The release of fission products from degraded UO2 fuel: Thermochemical aspects

Journal of Nuclear Materials ZOl(1993) ZOl(1993) 57-69 North-Holland

The release of fission products from degraded UO, fuel: Thermochemical aspects E.H.P. Cordfunke Netherlands

and R.J.M. Konings

Energy Research Foundation ECN, P.U. Box 1, 1755 ZG Petten, The Netherlands

This review describes the chemical aspects of the release of the important fission products. Their behaviour in the fuel, in the reactor coolant system (primary system), and during core-concrete interactions, is discussed. Special attention is given to thermochemical modeling to incorporate the results of recent studies on the acquisition of thermochemical data for relevant species by experimental and evaluation techniques.

1. Introduction

2. Historical perspective

Since the accident at the Three Mile Island (TM11 nuclear reactor in 1979, much attention has been given to the chemical behaviour of the fission products evolved from degrading fuel. In severe reactor accidents fission products are released from the fuel and are transported - in light water reactors (LWRsl - by steam/hydrogen mixtures flowing through the core. Hydrogen is produced by the reaction of steam with zirconium of the Zircaloy cladding. The release and transport behaviour of the fission products is affected by their thermochemical properties and the attenuating factors in the primary system. Thermochemical data are used in mechanistic computer codes, and it is evident that assessment of source terms in severe accidents require an adequate data base. Only then the fission product release rates from degraded fuel, the chemical forms of the volatile species and their interaction with aerosols and structural materials in the primary system can be defined. During the past ten years much progress has been made in understanding the physical and chemical behaviour of the fission product elements during accident situations. This paper emphazises how powerful thermochemistry is as a tool in predicting the stable chemical forms of the fission products and, consequently, their behaviour during release from degrading fuel, through the primary reactor coolant circuit, and into the ~ntainment building during severe reactor accidents.

Over the past 35 years a continually increasing understanding of nuclear accident phenomena has been obtained. Small-scale laborato~ release experiments formed the basis for the Reactor Safety Study, published in 1975 [l]. In this famous report accident sequences were defined and predictions of the release of fission products to the environment were made. Consequence calculations assumed that the released fission products species were transported unchanged to the containment, irrespective of the phenomena that might occur. However, after the accident at the Three Mile Island reactor in 1979, the validity of the methods for estimating the release of the fission products, particularly iodine, was questioned in the USA by scientists from Oak Ridge National Laborato~ and Los Alamos Scientific Laboratory. They suggested that in LWRs the chemical form of iodine that escapes from the fuel matrix is cesium iodide, which is not only much less volatile than the then accepted form I,, but also very soluble in water. As a result, a report was issued by the US Nuclear Regulatory Commission (NRC) in 1981 emphasizing iodine accident behaviour, as well as the fact that only a relatively few fission product species are responsible for the consequences of severe accidents [2]. Release rates for the important fission products as a function of temperature were given, and the crucial role played by the primary circuit in attenuating the release and modi~ng the chemical and physica forms of the fission products were emphasized. The

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0 1993 - Elsevier Science Publishers B.V. All rights reserved

report also concluded that “there is no way of predicting whether or not cesium at this stage has an opportunity to combine with iodine”, and that the major form of cesium in the fuel is cesium uranatc, Ca,UO,. Uncertainties with respect to the chemical form of iodine in the fuel, and to the release-from-fuel experiments have been the stimulus for many further stud& in the 1980s. The purpose of these studies was to obtain the experimental and theoretical data needed to reliably assess the consequences of nuclear reactor accidents. Since the release and transport behaviour of the fission products depend on their chemical forms, various attempts were made to define the most important chemical species, and the phenomena that occur during their transport to the environment. Postirradiation studies, notably by Kleykamp [3], have given valuable information on the chemical species present in the fuel during normal operation, whereas large-scale release experiments at Oak Ridge National Laboratory demonstrated significant differences from results of the smaller-scale work done previously [4,.5]. In addition, thermochemical databases have been developed in the late 1980s that permit the calculation of the chemical phenomena occuring during a severe reactor accident. Thus, a thermodynamic database has been made for the 167 chemical species in the VICTORIA code (primary circuit modelling code) [6], and for most of the chemical species that are formed during molten-corc/concretc interactions (VANESA code). On the initiative and with financial support of the Commission of the European Communities (CEC) a library of carefully evaluated thermodynamic data of fission products and reactor materials has been made by specialists from Harwell Laboratory (UK) and Netherlands Energy Research Foundation ECN at Pctten which was published in 1990 [7]. In 1087 an important initiative was taken which indicated the growing awareness of the major chemistry issues in severe accidents. The US National Research Council organized, on the initiative of the US NRC, a workshop at Captiva Island (December 1987) whcrc the main uncertaintics in the analysis of reactor accidents were identified, and experimental stud& were specified for study, notably: _ gaseous species : hydroxides, oxi-iodides, and ternary metal oxides. ~ condensed species: effect of concrete on fissionproduct volatility. Other important systems were also recommended at this meeting [8]. This initiative was followed by a second one, when

the C’ommission of the European Communities organized a specialists’ meeting at Ispra (l5- 17 January 1990) to rcvicw the current understanding of fission product chemistry during severe nuclear accidents in LWRS. AS a result 41 compounds/systems were identified that required improved thermodynamic data, such as hydroxides, mixed fission-product/bulk materials compounds, such as Cd12, and ternary systems (e.g. Cs2MnOJ and CszCdl,). The acquisition of thcrmochemical data for mixed fission-product/bulk materials compounds was seen as a priority: the surface area of aerosols generated from bulk reactor materials has been estimated to be two to three orders of magnitude greater than the arca of the primary circuit surfaces. However, few data were then available for compounds. such as iodides and tellurides of the control-rod conponcnts (cadmium, indium and silver). The proposed studies have been made both by measurement and by evaluation during the past two years in a joint effort of Winfrith Technology Centrc and Harwcll L,aboratory in the UK, and Netherlands Energy Research Foundation EC‘N at Petten with financial support of the Commission of the European Communities (CEC). The results obtained in thcsc studies have been published in detail in CEC reports [Y]. All of the data generated allow more accurate analysis and interpretation of the chemical phenomena to occur during severe nuclear accidents. A selection of the results obtained, has been made for presentation at this symposium [IO]. In the present review some cxampies have been chosen to illustrate the impact of new and improved thermodynamic data, including data ot new species. on consequence analysis.

3. Calculational

procedures

During fission of uranium more than 30 elements are formed. The initial states of the fission products arc individual atoms in interstitial positions in the UO, lattice. Depending on temperature and oxygen potential the atoms move more or less easily to stable states. either a gas bubble or a secondary phase [ll]. Above 1400 K sufficient lattice mobility exists allowing the fission products to dissolve in the U02 lattice (e.g. Zr or La) or to nucleate in gas bubbles. When the vapour pressure of a volatile fission product compound in the gas bubbles exceeds its equilibrium value, a condensed phase will be formed. The formation of these condensed phases is an important aspect of the bchaviour of the fission products in accident scquenccs, bccausc it controls the partial pressures of the fission product

E.H.P. Cordfunke, R.J.M. Konings / Release of fission products from degraded UO, fuel

compounds, and thus determines their fraction in the gaseous phase. At temperatures above 1400 K kinetic limitations will not play a predominant role and, hence, chemical reactions occuring during a severe reactor accident, will be extremely rapid. This enables us to calculate the chemical state of the fission products in the fuel, in the gap, and during transport after release from degrading fuel to different locations in the nuclear plant. The thermochemical data used here are taken from our data assessment [7], the recent CEC study [9,10], and (still) unpublished data. The approach for calculating the chemical equilibria is straightforward, and has been described in detail elsewhere [12,13]. Briefly, the chemical equilibria for a PWR have been calculated by using the ChemSage program [14], which calculates the equilibrium positions of a multiphase, multicomponent system by minimising the total Gibbs energy of the system. An average burn up of ca. 3% of the uranium atoms has been taken to calculate the quantities of the fission product elements present in the fuel (table 1). The chemical forms of the condensed phases in the fuel depend on the temperature and on the oxygen potential [12,13,15]. Under normal operation conditions, the slightly enriched UO, fuel of LWRs has a central temperature of 1500 K, and little deviation from the stoichiometry. This means that initially an oxygen potential of around -550 kJ mall’ is encountered which could increase to some extent during fis-

59

1500 1000 K Fig. 1. The radial temperature and oxygen-potential distribution in a LWR UO,-fuel pellet under normal operating conditions.

sion depending on the burn up when more oxygen is liberated by fission than is retained by the fission products (fig. 1). However, MacInnes and Winter 1161 recently suggested that the noble gases Kr and Xe are bound in defect clusters, and bind oxygen which lowers the oxygen potential. Calculations of the equilibria in the fuel should therefore be done with oxygen potentials over the entire range to be expected, thus from -700 to -400 kJ mall’.

4. The chemistry of the fission products in degrading fuel

Table 1 The amounts of fission products elements used in the calculation chemical constitution of the fuel and the gap, at a burnup of _ 3%, and the amount of uranium present in a fuel pin, after NUREG-0956 Fission product element

Amount, (10’ mol)

mol%

U Xe Kr cs Rb I Tc MO Zr Ru Ba Sr RE

132.95 4.92 0.400 2.444 0.430 0.242 0.492 4.014 4.896 2.560 1.109 1.351 7.832

95.5 0.64 0.05 0.32 0.06 0.03 0.06 0.52 0.64 0.33 0.14 0.18 1.02

Knowledge of the chemical forms of the fission products present in degrading fuel is necessary in understanding the release rates, and the subsequent reactions that may take place in the primary system. Of the fission products formed, only a relatively few species are important since they pose a radiological hazard when released to the environment. This is the case with the volatile fission products iodine, cesium and tellurium. The fission products ruthenium, barium, strontium, and the rare earths have chemical forms which are only volatile under the extreme conditions of a severe reactor accident when the core melts and subsequently reacts with the concrete basemat in the presence of steam. 4. I. Iodine / cesium / tellurium

Iodine is one of the most important radionuclides to consider in nuclear accident analysis. Its chemical form in the fuel has long been the subject of conflicting

E.H.P.

ho

Cordfunke,

R.J.M. Konings

/ Release of fission products from

results obtained by different authors [3]. Theoretical simulation techniques at atomistic level [17] and thcrmodynamic calculations predict that in the fuel the formation of CsI is favoured over solution of the individual atoms in the matrix. Careful release experiments at Oak Ridge National Laboratory support the calculations that CsI rather than elemental iodine is the dominant form released from the fuel under severe accidents [18]. The most favourablc sites where CsI compound formation can take place arc within the fission-gas bubbles, and at grain boundaries or cracks. It has been shown [19] that at low-burnup conditions (< 5000 MWd/t) CsI formation will not take place, whereas at higher burnups the bubble microstructure can be expected to serve as reaction sites for CsI compound formation. The reaction kinetics at 1500 K is calculated to be so fast that if the concentration criterion is satisfied, CsI formation will occur. The fission product cesium is formed in large excess to iodine, the molar ratio Cs/I being about 10 (table 2). According to the thermochemical calculations the excess cesium over iodine in the fuel will react with the fission product molybdenum, Cs,MoO, being formed. Deposits of CsI and Cs,MoO, have been observed on the fuel side of the cladding [3,20]. Since the oxygen potential for the MO/MOO, equilibrium at 1000 K has the value -407.4 kJ mol- ‘, molybdenum will be in its

drgruded

UO, fuel

elemental form below this value, in solid solution with other metals, such as Tc and Ru [3]. The cesium potential will thus be determined by the reactions 2Cs(g> + MO(S) + 20,(g)

--7)Cs,MoO,(s)

or

2Cs(g) + Moo,(s) At relatively 2CsI(I)

+ O,(g)

+ CsIMo04(s).

high oxygen potentials

+ Moo,(s)

+ O,(g)

the reaction

+ &Moo,(s)

+ I:(g)

( 3j as suggested

by Giitzmann [21], can occur. This would increase the iodine potential dramatically (fig. 3). However, this reaction can only be envisaged at oxygen potentials > - 172 kJ molt ’ which are normally not encountered in UO, fuel. The present thermochemical calculations indicate that under the conditions in the fuel cesium will not react with UO, to give a uranate, such as Cs,UO, or Cs $JO3.5, as suggested previously [2,12,13]. Instead, at low oxygen potentials ZrOZ in the fuel will react according to the reaction CszMo04(s)

+ ZrO,(s)

-+ Cs,ZrO,(s)

+ MO(S)

+ l$OZ(8). At

1000 K the

-10

threshold

is

(4)

- 535 kJ mol

MOO2

1500

K

‘. The

ON

-300

T -400

z E 1 Y -.

c:

aG c -500

1

\

I

-10

-5

1

0

-10

5

0

log pcs/atm

log p#tm

Fig. 2. Predominance

-5

area diagrams

of the Cs-Zr-MO-U-0

system at (a) 1000 K and (b) 1500 K.

5

E.H.P. Cordfunke, R.J.M. Konings / Release of fusion products from degraded VO, fuel

Table 2 Speciation of fission product cesium in the fuel (AGo, = - 500 kJ mol-‘) 1000 K

1500 K (mol%)

CSI Cs 2MOO, Cs,Te Cs,ZrO, Gases (mainly Cs and CsIl

9.9 49.8 40.3 0 0.004

9.3 0 40.3 45.0 5.4

with previous calculations is mainly due to improved data for the Cs,ZrO, phase 19,101.It should be noted that many of the chemical equilibria in UO, fuel are strongly temperature-dependent. Whereas at 1000 K Cs,ZrO, can be formed only at low oxygen potentials (< -535 kJ mol-l), a different situation is obtained at 1500 K (table 2). At this temperature Cs,MoO, will not be formed at the oxygen potentials which may exist in the fuel, the threshold now being -317 kJ mol-‘. The phase diagrams (fig. 2) illustrate the stability areas of the chemical forms of molybdenum. The formation of less-volatile Cs,MoO, or nonvolatile Cs,ZrO, is advantageous since it will reduce the volatility of cesium appreciably compared with elemental cesium or even cesium iodide, and hence its release. However, the volatility of molybdenum will be enhanced since cesium molybdate vaporizes congruently [22], and may as such provide a route for its transport to the primary circuit. The fission product tellurium is of considerable importance in risk assessment since it is a precursor of iodine, and is present in the core at levels similar to iodine. At oxygen potentials below = -440 kJ mol-’ cesium telluride, Cs,Te, can be formed according to the reaction difference

Te(2) + Cs,MoO,(s) -+ Cs,Te(s)

+ O,(g) + Moo,(s).

(5)

The formation of Cs,Te reduces the tellurium transport to the gap considerably since the gaseous Cs-Te species which are formed by incongruent vaporization of Cs,Te 1231, are less volatile than elemental tellurium, Te,(g), which is the main volatile tellurium species at higher oxygen potentials. The speciation of cesium, and hence its location within the fuel, is given in table 2 for two temperatures, 1000 and 1500 K.

61

4.2. Ruthenium Ruthenium is, under the conditions in the fuel, in elemental form and in solid solution with other metals, such as Tc, MO, and Pd, the so-called “white inclusions” [3]. Only at very high oxygen potentials (AGo* > - 133 kJ mall’) RuO,(s) can be formed. The ternary oxide Cs,RuO, may combine and influence, as with Cs,MoO,, the behaviour of two radiologically important fission products. The thermochemical properties of this volatile compound, including its vapour pressure, have been studied recently [22]. From the present calculations it follows that it will not be formed in the condensed phase under the conditions in the fuel, and only as a minor species in the gaseous phase (table 3).

4.3. Barium and strontium The fission products barium and strontium have identical chemical behaviour in many respects. The chemistry of the strontium compounds is, however, less well known. At low oxygen potentials and below 1000 K barium and strontium form the very stable zirconates BaZrO, and SrZrO, which are soluble into each other [24]. At temperatures higher than 1000 K the uranate BaUO, will be formed as well. The corresponding strontium compound SrUO, does not exist, and a compound with the net composition Sr,UO,, could be formed [25]. At oxygen potentials higher than - 400 kJ mall ’ BaMoO, appears to be more stable than BaZrO,, and,

Table 3 Vapour pressures of selected fission product species in the gap at a fuel surface temperature of 1000 K, and AGo2 = - 500 kJ mol-’ Species

Vapour pressure (bar)

I2 I CSI

1.5 x 10-22 6.9 x lo- l3

cs

Cs,MoO, Cs,RuO, Cs ,Te CsTe Te2 BaO Sr

2.0x10-3

1.5 x 10-4 5.5 x 10-s 7.5 x 10-34 7.oxlo-s 2.9x10W9 5.1 x lo- la 3.9x 1o-2’ 4.7x 1o-21

62

E.H. P. Cordfunke, R.J.M. Konings / Release of fission products from degraded UOz fuel

as a consequence, to

MOO, in excess will react according 0

BaZrO,(s)

+ Moo,(s)

+ +0,(g)

-

+ BaMoO,(s) + ZrO,(s).

(6)

This means that under these circumstances molybdenum is present as Cs,MoO,, BaMoO,, and MOO, (or MOO,). The solubility of BaUO, in BaZrO, is found to be negligible [24]. This implies that the complex “grey phase” is unlikely to contain much uranium, contrary to previous observations by Kleykamp [3]. The solubility of barium in UO, was found to be enhanced by the presence of smaller cations (e.g. Y’+) in the lattice [26]. It should also be noted that the speciation and hence the location of the fission product compounds within the fuel is temperature dependent. Postirradiation experiments may therefore not represent the actual locations of the compounds during operation or not give the actual compositions. The speciation may thus differ between the low-temperature hot cell examinations and the high-temperature accident conditions in the fuel.

5. Release from the fuel Below 1400 K the mobility of the fission products in the fuel is slow, the rate-determining step probably being the uranium-vacancy mobility in the UO, lattice. Theoretical calculations at atomistic level have helped to better understand the behaviour of the fission products in both normal operating and accident conditions [27]. This was done by calculating the trap site equilibria for the fission products with respect to the Frenkel and Schottky defects, and to the stoichiometry. From the results it has been concluded that in UO,_, the most favourable site at which the fission products can bc incorporated, is the neutral trivacancy which is formed by a uranium vacancy bound to hvo oxygen vacancies. The mechanism of diffusion of Xe in these traps can be explained in this way 1281, and is consistent with experimental data. In UO,+* the uranium vacancy is the trap site for all cations which explains the similarity in magnitude of the release rates of the noble gases, iodine, and cesium at high temperatures [2]. The diffusion coefficient D” increases dramatically with x in UOz+x at a constant temperature, e.g. by about a factor of lo5 between UOZ.,,,, and UO,.,” at 1770 K. This is predominantly due to an increased concentration of uranium-vacan-

,E

h

n

-5

0”

-10

I, -700

-500

A -300

-100

\Go,/kJ-molK’ Fig. 3. The vapour pressures of Iz, Te2 and Cs as a function of the oxygen potential in UO, fuel at 1000 K.

ties in UO,,,. Experimental studies [29] showed that _ as the fuel is oxidized - the xenon release increases significantly. The concentration of the volatile fission products species in the fuel-clad gap under normal operating conditions has been calculated with the improved thermodynamic data from the studies mentioned above. It has been assumed that the volatile species are in equilibrium with the LWR fuel of which the surface temperature is 1000 K. The main volatile species that can be expected to be present in the gap, are listed in table 3. From the results it can be concluded that the speciation of the volatile fission product species is strongly dependent on the oxygen potential of the fuel which controls the chemical equilibria in the fuel. This is particularly evident for the elemental species I, Tc,. and Cs, as illustrated in fig. 3. For instance, the tellurium potential rapidly drops at AC,,? < -440 kJ mol..’ due to the formation of Cs,Te. The same situation holds for I (formation of CsI) and Cs (formation of Cs,MoO, or Cs,ZrO,). It is evident that, in order to limit the amounts of elemental species in the gap, the stoichiometry of UO, should be in the vicinity of UO,.,, (AGo, = -550 kJ mol-‘1. At higher temperatures, under accident conditions, the fission product gases migrate to the grain boundaries where they precipitate into intergranular bubbles

63

E.H.P. Cordfunke, R.J.M. Konings / Release of fission products from degraded UO, fuel

and from which they will be released at a rate depending on the accident sequence. Postirradiation annealing is the usual technique of measuring gas release from solids. However, such measurements are by nature integral tests, and may not represent the actual release during accident conditions. There are a number of models to describe the release of the volatile fission products: (1) Semi-empirical models mostly based on correlations as given in NUREG-0772 [2]. It has been shown by Andriesse and Tanke, that the release process is dominated by diffusion in grains and that the release rate can be described by an Arrhenius-type rate equation [30]. The CORSOR-code [31], based on a similar correlation, treats the fission product release from overheated fuel as being solely due to temperature effects. The basic objection is that the experiments on which these correlations are based, cannot be simply extrapolated to the reactor conditions. (2) The FASTGRASS code developed at Argonne National Laboratory [32], is a detailed mechanistic model which accounts for diffusion of gases in grains, nucleation in bubbles, bubble migration, growth and interlinkage of fission gas bubbles on grain surfaces, as studied experimentally in detail by Matzke [33]. The FASTGRASS code contains models for swelling and cracking, and is valid at least to the point of nonlinear swelling, but not at extreme disruption conditions. (3) The VICTORIA code developed at Sandia National Laboratory [6], treats the grains in the fuel, open porosity, the fuel-clad gap, and bulk coolant. It performs equilibrium chemical calculations in each of these regions in each time step, and contains for that purpose a thermochemical database in which 167 chemical species have been included. This library will be extended to 250 species in the 1992-version. The VICTORIA code can therefore be considered as the state of the art for calculations of the release of fission products during the in-vessel stage of a severe LWR accident. The

importance

FASTGRASS

of mechanistic

or VICTORIA,

codes, such can be illustrated

as by

considering the release of cesium and tellurium. Originally, e.g. in NUREG-0772 [2], it was assumed that these volatile fission product species would be completely released from the fuel during in-vessel accident progress. However, from the analysis of the TM1 accident it is clear that a significant fraction of ccsium (20%) can be retained within the fuel [34], most likely as Cs,ZrO, as follows from the foregoing considerations.

Table 4 Speciation of volatile compounds in the Cs-Te system at 1000 K and AGo, = -300 and -500 kJ mol-’ Compound

Vapour pressure (bar) -500 kJ mol-’

-300 kJ mol-’

CS Cs ,Te CsTe

1.47x 10-4 7.04x 10-s 1.96~ 1O-9 1.15 x 1o-9 3.4 x10-l’ 4.17x1o-14 1.45x10-” 5.0 x10-‘8

3.65 x lo-” 4.40x lo-‘5 4.96x 1O-9 7.10x 10-25 2.17x lo-” 1.08X 10-S 9.5 x10-’ 5.24x1O-2

cs2

Cs,Te, CsTe, Cs,Te, Te2

According to NUREG-0772 the fission product tellurium has a release rate just below that of iodine and cesium, but its release is complicated by interaction with the inner surface of the Zircaloy cladding. It is initially held according to the reaction

Cs,Te,(g)

+ 3yZr(s)

+yZr,Te(s)

+xCs(g).

(7)

The vapour pressures of the Cs,Te, species strongly depend on the oxygen potential (table 4), as can be calculated from a recent study by Drowart and Smoes [23]. The reaction product formed, was assumed to be ZrTe, [3.5,36]. However, in a recent study [37] it was found that the compound ZrTe, is not stable under the conditions prevailing in the fuel element, and that, instead, a telluride with the composition Zr,Te is formed. It has a tetragonal unit cell, comparable with that of the same phase described by Matkovic et al. [38]. Thermodynamic properties of the Zr,Te phase are not available. Tellurium remains in the cladding until it is appreciably oxidized [39], which occurs above 1900 K. When oxidation of the Zircaloy clad exceeds 70%, the tellurium release, probably as SnTe(g) [40], approaches the values predicted in NUREG-0772. The degree of oxidation of the Zircaloy cladding is thus an essential parameter of the tellurium release. By contrast, the relatively large release of ruthenium from TM1 core debris is unexpected based on NUREG-0772. This can only be explained by high local oxygen potentials resulting in the formation of gaseous binary and ternary ruthenium oxides. These examples illustrate the need of accurate modelling of fission product release as is currently employed in safety analysis studies. The thermochemical databases form an essential part of these studies.

64

E.H.P. Cordfunke, R.J.M. Konings / Release of fission products from degraded UO, jitel

6. The primary system

After rupture of the cladding during a severe accident, the fission product compounds which exist in the vapour will be released into the primary system. Originally, no retention of the fission products in the primary system was assumed, for instance, in the WASH1400 Reactor Safety Study (1975) [I]. However, it is evident that during the transport process they will react with the aerosols in the cooler regions of the vessel and with the surface of the structural materials. The chemical reactions in the primary system to be expected, depend on the local temperature and the H-,/H,0 ratio (typical values range from 0.1 to 10). The reactor coolant system (RCS) or primary system is in many ways an extension of the core region, since its chemical features include gases and surfaces. It is a region of great chemical importance because gas and surface temperatures are quite hot at points near the core region, but decrease along the flow path. During the flow to cooler regions, significant changes in the physical and chemical composition of the transported fission products are likely. The TRAP-MELT code [41] describes the chemical retention of volatile fission product compounds, such as CsI, CsOH, and Te, due to reactions with the structural materials, like pipes and steam generators. The transport of the fission product compounds clearly depends on their chemical form as has been demonstrated, e.g. in the Falcon facility at Winfrith (UK) [42]. The Falcon experiment has been designed to study the transport of fission product vapours and aerosols in both the primary system and containment building to test the computer codes, and is as such complementary with large-scale integral experiments, like Phebus-FP (CEC) [43] or ACE (Advanced Containment Experiments) (EPRI). Spence and Wright [44] have calculated that the fission product source term to the containment building can be altered by several orders of magnitude as a result of such chemical changes. For instance, they predicted that tellurium can be trapped on stainless steel or Inconel surfaces, where it was combined with iron or nickel as tellurides, if subsequent heating does not revaporize the tellurium. Clearly, an understanding of these processes, and the development of relevant computer modelling, such as VICTORIA, can only be achieved by a detailed assessment of the chemical aspects. A comprehensive review in which the major uncertainties were indicated, has been given by Bowsher [45]. More specifically, the importance of interactions between the fission product vapours and bulkmatcrial aerosols (Ag-In-Cd control rod and boric acid)

,OQ-7,

.~,

_

I

/--,

.

.-

CSI

10 ’

CsOH

1000

500

2000

1500 T/K

Fig. 4. The cesium speciation as a function of temperature for p = 100 bar and molar ratios n(CsI)/n(H, + H,O) = 10 ” and n(H,)/n(H20)= I.

in determining the magnitude of containment, was emphasized. The cal data obtained in the CEC study to reduce these uncertainties. This here with some examples. 6.1. The interaction

with hydrogen

the release to the new thermochemi[9], were necessary will be illustrated

/steam

mixtures

During release from degrading fuel, Csl may react with steam to form CsOH and HI [46]: Cd(g)

+ HzO(g)

+ OH(g)

+ HI(g).

Depending on the fission product concentrations and gas temperatures either CsI (favoured at high concentrations: fission product/steam ratio > lo- ‘) or HI (lower concentration and high temperatures) will be predominantly present [47]. The speciation of cesium

10-s

10-a i -3.00

(CSOH)~

r -~~~ ~1~~~

-1.80

-0.60

0.60

csz 12

1 1

_

L

100

___J 300

lwWJH,O)

Fig. 5. The cesium speciation as a function of the Hr /H,O ratio for p = 100 bar, T = 1500 K and molar ratio n(CsI)/n(Hz+I~IZO)=10 -‘.

E.H.P. ~ord~~n~e, R.I.M. Konings / Release o~~sion prodwis from degraded UU, Fiji

can also be calculated in dependence of the temperature or H&I,0 ratio using the ChemSage programme. Typical results are given in figs. 4 and 5. The atmosphere to which the fuel rod is exposed during a severe LWR accident will not only affect the chemical forms and the extent of the release of the fission products but also their subsequent behaviour. For instance, once CsOH is formed, it may react with structural materials to form compounds, such as cesium silicate (Cs,Si,O,), cesium phosphate or manganate. These reactions would reduce the release of cesium to the ~ntainment appreciably. A model for the retention of cesium in the primary system due to the formation of Cs,Si,O, is given by Elrick and Sallach [48]. By contrast, CsI is very stable towards reactive surfaces, as follows from experiments by the same authors. An important role is also played by boron which is present as boric acid as an additive in the coolant water of PWRs or as B,C in control blades in BWRs. In steam atmosphere its volatility is enhanced due to HBO, species in the vapour phase where it may react with CsI and CsOH 1491: CsOH(g) + HBO,(g) CsI(g) + HBO,(g)

+ CsBO,(s)

-+ CsBO,(s)

+ H,O(g),

-t HI(g).

(9) (10)

Cesium borate formed is not reactive; moreover, it is less volatile than CsOH as follows from our vapour pressure measurements [SO]. The~odynamic calculations by Minato [Sl] have shown that for mole ratios B/(H, + H,O) greater than 5 X 10-3, HI is the predominant iodine species; below this value CsI predominates. It should be noted that a thermochemical treatment refers to the equilibrium state of the system, but in practice reaction kinetic limitations can exist. Cronenberg and Osetek 1471 showed that for the system iodine/cesium/steam/ hydrogen the times to establish equilibrium are of the order of tens of seconds when the fission product to steam molar ratios are less than 10e8. At fission product to steam ratios greater than lo-‘, equilibrium is attained more rapidly (< 10m4 s) and only thermodynamic limitation then prevails. 6.2. The interactions of CsZ and HZ with the (Ag, Zn, Cd} control materials For specific PWR conditions the influence of the control-rod components (Ag, In, Cd) must be taken into account since all three metals form stable iodides. The analysis of the control-rod materials in the TMI

65

reactor building indicated widely varying ratios of these elements from their initial ratios (80% Ag, 15% In, and 5% Cd). Larger fractions of Cd were volatilized and transported than for Ag or In. Recent studies at ECN, in collaboration with AEA Winfrith and AEA Harwell, have increased current knowledge of the iodides of the relevant metals significantly. The iodides of In and Cd [9,52-551 were studied in the solid and in the gas-phase and thermodynamic assessments for all three systems Ag-I [56], In-I IS], and Cd-I [9] have been made. The work carried out means a significant improvement in the thermodynamic database and will allow a more realistic calculation of the iodine chemistry in the primary circuit of PWRs to get insight into important issues like vapour condensation, vapour/ aerosol interaction, revolatilization and fission product trapping in water pools. For example, the formation of AgI can have extreme effects on the release of iodine: silver iodide has a negligible solubility in water whereas that of CdI, is high. As a consequence the formation of AgI will prevent the dissolution of iodine in the sumpwater and will thus act as a sink for the biohazardous iodine. The most important species that are involved in the transport of iodine from the core region to the primary circuit are CsI(g) and HI(g). The reactions with the aerosols or vapours formed by the control-rod materials (Ag, In, Cd), can be described by xCsI(g) + M(s) +xCs(g)

+ MI,(g),

xHI(g) + M(s) + (x/V&(g)

+ MI,(g).

(11) (12)

Simple thermochemical calculations for reactions between the iodine compounds and silver, show that CsI is thermodynamically stable but that HI(g) reacts with silver to form AgI. This effect has been demonstrated e~erimentally by Sallach et al. [57] who observed a reactive behaviour of HI towards solid silver between 670 and 930 K and not of CsI. Under accidents conditions, however, one has to consider the competition between the formation of the iodides of Ag, In, and Cd. In order to predict the chemical speciation of iodine in the presence of control rod material, thermochemical equilibrium calculations have been performed for both CsI and HI as iodine bearing species in the multicomponent/multiphase system Cs-I-Te-Ag-In-Cd-O-H. Typical results are shown in figs. 6 and 7. The calculations show that CsI is rather stable up to about 1000 K, above which temperature the fraction decreases to a constant value of about 10e3. Above 1000 K, In1 is the predominant vapour species, its mole

E.H.P. Cordfunke, R.J.M. Koniqs

IO’

as CsI or HI in the steam/ hydrogen gas flow from the core region into the primary system. In this respect. the role of boron is of great importance since the formation of CsBO,. as predicted by thermochemical calculations, may lead to the presence of HI (eq. (10)).

HI

:

Ag’

10'

1

<’ Cdl 10 ,06 5

/ Releuse of fission products from degraded UO, furl

---

500

.~r-l.I_..--i_i.-

1500

1000

:1 2000

‘UK

Fig. 6. The iodine speciation the system Cs-Ag-In-Cd-I-O-H ratios n(CsI)/n(H, +H,O)=

as a function of temperature for for p = 100 bar and molar 10-” and n(H2)/n(H,0)= 1.

being close to 1 up to about 1700 K. At temperatures above 1500 K the fractions of HI, AgI and Cd1 further increase. The formation of Cs,CdI, as suggested by Beard et al. [58], is thermodynamically not favoured. However, if HI is the iodine bearing species through which iodine is being transported, the speciation is quite different below 1000 K where the formation of CdI?(g) predominates. Above this temperature its stabilitity rapidly decreases in favour of InI( and the same equilibrium speciation is approached as for the CsI case. The above calculations show that in the cast of interaction of iodine vapours/aerosols with the (Ag, In, Cd) control-rod alloy, the speciation of iodine is highly different below 1000 K if iodine is transported fraction

T/K

Fig. 7. The iodine speciation the system Ag-In-Cd-I-O-H ratios n(HI)/n(H2 +H20)=

as a function of temperature for for p = 100 bar and molar lo-’ and n(H2)/n(H20)= 1.

7. Core-concrete

interactions

When. in the unlikely event of reactor accident progression the core has melted through the reactor vcsscl, interactions of the core and the molten structure materials with concrete surfaces in the containmcnt cavity will take place. These molten-core/ concrete interactions are the principal source of lowvolatility fission product releases to the containment. The high temperatures of this process (> 2300 K) cause thermal decomposition of the concrete which leads to large amounts of CO, and steam being rcIcased. These gases bubble through the molten core debris, and are reduced to CO and hydrogen by zirconium and iron in the melt. When the gas bubbles bursl through the melt surface, many of the vapour species will condense immediately in the cooler environment forming aerosol particles which will be transported to the containment atmosphere. Research programmes to investigate molten-core/ concrete interactions were already initiated in the USA and in Germany around 1975. After small-scale laboratory experiments, computer programmes were developed which, in turn, required verification by large-scale experiments. In the USA the CORCON-VANESA code was developed at SANDIA National Laboratory. and in Germany the WECHSL code at KfK (Karlsruhe). The molten-core/concrete interactions wcrc thoroughly investigated in the German experimental facility BETA [59] and in the SURC experiments at Sandia [60]. These studies showed that segregation occurs in an oxide and a metallic phase, the former containing fission products like Ba and Sr. the lattct Ru, Tc, and Pd. The CORCON-VANESA code contains a library oi thermochemical properties for about 125 chemical species that might be formed, mainly elements, oxides. and hydroxides [61]. These data are used to calculate the release of fission product species by vaporization from the melt into gas bubbles. The magnitude of the release has been a point of considerable disagreement. It is beyond the scope of this review to give a full discussion of the various areas of uncertainty. For instance, the extent to which the volatile fission products, that are predicted to be retained on coolant

E.H.P. Cordfunke, R.J.M. Konings / Release of fission products from degraded UO, fuel

61

Due to the formation of silicates in the melt during -

Ba

SOLGASMIX

core-concrete interactions, the releases of barium, strontium, and lanthanum depend strongly on the type

. “““. 0

15

Weigh

45

50

percent

silica

60

-

75

in concrete

Fig. 8. The release of the fission products Ba, Sr and La from ) Calculated core-concrete mixtures (after ref. [64]).(with SOLCASMIX; experimental results: ( n ) Ba, (0) Sr, (0) La.

system surface, will remain on these surfaces after vessel failure. Another problem concerns the fission product ruthenium, which is not predicted to be released at all in the reducing environment of the reactor coolant system. However, it could be much more volatile if exposed to oxygen in the containment, thus giving rise to the formation of volatile oxides [62] or even ruthenates. As a result, considerable differences between the observed and calculated releases of ruthenium have been obtained [63]. These uncertainties are mainly caused by the uncertainties associated with the thermal-hydraulic conditions after vessel failure. A major problem encountered in the analysis and modelling of core-concrete interactions, has also been the lack of accurate thermochemical data at the high temperatures involved, both of the condensed and the gaseous phases. For example, as is shown in fig. 8, calculations with the SOLGASMIX programme for the release of lanthanum by Roche et al. [64] gave results which were up to a factor of 40 higher than the measured values. Although the results for barium and strontium were much better, it is evident that thermochemical data for these fission product compounds need improvement. A recent assessment at ECN [65] shows considerable uncertainties in the thermochemical values of the gaseous oxides and hydroxides of barium and strontium. For the condensed phases the lack of accurate heat capacity data at temperatures above 1200 K causes uncertainties in the Gibbs energy functions.

of concrete (SiO, content). However, thermodynamic data of the pure silicates and zirconates are still poorly known from the literature. To fill in this gap, a comprehensive experimental study has been started at ECN (Petten) to obtain these properties, the first results for which have been published recently [66]. In the CORCON-VANESA code the core-concrete system is treated as a pseudo-binary ideal liquid and solid solution. This will certainly not represent the actual situation, and for this reason other models have been developed to describe the phase equilibria of complex phase systems, such as UO,-ZrO,-SiO,CaO-MgO-Al,O, formed during core-concrete interactions [67]. The development of such a model requires the optimization of available thermodynamic data. It can also be used to calculate the extent of release of the components of the system by vaporization. The results show that interactions in the melt lower the activities and, consequently, the vapour pressures of the species in the vapour phase.

8. Containment

chemistry

As long as the containment building remains intact, aerosols and vapours generated in the primary circuit or during the core-concrete interaction will be retained within the barriers of the reactor. Settling of aerosols and condensation of vapour species are the predominant physical processes affecting the fission product distribution. Chemical processes that have to be considered here are the solution of aerosols/ vapours in water pools and reactions of iodine compounds with (epoxy-l paints to form volatile organic species (e.g. CH,I). The iodine behaviour in the reactor containment is complex due to the different oxidation levels that are possible. Most of the iodine is released from the primary circuit as metal iodides. The remaining small amount of free iodine (< 5%) will be distributed between the sump water and the gas phase. The distribution coefficient depends on pH [7], and only small amounts of airborne iodine can be expected to be released out of the containment. It should be noted, however, that the ionizing radiation present will increase the I, fraction. The expected extent widely differs, however, between the various authors. The same holds for the amount of organic iodine formed in the reactor containment. It is, however, outside the

68

E.H.P. Cordfunke, R.J.M. Konings / R&use of fissiorr products from degraded UC), fupl

scope of the present review to treat in a detailed way the vast amount of information presently available on these items [68].

Report for Fission Product Release Test VI-4. NUREG/ CR-5481 (19911, (bl Iodine Chemical Forms in LWR Severe Accidents, NUREG/CR-5732 (1991). [h] US Nuclear Regulatory Commission. VICTORIA: A

Mechanistic

9. Conclusions

The following conclusions can be drawn from the present review: (1) The chemistry of the fission products in the fuel is reasonably well known and can be described adequately by thermodynamic models. Thermochemical calculations show that the release of the volatile fission products is determined by the chemical changes in the bonding of the radionuclides (e.g. CsI or CszMoOJ) as a function of temperature and oxygen potential. (2) Thermochemical models for the behaviour of the fission products cesium and iodine in the primary system have been improved substantially in recent years as a result of considerable improvement of the thermochemical database. The major area of uncertainty is formed by the treatment of interactions of aerosols and vapours with structural materials such as stainless steel and Inconel since the activities of the minor compounds in these alloys (Si, Mn, P) are poorly known [69] and surface diffusion will play an important role. (3) The analysis and modelling of the chemical interactions during core-concrete interactions is still subject to uncertainties since the thermochemical database for the relevant compounds and systems is poorly defined at present. Major improvements of the database are required in the coming years, in addition to better knowledge of kinetics as well as transport chemistry, to reduce the uncertainties to acceptable levels.

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