Fabrication of uniform ZrC coating layer for the coated fuel particle of the very high temperature reactor

Fuel for the very high temperature reactor is required to be used under severer irradiation conditions and higher operational reactor temperatures than those of present high temperature gas cooled reactors. Japan Atomic Energy Agency has developed zirconium carbide (ZrC)-coated fuel particles previously in laboratory scale which are expected to maintain their integrity at higher temperatures and burnup conditions than conventional silicon carbide-coated fuel particles. As one of the important R&D items, ZrC coating process development has been started in the year 2004 to determine the coating conditions to fabricate uniform structure of ZrC layers by using a new large-scale coater up to 0.2 kg batch. It was thought that excess carbon formed in the ZrC layer under the oscillation of coating temperature would cause non-uniformity of the ZrC layer. Finally, uniform ZrC coating layer has been fabricated successfully by adjusting the time constant of the coater and keeping the coating temperature at around 1400 °C.     

Design of a fuel element for a lead-cooled fast reactor

The options of a lead-cooled fast reactor (LFR) of the fourth generation (GEN-IV) reactor with the electric power of 600 MW are investigated in the ELSY Project. The fuel selection, design and optimization are important steps of the project. Three types of fuel are considered as candidates: highly enriched Pu-U mixed oxide (MOX) fuel for the first core, the MOX containing between 2.5% and 5.0% of the minor actinides (MA) for next core and Pu-U-MA nitride fuel as an advanced option. Reference fuel rods with claddings made of T91 ferrite-martensitic steel and two alternative fuel assembly designs (one uses a closed hexagonal wrapper and the other is an open square variant without wrapper) have been assessed. This study focuses on the core variant with the closed hexagonal fuel assemblies. Based on the neutronic parameters provided by Monte-Carlo modeling with MCNP5 and ALEPH codes, simulations have been carried out to assess the long-term thermal-mechanical behaviour of the hottest fuel rods. A modified version of the fuel performance code FEMAXI-SCK-1, adapted for fast neutron spectrum, new fuels, cladding materials and coolant, was utilized for these calculations. The obtained results show that the fuel rods can withstand more than four effective full power years under the normal operation conditions without pellet-cladding mechanical interaction (PCMI). In a variant with solid fuel pellets, a mild PCMI can appear during the fifth year, however, it remains at an acceptable level up to the end of operation when the peak fuel pellet burnup ∼80 MW d kg⁻¹ of heavy metal (HM) and the maximum clad damage of about 82 displacements per atom (dpa) are reached. Annular pellets permit to delay PCMI for about 1 year. Based on the results of this simulation, further steps are envisioned for the optimization of the fuel rod design, aiming at achieving the fuel burnup of 100 MW d kg⁻¹ of HM.     

Irradiation testing of matrix material for spherical HTR-10 fuel elements

In order to evaluate if the fuel elements and their components (matrix material and coated fuel particles) can meet the design requirements of 10 MW high temperature gas-cooled reactor (HTR-10), an irradiation testing with four spherical fuel elements, 60 matrix material specimens of 5 mm × 5 mm × 40 mm and 13,500 coated fuel particles was performed in Russian IVV-2M reactor from July 2000 to February 2003. The irradiation temperature was 1000 °C. The fast neutron fluence of matrix material specimens reached 1.3 × 10²¹ cm⁻² (E > 0.1 MeV). Post-irradiation examination contained the visual inspection, dimension measurement and determining the density, porosity, specific electrical resistance and bending strength. The irradiation results are given in this paper, and show that the matrix material for spherical HTR-10 fuel elements made from the domestic raw materials and fabricated by the quasi-isostatic room-temperature moulding process is suitable as a structural material for spherical HTR fuel elements.     

Irradiation induced dislocation loop and its influence on the hardening behavior of Fe-Cr alloys by an Fe ion irradiation

Nano indentation analysis and transmission electron microscopy observation were performed to investigate a microstructural evolution and its influence on the hardening behavior in Fe-Cr alloys after an irradiation with 8 MeV Fe⁴⁺ ions at room temperature. Nano indentation analysis shows that an irradiation induced hardening is generated more considerably in the Fe-15Cr alloy than in the Fe-5Cr alloy by the ion irradiation. TEM observation reveals a significant population of the a₀<1 0 0> dislocation loops in the Fe-15Cr alloy and an agglomeration of the 1/2a₀<1 1 1> dislocation loops in the Fe-5Cr alloy. The results indicate that the a₀<1 0 0> dislocation loops will act as stronger obstacles to a dislocation motion than 1/2a₀<1 1 1> dislocation loops.     

Fuel elements of the high temperature pebble bed reactor

This work is devoted to spherical fuel elements for the high temperature pebble bed reactor, their manufacture and the conditions which they must satisfy for use in a process-heat reactor with an average gas outlet temperature T_{G, out} of 950°C. The positive results known from the operation of the AVR with T_G,out = 950°C and from extensive irradiation tests of the THTR-300 element with BISO coated mixed-oxide particles, even beyond the range of design specifications, and possible damage mechanisms are described in detail. They show that a spherical fuel element already exists, for which only a short-term development is needed to produce a coolant temperature of 950°C in a process-heat reactor. Further developments will be characterized by the use of a pebble bed HTR for high conversion rates (c ≈ 0.95) or for average gas outlet temperatures of more than 950°C. At higher temperatures the increased demands, mainly with regard to the release of fission products, can be fulfilled through the application of TRISO-coated fuel particles and the doping of the fuel kernels with . The reprocessing programme for fuel elements in the Federal Republic of Germany is mentioned briefly.     

Tomographic atom probe characterization of the microstructure of a cold worked 316 austenitic stainless steel after neutron irradiation

For the first time, chemical analyses using Atom Probe Tomography were performed on a bolt made of cold worked 316 austenitic stainless steel, extracted from the internal structures of a pressurized water reactor after 17 years of reactor service. The irradiation temperature of these samples was 633 K and the irradiation dose was estimated to 12 dpa (7.81 × 10²⁵ neutrons.m⁻², E > 1 MeV). The samples were analysed with a laser assisted tomographic atom probe. These analyses have shown that neutron irradiation has a strong effect on the intragranular distribution of solute atoms. A high number density (6 × 10²³ m⁻³) of Ni-Si enriched and Cr-Fe depleted clusters was detected after irradiation. Mo and P segregations at the interfaces of these clusters were also observed. Finally, Si enriched atmospheres were seen.     

Selection of water-dispersible carbon black for fabrication of uranium oxicarbide microspheres

Fabrication of uranium oxicarbide microspheres, a component of TRISO fuel particles for high temperature nuclear power systems, is based on the internal gelation of uranium salts in the presence of carbon black. In order to obtain a high quality product, carbon black should remain dispersed during all phases of the gelation process. In this study, the surface and structural properties of several commercial carbon black materials, and the use of dispersing agents was examined with the goal of finding optimal conditions for stabilizing submicron-sized carbon black dispersions. Traditional methods for stabilizing dispersions, based on the use of dispersing agents, failed to stabilize carbon dispersions against large pH variations, typical for the internal gelation process. An alternate dispersing method was proposed, based on using surface-modified carbons functionalized with strongly ionized surface groups (sodium sulfonate). With a proper choice of surface modifiers, these advanced carbons disperse easily to particles in the range of 0.15-0.20 μm and the dispersions remain stable during the conditions of internal gelation.     

In-pile Xe diffusion coefficient in UO2 determined from the modeling of intragranular bubble growth and destruction under irradiation

Intragranular bubbles grow in the nuclear fuel by diffusion and precipitation of fission gases, mainly xenon; and are ultimately destroyed, under irradiation, by fission fragments. This article will attempt to determine the in-pile bubble distributions taking into account the evolution of the concentration profile around a bubble during its growth and the destruction process by fission fragments. From these distributions a relation between the bubble mean radius and the diffusion coefficient of xenon can be established, allowing the determination, from experimental measurements of intragranular bubble sizes, of the in-pile Xe diffusion coefficient in UO₂. The estimated activation energy (0.9 eV) is about one order of magnitude lower than the widely used value of 3.9 eV determined from out-of-pile experiments. This effect can be attributed to the presence of point defects created by the irradiation.     

Optical anisotropy measurements of TRISO nuclear fuel particle cross-sections: The method

The analysis of two-modulator generalized ellipsometry microscope (2-MGEM) data to extract information on the optical anisotropy of coated particle fuel layers is discussed. Using a high resolution modification to the 2-MGEM, it is possible to obtain generalized ellipsometry images of coating layer cross-sections with a pixel size of 2.5 μm and an optical resolution of ∼4 μm. The most important parameter that can be extracted from these ellipsometry images is the diattenuation, which can be directly related to the optical anisotropy factor (OAF or OPTAF) used in previous characterization studies of tristructural isotropic (TRISO) coated particles. Because high resolution images can be obtained, the data for each coating layer contains >6000 points, allowing considerable statistical analysis. This analysis has revealed that the diattenuation of the inner pyrocarbon (IPyC) and outer pyrocarbon (OPyC) coatings varies significantly throughout the layer. The 2-MGEM data can also be used to determine the principal axis angle of the pyrocarbon layers, which is nearly perpendicular to the TRISO radius (i.e., growth direction) and corresponds to the average orientation of the graphene planes.     

Formation of dislocation loops in silicon by ion irradiation for silicon light emitting diodes

We have studied the influence of the ion species, ion energy, fluence, irradiation temperature and post-implantation annealing on the formation of shallow dislocation loops in silicon, for fabrication of silicon light emitting diodes. The substrates used were (1 0 0) Si, implanted with 20-80 keV boron at room temperature and 75-175 keV silicon at 100 and 200 °C. The implanted fluences were from 5 × 10¹⁴ to 1 × 10¹⁵ ions/cm². After irradiation the samples were processed for 15 s to 20 min at 950 °C by rapid thermal annealing. Structural analysis of the samples was done by transmission electron microscopy and Rutherford backscattering spectrometry. In all irradiations the silicon substrates were not amorphized, and that resulted in the formation of extrinsic perfect and faulted dislocation loops with Burgers vectors a/2〈1 1 0〉 and a/3〈1 1 1〉, respectively, sitting in {1 1 1} habit planes. It was demonstrated that by varying the ion implantation parameters and post-irradiation annealing, it is possible to form various shapes, concentration and distribution of dislocation loops in silicon.     

Computer simulation of primary damage creation in displacement cascades in copper. I. Defect creation and cluster statistics

Atomic-scale computer simulation has been used to investigate the primary damage created by displacement cascades in copper over a wide range of temperature (100 K ? T ? 900 K) and primary knock-on atom energy (5 keV ? E_PKA ? 25 keV). A technique was introduced to improve computational efficiency and at least 20 cascades for each (E_PKA, T) pair were simulated in order to provide statistical reliability of the results. The total of almost 450 simulated cascades is the largest yet reported for this metal. The mean number of surviving point defects per cascade is only 15-20% of the NRT model value. It decreases with increasing T at fixed E_PKA and is proportional to (E_PKA)^1.1 at fixed T. A high proportion (60-80%) of self-interstitial atoms (SIAs) form clusters during the cascade process. The proportion of clustered vacancies is smaller and sensitive to T, falling from 30% to 60% for T ? 600 K to less than 20% when T = 900 K. The structure of clusters has been examined in detail. Vacancies cluster predominantly in stacking-fault-tetrahedron-type configurations. SIAs tend to form either glissile dislocation loops with Burgers vector b = 1/2<1 1 0> or sessile faulted Frank loops with b = 1/3<1 1 1>. Despite the fact that cascades at a given E_PKA and T exhibit a wide range of defect numbers and clustered fractions, there appears to be a correlation in the formation of vacancy clusters and SIA clusters in the same cascade. The size and spatial aspects of this are analysed in detail in part II [unpublished], where the stability of clusters when another cascade overlaps them is also investigated.     

Luminescence studies on gamma irradiated RbCl:Sm and RbBr:Smcrystals

This paper presents and discusses the results obtained on the thermoluminescence (TL) and other optical studies of gamma irradiated RbCl:Sm³⁺ and RbBr:Sm³⁺ crystals. Samarium when doped into the RbCl and RbBr is found to enter the host lattice in its trivalent state and act as electron trap during gamma irradiation, there by partially converting itself to Sm²⁺. The photoluminescence (PL) spectra of both RbCl and RbBr crystals doped with Sm exhibit the strong red/orange emissions of Sm corresponding to ⁴G_5/2 → ⁶H_9/2 (red) and ⁴G_5/2 → ⁶H_7/2 (orange) transitions. Z₃ centers are observed in RbBr on F bleaching subsequent to gamma irradiation and a TL glow peak attributable to the same has been identified. The thermoluminescence (TL) glow curve explains the defect annihilation process to be due to the mobilization of two different kinds of traps created as a result of exposure to high energy irradiation in both the cases. Spectral distribution under the thermoluminescence emission (TLE) and optically stimulated emission(OSL), support the idea that defect annihilation process to be due to thermal release of F electron in RbCl:Sm³⁺ and in RbBr:Sm³⁺ crystals. Both Sm³⁺ and Sm²⁺ emissions were observed in the thermoluminescence emission of the crystals.     

New concept of designing Pu and MA containing fuel for fast reactors

New type of metal base fuel element is suggested for fast reactors. Basic approach to fuel element development - separated operations of fabricating uranium meat fuel element and introducing into it Pu or MA dioxides powder, that results in minimizing dust forming operations in fuel element fabrication. According to new fuel element design a framework fuel element having a porous uranium alloy meat is filled with standard PuO₂ powder of <50 μm fractions prepared by pyrochemical or other methods. In this way a high uranium content fuel meat metallurgically bonded to cladding forms a heat conducting framework, pores of which contain PuO₂ powder. Framework fuel element having porous meat is fabricated by capillary impregnation method with the use of Zr eutectic matrix alloys, which provides metallurgical bond between fuel and cladding and protects it from interaction. As compared to MOX fuel the new one features high thermal conductivity, higher uranium content, hence, high conversion ratio does not interact with fuel cladding and is more environmentally clean. Its principle advantage is a simple production process that is easily realized remotely, feasibility of involving high background Pu and MA isotopes into closed nuclear fuel cycle at the minimal influence on environment.     

Effect of temperature on studtite stability: Thermogravimetry and differential scanning calorimetry investigations

The main objective of this work is the study of the influence of temperature on the stability of the uranyl peroxide tetrahydrate (UO₂O₂ · 4H₂O) studtite, which may form on the spent nuclear fuel surface as a secondary solid phase. Preliminary results on the synthesis of studtite in the laboratory at different temperatures have shown that the solid phases formed when mixing hydrogen peroxide and uranyl nitrate depends on temperature. Studtite is obtained at 298 K, meta-studtite (UO₂O₂ · 2H₂O) at 373 K, and meta-schoepite (UO₃ · nH₂O, with n < 2) at 423 K. Because of the temperature effect on the stability of uranyl peroxides, a thermogravimetric (TG) study of studtite has been performed. The main results obtained are that three transformations occur depending on temperature. At 403 K, studtite transforms to meta-studtite, at 504 K, meta-studtite transforms to meta-schoepite, and, finally, at 840 K, meta-schoepite transforms to U₃O₈. By means of the differential scanning calorimetry the molar enthalpies of the transformations occurring at 403 and 504 K have been determined to be −42 ± 10 and −46 ± 2 kJ mol⁻¹, respectively.     

Synthesis of uranium nitride by a mechanically induced gas-solid reaction

A high energy, low-temperature, ball-milling route was used to directly produce uranium nitride. Pure uranium metal particles (∼100 μm) were ball milled under a 420 kPa nitrogen atmosphere for 24 h at ambient temperature to yield phase pure U₂N₃ powder as confirmed by X-ray diffraction and energy dispersive spectroscopy. The median particle size was measured to be approximately 6 μm.     

A direct current glow discharge plasma source for inner surface modification of metallic tube

A dc glow discharge plasma source was developed for inner surface modification of metallic tubes with an inner diameter of 10 mm. A tungsten wire of 30 μm thick was stretched inside the tube to form coaxial electrodes. DC glow discharge plasma was generated inside the tube by applying a negative high dc voltage to the tube. It was found that the length of the cylindrical plasma bulk depends linearly on the applied voltage. The electron excitation temperature of Ar plasma was measured as 12830 ± 550 K by optical emission spectroscopy method. As a preliminary application, diamond-like carbon (DLC) films were deposited onto the inner surface of stainless steel tube of 100 mm in length and 10 mm in inner diameter by using CH₄/Ar mixture with 40% CH₄ at 40 Pa pressure. The chemical structure of the DLC film deposited on the substrate was analyzed by Raman spectroscopy. The integrated intensity ratio (I_D:I_G) was obtained as 1.62 from the Raman spectra. The thickness of the DLC film deposited on the substrate was estimated as 1.5 μm by scanning electron microscopy (SEM) observation.     

Post-annealing temperature dependence of blistering in high-fluence ion-implanted H in Si 〈1 0 0〉

This study examined the influence of post-annealing temperature on blister formation and growth in ion-implanted H in Si 〈1 0 0〉. Ion energy levels of 40 and 100 keV and fluences of 2 × 10¹⁶ and 5 × 10¹⁶ cm⁻² were investigated. Post-annealing treatments were performed using the furnace annealing (FA) method with temperatures ranging from 200 to 600 °C for a duration of 1 h. Raman scattering spectroscopy (RSS), optical microscopy (OM), secondary ion mass spectrometry (SIMS), atomic force microscopy (AFM), and cross-sectional transmission electron microscopy (XTEM) were employed to explore the mechanisms behind the smart cut technique. The results revealed that variations among the transformation of the VH₃ (or V₂H₆) defect complex phase into the Si(1 0 0):H bonding configuration phase (RSS results), the appearance of optically detectable blisters and craters (OM results), the average depth of craters (AFM results), the trapping of hydrogen atoms and gettering of oxygen atoms (SIMS results), and the damaged microstructures (XTEM results) against post-annealing temperature were in close correspondence. It was also found that the optimal post-annealing temperature for blister formation and growth was 550 °C.     

Short-term irradiation behavior of minor actinide doped uranium plutonium mixed oxide fuels irradiated in an experimental fast reactor

A fuel irradiation program is being conducted using the experimental fast reactor ‘Joyo’. Two short-term irradiation tests in the program were completed in 2006 using a uranium and plutonium mixed oxide fuel which contains minor actinides (MA-MOX fuel). The objective of the tests is the investigation of early thermal behavior of MA-MOX fuel such as fuel restructuring and redistribution of minor actinides. Three fuel pins which contained MA-MOX: 2% neptunium and 2% americium doped uranium plutonium mixed oxide (Am,Pu,Np,U)O_2−x fuel were supplied for testing. The first test was conducted with high-linear heating rate of approximately 430 W cm⁻¹ for only 10 min. After the first test, one fuel pin was removed for examinations. Then the second test was conducted with the remaining two pins at nearly the same linear power for 24 h. In these tests, two oxygen-to-metal molar ratios were used for fuel pellets as a test parameter. Non-destructive and destructive post-irradiation examinations results are discussed with early on the behavior of the fuel during irradiation.     

Ion beam induced defects in solids studied by optical techniques

Optical methods can provide important insights into the mechanisms and consequences of ion beam interactions with solids. This is illustrated by four distinctly different systems.X- and Y-cut LiNbO₃ crystals implanted with 8 MeV Au³⁺ ions with a fluence of 1 × 10¹⁷ ions/cm² result in gold nanoparticle formation during high temperature annealing. Optical extinction curves simulated by the Mie theory provide the average nanoparticle sizes. TEM studies are in reasonable agreement and confirm a near-spherical nanoparticle shape but with surface facets. Large temperature differences in the nanoparticle creation in the X- and Y-cut crystals are explained by recrystallisation of the initially amorphised regions so as to recreate the prior crystal structure and to result in anisotropic diffusion of the implanted gold.Defect formation in alkali halides using ion beam irradiation has provided new information. Radiation-hard CsI crystals bombarded with 1 MeV protons at 300 K successfully produce F-type centres and V-centres having the structure as identified by optical absorption and Raman studies. The results are discussed in relation to the formation of interstitial iodine aggregates of various types in alkali iodides. Depth profiling of and aggregates created in RbI bombarded with 13.6 MeV/A argon ions at 300 K is discussed.The recrystallisation of an amorphous silicon layer created in crystalline silicon bombarded with 100 keV carbon ions with a fluence of 5 × 10¹⁷ ions/cm² during subsequent high temperature annealing is studied by Raman and Brillouin light scattering.Irradiation of tin-doped indium oxide (ITO) films with 1 MeV protons with fluences from 1 × 10¹⁵ to 250 × 10¹⁵ ions/cm⁻² induces visible darkening over a broad spectral region that shows three stages of development. This is attributed to the formation of defect clusters by a model of defect growth and also high fluence optical absorption studies. X-ray diffraction studies show evidence of a strained lattice after the proton bombardment and recovery after long period storage. The effects are attributed to the annealing of the defects produced.     

Individual pebble temperature peaking factor due to local pebble arrangement in a pebble bed reactor core

Scientists at the German AVR pebble bed nuclear reactor discovered that the surface temperature of some of the pebbles in the AVR core were at least 200 K higher than previously predicted by reactor core analysis calculations. The goal of this research paper is to determine whether a similar unexpected fuel temperature increase of 200 K can be attributed solely or mostly to elevated power production resulting from exceptional configurations of pebbles. If it were caused by excessive pebble-to-pebble local power peaking, there could be implications for the need for core physics monitoring which is not now being considered for pebble bed reactors. The PBMR-400 core design was used as the basis for evaluating pebble bed reactor safety. Through exhaustive Monte Carlo modeling of a PBMR-400 pebble environment, no simple pebble-to-pebble burn-up conditions were found to cause a sufficiently high local power peaking to lead to a 200 K temperature increase. Simple thermal hydraulics analysis was performed which showed that a significant core coolant flow anomalies such as higher than expected core bypass flows, local pebble flow variation or even local flow blockage would be needed to account for such an increase in fuel temperature. The identified worst case scenarios are presented and discussed in detail. The conclusion of this work is that the stochastic nature of the pebble bed cannot lead to highly elevated fuel temperatures but rather local or core-wide coolant flow reductions are the likely cause.