Comprehensive nuclear materials 3 08 advanced concepts in TRISO fuel

Comprehensive nuclear materials 3 08 advanced concepts in TRISO fuel Comprehensive nuclear materials 3 08 advanced concepts in TRISO fuel Comprehensive nuclear materials 3 08 advanced concepts in TRISO fuel Comprehensive nuclear materials 3 08 advanced concepts in TRISO fuel Comprehensive nuclear materials 3 08 advanced concepts in TRISO fuel Comprehensive nuclear materials 3 08 advanced concepts in TRISO fuel Comprehensive nuclear materials 3 08 advanced concepts in TRISO fuel

K Minato and T Ogawa

Japan Atomic Energy Agency, Tokai-mura, Ibaraki, Japan

ß 2012 Elsevier Ltd All rights reserved.

3.08.2.3 Characterization Techniques for ZrC-Coated Particle Fuel 219

3.08.2.4.2 Resistance to chemical attack by fission products 221

3.08.3.1 Designs of ZrC-Containing TRISO-Coated Particle Fuel 2273.08.3.2 Performance of ZrC-Containing TRISO-Coated Particle Fuel 229

3.08.4.1 Designs of SiC-Containing TRISO-Coated Particle Fuel 2313.08.4.2 Fabrication of SiC-Containing TRISO-Coated Particle Fuel 2323.08.4.3 Performance of SiC-Containing TRISO-Coated Particle Fuel 233

Abbreviations

DB-MHR Deep-burn modular helium reactor

DBa(ZrC) Diffusion coefficient for Ba in the ZrC

GAC General Atomic Company

GFR Gas fast reactor HFIR High Flux Isotope Reactor HTGR High-temperature gas-cooled reactor ICP-AES Inductively coupled plasma-atomic

emission spectrometry IPyC Inner dense PyC JAEA Japan Atomic Energy Agency JAERI Japan Atomic Energy Research

Institute JMTR Japan Materials Testing Reactor JRR-2 Japan Research Reactor-2 LANL Los Alamos National Laboratory LASL Los Alamos Scientific Laboratory LMFBR Liquid metal fast breeder reactor MTS Methyltrichlorosilane

OPyC Outer dense PyC

215

ORR Oak Ridge Research Reactor

PyC Pyrolytic carbon

R/B Release-to-birth ratio

VHTR Very-high-temperature reactor

3.08.1 Introduction

The TRISO-coated fuel particle consists of a

micro-spherical fuel kernel and coating layers of porous

pyrolytic carbon (PyC), inner dense PyC (IPyC),

silicon carbide (SiC), and outer dense PyC (OPyC)

The chemical form of the fuel kernel can be oxide,

carbide, or a mixture of the two The function of

these coating layers is to retain fission products

within the particle The porous PyC coating layer,

called the buffer layer, attenuates fission recoils and

provides void volume for gaseous fission products

and carbon monoxide in the cases of oxide and

oxy-carbide fuels The IPyC coating layer acts as a

con-tainment to gases during irradiation and protects

the fuel kernel from the reaction with the coating

gases during the SiC coating process The SiC

coat-ing layer provides mechanical strength for the

parti-cle and acts as a barrier to the diffusion of metallic

fission products, which diffuse easily through the

IPyC layer The OPyC coating layer protects the

SiC coating layer mechanically

The recent interest in the coated particle fuel

concept includes its application outside the past

experience of the high-temperature gas-cooled

reac-tor (HTGR)1:

1 Very-high-temperature reactor (VHTR) with a

gas outlet temperature of 1273 K for supplying

both the electricity and the process heat for

hydro-gen production, as proposed in the Generation-IV

International Forum.2

2 Actinide burning in deep-burn modular helium

reactor (DB-MHR) with fuel kernels consisting of

high concentrations of transuranium elements.3,4

3 Advanced gas fast reactor (GFR) with nitride fuel,

which aims at a more improved performance

com-pared with the conventional liquid metal fast

breeder reactor (LMFBR) and/or the efficient

actinide burning.1,5

Although SiC has excellent properties, it gradually

loses its mechanical integrity at very high

tempera-tures, especially >1973 K.6–8The annealing

tempera-ture during fuel element fabrication is limited to

2073 K for 1 h Higher temperatures should result

in a porous structure due to the b-SiC to a-SiCtransformation and the thermal dissociation of SiC,9which leads to an extensive release of fission productsfrom the TRISO-coated fuel particles The fuel tem-peratures were limited to well below 1973 K duringthe design-basis accidents in HTGR designs.10–12Chemical interaction of the SiC coating layer withfission products is one of the possible performancelimitations of the TRISO-coated fuel particles Thefuel performance of TRISO is described inChapter

3.07, TRISO-Coated Particle Fuel Performance.The fission product of palladium is known to reactwith the SiC layer Corrosion of the SiC layer couldlead to fracture of the coating layers or provide alocalized fast diffusion path, which degrades thefission-product retention capability within the parti-cle Since the fission yield for palladium from239Pu isabout tenfold that from235U,13a careful particle fueldesign should be made in the actinide burning.The PyC layer develops gas permeability withincreasing fast neutron doses Intactness of the IPyClayer is crucial in keeping the integrity of the SiClayer in the oxide fuel When the IPyC layer fails, ordevelops gas permeability, the SiC layer will reactwith CO gas to form volatile SiO.14 The PyC layerwill also develop anisotropy above 2173 K, which isdeleterious to its irradiation behavior

To improve the high-temperature performance ofthe TRISO-coated fuel particles, a new material otherthan SiC is needed Zirconium carbide is a candidate,and ZrC-coated particles, where the SiC layer wasreplaced by a ZrC layer, have been tested New con-figurations of the coating layers, with a layer contain-ing SiC or ZrC added to the TRISO coating, havebeen proposed and tested to improve the chemicalstability of the TRISO-coated particles For the appli-cation to the fast reactor fuel, TiN coating layers havebeen proposed and tested instead of PyC layers.The following sections summarize the designsand the research and development of the advancedconcepts in TRISO fuel: (1) ZrC-coated particlefuel, (2) ZrC-containing TRISO-coated particle fuel,(3) SiC-containing TRISO-coated particle fuel, and(4) TiN-coated particle fuel

3.08.2 ZrC-Coated Particle Fuel3.08.2.1 Designs of ZrC-Coated Particle FuelZirconium carbide (ZrC) is known as a refractoryand chemically stable compound, which melts eutec-tically with carbon at 3123 K The properties of

ZrC are summarized in Chapter 2.13, Properties

and Characteristics of ZrC To improve the

high-temperature stability, the resistance to chemical

attack by fission products, and the retention of

fission products, the ZrC coating layer is a

candi-date that can replace the SiC coating layer of

the TRISO-coated fuel particle; the resulting

parti-cle is termed a ZrC-TRISO-coated fuel partiparti-cle

The apparent drawback of the ZrC-TRISO coating

may be that ZrC does not withstand the oxidation in

such an accident as a massive air-ingress accident

though it is highly hypothetical in the modern

HTGR designs

Historically, several coating designs have been

tested in the United States15,16: (1)

ZrC-TRISO-coated particles, (2) ZrC-TRISO type ZrC-TRISO-coated

particles without OPyC layer, (3) ZrC-coated

parti-cles with doped OPyC layer, and (4)

ZrC-coated particles with graded C–ZrC layer(s) In

the graded C–ZrC layer, the compositions were

changed gradually from the pure PyC through

the ZrC with excess carbon and into the pure ZrC

The graded layer was applied to either the inside

or the outside surface of the ZrC layer Propylene

was used to produce the pure PyC and to provide

the carbon for the graded portion of the codeposited

carbon and ZrC

ZrC-coated fuel particles are being developed in

Japan17 since the early 1970s The ZrC-coated fuel

particles at the early stage of the development were

characterized by a thick ZrC layer with a composition

of C/Zr> 1.0 and by the absence of the OPyC layer

A ZrC layer of this kind was called

‘zirconium-carbal-loy,’ meaning ZrC–C alloy Later, it was found that

the retention of metal fission products, especially

90

Sr, by the zirconium-carballoy was rather poor,18

presumably owing to a short circuit through the free

carbon phase It was also felt from the irradiation

experiences that the presence of the OPyC layer

was essential for the mechanical integrity of the

coated fuel particle The emphasis was, therefore,

placed on the development of ZrC-TRISO-coated

particles with the stoichiometric composition of

C/Zr¼ 1.0

Although most of the reported work on the use of

ZrC in coated fuel particles has been directed toward

the development of a replacement for the SiC barrier

layer, ZrC was also tested as a fission product and

oxygen getter,19,20in which ZrC was deposited over

the fuel kernel or dispersed throughout the buffer

layer These types of the coated particle fuels are

described inSection 3.08.3

3.08.2.2 Fabrication of ZrC-CoatedParticle Fuel

The coating layers of ZrC and ZrC–C were produced

by chemical vapor deposition, in which the pyrolyticreaction of zirconium halide with hydrocarbon in thepresence of hydrogen was used in principle Mainly,two different processes have been developed in sup-plying zirconium halide to the coater: (1) using ZrCl4

powder and (2) usingin situ generation of zirconiumhalide vapor

The chemical vapor deposition of ZrC has beenstudied using a gas mixture of CH4, H2, ZrCl4, and Ar

at Los Alamos Scientific Laboratory (LASL; now LosAlamos National Laboratory, LANL).16,21–23 A keydevelopment in the ZrC coating project proved to bethe ZrCl4powder feeder for metering ZrCl4into thecoater ZrCl4 is a solid at room temperature andsublimes at 625 K In this process, hygroscopicZrCl4powder was supplied from the powder feeder,whose rate was controlled by the auger speed andmetered by the output of the load cell on which thepowder feeder was hung The powder was swept by

Ar to the coater base where it was mixed with theother coating gases supplied from a gas manifold TheZrCl4powder in the gas stream was vaporized in thecoater base before entering the coating chamber.22

Figure 1 shows an apparatus for ZrC deposition bythe process using ZrCl4.22

The effects of varying CH4and H2concentrationsand particle bed area on the coating rate, the appear-ance, and the composition of the ZrC were studiedusing the ZrCl4 powder feeder.22 Increases in CH4

Argon

Mixing chamber

Gas manifold

Induction heater

Coater

ZrCI4 powder Auger

Drive motor

Figure 1 Experimental apparatus for ZrC deposition by the process using ZrCl 4 Reproduced from Wagner, P.; Wahman, L A.; White, R W.; et al J Nucl Mater 1976, 62, 221–228.

and H2concentrations were found to be effective in

increasing the linear coating rate of ZrC Increases in

the ratio of CH4to ZrCl4in the coating gas resulted

in a decreased metallic appearance of the coating and

an increase in the C/Zr in the deposit Increases in

H2inhibited these effects

The ZrC coating layers were prepared using a gas

mixture of C3H6, H2, ZrCl4, and Ar with the same

coater and ZrCl4 powder feeder as described

ear-lier.23 In general, ZrC coating layers made with

CH4and C3H6were similar and were affected

simi-larly by variations in the hydrocarbon and hydrogen

concentrations The coating layers produced using

C3H6 were more sensitive to changes in hydrogen

concentration than those produced using CH4

The coating processes based on thein situ

genera-tion of zirconium halide vapor are developed at Japan

Atomic Energy Research Institute (JAERI; now Japan

Atomic Energy Agency, JAEA) to avoid the handling

of highly hygroscopic halide powder.17Several

pro-cesses were studied: the chloride process,24the

meth-ylene dichloride process,25 the iodide process,26,27

and the bromide process.28–31Among these processes,

the bromide process proved to be the most convenient

and reliable; in this process, ZrBr4 is produced by

reacting bromine with zirconium sponge inside the

coater ZrBr4is preferred to ZrCl4since the reaction

of excess chlorine with hydrogen is a potential

explo-sion hazard

Figure 2shows an apparatus for ZrC deposition

by the bromide process.32 In this process, bromine,

which is liquid at room temperature, was carried by

argon onto the heated zirconium sponge beneath the

spouting nozzle and reacted to generate ZrBr4vapor,

which was mixed with the other coating gases of

CH4and H2before entering the chamber Propylene

could be used instead of methane The flow rate of

ZrBr4was controlled successfully by controlling the

flow rate of Ar passing through liquid bromine at

273 K and maintaining the temperature of zirconium

sponge at 873 K

Along with the deposition experiments of ZrC by

the bromide process, thermochemical analyses were

performed to find the optimum deposition

condi-tion.31The multiphase equilibrium in the system was

analyzed based on the minimization of the total free

energy of the system The analyses predicted that the

ZrC monophase region exists in a wide composition

range of the feed gas mixture It was concluded from

the analyses that the deposition rate could be

con-trolled through the methane flow rate, and the

com-position of the deposit through the ZrBr flow rate in

the presence of excess hydrogen The experimentalresults agreed with the predicted results By adjustingthe deposition condition, stoichiometric ZrC layerswere obtained with the bromide process

Recently, a new ZrC coater was installed andZrC coating experiments were carried out at JAEA.33The ZrC coater was designed with the maximum batchsize of 0.2 kg, which is about 10 times larger than theprevious one The ZrC coater mainly consists of thegas supply equipment, the coater, and the off-gas com-bustion equipment The coater is composed of thelower and the upper heaters with in-line configuration

Quartz

Carbon wool Graphite

Induction coil

The lower one is for the reaction of bromine with

zirconium sponge and the upper one, for the chemical

vapor deposition of ZrC at 1873 K in the maximum

The off-gas treatment equipment removes soot,

hydro-gen bromide, and residual hydrohydro-gen

3.08.2.3 Characterization Techniques for

ZrC-Coated Particle Fuel

It is important to characterize the key basic

proper-ties of the coating layers that are critical to the fuel

performance Although most of the characterization

techniques used for the ordinary TRISO-coated

par-ticle fuels can be applied to the ZrC-coated parpar-ticle

fuels, some techniques have to be developed

primar-ily for the ZrC-coated particle fuels

In the case of the ordinary TRISO-coated fuel

particles, the PyC layers are burnt off to recover the

SiC fragments for characterization, such as density,

composition, and strength measurements However,

it is almost impossible to separate the ZrC from the

PyC layers by the same method since ZrC, in contrast

with SiC, does not form a protective oxide layer,

resulting in oxidation of ZrC to ZrO2when exposed

to air at high temperatures To begin with, a method

of obtaining fragments of the ZrC coating layer from

the ZrC-coated particles is needed

The plasma oxidation method was developed to

obtain the ZrC fragments from the coating layers

con-taining the PyC.34The difference in the oxidation rates

between PyC and ZrC is very large In this method, the

samples were set in the plasma oxidation apparatus,

where low-pressure oxygen was ionized by high

fre-quency induction coupling Plasma reaction was

moni-tored by a color analyzer and an optical power meter

The color changed from the pale violet of pure oxygen

to pale blue during vigorous oxidation of free carbon,

and again to pale violet when the PyC was completely

removed and a very thin oxide scale was formed on the

ZrC The brightness also changed dramatically during

the reaction The obtained ZrC fragments were

exam-ined by Raman spectroscopy and X-ray diffraction

It was confirmed that the bulk of the ZrC remained

unaffected by the plasma oxidation.35

The physical grinding technique was also

devel-oped to obtain the ZrC fragments.33In this technique,

quartz powder having the Mohs hardness of 7 was

used since the Mohs hardness for ZrC is about 8–9

and that for graphite is 6 The fragments of the

combined layers of ZrC and PyC were ground with

quartz powder After grinding, the fragments of the

ZrC layers without PyC were separated from the rest

in liquid tetrabromoethane (C2H2Br4) by the densitydifference The specific gravity of tetrabromoethane

is 2.965 Mg m3.The density of the SiC layers is measured bythe sink-float technique or a liquid gradient column,

in which a liquid having the same density as thesample is needed The density of the SiC layers

is around 3.21 Mg m3, and a liquid mixture ofmethylene iodide (CH2I2) having a density of3.325 Mg m3and benzene (C6H6) having a density

of 0.8785 Mg m3, for example, is used for the surement of density In the case of the ZrC densitymeasurement, no suitable liquid is present since thedensity of the ZrC layers is around 6.6 Mg m3, and

mea-so, other techniques are needed Gas pycnometry,which required at least 100 mg of the samples,33wasdeveloped for the ZrC density measurement Thestoichiometry of the ZrC layer affects the thermalconductivity,36fission product retention,18,37etc Theproperties of ZrC are summarized inChapter2.13,Properties and Characteristics of ZrC Analysis ofthe free carbon is important for controlling the quality

of the ZrC coating Plasma oxidation with emissionmonitoring was also applied to the quantitative analysis

of the free carbon in ZrC powder.38The emission wasmonitored with an optical color analyzer and was cali-brated with standard samples of ZrO2+ C mixtures.The oxidation rates of the free and the combinedcarbons are so different that it is possible to estimatethe amount of the former from the emission Withpowdered ZrC of about 10 mg, free carbon of<1 wt%could be easily determined Without this method, thecomposition of the ZrC was estimated by burning theZrC and PyC together, weighing ZrO2and CO2, andthen subtracting the contribution of PyC from the totalamount of CO2.34

A new method to measure ZrC stoichiometry wasdeveloped using the infrared light absorption duringcombustion in oxygen and the inductively coupledplasma-atomic emission spectrometry (ICP-AES).With this method, accuracy of the C/Zr atom ratiowas on the order of 0.01.33

The defective SiC layer fraction is a very tant indicator to show the quality of the ordinaryTRISO-coated particle fuels This is measured bythe burn-leach method, where the fuel compact

impor-or the coated fuel particles are heated at 1173 K inair to oxidize the graphite matrix of the compact andthe OPyC layers, followed by the acid leaching

of the exposed uranium The defective SiC coatinglayer exposes uranium during burning However, thismethod cannot be applied to the ZrC-coated particle

fuels since the intact ZrC layer is oxidized and exposes

the actinide oxides to be leached by nitric acid solution

A new method of measuring the defective ZrC layer

fraction is needed

3.08.2.4 Performance of ZrC-Coated

Particle Fuel

3.08.2.4.1 Irradiation performance

Although systematic irradiation experiments on the

ZrC-coated fuel particles have not been completed,

some promising data have been obtained; the ZrC

layer is less susceptible to chemical attack by fission

products and fuel kernels, and the ZrC-coated fuel

particles perform better than the ordinary

TRISO-coated particles at high temperatures, especially above

1873 K Some early irradiation tests showed poor

per-formance of the ZrC-coated particles, which may be

attributed to the fact that the fabrication conditions

had not been optimized

Irradiation tests on several coating designs of the

ZrC-coated particles were carried out in the High Flux

Isotope Reactor (HFIR) and the Oak Ridge Research

Reactor (ORR).39–45 The irradiated ZrC-coated fuel

particles, which were made at LASL, were (1)

ZrC-TRISO-coated particles, (2) ZrC-TRISO type coated

particles without OPyC layer, (3) ZrC-coated particles

with ZrC-doped OPyC layer, and (4) ZrC-coated

particles with graded C–ZrC layer(s) Prior to the

tests on the fuel-kerneled coated particles,

carbon-kerneled inert coated particles were tested to

deter-mine the stability of the ZrC at temperatures of 1173

and 1473 K and fast neutron fluencies of 3.5 1025

to10.7 1025

m2.39,40 The irradiation results of the

inert coated particles were encouraging In the test of

the fuel-kerneled particles, it was found that the graded

coatings were cracked and it was postulated that the

cracking was associated with the low PyC deposition

rate and was not related to the ZrC.41 The

ceramo-graphic examination showed that the performance of

the ZrC-coated particles in the HRB-12 capsule

appeared to be poor in comparison with the ordinaryTRISO-coated particles However, there was no evi-dence of palladium attack on any of the ZrC layers.42The ZrC-TRISO-coated UC2, UO2, and UCxOyparticles made at General Atomic Company (GAC;now General Atomics, GA) were irradiated inHFIR.43–45 The irradiation conditions of the ZrC-TRISO-coated fuel particles in the United States arelisted inTable 1.42–45The performance of the ZrC-TRISO-coated particles was very favorable But therewas evidence that the ZrC-TRISO-coated UC2parti-cles in the HRB-7 and HRB-8 capsules might not be aseffective in retaining fission products as the ordinaryTRISO-coated particles; the electron probe micro-analysis showed the rare-earth fission products on theoutside of the ZrC coating, whereas all cesium wasretained within the coating.43The ZrC coating layers

in the ZrC-TRISO-coated UO2and UC2particles inthe HRB-15A capsule suffered no fission productattack, while the ordinary TRISO-coated fuel particlesshowed some degree of SiC-fission product interac-tion The ZrC-TRISO-coated UC2particles, however,showed poor retention of silver and europium, withgreat variety from particle to particle in this respect.44The irradiation tests on the ZrC-coated UO2par-ticles characterized by the zirconium-carballoy layerwithout the OPyC made at JAERI were carried out inthe Japan Materials Testing Reactor (JMTR).17 Theparticles in the 73F-12A and 73F-13A capsules expe-rienced very high temperatures exceeding 1873 K,where the failure fractions of the ZrC-coated parti-cles were rather high though they were on par withthose of the ordinary TRISO-coated particles orbelow The ceramographic examination revealed nointeraction between the UO2 kernel and the zirco-nium-carballoy layer when they came into contactwith each other due to the kernel migration, showingbetter performance of the zirconium-carballoy layerthan the SiC layer against chemical attack

The ZrC-TRISO-coated UO2 particles made atJAERI were irradiated in JMTR and the JapanTable 1 Irradiation tests of ZrC-TRISO-coated fuel particles in the United States

Capsule Fuel kernel Temperature (K) Burnup (% FIMA) Fast neutron fluence (m 2 , E > 29 fJ) Reference

Research Reactor-2 ( JRR-2) The irradiation

condi-tions are summarized in Table 2.46,47 The

release-to-birth ratio (R/B) of 88Kr was measured during

irradiation in the 80F-4A capsule, which showed no

through-coating failure of the ZrC-TRISO-coated

UO2 particles In the postirradiation examination,

no coating failure was detected by stereomicroscopy,

X-ray microradiography, and acid leaching The

ZrC-TRISO-coated UO2 particles in the VOF-14H

capsule were irradiated at 1873 K in a steep

tempera-ture gradient of 15 K mm1 The ceramographs of the

particles from the VOF-14H showed carbon deposits

at the colder end of the fuel kernel accompanied

by kernel migration up the temperature gradient In

the ordinary TRISO-coated fuel particles, palladium

attack of the SiC layer was occasionally found at the

colder side of the particle.48 However, there was no

indication of coating deterioration in the ZrC layer

In the 88F-3A capsule, the ZrC-TRISO-coated

UO2 particles and ordinary TRISO-coated UO2

particles were irradiated under identical

condi-tions.47 The postirradiation measurement of the

through-coating failure fractions revealed better diation performance of the ZrC-TRISO-coated fuelparticles.Figure 3shows a typical optical micrograph

irra-of the polished cross-section irra-of the coated fuel particle after irradiation.47Optical micros-copy and electron probe microanalysis on the polishedcross-section of the ZrC-TRISO-coated fuel particlesrevealed no interaction of palladium with the ZrCcoating layer or accumulation of palladium at theinner surface of the ZrC coating layer, whereas severecorrosion of the SiC coating layer was observed in theordinary TRISO-coated fuel particles

ZrC-TRISO-3.08.2.4.2 Resistance to chemical attack byfission products

The better performance of the ZrC coating layerthan the SiC coating layer against chemical attack

by fission product palladium has been demonstrated

in out-of-reactor experiments and irradiation tests.The out-of-reactor experiments of the chemicalreactions of ZrC and SiC with palladium wereperformed, where the ZrC-TRISO and ordinaryTRISO-coated particles were heated in either palla-dium powder or vapor.49The coating layers would beattacked by the fission product palladium from inside

in irradiation, while palladium was supplied fromoutside the particles in the out-of-reactor experi-ments to simulate the situation The experiments onthe reactions in the mixture of SiC, ZrC, Pd, and C,and the reaction of ZrC with Ag–Pd alloy were alsostudied Reaction morphology was observed by cer-amography and the reaction products were identified

by X-ray diffraction and electron probe sis When the ZrC-TRISO-coated particles wereheated in the palladium powder, ZrPd3and C wereformed However, no reaction was found on the ZrC-TRISO-coated particles heated in the palladiumvapor at 1830–2150 K, whereas the SiC layers wereattacked severely It was revealed that ZrC did reactwith palladium at a sufficiently high palladium activ-ity, but the reaction could not occur at a low palla-dium activity, such as in the fuel particles.49

microanaly-Table 2 Irradiation tests of ZrC-TRISO-coated fuel particles in Japan

Capsule Fuel kernel Temperature (K) Burnup (% FIMA) Fast neutron fluence (m 2 , E > 29 fJ) Reference

particle after irradiation at 1673–1923 K to 4.5% FIMA.

Reproduced from Minato, K.; Ogawa, T.; Sawa, K.; et al.

Nucl Technol 2000, 130, 272–281.

As described briefly in Section 3.08.2.4.1, the

comparatively better performance of the ZrC coating

layer than the SiC coating layer against chemical

attack by fission product palladium was confirmed

in the irradiation tests For example, the

ZrC-TRISO-coated UC4.6O1.1 particles irradiated at

1523 K to 86% FIMA in the HRB-12 capsule had

no evidence of palladium attack on the ZrC layers,

and the ZrC-TRISO-coated UO2and UC2particles

irradiated at 1328–1403 K to 24.9–28.8% FIMA in

the HRB-15A capsule suffered no fission product

attack, while the ordinary TRISO-coated particles

showed effects of the SiC-fission product interaction

The irradiation tests described earlier that were

carried out in the United States were characterized

by low irradiation temperatures and high burnup

compared with those in Japan The irradiation test

of the ZrC-TRISO-coated UO2particles in the

88F-3A capsule in Japan is a good example, where the

irradiation temperature was 1673–1923 K and the

burnup was 4.5% FIMA.47

Optical microscopy and electron probe

microanal-ysis on the polished cross-section of the

ZrC-TRISO-coated particles irradiated in the 88F-3A

capsule revealed no interaction of palladium with

the ZrC coating layer or accumulation of palladium

at the inner surface of the ZrC coating layer, as shown

inFigure 4(a).47Optical microscopy on the polished

cross-section of the ordinary TRISO-coated fuel

par-ticles irradiated under identical conditions, on the

other hand, showed severe corrosion of the SiC

coat-ing layer.Figure 4(b)shows an example,47which is a

typical feature of the corrosion of the SiC coating

layer by the fission product palladium.48,50

The fission product behavior inside the IPyC

coating layer should be the same regardless of the

ZrC or SiC coating layer as long as the IPyC coating

layer is intact It is reasonable to assume that the

fission product palladium is released from the kernel

to the ZrC coating layer in a fashion similar to its

release to the SiC coating layer in the ordinary

TRISO-coated fuel particles According to the

out-of-reactor experiment, the reaction of ZrC with

palladium will occur when the concentration of

pal-ladium is sufficient.49A probable explanation of the

absence of corrosion on the ZrC coating layer is that

palladium was not stopped by the ZrC coating layer

and never reached a concentration on the surface of

the coating layer to cause the corrosion.47

No data are available on the release behavior of

palladium from the ZrC-TRISO-coated fuel

parti-cles, but some data that may be relevant are available

for ruthenium It has been reported that rutheniumwas released from the ZrC-TRISO-coated fuel par-ticles during postirradiation heating tests at 1873,

2073, and 2273 K,51,52 while no ruthenium releasewas reported under similar conditions from the ordi-nary TRISO-coated fuel particles.53,54In addition, inpostirradiation examinations, ruthenium was some-times found at the inner surface of the SiC coatinglayer of the ordinary TRISO-coated fuel particles,indicating that it does not easily diffuse through theSiC layer.48

3.08.2.4.3 High-temperature stabilityThe better mechanical integrity of the ZrC-TRISO-coated fuel particles when compared with the ordinaryTRISO-coated fuel particles at high temperatures hasbeen revealed in the out-of-reactor heating experiments

(a)

(b)

IPyC ZrC

The out-of-reactor high-temperature heating

experiments on unirradiated ZrC-TRISO-coated

par-ticles, together with the ordinary TRISO-coated

particles, were carried out in a vertical graphite

resis-tance furnace contained in a stainless steel

water-cooled jacket.55The particles were heated at 2073 K

for 1 h to simulate the annealing effect during

com-pact fabrication, and then at the desired temperatures

for 1 h The ZrC-TRISO-coated UO2particles

with-stood the heating at 2723 K for 1 h, though more

than half of the particles failed at 2773 K within 1 h

The ZrC-TRISO coating layers were expanded

plastically leaving a large gap between the kernel and

the buffer layer by a large internal pressure of CO,

while the SiC coating layers decomposed to lose their

mechanical integrity The limiting factor of the

sta-bility of ZrC-TRISO-coated UO2particles is not the

chemistry of ZrC but that of the system enclosed

by the ZrC layer Although the ZrC itself was stable

up to the eutectic point of ZrC–C at about 3123 K,56

the ZrC-TRISO particles could not withstand the

heating above 2773 K Failure was induced by large

internal pressures.55

The postirradiation heating test of the

ZrC-TRISO-coated UO2particles was performed at a rate

of 1 K min1to the maximum temperature of 2673 K to

clarify the high-temperature stability of the particles.46

The particles were sampled from an irradiated fuel

compact at 1373 K to 4% FIMA after its electrolytic

disintegration A total of 101 particles were heated

individually, placed in holes of two graphite disks in a

cold walled furnace with a graphite heater During

heating, the radioactivity in flowing helium gas was

monitored with an ionization chamber The activity

was due mostly to85Kr When a through-coating

fail-ure or a pressfail-ure-vessel failfail-ure occurs, the activity of

fission gas retained within the particle is released

No failure was detected during the heat-up stage

An activity burst occurred only after keeping the

particles at 2673 K for about 6000 s (100 min) The

activity burst corresponded to one particle failure

among the 101 particles heated This interpretation

was confirmed by X-ray microradiographs of the

par-ticles after heating Figure 546 compares the failure

fractions of the ZrC-TRISO-coated UO2 particles

with the ordinary TRISO-coated UO2 particles.57

Under this heating condition, most of the ordinary

TRISO-coated particles would fail, as shown in

Figure 5, where the dashed line gives a model

predic-tion for the ordinary TRISO-coated particles with

the same dimensions as the ZrC-TRISO-coated

par-ticles in this study.46

The different behavior of the ZrC-TRISO-coated

UO2 particles compared with that of the ordinaryTRISO-coated UO2 particles at high temperatureswas discussed with the ceramographs of the particlesafter heating.Figure 6shows a ceramograph of the

1900 0

Calculated (SiC) Experimental (ZrC)

Figure 5 Comparison of failure fractions of the ZrC-TRISO-coated UO 2 particles with the ordinary TRISO-coated UO 2 particles Reproduced from Ogawa, T.; Fukuda, K.; Kashimura, S.; et al J Am Ceram Soc.

1992, 75, 2985–2990.

Figure 6 A ceramograph of the ZrC-TRISO-coated UO 2

particle that survived the postirradiation heating at 2673 K Reproduced from Ogawa, T.; Fukuda, K.; Kashimura, S.;

et al J Am Ceram Soc 1992, 75, 2985–2990.

ZrC-TRISO-coated UO2 particle that survived the

postirradiation heating at 2673 K.46 The OPyC and

ZrC layers expanded, while the IPyC layer did not

There was a significant difference in the behavior

of the ZrC coating layer compared with that of the

SiC at high temperatures The ZrC can sustain a very

large strain, whereas the SiC is brittle in nature The

high plasticity is explained by the fact that resistance

of ZrC crystal lattice to the dislocation motion

becomes very weak above 2473 K.58In these heating

tests, the particles were heated in a loose condition

without mechanical support from the surrounding

graphite matrix of the fuel compact The presence

of the graphite matrix could offset the coating

expan-sion and would further reinforce the integrity of the

ZrC-TRISO-coated particles.46

Isothermal postirradiation heating tests were also

performed to study the coating integrity and fission

product retention of the ZrC-TRISO-coated UO2

particles Three tests were carried out at 1873 Kfor 4500 h, at 2073 K for 3000 h, and at 2273 K for

100 h in a cold-wall furnace with a graphiteheater.51,52For each test, 100 particles were sampledfrom fuel compacts irradiated at 1173 K to 1.5%FIMA During all the heating tests, no through-coating failure was detected by the 85Kr releasemonitoring The X-ray microradiography on thecoated particles after the heating tests revealed noOPyC failure, which confirmed the results of thegas release monitoring

Typical polished cross-sections of the TRISO-coated fuel particles after the heating testsare shown in Figure 7.51,52No failure was observed

ZrC-in the ZrC and OPyC coatZrC-ing layers of the particlesheated at 1873 K for 4500 h, as shown inFigure 7(a)and7(b) In some particles, the IPyC coating layers

Figure 7 Ceramographs of the ZrC-TRISO-coated UO 2 particles after the postirradiation heating tests; (a) and (b) at

1873 K for 4500 h, (c) and (d) at 2073 K for 3000 h, and (e) and (f) at 2273 K for 100 h Adapted from Minato, K.; Ogawa, T.; Fukuda, K.; et al J Nucl Mater 1995, 224, 85–92; Minato, K.; Ogawa, T.; Fukuda, K.; et al J Nucl Mater 1997, 249, 142–149.

were cracked in the radial direction No palladium

attack or thermal degradation of ZrC was observed.51

On the polished cross-sections of the particles

heated at 2073 K for 3000 h, no failure of the ZrC

and OPyC coating layers was observed, as shown in

Figure 7(c) and 7(d) However, some degradation of

the ZrC coating layer seemed to have occurred The

inner and outer surfaces of the ZrC coating layers in

most of the particles heated at 2073 K were not

smooth In some particles, the IPyC coating layers

were cracked in the radial direction, where about a

quarter of the thickness of the ZrC coating layers

seemed to have been attacked along the grain

boundaries.52

The ceramography on the ZrC-TRISO-coated

fuel particles after heating at 2273 K for 100 h

revealed that the ZrC coating layers as well as the

IPyC coating layers were damaged in most of the

particles observed, as shown inFigure 7(e) and 7(f )

The ZrC coating layers were damaged through the

thickness Based on the results of the ceramographic

examination, the electron probe microanalysis, and

the thermodynamic analysis, the observed

deteriora-tion of the ZrC-TRISO-coated fuel particles was

attributed to the reaction of ZrC with CO gas caused

by the failure of the IPyC coating layer.52

In the case of the ordinary TRISO-coated fuel

particles, fission gas release was observed in the

post-irradiation heating of fuel elements at 2073 K for 100

or 200 h.53,54 Although the number of the

ZrC-TRISO-coated particles tested was small compared

with that on the fuel element, it is probably safe to say

that the ZrC-TRISO-coated fuel particles have higher

capability of fission gas retention than the ordinary

TRISO-coated fuel particles at high temperatures

3.08.2.4.4 Retention of fission products

Retention of the fission product cesium by the ZrC

coating layer has been demonstrated to be better than

that by the SiC coating layer though the data for the

other elements are limited compared with those for

the SiC coating layer

The diffusion coefficients for strontium and

barium in the ZrC coating layer were obtained in

strontium soaking experiments and postactivation

annealing experiments.59 The diffusion coefficient

for Sr in the ZrC coating layer,DSr(ZrC), was

esti-mated to be 2 1018m2s1at 1673 K and that for

Ba, DBa(ZrC), was estimated to be 2.9 1018 to

4.6 1018m2s1at 1673 K The retention of these

elements by the ZrC coating layer was better than

that by the SiC coating layer

The diffusion coefficients for silver, barium, methium, and cerium in the ZrC coating layer wereevaluated from annealing experiments.60Based on thedistribution of the nuclides in the ZrC measured byremoving the ZrC stepwise, the diffusion coefficient

pro-DBa(ZrC) was estimated to be 1.3 1017m2s1and

DCe(ZrC) was estimated to be 6.4 1018m2s1 at

1773 K The ZrC coating layer showed better tion of these elements, though the characteristics ofthe ZrC coating layer were not reported

reten-The postirradiation heating tests of the TRISO-coated UO2 particles were performed at

ZrC-1873 K for 4500 h, at 2073 K for 3000 h, and at 2273 Kfor 100 h, to study the release behavior of the fissionproducts.51,52 For each heating test, 100 of the ZrC-TRISO-coated particles, which had been irradiated at

1173 K to 1.5% FIMA, were used The furnace in a hotcell was composed of a graphite heater, a graphitesample holder, graphite holder disks, and carbon insu-lators within a stainless steel vessel The coated fuelparticles were placed individually in the holes ofthe graphite disks Each heating test was divided intoseveral time steps At the end of each time step,the graphite components and the carbon insulatorswere replaced by new ones to measure the releasedmetallic fission products by g-ray spectrometry Thefission gas release monitoring during the tests andthe X-ray microradiography after the tests revealedthat no through-coating failure occurred in the tests.The measured fractional releases of 137Cs areshown in Figure 8 as a function of heating time.52The calculated fractional release of 137Cs from theordinary TRISO-coated particles at 1873 K is alsopresented in the figure for comparison This curvewas drawn based on the effective diffusion coefficient

of137Cs in the SiC coating layer61and the particledimensions The fractional release of137Cs was found

to be below 1 103even after heating at 1873 K for

4500 h or at 2073 K for 3000 h, whereas it was morethan 1 101after heating at 2273 K for 100 h Thesudden increase in the fractional release at 2273 K wasprobably attributed to the degradation of the ZrCcoating layer observed in the ceramography Thehigh cesium retention of the ZrC-TRISO-coatedfuel particles was confirmed to 2073 K

Based on a diffusion model, where a fuel kernelwith a single coating layer was assumed, the effectivediffusion coefficients for 137Cs in the ZrC coatinglayer, DCs(ZrC), were evaluated to be between

1 1018 and 5 1018m2s1 at 1873 K, andbetween 2 1018and 1 1017m2s1at 2073 K.52The present value forD (ZrC) at 1873 K was more