CEA and AREVA R&D on HTR fuel fabrication and presentation of the CAPRI experimental manufacturing line

In the framework of the French V/HTR fuel development and qualification program, the Commissariat à l’Energie Atomique (CEA) and AREVA are conducting R&D projects covering the mastering of UO₂ coated particle and fuel compact fabrication technology. To fulfill this task, a review of past knowledge, of existing technologies and a preliminary laboratory-scale work program have been conducted with the aim of retrieving the know-how on HTR coated particle and compact manufacture:

• The different stages of UO₂ kernel fabrication GSP process have been reviewed, reproduced and improved.

• The experimental conditions for the chemical vapor deposition of coatings have been defined on dummy kernels and development of innovative characterization methods has been carried out.

• Former CERCA compacting process has been reviewed and updated.

In parallel, an experimental manufacturing line for coated particles, named GAIA, and a compacting line based on former CERCA compacting experience have been designed, constructed and are in operation since early 2005 at CEA Cadarache and CERCA Romans, respectively. These two facilities constitute the CAPRI line (CEA and AREVA PRoduction Integrated line).The major objectives of the CAPRI line are:

• to recover and validate past knowledge,

• to produce representative HTR TRISO fuel meeting industrial standards,

• to permit the optimization of reference fabrication processes for kernels and coatings defined previously at a laboratory-scale and the investigation of alternative and innovative fuel design (UCO kernel, ZrC coating),

• to test alternative compact process options and

• to fabricate and characterize fuel required for irradiation and qualification purpose.

This paper presents the status of progress of R&D conducted on HTR fuel particles and compact manufacture by early 2005 and the potential of the laboratory-scale HTR fuel CAPRI line.     

Diameter deviation calculation for UO2 kernel by sol-gel process

UO₂ kernels for HTR (High Temperature Gas-cooled Reactor) fuel element are produced using a sol-gel method. The size of UO₂ kernels was controlled strictly by minimizing the dispersion of the UO₂ kernel diameter during the casting process. A relationship between the filling gas tank pressure and solution level in the tank was developed to keep a constant solution mass flow rate in the nozzles. Based on equation of flow energy conservation, a formula about jet velocity, solution level and gas tank pressure on solution was established. Using the formula, the minimum pressure and maximum pressure were calculated. During the casting process, the decline of solution level will lead to the decreasing of jet velocity and the solution mass flow rate that can cause the increasing of diameter deviation. By compensating the pressure on the liquid gradually during the casting process, the flow fluency was kept in an acceptable range stably, the standard deviation of kernel diameter was less than 12 μm.     

The AREVA HTR fuel cycle: An analysis of technical issues and potential industrial solutions

The AREVA high temperature reactor (HTR) is a modular 600 MWt high temperature gas-cooled reactor that provides up to 950 °C process helium for power generation and/or hydrogen production. Energy facilities based on this technology will consist of up to four AREVA HTR modules achieving high thermodynamic efficiency with a total possible electrical generation capability of 1200 MWe. At the heart of the AREVA HTR is the TRISO-coated low-enriched uranium (LEU) fuel which is assembled into fuel compacts that are inserted into prismatic graphite fuel elements.The intent of this paper is to examine the AREVA HTR fuel and fuel cycle from two perspectives. First, from the “front-end” perspective, the available fuel cycle-related options are examined along with the basis for the subsequent choice of fuel type and cycle. Second, from the “back-end” perspective, the generation and management of spent fuel and graphite waste is examined along with the strategies and options available for disposition or disposal.The AREVA HTR has the inherent flexibility to accommodate many fuel types and to permit full cost-effective optimization. The reasons for this flexibility are presented and the main advantages and disadvantages of the various fuels cycle are discussed. The reference fuel cycle for the AREVA HTR is then introduced and its main features are given. Additionally, non-proliferation attributes of the AREVA HTR technology are also examined.The amount of spent fuel and graphite waste generated by the AREVA HTR is governed by the burn-up capability of the fuel and the longevity of the adjacent graphite reflector blocks. As currently envisioned, the AREVA HTR will need to be refueled every 18 months, with 50% of the core being replaced. Additionally, the graphite reflector blocks will need to be replaced at regular intervals. Given a 60-year design life, approximately 20,000 spent fuel elements and 10,000 graphite reflectors blocks will be generated. The near-term and long-term options for dealing with these waste streams are examined and, as with the front end of the fuel cycle, a reference strategy for spent fuel and graphite waste proposed.     

A non destructive technique for determination of U and Pu content in MOX fuel pins for fast reactors

MOX fuel pins containing both U²³³O₂ and PuO₂ have been fabricated for making an experimental subassembly for irradiation in Fast Breeder Test reactor (FBTR) at Kalpakkam, India. This unique composition of the fuel pin is chosen to simulate the thermo-mechanical conditions of the upcoming Prototype Fast Breeder Reactor (PFBR) in the existing Fast Breeder Test Reactor. Since the fertile matrix is natural UO_2, it was difficult to monitor the percentage of U²³³O₂ through chemical methods and neutron assay methods. During the fabrication of MOX fuel pins at Advanced Fuel Fabrication Facility; Bhabha Atomic Research Centre, Tarapur, Passive Gamma Scanning (PGS) was employed as one of the characterisation tools for verifying the fuel composition. PGS was found to be effective in estimating the percentage composition of both U²³³O₂ and PuO₂ and also in ensuring the uniform distribution of the fissile material in MOX fuel pins. PGS is also found effective in monitoring the correct loading of natural UO₂ insulation pellets and MOX fuel pellets in welded MOX pins.     

Fission-product behaviour in irradiated TRISO-coated particles: Results of the HFR-EU1bis experiment and their interpretation

It is important to understand fission-product (FP) and kernel micro-structure evolution in TRISO-coated fuel particles. FP behaviour, while central to severe-accident evaluation, impacts: evolution of the kernel oxygen potential governing in turn carbon oxidation (amoeba effect and pressurization); particle pressurization through fission-gas release from the kernel; and coating mechanical resistance via reaction with some FPs (Pd, Cs, Sr). The HFR-Eu1bis experiment irradiated five HTR fuel pebbles containing TRISO-coated UO₂ particles and went beyond current HTR specifications (e.g., central temperature of 1523 K). This study presents ceramographic and EPMA examinations of irradiated urania kernels and coatings. Significant evolutions of the kernel (grain structure, porosity, metallic-inclusion size, intergranular bubbles) as a function of temperature are shown. Results concerning FP migration are presented, e.g., significant xenon, caesium and palladium release from the kernel, molybdenum and ruthenium mainly present in metallic precipitates. The observed FP and micro-structural evolutions are interpreted and explanations proposed. The effect of high flux rate and high temperature on fission-gas behaviour, grain-size evolution and kernel swelling is discussed. Furthermore, Cs, Mo and Zr behaviour is interpreted in connection with oxygen-potential. This paper shows that combining state-of-the-art post-irradiation examination and state-of-the-art modelling fundamentally improves understanding of HTR fuel behaviour.     

High and very high temperature reactor research for multipurpose energy applications

Ten years ago, the European High Temperature Reactor (HTR) Technology Network (HTR-TN) launched a programme for developing HTR Technology, which expanded so far through 4 successive Euratom Framework Programmes. Many projects have been performed - in particular the RAPHAEL project in the 6th Euratom Framework Programme and presently ARCHER in the 7th - in line with the Network strategy that identified cogeneration of process heat and power as the main specific mission of HTR. HTR can indeed address the growing energy needs of industry presently fully relying on fossil fuel combustion with a CO₂-lean generation technology, thanks to its high operating temperature and to its unique flexibility obtained from its large thermal inertia and its low power.Relying on the legacy of the former European leadership in HTR technology, this programme has addressed specific developments required for industrial process heat applications and for increasing HTR performances (higher temperatures and fuel burn-up). Decisive achievements have been obtained concerning fuel manufacturing and irradiation behaviour, key components and their materials, safety, computer code validation and specific HTR waste (fuel and graphite) management. Key experiments have been performed or are still ongoing: irradiation of graphite, fuel and vessel materials and the corresponding post-irradiation examinations, safety tests and isotopic analyses; thermal-hydraulic tests of an Intermediate Heat Exchanger mock-up in helium; air ingress experiments for a block type core, etc. Through Euratom participation in the Generation IV International Forum (GIF), these achievements contribute to international cooperation.HTR-TN strategy has been recently integrated by the “Sustainable Nuclear Energy Technology Platform” (SNE-TP) as one of the 3 “pillars” of its global nuclear strategy. It is also in line with the orientations and the timing of the “Strategic Energy Technology Plan (SET-Plan)” for the development of CO₂-lean energy technologies, and thus strengthens the nuclear option in a future European energy mix.Nuclear cogeneration for industrial process heat applications is a major innovation and a major challenge, requiring large-scale demonstration to prove its industrial viability. To enable demonstration, it is necessary not only to develop an appropriate nuclear heat source, but also to develop coupling technologies and to adapt industrial processes to the coupling with a HTR. This requires a close partnership between the conventional and the nuclear technology holders as the base of a Nuclear Cogeneration Industrial Initiative.Recently the project EUROPAIRS initiated by HTR-TN together with process heat user industries has set the bases of such a strategic partnership.     

HTR fuel coating separate effect test PYCASSO

The “analytical” PYCASSO (PYrocarbon irradiation for Creep and Swelling/Shrinkage of Objects) irradiations focus on determining the effects of neutron irradiation in the temperature range of 900-1100 °C, excluding effects due to the presence of fuel, such as pressurization or chemical attack by fission products. These irradiations can therefore be considered separate effect tests, where only the influence of neutron fluence and temperature on coatings and coating combinations is investigated.For this purpose dedicated particles have been manufactured consisting of surrogate kernels (ZrO₂ and Al₂O₃) with different types of PyC/SiC/ZrC coatings and coating combinations. All specimens delivered have been extensively characterized, such that even potentially small changes due to the irradiation in dimensions, microstructure and density can be determined accurately after irradiation.Partners involved in this irradiation are CEA (France), JAEA (Japan) and KAERI (South Korea). The PYCASSO irradiations take place in the High Flux Reactor (HFR) in Petten, and are coordinated by NRG (The Netherlands). The partnership for PYCASSO was initiated by the RAPHAEL (V)HTR European 6th Framework Program and is integrated in the Generation IV International Forum VHTR Fuel and Fuel Cycle project.     

Behaviour of spent HTR fuel elements in aquatic phases of repository host rock formations

One back-end option for spent HTR fuel elements proposed for future HTR fuel cycles in the EC is an open fuel cycle with direct disposal of conditioned or non-conditioned fuel elements. This option has already been chosen in Germany due to the political decision to terminate the use of HTR technology. First integral leaching investigations at Research Centre Juelich on the behaviour of spent HTR fuel in salt brines, typical of accident scenarios in a repository in salt, proved that the main part of the radionuclide inventory cannot be mobilised as long as the coated particles do not fail. However, such experiments will not lead to a useful model for performance assessment calculations, because a failure of the coatings by corrosion will not occur during experimental times of a few years. In order to get a robust and realistic model for the long-term behaviour in aqueous phases of host rock systems, it is necessary to understand the barrier function of the different parts of an HTR fuel element, i.e. the matrix graphite, the different coating materials, and the fuel kernel.Therefore, our attention is focused on understanding and modelling the barrier performance of the different parts of an HTR fuel element with respect to their barrier function, and on the development of an overall model for performance assessment. In order to understand this behaviour, it is necessary to start with investigations of unirradiated material, and to proceed with experiments with external gamma irradiation to determine the effects of oxidising radiolysis species. Further experiments with irradiated material have to be performed to investigate the influence of the irradiation damage, and finally an investigation has to be made of the irradiated material plus additional gamma irradiation. Experimental data are now available for the diffusive transport of radionuclides in the water-saturated graphite pore system, the corrosion rates of unirradiated graphite with and without external gamma irradiation and unirradiated and irradiated silicon carbide, and for the dissolution rates of UO₂ and (Th,U)O₂ fuel kernels with and without external gamma irradiation. All investigations were performed in aquatic phases from salt, granite, and clay host rock.     

Next generation fuel irradiation capability in the High Flux Reactor Petten

This paper describes selected equipment and expertise on fuel irradiation testing at the High Flux Reactor (HFR) in Petten, The Netherlands. The reactor went critical in 1961 and holds an operating license up to at least 2015. While HFR has initially focused on Light Water Reactor fuel and materials, it also played a decisive role since the 1970s in the German High Temperature Reactor (HTR) development program. A variety of tests related to fast reactor development in Europe were carried out for next generation fuel and materials, in particular for Very High Temperature Reactor (V/HTR) fuel, fuel for closed fuel cycles (U-Pu and Th-U fuel cycle) and transmutation, as well as for other innovative fuel types. The HFR constitutes a significant European infrastructure tool for the development of next generation reactors. Experimental facilities addressed include V/HTR fuel tests, a coated particle irradiation rig, and tests on fast reactor, transmutation and thorium fuel. The rationales for these tests are given, results are provided and further work is outlined.     

Requirements for high temperature reactor fuel particle design assessment

In the framework of a large Research and Development programme devoted to High Temperature Reactors (HTR) and set up in the CEA from 2000 on, we will address ourselves to the issue of coated fuel performance and design. Although HTR fuel main features have been established for a long time, we need today to reassess the fuel design to make sure that it meets the requirements linked to the most recent projects of High Temperature Reactors. Thus, in collaboration with Framatome and in connection with the Gas Turbine - Modular Helium Reactor (GT-MHR) international project, we are planning to perform parametric thermal and mechanical studies, regarding different particle design options (kernel diameter, layers composition and thickness) and seeking optima concerning particle leak tightness and fission product retention. But to initiate such studies, we have first of all to define the design bases and the requirements for HTR fuel, in terms of kernel composition (fissile element, oxide stoechiometry, enrichment), particle and compact geometry (dimensions, particle volume fraction in the graphite matrix), power density, cooling gas temperature and irradiation conditions (burnup, fast fluence).     

TRU self-recycling in a high temperature modular helium reactor

The Deep Burn Project is developing high burnup fuel based on Ceramically Coated (TRISO) particles, for use in the management of spent fuel Transuranics. This paper evaluates the TRU deep-burn in a High Temperature Reactor (HTR) that recycles its own transuranic production. The DB-HTR is loaded with standard LEU fresh fuel and the self-generated TRUs are recycled into the same core (after reprocessing of the original spent fuel). This mode of operation is called self-recycling (SR-HTR). The final spent fuel of the SR-HTR can be disposed of in a final repository, or recycled again.In this study, a single recycling of the self-generated TRUs is considered. The UO₂ fuel kernel is 12% uranium enrichment and the diameter of the kernel is 500 μm. TRISO packing fraction of UO₂ fuel compact is 26%. In the SR-HTR fuel cycle, it is assumed that the spent UO₂ fuel is reprocessed with conventional technology and the recovered TRUs are fabricated into Deep Burn TRISO fuel. The diameter of 200 μm is used for the TRU fuel kernel. A typical coating thickness is used. The core performance is evaluated for an equilibrium cycle, which is obtained by cycle-wise depletion calculations. From the analysis results, the equilibrium cycle lengths of Case 1 (5-ring fuel block SR-HTR) and Case 2 (4-ring fuel block SR-HTR) are 487 and 450 EFPDs (effective full power days), respectively. And the UO₂ fuel discharge burnups of Case 1 and Case 2 are 10.3% and 10.1%, respectively. Also, the TRU discharge burnups of Case 1 and Case 2 are 64.7% and 63.5%, respectively, which is considered extremely high. The fissile (Pu-239 and Pu-241) content of the self-generated TRU vector is about 52%. The deep-burning of TRU in SR-HTR is partly due to the efficient conversion of Pu-240 to Pu-241, which is boosted by the uranium fuel in SR-HTR. It is also observed that the power distribution is quite flat within the uranium fuel zone. The lower power density in TRU fuel is because the TRU burnup is very high. Also, it is found that transmutation of Pu-239 is near complete in SR-HTR and that of Pu-241 is extremely high in all cases. The decay heat of the SR-HTR core is very similar to the UO₂-only core. However, accumulation of the minor actinides is not avoidable in the SR-HTR core. The extreme high burnup of the Deep Burn fuel greatly reduces the amount of heat producing isotopes that could be problematic in spent fuel repositories (like Pu-238).     

Determination of Heats of Formation for Solid and Liquid Lithium Metasilicates by Vaporization Method

In-reactor experiments were performed in Nuclear Safety Research Reactor of Japan Atomic Energy Research Institute to study the failure behavior of stainless steel clad fuel rods under a simulated reactivity initiated accident (RIA) condition. A single test fuel rod with stainless steel cladding was contained in a capsule filled with water at room temperature and atmospheric pressure and irradiated by pulsing power simulating an RIA. It was revealed through the experiments that the failure mechanism of the stainless steel clad fuel rod was cladding melting, which was different from oxygen-induced embrittlement observed in the Zircaloy clad fuel rod in the same test condition, and the failure threshold energy was determined to be about 240cal/g·UO₂ (–1,000 kJ/kg·UO₂), which was about 20 cal/g·UO₂ (–85 kJ/kg·UO₂) lower than that of the Zircaloy clad fuel rod. It was also found that the mechanical energy was generated by explosive vaporization of coolant due to molten fuel-coolant interaction as a consequence of the fuel rod failure accompanying fuel pellet fragmentation at an energy deposition of nearly 380 cal/g·UO₂ (–1,600 kJ/kg·UO₂) or more.     

Behavior of Coated Fuel Particle of High-Temperature Gas-Cooled Reactor under Reactivity-Initiated Accident Conditions

In order to clarify the failure mechanism and determine the failure limit of the High-Temperature Gascooled Reactor (HTGR) fuel under reactivity-initiated accident (RIA) conditions, pulse irradiations were performed with unirradiated coated fuel particles at the Nuclear Safety Research Reactor (NSRR). The energy deposition ranged from 0.578 to 1.869 kJ/gUO₂, in the pulse irradiations and the estimated peak temperature at the center of the fuel particle ranged from 1,510 to 3,950 K. Detailed examinations after the pulse irradiations showed that the coated fuel particles failed above 1.40 kJ/gUO₂, where the peak fuel temperature reached over the melting point of UO₂ fuel. It was concluded that the coated fuel particle was failed by the mechanical interaction between the melted and swelled fuel kernel and the coating layer under RIA conditions.     

Testing of high heat flux components manufactured by ENEA for ITER divertor

ENEA is involved in the International Thermonuclear Experimental Reactor (ITER) R&D activities and in particular in the manufacturing of high heat flux plasma-facing components, such as the divertor targets. During the last years ENEA has manufactured actively cooled mock-ups by using different technologies, namely brazing, diffusion bonding and HIPping. A new manufacturing process that combines two main techniques PBC (Pre-Brazed Casting) and the HRP (Hot Radial Pressing) has been set up and widely tested.A full monoblock medium scale vertical target, having a straight CFC armoured part and a curved W armoured part, was manufactured using this process.The ultrasonic method was used for the non-destructive examinations performed during the manufacturing of the component, from the monoblock preparation up to the final mock-up assembling. The component was also examined by thermography on SATIR facility (CEA, France), afterwards it was thermal fatigue tested at FE200 (200 kW electron beam facility, CEA/AREVA France).The successful results of the thermal fatigue testing performed according the ITER requirements (10 MW/m², 3000 cycles of 10 s on both CFC and W part, then 20/15 MW/m², 2000 cycles of 10 s on CFC/W part, respectively) have confirmed that the developed process can be considerate a candidate for the manufacturing of monoblock divertor components. Furthermore, a 35-MW/m² Critical Heat Flux was measured at relevant thermal–hydraulics conditions at the end of the testing campaign.This paper reports the manufacturing route, the thermal fatigue testing results, the pre and post non-destructive examination and the destructive examination performed on the ITER vertical target medium scale mock-up.These activities were performed in the frame of EFDA contracts (04-1218 with CEA, 93-851 JN with AREVA and 03-1054 with ENEA).     

PECITIS-II, a computer program to predict the performance of collapsible clad UO2 fuel elements

The codes devised and used in India for the design of fuel for their Pressurized Heavy Water Reactor (PHWR) programme are described. The scheme includes the use of collapsible fuel cladding for improved neutron economy.This code is made with reference to collapsible clad UO₂ fuel elements. This evaluates sheath strain and fission gas pressure. The fuel expansion is calculated by a two zone model which assumes that above a certain temperature the UO₂ deforms plastically and below that temperature it cracks radially and behaves as an elastic solid; the plastic core is under compression. The pellet clad gap conductance is calculated by using a modified Ross and Stoute model considering the effects of fuel and clad thermal expansion, fission gas release, dilution of filler gas and irradiation swelling. Stress relaxation of the sheath and its effect on fuel sheath contact pressure is also considered for arriving at the end result.     

Development of Impregnated Agglomerate Pelletization (IAP) process for fabrication of (Th,U)O2 mixed oxide pellets

Impregnated Agglomerate Pelletization (IAP) technique has been developed at Advanced Fuel Fabrication Facility (AFFF), BARC, Tarapur, for manufacturing (Th,²³³U)O₂ mixed oxide fuel pellets, which are remotely fabricated in hot cell or shielded glove box facilities to reduce man-rem problem associated with ²³²U daughter radionuclides. This technique is being investigated to fabricate the fuel for Indian Advanced Heavy Water Reactor (AHWR). In the IAP process, ThO₂ is converted to free flowing spheroids by powder extrusion route in an unshielded facility which are then coated with uranyl nitrate solution in a shielded facility. The dried coated agglomerate is finally compacted and then sintered in oxidizing/reducing atmosphere to obtain high density (Th,U)O₂ pellets. In this study, fabrication of (Th,U)O₂ mixed oxide pellets containing 3–5 wt.% UO₂ was carried out by IAP process. The pellets obtained were characterized using optical microscopy, XRD and alpha autoradiography. The results obtained were compared with the results for the pellets fabricated by other routes such as Coated Agglomerate Pelletization (CAP) and Powder Oxide Pelletization (POP) route.     

Measurements of Doppler Effect of Coated Particle Fuel Rod in SHE-14 Core Using Sample Heating Device

Reactivity decrease due to temperature rise of a single fuel rod sample was measured in the SHE-14 core using a sample heating device with purpose to verify the calculation accuracy of the Doppler effect for resonance neutron absorption in a Very High Temperature Reactor. The fuel rod sample was a stuck of coated particle fuel compacts containing 4% enriched UO₂ kernels, and it was heated up to about 700°C in a sample heating tube which was placed along the axis of the core.

The difference of reactivity decrease between the two same size samples of fuel rod and graphite rod due to temperature rise can be interpreted as the increased resonance neutron capture of ²³⁸U, i.e. Doppler effect. The SRAC code system was applied to the Doppler effect calculation where the collision probability method was used in the cell calculation and the one-dimensional, multi-group diffusion approximations were adopted in the core calculation. The results gave a ratio of the calculated to the measured Doppler effect of 0.93.     

Nuclear heat for hydrogen production: Coupling a very high/high temperature reactor to a hydrogen production plant

Hydrogen has been dubbed the fuel of the future. As fossil fuel reserves become depleted and greenhouse gas emissions are reduced inline with the Kyoto protocol, alternative energy sources and vectors, such as hydrogen, must be developed. Hydrogen produced from water splitting, as opposed to from hydrocarbons, has the potential to be a carbon neutral energy solution. There are several methods to extract hydrogen from water, three leading candidates being high temperature electrolysis, the SI thermochemical cycle and the HyS hybrid thermochemical cycle. All three of these processes involve a section requiring very high temperatures. The Very High Temperature Reactor (VHTR), a gas cooled Generation IV reactor, is ideally suited for providing this high temperature heat. Nuclear hydrogen production is being investigated around the world. The four leading consortiums are the Japan Atomic Energy Agency (JAEA), PBMR/Westinghouse, GA, and AREVA NP/CEA/EDF. There are also many smaller R&D efforts focussing on the development of particular materials and components and on process flowsheeting.A nuclear hydrogen plant involves four key pieces of equipment: the VHTR, the hydrogen production plant (HPP), the intermediate heat exchanger (IHX) and the power conversion system (PCS). The choice of all four items varies dramatically between programmes. Both pebble bed and prismatic fuel block VHTRs are being developed, which can be directly or indirectly coupled to a HPP and PCS placed either in series or parallel. Either a Rankine steam cycle or a Brayton gas turbine cycle can be employed in the PCS. This report details the choices made and research being carried out around the world.Predicted process efficiencies and plant costs are currently at a preliminary stage and are very similar, regardless of the options chosen. The cost of hydrogen produced from water splitting using nuclear technologies is around $2/kg H₂. This is competitive with hydrogen produced by other methods, particularly if carbon emissions are regulated and costed. The technological feasibility and testing of key components will be one of the determining factors in plant viability.     

A fabrication technique for a UO2 pellet consisting of UO2 grains and a continuous W channel on the grain boundary

A new fabrication process of UO₂-W composite fuel has been studied in order to improve the thermal conductivity of the UO₂ pellet by the addition of a small amount of W. A fabrication process was designed from the phase equilibria among tungsten, tungsten oxides and UO₂. The conventionally sintered UO₂ pellet which contains W particles is heat-treated in an oxidizing gas and then in a reducing gas. In the oxidizing heat-treatment W particles are oxidized and liquid tungsten oxide penetrates within the UO₂ grain boundary, and in the reducing heat-treatment liquid oxide is transformed to solid tungsten which forms a continuous channel along the UO₂ grain boundary. This developed technique can provide a continuous W channel covering UO₂ grains for a UO₂-W composite fuel even with a small amount of a metal phase - below 6 vol.%. The thermal diffusivity of the UO₂-6 vol.%W cermet composite increases by about 80% when compared with that of a pure UO₂ pellet.     

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.