Mechanical properties at sub-microscale and macroscale of polycrystalline uranium mononitride

The indentation hardness, Vickers hardness, fracture toughness, and Young’s modulus of polycrystalline uranium mononitride (UN) at sub-microscale and macroscale were evaluated by an indentation test, Vickers hardness test, and the ultrasonic pulse echo method. The Young modulus and Vickers hardness were in good agreement with the literature values. The fracture toughness of UN was about three times that of UO₂. In addition, we revealed the indentation size effect on the indentation hardness of UN.     

Microstructural disorder and the mechanical properties of carbon fibres

The effect of fast neutron irradiation on the mechanical and microstructural properties of carbon fibres heat-treated to various stages of graphitization has been investigated. Changes in selected crystallinity parameters were used to monitor the microstructural disorder induced in each fibre type by three different irradiation doses. Associated changes in fracture strength and Young's modulus are related to the dependence of dislocation mobility on short-range microstructural damage.     

Physical properties of europium sesquioxide

Europium sesquioxide (Eu₂O₃) is a neutron-absorbing material of potential use in reactor control rods and is being evaluated for use in fast reactors. This paper presents the results of physical and mechanical property measurements performed on unirradiated europia. The material exists in two useful crystallographic forms. Both the monoclinic form and a cubic variety, stabilized by the addition of 17 wt.% molybdenum trioxide (MoO₃), have been examined. The properties reported are density, specific heat, thermal diffusivity and conductivity, thermal expansivity, Young's modulus, and strength. The data are compared with similar information in the literature.     

Effect of irradiation on the mechanical properties of isotropic pyrolytic carbons

The effect of fast-neutron irradiation on the mechanical properties of a wide range of isotropic pyrolytic carbons deposited in a bed of fluidized particles has been studied in three-point bending. For poorly crystalline isotropic carbons, large increases in the Young's modulus with increasing fast-neutron fluence were observed at irradiation temperatures below about 1200 °C. At 1400 and 1600 °C, increases in the modulus also occurred, but the increases were not as large as at the lower temperatures. Irradiation had little effect on the fracture stresses of these carbons. An as-deposited crystalline carbon irradiated along with the poorly crystalline carbons had smaller increases in the Young's modulus. The mechanical properties are discussed in terms of structural changes of the carbons during irradiation.     

Irradiation damage in reaction-bonded silicon carbide

Reaction-bonded SiC loses nearly 50% of its fracture strength when exposed to neutron irradiation. Young's modulus also decreases rapidly. The damage occurs soon after exposure and levels out with continuing exposure up to $\text{3 × 10}{}^{21}\text{n/cm}{}^2$. Both α- and β-SiC undergo a nearly Isotropic expansion that saturates with increasing irradiation. Silicon undergoes a very small expansion and mechanical property degradation. The strength reduction in SiC is explained in terms of differential lattice expansions between SiC and Si which result in misfit strains and subsequent crack growth and critical flaw size extension.     

Effect of Nd and Pr addition on the thermal and mechanical properties of (U, Ce)O2

The thermal conductivity, Young’s modulus, and hardness of (U_0.65−xCe_0.3Pr_0.05Nd_x)O₂ (x = 0.01, 0.08, 0.12) were evaluated and the effect of Pr and Nd addition on the properties of (U, Ce)O₂ were studied. The polycrystalline high-density pellets were prepared with solid state reactions of UO₂, CeO₂, Pr₂O₃, and Nd₂O₃. We confirmed that all Ce, Pr, and Nd dissolved in UO₂ and formed solid solutions of (U, Ce, Pr, Nd)O₂. We revealed that the thermal conductivity of (U_0.65−xCe_0.3Pr_0.05Nd_x)O₂ (x = 0.12) was up to 25% lower than that of x = 0.01 at room temperature. The Young’s modulus of (U_0.65−xCe_0.3Pr_0.05Nd_x)O₂ decreased with x, whereas the hardness values were constant in the investigated x range.     

Prediction of minimum UO2 particle size based on thermal stress initiated fracture model

An analytic study was employed to determine the minimum UO₂ particle size that could survive fragmentation induced by thermal stresses in a UO₂---Na fuel-coolant interaction (FCI), based on a brittle fracture mechanics approach. Solid and liquid UO₂ droplets were considered, with perfect wetting by the sodium or finite heat transfer coefficient. The analysis indicated that particles below the range of 50 μm in radius could survive an FCI under the most severe temperature conditions without thermal stress fragmentation, and seemed to verify the experimental observations as to the range of the minimum particle size due to thermal stress fragmentation by FCI. The basic complexities in fracture mechanics make further investigation in this area interesting but not necessarily fruitful for the immediate future.     

Experimental studies of mechanical properties of solid zirconium hydrides

Zirconium alloy Zr-2.5Nb has been hydrided to ZrH_x (x = 1.15-2.0), and studied using microhardness and unconfined and confined compression techniques. At room temperature, results on Young’s modulus and yield strength of solid hydrides show that these mechanical properties remain about the same as the original zirconium alloy for hydrogen compositions up to about ZrH_1.5. The levels of these properties start to drop when δ hydride becomes the major phase and reaches a minimum for the ε hydride phase. Between room temperature and 300 °C, Young’s modulus of solid hydrides decreases with temperature at about the same rate as it does for the original zirconium alloy.     

Effect of Temperature and Deformation Rate on Fracture Strength of Sintered Uranium Dioxide

The fracture strength of two kinds of UO₂ specimens possessing pores of different maximum sizes (60 and 140 μm) was measured in the range of room temperature ? 1,300°C by means of diametral compression testing. The fracture strength thus obtained proved to be smaller than any of the values reported by previous authors who mainly used bending tests. Finite element analysis showed that the method used in the present study should logically yield results close to the true tensile fracture strength. The descrepancies noted with the results reported from the other studies were attributable to the differences in the methods used for the measurements.

The fracture strength was found to remain almost constant in the relatively low temperature region (R.T.–800°C) beyond which the value increased with temperature (intermediate temperature region of 1,000–1,300°C). Electron-microscopic observations of the fractured surface indicated that the brittle-to-ductile transition temperature (T_c ) was situated between 800 and 1,000°C when the strain was applied slowly. Raising the strain rate proved to affect both fracture strength and T_c . These dependences of temperature and strain rate on the fracture strength are explained from the relation between dislocation velocity and deformation rate. Griffith's theory is cited to describe the relation between the largest pore size and fracture strength.     

Mechanical properties and XRD studies of silicon carbide inert matrix fuel fabricated by a low temperature polymer precursor route

The mechanical properties of silicon carbide (SiC) inert matrix fuel (IMF) pellets fabricated by a low temperature (1050 °C) polymer precursor route were evaluated at room temperature. The Vickers hardness was mainly related to the chemical bonding strength between the amorphous SiC phase and the β-SiC particles. The biaxial fracture strength with pre-notch and fracture toughness were found to be mostly controlled by the pellet density. The maximum Vickers hardness, biaxial fracture strength with pre-notch and fracture toughness achieved were 5.6 GPa, 201 MPa and 2.9 MPa m^1/2 respectively. These values appear to be superior to the reference MOX or UO₂ fuels. Excellent thermal shock resistance for the fabricated SiC IMF was proven and the values were compared to conventional UO₂ pellets. XRD studies showed that ceria (PuO₂ surrogate) chemically reacted with the polymer precursor during sintering, forming cerium oxysilicate. Whether PuO₂ will chemically react in a similar manner remains unclear.     

Investigation on high temperature vapor pressure of UO2 containing simulated fission-product elements

During the hypothetical core disruptive accident (HCDA) of a fast breeder reactor (FBR), the temperature of the fuel would rise above 3000 K. The experimental data concerning the saturated fuel vapor pressure are necessary for the analysis of the HCDA. In this study, the UO₂ containing Cs, Ba, Ag, or Sn was used to simulate the irradiated fuel in the FBR.The saturated vapor pressure of pure UO₂ and UO₂ containing Cs, Ba, Ag, or Sn at 3000 to 5000 K was measured dynamically with a pulse laser and a torsion pendulum. The surface of a specimen on the pendulum was heated to eject vapor by the injection of a giant pulse ruby laser beam. The pressure of the ejected vapor was measured by both the maximum rotation angle of the pendulum and the duration of vapor ejection. The saturated vapor pressure was theoretically calculated by using the ejected vapor pressure. The surface temperature of the specimen was estimated from the irradiated energy density measured with a laser energy meter.The saturated vapor pressure of UO₂ at 3640 to 5880 K measured in this study was near the extrapolated value of Ackermann's low temperature data. The vapor pressure of UO₂ containing Cs, Ba, Ag or Sn was higher than that of UO₂. The saturated vapor pressure of UO₂ and a solid fission products system was calculated by using these experimental data.     

Crushing behaviour of high temperature reactor coated fuel particles

This paper examines in detail the crushing behaviour of high-temperature reactor fuel particles with pyrolytic carbon or silicon carbide outer coatings and discusses their failure mechanisms, in an attempt to relate crushing failure loads to coating strengths, and provide a simple, quick testing technique for quality control or performance assessment. Failure occurs by a series of mechanisms, in varying sequence, initiated by Hertzian cracking. Because the first event detected in a crushing test load/deflection curve is not the formation but the propagation of the Hertzian crack, the crushing load cannot be related to the coating strength; instead, it is governed by the fracture surface energy of the outer coating. A crushing test is therefore not a suitable technique for measurement of the strength of particle coatings. However, through measurement of the size of the contact surface, reliable estimates of the Young's modulus of the outer coating can be made by application of the Hertz theory of contact.     

Thermophysical properties of (U,Y)O2

We prepared polycrystalline pellets of (U,Y)O₂, containing YO_1.5 up to 11 mol.%. We performed indentation tests on the pellets, and evaluated the Young’s modulus and hardness. We measured the heat capacity and the thermal diffusivity, and evaluated the thermal conductivity. We succeeded in evaluating the effect of Y content on the thermophysical properties of (U,Y)O₂. We revealed that the Young’s modulus, hardness, and thermal conductivity of (U,Y)O₂ decreased with increasing the Y content.     

Thermal stress initiated fracture as a fragmentation mechanism in the UO2-sodium fuel-coolant interaction

An analytic study was carried out to determine the applicability of the concept of thermal stress fragmentation to the UO₂-Na fuel-coolant interaction. Major emphasis was put on the fracture mechanics approach to assess whether or not the solidifying UO₂ would fracture under the thermally-induced stresses. It was found that the stress levels were sufficient to generate K_I values substantially in excess of the UO₂ fracture toughness K_IC. Thus, rapid instantaneous propagation of inherent flaws is anticipated.     

Thermal and mechanical properties of (U,Er)O2

Erbium is considered as a slow burnable poison suitable for use in light water reactors (LWRs). Addition of a small amount of Er₂O₃ to all UO₂ pellets will make it possible to develop super high burnup fuels in Japanese nuclear facilities which are now under the restriction of the upper limit of ²³⁵U enrichment. When utilizing the (U,Er)O₂ fuels, it is very important to understand the thermal and mechanical properties. Here we show the characterization results of (U_1−xEr_x)O₂ (0 ? x ? 0.1). We measured their thermal and mechanical properties and investigated the effect of Er addition on these properties of (U,Er)O₂. All Er completely dissolved in UO₂, and the lattice parameter decreased linearly with the Er content. Both the thermal conductivity and Young’s modulus of (U,Er)O₂ decreased with the Er content. These results would be useful for us in evaluating the performance of the (U,Er)O₂ fuels in LWRs.     

Change in electrical resistivity of nuclear graphite during compression tests and a model for its deformation and fracture mechanism

The results of the resistivity changes during compression of some nuclear graphites are summarized in order to cast light on the fracture mechanism of the materials; data on pyrolytic graphite and amorphous carbon are also taken into account. It is found that all the graphites investigated show an abrupt increase in resistivity when the applied stress increases to about a half of the fracture stress. Above this stress the non-linearity of the stress-strain curve becomes more pronounced and the formation and growth of optically resolvable cracks occur. A model based on the deformation of cracks and pores on the basal plane is proposed for explaining the change in resistivity, and is supported by measurements of the effect of pre-stressing on the Young's modulus, thermal expansion, mercury porosimetry and Knoop micro-hardness of the material.     

Break up time of fragmentating-solidifying UO2 spheres when quenched in sodium

When molten UO₂ is quenched in sodium, a sand-like debris results containing about 80% of fractured particles and 20% of smooth particles and spheres. The production of the fractured particles is normally explained by the thermal stress fragmentation model. Previously brittle fracture mechanics was applied to the complete solid shell of a freezing UO₂ drop, i.e. where 954°C < T < 2850°C; a calculation of fragmentation time was not possible. In this contribution the solid shell is continuously subdivided in a plastic or ductile layer for 1300°C < T < 2850°C and a brittle one for 945°C < T < 1300°C. Cracking occurs in the brittle layer only. In the present model a layer of a predescribed depth is assumed to ablate instantaneously, when the temperature reaches the transition point of elastic of ductile behavior (T = 1300°C) at its inner boundary. A new layer is formed within a time step, governed by the heat conduction equation. The discontinuous ablation process is thus related to the continuous progression of the solidification front. A calculation of the fragmentation time is possible: in principle it comprehends the summation of a large number of time steps for the formation of brittle layers. The thickness of the cracked brittle layer is parametrized to 20, 10, 5 and 1 μm. The concept of instantaneous ablation was suggested by the experience that the violent boiling forces of sodium are very effective on the UO₂ surface. The introduction of these minor changes makes the thermal stress model more realistic, because it can explain now, why UO₂ does not fragment in argon and water. The fragmentation time assessed for a UO₂ drop of 7.2 mm diameter in sodium, brittle layer 10 μm, is 250 ms.     

Elastic constants of zirconium alloys

The dynamic elastic moduli of Zircaloy-2, Zr-1.15 wt% Ci-0.1 wt% Fe and Zr-2.5 wt% Nb have been determined over the temperature range 293–773 K. Young's modulus and shear modulus decreased linearly with increasing temperature. Poisson's ratio decreased with increasing temperature for Zircaloy-2 and Zr-2.5 wt% Nb but increased for Zr-1.15 wt% Cr-0.1 wt% Fe. The results have been compared with previous values determined both by static and dynamic techniques and with polycrystalline constants computed from single crystal elastic constants. The difference in behaviour between the three alloys is not due to differences in alloy composition but to texture effects. A relationship is derived between texture and elastic constants.     

Low cost, proliferation resistant, uranium–thorium dioxide fuels for light water reactors

Our objective is to develop a fuel for the existing light water reactors (LWRs) that, (a) is less expensive to fabricate than the current uranium-dioxide (UO₂) fuel; (b) allows longer refueling cycles and higher sustainable plant capacity factors; (c) is very resistant to nuclear weapon-material proliferation; (d) results in a more stable and insoluble waste form; and (e) generates less high level waste. This paper presents the results of our initial investigation of a LWR fuel consisting of mixed thorium dioxide and uranium dioxide (ThO₂–UO₂). Our calculations using the SCALE 4.4 and MOCUP code systems indicate that the mixed ThO₂–UO₂ fuel, with about 6 wt.% of the total heavy metal U-235, could be burned to 72 MW day kg⁻¹ (megawatt thermal days per kilogram) using 30 wt.% UO₂ and the balance ThO₂. The ThO₂–UO₂ cores can also be burned to about 87 MW day kg⁻¹ using 35 wt.% UO₂ and 65% ThO₂with an initial enrichment of about 7 wt.% of the total heavy metal fissile material. Economic analyses indicate that the ThO₂–UO₂ fuel will require less separative work and less total heavy metal (thorium and uranium) feedstock. At reasonable future costs for raw materials and separative work, the cost of the ThO₂–UO₂ fuel is about 9% less than uranium fuel burned to 72 MW day kg⁻¹. Because ThO₂–UO₂ fuel will operate somewhat cooler, and retain within the fuel more of the fission products, especially the gasses, ThO₂–UO₂ fuel can probably be operated successfully to higher burnups than UO₂ fuel. This will allow for longer refueling cycles and better plant capacity factors. The uranium in our calculations remained below 20 wt.% total fissile fraction throughout the cycle, making it unusable for weapons. Total plutonium production per MW day was a factor of 3.2 less in the ThO₂–UO₂ fuel than in the conventional UO₂ fuel burned to 45 MW day kg⁻¹. Pu-239 production per MW day was a factor of about 4 less in the ThO₂–UO₂ fuel than in the conventional fuel. The plutonium produced was high in Pu-238, leading to a decay heat about three times greater than that from plutonium derived from conventional fuel burned to 45 MW day kg⁻¹ and 20 times greater than weapons grade plutonium. This will make fabrication of a weapon more difficult. Spontaneous neutron production from the plutonium in the ThO₂–UO₂ fuel was about 50% greater than that from conventional fuel and ten times greater than that from weapons grade plutonium. High spontaneous neutron production drastically limits the probable yield of a crude weapon. Because ThO₂ is the highest oxide of thorium while UO₂ can be oxidized further to U₃O₈ or UO₃, ThO₂–UO₂ fuel appears to be a superior waste form if the spent fuel is to be exposed ever to air or oxygenated water. And, finally, use of higher burnup fuel will result in proportionally fewer spent fuel bundles to handle, store, ship, and permanently dispose of.