E+05 4.768E+05 4.173E45 3.19M451.076E+05 0.42 _O C8 ll?R 3.944E+05 3.479E45 3.142E45 2.837E45 2.562E45 2.090E+051.390E4 ~ 49 IN !!5M 1.240EW !.240E461.247E+41.24E+M !.244E461.239E+M 1.22 SEW l.19E461.17 l-491# !!7 6.535E45 6.477E45 6.31E45 6 ellE45 5.795E+05 5.13E+05 3.UE45 2.163E+051 O IN !!?R t.271E+05 8.19145 7.975E+05 7.659E+05 7.27tE+05 6.411E45 4.654E45 2.61 49 IN !!8.1.0t!E+06 7.172E45 4.74M45 3.137E+05 2.075E45 9.07E441.73M441.454E 03 50 IN 123 2.490E+05 2.490E+05 2.409E+05 2.499E45 2.409E45 2.400E+05 2.487E+0 l 50 IN 12361.051E46 6.309E45 3.755E45 2.235E+051.331E+05 4.714E+04 5.910E+03 50 IN 125 9.147E+05 9.334E45 9.120E45 9.lME45 9.MM+05 9.0ME+05 9.0!!E45 0.93 50 IN 127 4.62E+M 3.925E+M 3.320E+M 2.821E+M 2.392E41.720E+M t.II7E45 3. l' 50 IN 120 ' !.!!7E+07 7.tSM46 5.52E+M 3.lt0E+06 2.720E46,1.34tE46 3.292E+05 3.972E44 9.700E43 2.3 $1 $t 122 1.171E451.165E+M 1.159E+051.!!H451.14E451.134E+051.!!0E451.075E l 5190124 7.445E+04 7.444E+04 7.442E44 7.440E+M 7.430E44 7.435E+M 7.42E44 E45 51 II 125 9.55tE+05 9.559E+05 9.559E45 9.359E+05 9.559E45 9.559E+05 9.559E+ \\ 51 It 126 7 !!9E+ M 7.!!!E+04 7.10 N 44 7.094E+04 7.086E 44 7.070E 44 7.037E+M 51 $t 127 7.310E46 7.300EW 7.267E+M 7.27tE46 7.253E+M 7.214E46 7.12E+M .(- 51 It 129R 1.212E+07 9.33E46 4.675E+M 4.707E 5190129 2.190E47 2.049E+071.091E471.746E471.611E471.372E+07 9.954 5190130 7.12E+M 4.237E+M 2.519E+M 1.49E+M 0.900E45 3.149E45 3.937E 5.703E47 2.350E47 9.517E46 3.tSH461.56M+h 2.559E45 6.575E42 3.03 l 521E 127 7.230E+M 7.23E+M 7.229E46 7.22E46 7.226E+M 7.222E+06 7.2M 51 SI 131 52 TE 127R 9.530E+05 9.538E45 9.53tE45 9.539E405 9.!39E45 9.539E45 i' 52 TE !?9 2.157E47 2.131E47 2.07E+07 2.002E47.1.91M471.726E471.359E [, 52 7E 12th 3.244E+M 3.243E+M 3.24M+M 3.243E46 3.242E46 3.241E46 3 [, 6.052E47 4.647E+07 2.910E+071.HM+07 9.832E+06 3.96ti+M 2.080E < 52 TE 131R 9.te0E+M 9.UM+M 9.677E46 9.57E406 9.462E+M 9.247 52 TE 131 52 TE 132. 9.H7E47 9.65tE+07 9.61E+07 9.57X+07 9.531E+07 9.447 52 TE 133 8.162E+07 2.140E+07 6.981E+M 3.342E+M 2.023E+M 9.M 52 TE 133R 5.133E+07 3.530E47 2.42SE+071 53 1 129 2.194E 4 0 2.194E+00 2.194E 4 0 2.194E+00 2.194E+00 52 TE 134 53 1130 2.200E46 2.155E+M 2.097E46 2.039E+M !.98X461.074 6.83M47 6.831E47 6.82M+07 6.820E47 6.812E+07 6.792E47 53 1132 9.M9E47 9.825E47 9.799E47 9.770E+07 9.740E47 9.67 53 1 131 53 1133 1.395E+081.38E461.349Edt 1.35M 53 1 135 ' l.303E+091.237t+081.174E+081.!!4E491.05M48 9.51E47 7.719547 5 I q .y 4+* s =t CINE VANKit P0uEA 87ATIM OR10EN INVENTORIES (CUR 1 Ell AF7tn INUTDONE FROR A 7Yf1 CAL FULL POWER OPERA 7!NE CYCLE INU780NN 4.5MP,

1. M 1.5NR

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ML20128B200

Contents

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1.0 INTRODUCTION

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ML20128B200

Text

1.0 INTRODUCTION

1.0 INTRODUCTION

SUMMARY

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