Comprehensive nuclear materials 2 15 uranium oxide and MOX production

Comprehensive nuclear materials 2 15 uranium oxide and MOX production Comprehensive nuclear materials 2 15 uranium oxide and MOX production Comprehensive nuclear materials 2 15 uranium oxide and MOX production Comprehensive nuclear materials 2 15 uranium oxide and MOX production Comprehensive nuclear materials 2 15 uranium oxide and MOX production

T Abe and K Asakura

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

ß 2012 Elsevier Ltd All rights reserved.

2.15.2.2 Nuclear Characteristics of Uranium and Plutonium Isotopes 398

393

ABWR Advanced boiling water reactor

ADU Ammonium diuranate

AGR Advanced gas cooled reactor

ATALANTE Atelier Alpha et Laboratoires

d’ Analyses des Transuraniens et

d’Etudes de retraitement, France

AUC Ammonium uranyl carbonate

AUPuC Ammonium uranyl plutonyl carbonate

BN Belgonucle´aire, Belgium

BNFL British Nuclear Fuels plc, United

Kingdom

BWR Boiling water reactor

CANDU CANadian Deuterium Uranium reactor

CFCa Complexe de Fabrication de

Cadarache, France

COCA Cobroyage (co-milling) Cadarache

COEX CO-EXtraction

DNB Departure from nucleate boiling

DOVITA Dry reprocessing, Oxide fuel,

Vibropac, Integral, Transmutation of

Actinides

FBR Fast breeder reactor

FR Fast reactor

HTR High-temperature reactor

HWR Heavy water reactor

IDR Integrated dry route

ITU Institute for Transuranium Elements,

Germany

JAEA Japan Atomic Energy Agency, Japan

LEFCA Laboratoire d’Etudes et de

MIMAS Micronized master blend

MOX Mixed oxide of uranium and

plutonium

O/M ratio Oxygen-to-metal ratio

OCOM Optimized CO-Milling PCI Pellet–cladding interaction PCMI Pellet–cladding mechanical

interaction PFFF Plutonium Fuel Fabrication Facility,

Japan PFPF Plutonium Fuel Production Facility,

Japan PVA Polyethylene glycol or polyvinyl alcohol PWR Pressurized water reactor

R&D Research and development RIAR Research Institute of Atomic

Reactors, Russia SBR Short binderless route SCK
CEN Studiecentrum voor Kernenergie –

Centre d’Etude de l’e´nergie Nucle´aire, Belgium SEM Scanning electron microscope tHM Tons of heavy metal

TIG Tungsten inert gas UKAEA United Kingdom Atomic Energy

Authority, United Kingdom VHTR Very high-temperature reactor VVER Vodo-Vodyanoi Energetichesky

Reaktor (Russian type PWR)

% TD Theoretical density ratio

Symbols

A Mass number

D Pu Diffusion coefficient of plutonium

s afast Fast neutron absorption cross-section

s athermal Thermal neutron absorption cross-section

s ffast Fast neutron fission cross-section

s fthermal Thermal neutron fission cross-section

Almost all the commercial nuclear power plantsoperating currently utilize uranium oxide fuel Thesereactors, sometimes referred to as Generation II orGeneration III reactors, produce15% of the world’s

electricity supply Production of the uranium oxide

fuel required for these reactors is a mature industry

and it annually requires more than 68 000 tU.1

Fuel design differs according to the reactor types,

which include the advanced gas cooled reactors

(AGRs), pressurized water reactors (PWRs), boiling

water reactors (BWRs), PWRs developed in the

for-mer Soviet Union (Vodo-Vodyanoi Energetichesky

Reaktor, VVERs), and CANadian Deuterium Uranium

(CANDU) reactors There are some differences in the

production processes to fit each fuel design

Plutonium utilization within the closed fuel cycle

is essential to utilize natural uranium resources

effi-ciently Plutonium recycling demonstrations have been

conducted in light water reactors (LWRs) and heavy

water reactors (HWRs).2Industrial utilization of MOX

in LWRs has commenced in some countries

The use of MOX in fast neutron reactors has many

attractive features Plutonium breeding in fast

breeder reactors (FBRs) leads to drastically increased

energy output from uranium resources Nuclide

transmutation by fast neutrons to incinerate minor

actinides (MAs) has the potential to reduce the

long-term radio-toxicity of spent nuclear fuel

Characteristics

Properties of Oxides

The starting material for oxide fuel production is oxide

powder It is fed to a powder preparation process and

then to a pelletizing process to get powder compacts,

which are called green pellets The green pellets

undergo a dewaxing and sintering process to get

sin-tered oxide pellets Certain characteristics of the oxide

powder and the sintered pellets are very important for

fuel production A brief summary of their important

characteristics is presented in this section As a

com-prehensive review of the characteristics of actinide

oxide has been given in Chapter 2.02,

Thermody-namic and Thermophysical Properties of the

Acti-nide Oxides, most of the data presented here are those

dealt with in Chapter 2.02, Thermodynamic and

Thermophysical Properties of the Actinide Oxides

2.15.2.1.1 Basic properties

2.15.2.1.1.1 Crystal structure

The phase diagrams and crystal structures of

uranium oxide and MOX have been described in

Sections 9.1.1, 9.1.2, and 9.1.3 These oxides exhibit

the fluorite or CaF2structure MOX is a substitutionalsolid solution in which U-cations of UO2, as MOXbase material, are substituted for Pu-cations There

is complete substitutional solid solubility betweenUO2 and PuO2 As mentioned in Section 9.1.2.7,phase separation into two fcc phases occurs in MOXwith a plutonium content exceeding 30% in thehypostoichiometric region

Uranium oxide can become a metric type oxide (UO2þx) at room temperaturewhile MOX can become both a hyperstoichiometrictype and a hypostoichiometric type (MO2x) oxide

hyperstoichio-at room temperhyperstoichio-ature This is because uranium canexist in an oxide as ions with valences of 4þ, 5þ, and6þ and plutonium can exist in an oxide as ions withvalences of 3þ and 4þ due to the oxygen potential

in the atmosphere Therefore, the oxygen-to-metal(O/M) ratio regions in which the single phaseMOX exists vary according to the plutonium con-tent of MOX

2.15.2.1.1.2 Oxygen potential

Oxygen potential is an important property for trolling certain properties related to oxide fuel fabrica-tion such as variations in density and O/M ratio

con-As mentioned in Section 9.1.4.3.2, the oxygenpotentials of uranium oxide and MOX increase with

an increase in temperature and plutonium content Inaddition, these potentials increase with an increase inO/M ratio and they increase rapidly, especially nearthe stoichiometric region (refer toFigures 22 and 23

in Section 9.1.4.3.2) In the case of (U, Gd)O2x,the oxygen potential increases with an increase in

of dry solids.8,9 The powder flowabilities of wave heating denitrated MOX (MH-MOX) powderand ammonium diuranate (ADU) powder have beenevaluated on the basis of Carr indices both beforeand after granulation.10,11

micro-2.15.2.1.2.2 Effective thermal conductivity

The temperature of MOX powder increases by self heat

generation of plutonium by a-decay when the powder

is kept in the fuel fabrication process In a MOX fuel

fabrication plant, the temperature increase in MOX

powder should be prevented because the excessive

temperature increase of MOX powder may possibly

cause changes in powder characteristics (e.g., O/M

ratio variation), degradation of additives (e.g., lubricant

agents), and overheating of equipment in the

fabri-cation process An example of a preventive measure

against the temperature increase of MOX powder is

the use of a storage vessel that has radiator plates

The effective thermal conductivity of MOX powder

is important for estimating its temperature distribution

The effective thermal conductivity of a powder can be

defined as the combination of thermal conductivities of

powder particles and the atmospheric gas because the

volume fraction of the atmosphere gas in the total

volume is large In addition, particle shapes, mean

particle size, specific surface area, and O/M ratio of

powder particles influence the effective thermal

con-ductivity of the powder.12Figure 1shows the effective

thermal conductivities of various MOX powders as

functions of O/M ratio and bulk density.12

2.15.2.1.3 Sintered oxide pellet

2.15.2.1.3.1 Sintering process

During the sintering process, MOX powder compacts

are subjected to high temperature for a few hours

under a controlled atmosphere to improve theirmechanical strength The powder compact is com-posed of individual grains separated by 35–50 vol.%porosity During sintering, the following majorchanges commonly occur: an increase in grain size,and changes in pore shape, pore size, and pore num-ber In the early stages of sintering, the powder par-ticles begin to mutually bond In the middle stage,grain growth, disappearance of pores, and formation

of closed pores occur The pellet densification ceeds according to the shape change from a pointcontact to a face contact between grains In thelast stage, disappearance of the closed pores occurs.The diffusion of uranium, plutonium, and oxygen,the evaporation–condensation process of their com-pounds, the grain growth process, the pore migrationprocess, and the pore disappearance processes areimportant for understanding the process of sintering

pro-To obtain pellets with high mechanical strength anddensity, it is desirable to eliminate as much porosity

as possible

Diffusion coefficients of these elements are neededfor evaluating the sintering behavior (e.g., volumeshrinkage in the fuel fabrication technology).Section9.1.6.1 shows that the oxygen self diffusion coeffi-cients of actinide oxides increase with increasingdeviation from stoichiometry near the stoichiometricregion and that the diffusion coefficients of cations

in hyperstoichiometric actinide oxides increase tically with deviation from stoichiometry It wasshown that the diffusion coefficient of plutonium in(U0.8 Pu0.2)O2xhas the lowest value near the stoi-chiometric region and it increases significantly with

dras-an increase in deviation from stoichiometry13 (see

Figure 2)

Vapor species of oxide fuel and its vapor pressureare required to assess the redistribution of elements,pore migration, and fuel restructuring The O/Mratio dependencies of vapor pressures in the vaporspecies of uranium oxide, plutonium oxide, andMOX are shown in Figures 26 and 27 of Section9.1.5 The vapor pressures of each of these specieshave a large dependency on the O/M ratio and theirbehavior is different in each vapor species

Temperatures used during dewaxing and sinteringare very important factors in the fabrication process.The Hu˝ttig and Tamman temperatures, which aredefined as the start temperatures for surface diffusionand volume diffusion of powder particles, respec-tively, are provided for establishing temperatures fordewaxing and sintering These temperatures can beeasily calculated using melting point temperature

Figure 1 Effective thermal conductivities of mixed oxide

of uranium and plutonium powders Reproduced from

Takeuchi, K.; Kato, M.; Sunaoshi, T.; Aono, S.; Kashimura,

M J Nucl Mater 2009, 385, 103–107.

2.15.2.1.3.2 Effects of O/M ratio on physical

properties of sintered oxide pellet

Most of the physical properties of oxide fuel such as

lattice parameter, diffusion coefficient, and thermal

conductivity are affected by the O/M ratio

The lattice parameter is needed for calculation of

the theoretical density (TD) ratio in the fuel

fabrica-tion process The thermal expansion coefficient, which

is defined as the temperature dependency of the

lattice parameter, is also an important thermophysical

property in fuel design when the variation in heat

transport between the fuel and the cladding tube by

thermal expansion of the fuel pellets and the stress to

the cladding tube by fuel pellets under irradiation are

evaluated

The lattice parameters and thermal expansion

coefficients of actinide dioxides are summarized

in Table 2 in Section 9.1.3.1 As mentioned in

Section 9.1.3.1.2, the dependency of the lattice

parameter of stoichiometric mixed oxides on their

chemical composition usually obeys Vegard’s law The

lattice parameter of MOX fuel decreases with an

increase in the plutonium content In the

hypostoi-chiometric region, the lattice parameter of MOX fuel

increases with a decrease in O/M ratio In addition,

Leyva et al.14

showed that the lattice parameter of

(U, Gd)O2decreases with an increase in Gd content

As mentioned in Section 9.1.3.1.2, Vegard’s law isapplied to the evaluation of lattice parameters as afunction of composition and temperature in manycases (refer to Figure 13 in Section 9.1.3.1.2)

It means that the thermal expansion coefficient ofMOX fuel is independent of plutonium content.Martin15 showed that the thermal expansion coeffi-cient of MOX fuel tends to increase with an increase

in deviation from stoichiometry in the metric region

hypostoichio-The melting point of oxide fuel is one ofthe most important thermophysical properties forfuel design and performance analyses As the chemi-cal composition and the O/M ratio of the oxide fuelchange the melting point of the fuel itself, fuel designand performance analysis should be done in consid-eration of not only the chemical composition atthe time of fuel fabrication but also its variationsubsequent to nuclear transmutation during reactoroperation In addition, the melting point is also used

in the estimation of sintering temperature, as tioned before

men-Section 9.1.2 shows that the melting point ofuranium oxide has its largest value near the stoichio-metric region and the melting point decreases with

an increase in deviation from stoichiometry (refer

toFigure 1inSection 9.1.2.1) Further, the meltingpoint of stoichiometric MOX decreases with anincrease in plutonium content (refer toFigure 7 inSection 9.1.2.7) In the hypostoichiometric MOX,the melting point of MOX fuel increases with adecrease in O/M ratio.16 Beals et al.17

studied theUO2–GdO1.5system at high temperatures and showedthat the melting point of Gd bearing UO2decreaseswith an increase in Gd content

During reactor operation, the heat generated inthe oxide fuel pellets flows from the central hightemperature region to the low temperature periphery

of the pellets, and consequently thermal equilibrium

is achieved in the pellets To evaluate the perature distribution when thermal equilibrium isreached, thermal conductivity is one of the mostimportant thermophysical properties As thermalconductivity is a function of O/M ratio, density,chemical composition, and so on, the variation inchemical composition that occurs during reactoroperation should be noted, along with the evaluation

tem-of the melting point, as mentioned before

As mentioned in Section 9.1.6.2, thermal ductivities of oxide fuel decrease with an increase

con-in temperature up to 1600–1800 K but con-increase with

an increase in temperature beyond this range

Figure 2 Dependence of Pu-diffusion coefficient, D Pu , in

(U 0.8 Pu 0.2 )O2xon oxygen partial pressure at 1773 K The

oxygen partial pressure was controlled using H 2 /H 2 O mixed

gas and CO/CO 2 mixed gas The high oxygen partial

pressures correspond to MO 2.07 , the low oxygen partial

pressures correspond to MO 1.92 Reproduced from Matzke,

H J J Nucl Mater 1983, 114, 121–135.

(refer to Figures 33 and 34 inSection 9.1.6.2) The

factors which heavily influence the thermal

conduc-tivity are O/M ratio and fuel density Thermal

con-ductivity decreases significantly with an increase in

deviation from stoichiometry and with a decrease in

density In addition, the thermal conductivity of a

gadolinium-bearing uranium oxide decreases

signifi-cantly with an increase in Gd content.18,19

2.15.2.1.3.3 Solubility in nitric acid solution

When the nuclear fuel cycle is considered, the

disso-lution of oxide fuel is the essential first step in

aqueous reprocessing The solubility and dissolutionrate of oxide fuel in nitric acid solution are importantparameters related to the capabilities of the reproces-sing process Generally, it has been supposed that thedissolution of MOX fuel decreases with an increase inthe plutonium content The maximum plutoniumcontent of MOX driver fuel for fast reactors hasbeen limited to about 30%, from the viewpoint ofsolubility in nitric acid solution

There have been many studies on the solubility ofoxide fuel in nitric acid solution.20–23 From theresults of these studies, it has been supposed thatthe factors affecting the dissolution rate of MOXare the fuel fabrication conditions (homogeneity ofthe admixture of UO2and PuO2, sintering conditionsand plutonium content, etc.) and the fuel dissolutionconditions (nitric acid concentration, solution tem-perature, dissolution time, etc.) (seeFigure 3)

2.15.2.2 Nuclear Characteristics ofUranium and Plutonium IsotopesPlutonium is an isotopically composition-variablematerial and the variation is attributable to its genera-tion reaction in LWR fuel, the initial uranium enrich-ment and burn-up of the LWR fuel, and so forth Itneeds various methodologies and much prudence in itshandling because its nuclear properties differ notice-ably from one isotope (nuclide) to another.Table 124,25

summarizes the principal nuclear properties of typicalnuclides in MOX fuel, including uranium isotopes

A material with high content of238Pu is more calorificowing to its decay mode (a) and short life Therefore,the content of238Pu would be the limiting factor forhandling batch sizes in a fabrication process 241Pu,which also has a short life, causes alteration in theisotopic composition even during a relatively shortperiod, for example, during storage after fuel

Pu: 0.5%, coprecipitated Pu: 5%, coprecipitated

Pu: 5%, mechanical blend Pu: 17.8%, mechanical blend

Pu: 20%, mechanical blend Pu: 35%, coprecipitated

Figure 3 Dissolution rate of mixed oxide of uranium and

plutonium with various Pu contents as a function of the

nitric acid concentration Reproduced from Oak Ridge

National Laboratory Dissolution of high-density UO 2 ,

PuO 2 , and UO 2 –PuO 2 pellets in inorganic acids,

ORNL-3695; Oak Ridge National Laboratory: Oak Ridge,

TN, 1965.

Table 1 Half lives and typical reaction cross sections of isotopes in MOX fuel

Nuclide Half life (year) Cross-section (barn, 1028m 2 ) Specific power from decay (W kg1)

s athermal s fthermal s afast s ffast235

fabrication but before loading into a reactor Besides

the above, neutron reaction cross-sections are

com-pletely different in isotopes and reactor types Taking

such variations in the cross-sections into consideration,

MOX fuel is prepared, in view of plutonium content,

to secure sufficient in-core reactivity.26

The nuclear characteristics of uranium and

pluto-nium are needed for the evaluation of radiation

exposure during the fuel fabrication process In

par-ticular, the short life of a nuclide merits attention with

regard to exposure to radiation All isotopes listed

in Table 1 are a-emitters, especially 238Pu, which

has highly significant a-radioactivity 241Am, which

is adjunct to241Pu, is also a strong a-emitter These

two nuclides also give off strong g-ray emissions

fol-lowing their a-decay The major sources of neutrons

are the even-A (mass number) plutonium isotopes

such as238Pu,240Pu, and242Pu because of their high

probability for spontaneous fission In addition,

espe-cially in oxide fuels such as MOX fuel, a-particle

bombardment of oxygen isotopes is an important

factor that determines neutron emission 238Pu and

241

Am have a higher specific (per unit mass) influence

on this reaction than other nuclides because of their

large a-ray emission rates, as mentioned above In

addition, these two nuclides have a somewhat higher

Q-value (a-ray energy) for decay and this

increas-ingly affects the neutron production rate

Turning to the topic of safeguards, the large

neu-tron yield by spontaneous fission from the MOX fuel

is utilized for a neutron coincidence counting method

for inventory verification This method uses the fact

that neutrons from spontaneous fission or induced

fission are essentially emitted simultaneously This

measurement can be made in the presence of

neu-trons from room background or (a, n) reactions

because these neutrons are noncoincident, or

ran-dom, in their arrival times The detection signals of

these neutrons are analyzed and plutonium isotopes

are determined by their quantity

Burnable poison suppresses initial fuel

reactiv-ity during fuel life and compensates fuel reactivreactiv-ity

with the gradual reduction in burnable poison with

burn-up Consequently, the fuel burn-up reactivity

is lowered and this lowered reactivity leads to an

extended operation cycle period Burnable poison

is often mixed into oxide fuel Gadolinium is a

typical one; it has a variety of stable and substable

isotopes and some of them (155,157Gd) have large

thermal capture cross-sections They are used in

the form of a sesqui-oxide compound, gadolinia,

in oxide fuel

2.15.3.1.1 Basic structural design

In LWRs and FBRs, a number of fuel rods are formedinto a fuel assembly The fuel rod is a barrier (con-tainment) for fission products; it has a circular cross-section that is suited for withstanding the primarypressure stress due to the external pressure of thecoolant and the increase in internal pressure by fissiongas release An axial stack of cylindrical fuel pellets isencased in a cladding tube, both ends of which arewelded shut with plugs A gas plenum is located atthe top part of the rod, in most cases, to form a freespace volume that can accommodate internal gas.Helium gas fills the free space at atmospheric pressure

or at a given pressure A hold-down spring, located inthe gas plenum, maintains the fuel stack in placeduring shipment and handling UO2insulator pelletsare inserted at both ends of the fuel stack, in somefuel designs, to thermally isolate metallic parts such asthe end plug and the hold-down spring

2.15.3.1.2 Fuel rods for LWRs

Table 2 summarizes LWR fuel rod design fications.30 LWR UO2 fuel rods contain denselow-enrichment UO2 pellets in a zirconium alloycladding; they are operated at a low linear heatrate with centerline temperatures normally below

speci-1400C The fuel pellets of the VVER have a smallcentral hole (1.2–1.4 mm in diameter)

Fission gas release is low under these conditionsand no large gas plenum is needed Burnable absorberfuel rods containing UO2–Gd2O3pellets are located

in some part of the fuel assemblies of LWRs toflatten reactivity change throughout the reactor oper-ation cycle

Great efforts have been made in LWR fuel roddesign in order to achieve the following good perfor-mance features: high burn-up, long operation cycle,good economy, and high reliability Toward achievingthese ends, many modifications have been made,such as the development of high-density UO2pellets,axial blankets for reducing neutron leakage, ZrB2integral burnable absorber, high Gd content UO2–Gd2O3pellets, corrosion-resistant cladding materials,and optimization of helium pressure and plenumlength in the rod designs

LWR MOX fuel rods contain MOX pellets thathave a low plutonium content As the plutoniumconcentration is low, their irradiation behavior issimilar to that of LWR UO2fuel rods No additional

problems are apparent, with the possible exception

of higher gas release and therefore an increase in

rod internal pressure at high burn-up Power

degra-dation with burn-up is less in the MOX fuel than

in UO2 fuel because of the neutronic properties of

the plutonium isotopes and thus MOX fuel is

irra-diated at higher power later in its life, releasing more

fission gases In addition, the slightly lower thermal

conductivity of MOX may give rise to higher fuel

temperatures, resulting in higher fission gas release

Design changes, such as lowering the helium filling

pressure, increasing the plenum volume, and/or

decreasing the fuel stack length in the rod, are

applied to accommodate higher gas release in MOX

fuel rods

2.15.3.1.3 Fuel rods for CANDU reactors

and AGRs

CANDU reactors and AGRs generally have fuel rod

design specifications similar to those of LWRs The

CANDU reactors use natural uranium oxide or slightly

enriched uranium oxide contained within a thin

Zircaloy clad, and design burn-up is lower than that of

LWRs In AGR fuel rods, uranium dioxide pellets,

enriched to about 3%, are encased in a stainless steel

clad Fuel bundles of both the reactors have circular,

cylindrical shapes to fit in the pressure tube of CANDU

reactors or in the graphite sleeve of AGRs The fuel rod

diameter differs according to the number of fuel rods

per bundle Typical CANDU fuel rod design

specifica-tions for a 28-rod bundle are presented inTable 2.30

The overall fuel rod lengths of both the reactor types

are much shorter than those of LWRs in order tofit their fuel assembly design which enables on-loadrefueling

2.15.3.1.4 Fuel rods for FBRsFBR fuel rods contain MOX pellets having highplutonium content, with the exception of RussianFBRs, BN-350, and BN-600 in which high enrich-ment UO2fuel pellets have been mostly used Fuelpellets of less than 8 mm diameter are encased in astainless steel cladding; they operate at a high linearheat rate with centerline temperatures of around

2000C or higher Under these conditions, fissiongas release is typically high (>80%) and a verylarge plenum is included to limit gas pressure Thegas plenum is located at the bottom of the rod insome fuel designs, aimed at minimizing plenumlength, thanks to the lower gas temperature at thebottom of the rod Upper and lower sections of thedepleted UO2 pellets are included for breeding.Pellet-smeared density is set not to exceed a crite-rion that is formulated as a function of burn-up

to avoid fuel–cladding mechanical interaction athigh burn-up; high-density annular pellets or low-density solid pellets are used; the former lower thefuel centerline temperature allowing a higher linearheat rate.31

2.15.3.1.5 Fissile content of oxide pelletsThe same U enrichment is used throughout a givenPWR fuel assembly, but the core usually containsseveral levels of enrichment arranged to give uniformpower distribution In contrast, BWR fuel rods have

Table 2 Summary of fuel rod design specifications for LWRs and CANDU reactors

a Partial length rod.

b Mitsubishi developed alloy.

Source: Tarlton, S., Ed Nucl Eng Int 2008, 53, 26–36.

several axial segments with different enrichments and

a BWR fuel assembly has several different rods with

different enrichments Thus, there are a variety of

UO2pellets with different U enrichments depending

on reactor design; the enrichments are within 5%

which is due to the limits of fuel fabrication facilities

and fuel shipments

For current LWR MOX fuels, depleted uranium

(0.2–0.3% 235U), which is obtained in the form of

tails from the enrichment process, is coupled with

plutonium because there are economic incentives

to concentrate as much plutonium in as few fuel

assemblies as possible as it conserves the expensive

fabrication cost of MOX fuel As the quality of

plutonium, from a neutronic aspect, varies with the

isotope composition of plutonium, the specification

of the plutonium content of LWR MOX fuel is

affected by the quality of plutonium Total plutonium

concentrations of 7.5% are considered to be

equiva-lent to U enrichments of 4.0–4.3% for the current

usual plutonium that is recycled from spent LWR

UO2fuel.2

To determine plutonium content of FBR MOX

fuel, equivalent239Pu (239Pu/(Uþ Pu)) is used The

actual plutonium content for a given batch is

obtained by a calculation that uses the neutronic

equivalent coefficient of each isotope and the isotope

composition of plutonium to be used for the batch

241

Am, a daughter product of241Pu, is considered in

the calculation as well The specification for

equiva-lent 239Pu (239Pu/(Uþ Pu)) is relatively low for a

large size core; equivalent 239Pu is 12–15% for the

SUPERPHENIX (1200 GWe),2814 –22% for MONJU

(280 GWe)

2.15.3.2.1 PWR UO2fuel assembly

Figure 432shows an example of a PWR fuel

bly PWRs have 197–230 mm square, ductless

assem-blies that traverse the full 2635–4550 mm height of

the core They comprise a basic support structure

of unfueled zirconium alloy guide tubes attached to

the top- and bottom-end fittings, an array of 14 14

to 18 18 fuel elements (minus the number of guide

tubes), and several axially spaced grids that hold

the array together About half of the assemblies have

rod control clusters attached at their upper end;

these consist of 18–24 slender stainless-steel-clad

absorber rods of AgInCd alloy or B4C, individually

located in the guide tubes The absorber rods are

withdrawn for startup and are repositioned after

refueling; the reactor is controlled at power by alteringthe concentration of an absorber (boric acid) in thecoolant The bottom-end fitting is located on the coregrid plate and the assembly is spring loaded against ahold-down system to compensate for differentialexpansion or growth during irradiation

Fine control is obtained by incorporating a able poison like Gd2O3 in some of the elements, inwhich it is admixed with UO2in the core region, andwith the upper and lower sections of natural UO2 Byminimizing power changes in this manner, the inci-dence of pellet–clad interaction (PCI) failures can bekept to very low, acceptable values Various improve-ments in fuel assembly design have been adopted Toimprove reliability, for instance, debris filtering wasadopted in the structural design of the bottom part ofthe fuel assembly, the grid structure design was mod-ified against fretting corrosion, and an intermediateflow mixer grid was added to enhance the margin

burn-to depart from nucleate boiling (DNB) Zirconiumalloy grids for better neutronics, optimized distribu-tion of fissile and fertile materials, and a burnablepoison to improve fuel cycle economy and to extendreactor cycle length were all introduced for econ-omy in the current assembly designs, as also theremovable top nozzle to reduce operation and main-tenance costs

Filter Bottom nozzle Figure 4 Example pressurized water reactor fuel assembly design of the 17  17 – 24 type with a fuel assembly averaged U enrichment of 3.9% Reproduced from http://www.mhi.co.jp/en/index.html

2.15.3.2.2 BWR UO2fuel assembly

Figure 530shows some examples of BWR fuel

assem-blies BWRs have 110–140 mm square full-core

height assemblies which, unlike their PWR

counter-parts, are contained within thick-walled channel

boxes of zirconium alloy They contain arrays of

6 6 to 10  10 fuel elements, usually with eight

elements acting as tie rods that screw into upper

and lower tie plates Some of the element

posi-tions are occupied by unfueled water-filled tubes

(called water rods) or water channels and are used

to control local flux peaking Element separation is

maintained by grid spacers that are attached to the

water rods and evenly distributed along the entire

length The square duct is attached to a top-end

fixture, relative to which the remainder of the

sub-assembly may slide The bottom-end fitting has a

mechanized orifice to control flow in the

subassem-bly and this is located in the core grid plate The

upper end fixture has a handle for loading and

unloading against which the hold-down bars rest to

prevent levitation

There are no absorber elements in BWR

assem-blies and reactor control is achieved by having

cruciform-shaped absorber blades throughout the

core which move vertically in the clearance between

sets of four subassemblies Power peaking is mized on the local scale by having fuel elements withdifferent enrichments and burnable poisons (gener-ally Gd2O3) dispersed within each assembly Variousfuel design improvements have been adopted, such as

mini-a debris-filtering structure for better relimini-ability, mized distribution of water channels, fissile materialwith partial length fuel rods and burnable poison use

opti-to improve fuel cycle economy and opti-to extend reacopti-torcycle length

2.15.3.2.3 VVER fuel assembly

Figure 630shows an example of a VVER fuel bly The VVER uses hexagonal fuel assemblies of3200–4690 mm length and 145–235 mm width Theassembly is used such that it is contained in a hexag-onal shroud, but shroudless assemblies are availablefor the VVER-1000.30

assem-2.15.3.2.4 CANDU reactor fuel

Figure 730 shows an example of a CANDU fuelbundle Twelve fuel bundles fit within each fuel chan-nel that is horizontally aligned in the reactor core

2.15.3.2.5 AGR fuelAGR fuel assemblies typically have 36 rodscontained within a graphite sleeve Twenty fuelassemblies are placed in a skip inside a flask

2.15.3.2.6 LWR MOX fuel assemblyPlutonium recycling has so far been limited to partialloading in LWR cores A primary design target of theMOX fuel assembly is compatibility with the UO2standard fuel assembly In the neutronic design forpartial loading of LWR cores, significant thermalneutron flux gradients at the interfaces between theMOX and UO2fuel assemblies have to be considered.The increase in thermal neutron flux in the direction

of an adjacent UO2assembly is addressed by a tion in the plutonium content of the MOX fuel rods

grada-at the edges and corners of the fuel assembly Thereare three typical rod types for PWR MOX fuelassemblies Optimized BWR fuel assemblies aremore heterogeneous: wider water gaps and largerwater structures within a BWR fuel assembly result

in MOX fuel assembly designs with an increase inthe number of different rod types Examples of MOXfuel assembly designs are shown inFigure 8.2Thereare plans for recycling weapons grade plutonium inPWRs in the United States.33

GNF GNF2

Figure 5 Example boiling water reactor fuel assemblies.

Reproduced from Tarlton, S., Ed Nucl Eng Int 2008, 53,

26–36.

The 100% MOX cores permit an increase in the

amount of plutonium under irradiation at a reduced

level of heterogeneity of the core An advanced

boil-ing water reactor (ABWR) to be constructed in

Ohma, Japan, will be the first plant with an in-built

100% MOX core capability

2.15.3.2.7 FBR fuel assembly

Figure 92shows an example of an FBR fuel assembly.FBR fuel assemblies have a hexagonal fuel rodarrangement with small gaps provided by a wirespacer, helically wound around each of the fuel pins

or by hexagonal grid spacers The fuel bundle is

Hold-down spring plunger

Removable top nozzle

Zircaloy shroud tube

Zircaloy spacer grid

Bottom nozzle

Inconel bottom grid

Debris filter flow plate

Water rod

Enriched

UO2 fuel pellets

Fuel rod

Removable top nozzle

Natural uranium axial blanket Zirc-4 cladding

Zirconium diboride integral fuel bundle absorber

Enriched

UO2 fuel pellets

Natural/

depleted/

ORP UO2axial blanket

UO2 +

Gd2O3neutron absorber, (in selected rods)

Low pressure drop Zirc-4 mid grid with mixing vanes

Plenum spring

Plenum spring

Figure 6 Example Westinghouse VVER-1000 fuel assembly Reproduced from Tarlton, S., Ed Nucl Eng Int 2008,

53, 26–36.

encased in a wrapper tube, in order to form a sodium

flow channel for efficient cooling and to prevent fuel

failure propagation during an accident

Austenitic or ferritic steels or nickel alloys are

selected as materials for structural components because

of their good compatibility with sodium and their

ability to cope with high temperatures and high levels

of fast neutron exposure These features of FBR fuel

assembly design result from the unique design

require-ments of the FBRs, including the hard neutron energy

spectrum, compact core size, high power density, high

burn-up, high temperature, and plutonium breeding

The fuel structure and actual fuel design vary with the

reactor scale, design targets, and the design

methodol-ogy Table 3 summarizes the fuel assembly design

specifications of the SUPERPHENIX, BN-600, and

MONJU.34

Uranium oxide has become the primary fuel for the

nuclear power industry today As of April 2010, there

are some 438 commercial nuclear power reactorsoperating in 30 countries, with a total capacity of

374 000 MWe.1 Most of these reactors are of theLWRs, AGRs, or the CANDU reactor types, andthey are fuelled with sintered pellets of UO2contain-ing natural or slightly enriched uranium

Prior to UO2pellet fabrication, the enriched uraniumfeed, UF6, is converted to UO2 powder Although anumber of conversion processes have been developed,only three are used on an industrial scale today Two

of these are wet processes: ADU and ammoniumuranyl carbonate (AUC) and the third is a dry process.The selected conversion process and its processparameters strongly influence the characteristics ofUO2powder and the resulting UO2pellets

2.15.4.1.1 ADU processThe ADU process has been widely used for manyyears It uses ADU as an intermediate product in

* 6 components

* 37 rods -Type w1 : 1 -Type w2 : 6 -Type w3 : 12 -Type w4 : 12 -Type w6 : 6

Bearing pad Sheath End plate

UO2 pellet Spacer pad End plug 3

1 2 3 4 5 6

6 4

w1 w2 w3 w4

w6

1 2 5

a two-step process First, UF6is vaporized and injected

into an ammonia solution UF6 hydrolyzes and

precipitates as ammonium diuranate (NH4)2U2O7

The ADU precipitate is collected on filters and dried

to get the ADU powder

þ 3H2OSecondly, the ADU powder is calcined and then

reduced to UO2with hydrogen

ðNH4Þ2U2O7þ 2H2! 2UO2þ 2NH3þ 3H2O

The properties of the resulting UO2are strongly

depen-dent on the processing parameters of precipitation,

calcinations, and reduction and equally on materialcontents, and reacting temperatures For example,the amount of NH3 is critical in the precipitationstep: too much will yield gelatinous ADU which isdifficult to filter; if there is too little then the result-ing UO2powder will be difficult to press and sinterinto pellets

to the following equation:

! ðNH4Þ4fUO2ðCO3Þ3g þ 6NH4F

Fuel rod, 3.7 wt% Pu Fuel rod, 5.2 wt% Pu Fuel rod, 8.2 wt% Pu Guide tube

Instrumentation tube

Water channel

2.5 wt% 235 U 7.0 wt% Pu

7.0 wt% Pu (part length) 8.2 wt% Pu 8.2 wt% Pu (part length) 3.95 wt% 235 U + 1.25 wt% Gd2O3

2.8 wt% Pu 3.8 wt% Pu 5.1 wt% Pu 5.1 wt% Pu (part length)

Figure 8 Example light water reactor mixed oxide of uranium and plutonium fuel assemblies The upper is pressurized water reactor design of the 17  17 – 24 type with a fuel assembly averaged plutonium concentration of 7.2% Pu The lower is boiling water reactor design of the 10  10 – 9Q type with a fuel assembly averaged plutonium concentration of 5.4 wt% Pu Reproduced from IAEA Status and Advances in MOX Fuel Technology; Technical Reports Series No 415; IAEA: Vienna, 2003.

The AUC precipitates in the form of yellow single

crystals The grain size depends on the precipitation

conditions Instead of UF6, uranyl nitrate solution can

also be used as a feed material

The AUC precipitate is filtrated and washed with

a solution of ammonium carbonate and methyl

alco-hol Then, the AUC powder is pneumatically

trans-ferred to a fluidized-bed furnace, decomposed, and

reduced to UO2 with hydrogen according to the

The resulting UO2 powder is made chemicallystable by a slight oxidation to about UO2.10

2.15.4.1.3 Dry process38The dry process was developed in the late 1960s and

is widely used today UF6is vaporized from steam orhot-water-heated vaporizing baths, and vaporized UF6

is introduced into the feed end of a rotating kiln.Here, it meets and reacts with superheated steam togive a plume of uranyl fluoride (UO2F2) UO2F2

Fuel assembly

Handling head

Fuel pin

Top end plug

Tag gas capsule Cladding

A –A cross-section

Upper spacer pad

Middle spacer pad

Wrapper tube

A

Fuel pin Wire spacer

Lower spacer pad

Entrance nozzle A

Figure 9 Example fast breeder reactor mixed oxide of uranium and plutonium fuel assembly design of MONJU.

Reproduced from IAEA Status and Advances in MOX Fuel Technology; Technical Reports Series No 415; IAEA:

Vienna, 2003.

passes down the kiln where it meets with a

counter-current flow of steam and hydrogen and is converted

to UO2 powder The reaction sequence follows the

equations below

4UO2F2þ 2H2Oþ 2H2! U3O8þ UO2þ 8HF

U3O8þ 2H2! 3UO2þ 2H2O

The UO2powder resulting from dry processes is of

low bulk density and fine particle size Therefore,

granulation before pressing and the employment of a

pore former process are usual during the pellet

fabri-cation process

A dry process has preferable advantages: the

pro-cess is simple and the equipment is compact; the

criticality limitation is less required; and liquid

waste treatment is not necessary

2.15.4.2 UO2Pellet Production

The flow sheet for UO2pellet production is shown in

Figure 10 The UO2pellet fabrication process

con-sists of mixing the UO2 powder with additives such

as binder, lubricant and pore former materials,

gran-ulating to form free-flowing particles, compaction in

an automatic press, heating to remove the additives,

sintering in a controlled atmosphere, and grinding to

a final diameter The process varies slightly according

to the nature of the starting UO2powder

2.15.4.2.1 Powder preparation

In the pelletizing process, UO2powder must be filled

easily and consistently into dies UO2 powder from

the AUC process is free-flowing and can be pressed

without granulation Usually it is mixed with a smallamount of U3O8 to control the density and poredistribution of the pellets The fine particle size ofthe integrated dry route (IDR) powders preventsthem from being free-flowing when produced; thesepowders are therefore prepressed into briquettes,fractured, sieved to produce granules, and a dry

Table 3 Summary of fuel assembly design data of SUPERPHENIX, BN-600 and MONJU

Source: IAEA Fast Reactor Database 2006 Update, IAEA-TECDOC-1531; IAEA: Vienna, Austria, 2006.