Fuel Fabrication Research Articles

04/00739 Aspects of fabrication of curium-based fuels and targets: Pillon, S. et al. Journal of Nuclear Materials, 2003, 320, (1–2), 36–43

04/00739 Aspects of fabrication of curium-based fuels and targets: Pillon, S. et al. Journal of Nuclear Materials, 2003, 320, (1–2), 36–43

04/00968 Design and fabrication of specific ceramic-metallic fuels and targets: Fernández, A. et al. Journal of Nuclear Materials, 2003, 319, (1), 44–50

04/00968 Design and fabrication of specific ceramic-metallic fuels and targets: Fernández, A. et al. Journal of Nuclear Materials, 2003, 319, (1), 44–50

Introduction to Nuclear Fuel Engineering ; Focused on LWR Fuel ― (7) LWR Fuel Fabrication

Introduction to Nuclear Fuel Engineering ; Focused on LWR Fuel ― (7) LWR Fuel Fabrication

The fabrication and irradiation of plutonium-containing inert matrix fuels for the ‘Once Though Then Out’ experiment

Seven plutonium containing inert matrix fuel (IMF) capsules were prepared for an irradiation experiment in the high flux reactor. In the irradiation experiment, named once-through-then-out, both spinel-based and zirconia-based targets are irradiated up to a plutonium burnup of about 200 GW d m −3, corresponding to a Pu depletion of 50–60%. Both micro and macrodispersed spinel targets are tested, the zirconia targets are of the homogeneous type. In this paper, the fabrication routes of the IMF pellets for this irradiation are described in detail. The results of the dimensional measurements, density measurements, ceramographies and X-ray images of the samples are given. Some preliminary results of the irradiation are included.

Aspects of fabrication of curium-based fuels and targets

The limited fabrication possibilities will be decisive in the design of curium fuels and targets. It appears that there are significant advantages to be gained by separating the Am and Cm at the reprocessing stage, simply to concentrate the problems caused by Cm and limit its mass flow. Vibrocompaction of particles or spheres is probably the most suitable consolidation process, as it reduces the number of fabrication steps and reduces the concomitant controls and scraps occurring in pelletization routes for mixed fuel and targets. The infiltration process is the most promising method to convert aqueous Cm solutions to solid material, but sol–gel and other controlled precipitation for granulate production should not be neglected. Storage of the solidified Cm with a delayed introduction into the reactor for Plutonium transmutation is the most promising concept.

Fabrication of nitride fuels for transmutation of minor actinides

At the Japan Atomic Energy Research Institute, the concept of the transmutation of minor actinides (MA: Np, Am and Cm) with accelerator-driven systems is being studied. The MA nitride fuel has been chosen as a candidate because of the possible mutual solubility among the actinide mononitrides and excellent thermal properties besides supporting hard neutron spectrum. MA nitrides of NpN, (Np, Pu)N, (Np, U)N, AmN, (Am, Y)N, (Am, Zr)N and (Cm, Pu)N were prepared from the oxides by the carbothermic reduction method. The prepared MA nitrides were examined by X-ray diffraction and the contents of impurities of oxygen and carbon were measured. The fabrication conditions for MA nitrides were improved so as to reduce the impurity contents. For an irradiation test of U-free nitride fuels, pellets of (Pu, Zr)N and PuN + TiN were prepared and a He-bonded fuel pin was fabricated. The irradiation test started in May 2002 and will go on for two years in the Japan Materials Testing Reactor.

Fabrication of transmutation fuels and targets: the ECRIX and CAMIX-COCHIX experience

This paper describes the developments and results of the fabrication technology of target materials prepared for the ECRIX and CAMIX–COCHIX experiments concerning the transmutation of americium in uranium-free targets. In these irradiation experiments, planned for the PHENIX reactor, several target concepts will be tested: microdispersed composites of AmO 1.6 in MgO (ECRIX), micro- as well as macrodispered composites of (Am,Y,Zr)O 2− x in MgO and homogeneous (Am,Y,Zr)O 2− x (CAMIX–COCHIX) material. Results of the completed fabrication campaign of ECRIX in the Atalante facility of CEA and the pre-qualification tests (with Ce and Pu) for CAMIX–COCHIX at ITU are given.

Equipment Arrangements for DUPIC Nuclear Fuel Fabrication and Their Remote Operation and Maintenance in Hot-Cell

DUPIC (Direct Use of spent PWR fuel In CANDU reactors) nuclear fuel that reuses spent PWR (Pressurized Water Reactor) fuel as raw material is being developed at KAERI (Korea Atomic Energy Research Institute). All DUPIC nuclear fuel fabrication processes that mainly consist of powder preparation, pelletization and rod fabrication are conducted in the completely shielded M6 hot-cell of the IMEF (Irradiated Material Examination Facility) at KAERI because of the nature of the high radioactivity of spent PWR fuel. Various types of equipment specially designed for DUPIC fuel fabrication are operated and maintained in a remote manner. This paper presents the in-cell arrangements of all the equipment used for DUPIC nuclear fuel fabrication in the confined hot-cell facility and their remote operation and maintenance in situ.

Transmutation of Actinides

An advanced process has been developed on a laboratory scale at the Institute for Transuranium Elements (ITU) for the fabrication of transmutation fuels and targets. The process consists of the fabrication of highly porous beads using a sol–gel method, their infiltration by a nitrate solution of the relevant actinide, and calcination, compaction, and sintering to give the final product pellets. This work presents the results obtained in the qualification of this process whereby dense partially‐yttria‐stabilized zirconia pellets with high contents of cerium and plutonium have been fabricated.

Economic Analysis on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors—I: DUPIC Fuel Fabrication Cost

A preliminary conceptual design of a Direct Use of spent Pressurized water reactor (PWR) fuel In Canada deuterium uranium (CANDU) reactors (DUPIC) fuel fabrication plant was studied, which annually converts spent PWR fuel of 400 tonnes heavy element (HE) into CANDU fuel. The capital and operating costs were estimated from the viewpoint of conceptual design. Assuming that the annual discount rate is 5% during the construction (5 yr) and operation period (40 yr) and contingency is 25% of the capital cost, the levelized unit cost (LUC) of DUPIC fuel fabrication was estimated to be 616 $/kg HE, which is mostly governed by annual operation and maintenance costs that correspond to 63% of LUC. Among the operation and maintenance cost components being considered, the waste disposal cost has the dominant effect on LUC (~49%). From sensitivity analyses of production capacity, discount rate, and contingency, it was found that the production capacity of the plant is the major parameter that affects the LUC.

Hot-Cell Shielding System for High Power Transmission in DUPIC Fuel Fabrication

This paper presents a newly designed hot-cell shielding system for use in the development of DUPIC (Direct Use of spent PWR fuel In CANDU reactors) fuel at KAERI (Korea Atomic Energy Research Institute). This hot-cell shielding system that was designed to transmit high power to sintering furnace in-cell from the out-of-cell through a thick cell wall has three subsystems - a steel shield plug with embedded spiral cooling line, stepped copper bus bars, and a shielding lead block. The dose-equivalent rates of the hot-cell shielding system and of the apertures between this system and the hot-cell wall were calculated. Calculated results were compared with the allowable dose limit of the existing hot-cell. Experiments for examining the temperature changes of the shielding system developed during normal furnace operation were also carried out. Finally, gamma-ray radiation survey experiments were conducted by Co-60 source. It is demonstrated that, from both calculated and experimental results, the newly designed hot-cell shielding system meets all the shielding requirements of the existing hot-cell facility, enabling high power transmission to the in-cell sintering furnace.

Process Modeling of Plutonium Conversion and Mixed-Oxide Fuel Fabrication for Plutonium Disposition

In an effort to reduce the global stockpile of nuclear explosive devices, ~50 tonnes of weapons-grade plutonium have been declared surplus to national security needs by the United States. This surplus, located at six sites within the U.S. Department of Energy complex (the Hanford Site, Idaho National Engineering and Environmental Laboratory, Los Alamos National Laboratory, the Pantex Plant, the Rocky Flats Environmental Technology Site, and the Savannah River Site) must now be rendered unattractive for use in nuclear weapons. The goal is that this drive will be concurrent with similar activities in Russia. One method currently under investigation is the conversion of the plutonium metal into mixed-oxide (MOX) reactor fuel. Approximately 35 tonnes of the surplus plutonium is in a form suitable for fabrication into MOX fuel. This fuel would be used in currently operating reactors for power production.Two processes are currently under consideration for the disposition of the 35 tonnes of surplus plutonium through its conversion into fuel for power production. These processes are the Advanced Recovery and Integrated Extraction System (ARIES) process, by which plutonium metal is converted into a powdered oxide form, and MOX fuel fabrication, where the oxide powder is combined with uranium oxide powder to form ceramic fuel. Because it is envisioned that plutonium disposition will occur concurrently in the United States and Russia, the timely disposition of the plutonium is deemed important to national security. However, the need for quick disposition must be tempered by cost considerations and constraints on the reactors that will ultimately use the fuel. This study was undertaken to determine the optimal size for both the pit conversion and MOX fabrication facilities, whereby the 35 tonnes of plutonium metal will be converted into fuel and burned for power. Proper sizing of the facilities will help avoid unnecessary delays and excessive costs and thus is important in the success of the disposition mission.The bounding conditions used were a plutonium concentration of 3 to 7%, a burnup of 20,000 to 40,000 MWd/tonnes HM, a core fraction of 0.1 to 0.4, and the number of reactors ranging from 2 to 6. Using these boundary conditions, the optimal plutonium concentration was found to be 7%. This resulted in an optimal throughput ranging from 2000 to 5000 kg/yr of plutonium. The data showed minimal costs (based solely on facility size and required manpower) resulting from throughputs in this range, at 3840, 2779, and 3497 kg/yr of plutonium, which resulted in a facility lifetime of 9.1, 12.6, and 10.0 yr, respectively.

Preliminary fabrication and characterisation of inert matrix and thoria fuels for plutonium disposition in light water reactors

The plutonium disposition is presently acknowledged as a most urgent issue at the world level. Inert matrix and thoria fuel concepts for Pu burning in LWRs show good potential in providing effective and ultimate solutions to this issue. In non-fertile (U-free) inert matrix fuel, plutonium oxide is diluted within inert oxides such as stabilised ZrO 2, Al 2O 3, MgO or MgAl 2O 4. Thoria addition, which helps improve neutronic characteristics of inert fuels, appears as a promising variant of U-free fuel. In the context of an R&D activity aimed at assessing the feasibility of the fuel concept above, simulated fuel pellets have been produced both from dry-powder metallurgy and the sol–gel route. Results show that they can be fabricated by matching basic nuclear grade specifications such as the required geometry, density and microstructure. Some characterisation testing dealing with thermo-physical properties, ion irradiation damage and solubility also have been started. Results from thermo-physical measurements at room temperature have been achieved. A main feature stemming from solubility testing outcomes is a very high chemical stability which should render the fuel strongly diversion resistant and suitable for direct final disposal in deep geological repository (once-through solution).

An Electrocatalyzed Process for Leaching of Plutonium from Active Gauntlets and Subsequent Separation by Solvent Extraction

An electrocatalyzed process has been described for the dissolution of plutonium from the contaminated neoprene gauntlets used during the fabrication of plutonium-bearing fuels. Agents Ag(II) and Ce(IV) were used as catalysts. The results indicated that it was possible to leach out plutonium satisfactorily from such contaminated neoprene gauntlets. The problems involved in processing with Ag(II) and the limitations of the process are also discussed. Further, plutonium was separated from the leached solution by solvent extraction using di-2-ethylhexyl phosphoric acid as the extractant.

98/02152 An enhanced Hopfield model for economic dispatch considering prohibited zones

Presently the installed capacity of electricity generation in India is 160 GWe. The integrated ene rgy policy aims at an installed capacity of 778 GWe by 2031-32 to achieve per capita electricity consumption of 2700 kWh/year as against the 700 kWh/year. The share of nuclear power in India is around 3% presently. The increase in installed capacity is possible by increasing the share of nuclear power. India has opted for a unique three-stage power programme based on closed nuclear fuel cycle, which provides a multiplier effect through breeding. Consequent to the 123 agreement and clearance from Nuclear Supplier‟s Group (NSG) for international cooperation in the field of nuclear energy, the reactor technology options are wide open. Some of the technologies which are being pursued are: Pressurized Heavy Water Reactors (PHWRs) fueled with domestic Natural Uranium (NU), PHWRs fueled with imported Natural Uranium/Slightly Enriched Uranium (SEU), Light Water Reactors (LWRs) procured from abroad using imported Lightly Enriched Uranium (LEU) fuel, PHWRs with Reprocessed Uranium (RU) obtained from reprocessing spent fuel of LWRs, indigenous Pressurized Water Reactors (PWRs) and Fast Breeder Reactors (FBRs) using MOX/metallic fuel. The increase in the share of nuclear power through above technologies require setting up of fuel fabrication facilities such as: new PHWR Fuel Fabrication with indigenous sources, joint collaboration with foreign countries for fabricating fuel for imported LWRs, setting up PHWR fuel fabrication facilities with SEU/RU under international safeguards, setting up of enrichment and fuel fab rication facilities for indigenous PWRs and a series of fast reactor fuel fabrication facilities for fabricating fuel for FBRs. In addition to fuel fabrication facilities, facilities for manufacturing zirconium alloy and stainless steel structurals, tubes and components are also required to be set up. The paper gives in detail the emerging nuclear fuel fabrication activities in India.

Iran's nuclear procurement program: How close to the bomb?

(1997). Iran's nuclear procurement program: How close to the bomb? The Nonproliferation Review: Vol. 5, No. 1, pp. 123-135.

Thermodynamics of carbothermic synthesis of actinide mononitrides

Carbothermic synthesis will be further applied to the fabrication of nitride fuels containing minor actinides (MA) such as neptunium, americium and curium. A thorough understanding of the carbothermic synthesis of UN will be beneficial in the development of the MA-containing fuels. Thermodynamic analysis was carried out for conditions of practical interest in order to better understand the recent fabrication experiences. Two types of solution phases, oxynitride and carbonitride phases, were taken into account. The PuNO ternary isotherm was assessed for the modelling of M(C, N, O). With the understanding of the UN synthesis, the fabrication problems of Am-containing nitrides are discussed.

The measurement of beta dose and gamma air kerma rates in the Ribble Estuary

As a by-product of the fabrication of uranium fuels, BNFL, Springfields are authorised to discharge beta emitting rardionuclides, principally into the Ribble Estuary. The accumulation of in the estuarine sediments leads to enhanced beta dose rates not seen elsewhere in the UK. The methods used to determine the beta dose rate and calculate organ and effective doses are presented. The variation in beta dose and gamma air kerma rates in a small area were investigated and recommendations are made as to the number of replicates needed to reduce the error to of the mean. An example of the dose to a walker on the intertidal areas of the Ribble Estuary is calculated by the detailed methods. The effective dose was , of which 86% was derived from gamma irradiation. Doses to hands contaminated by estuarine sediments are also low .

Use of powder metallurgy for fabrication of aluminium-uranium composite reactor fuels: P.V. Hedge et al (Atomic Research Centre, Bombay, India), Trans. PM Assoc. of India, Vol 20, 1993, 27–31

Use of powder metallurgy for fabrication of aluminium-uranium composite reactor fuels: P.V. Hedge et al (Atomic Research Centre, Bombay, India), Trans. PM Assoc. of India, Vol 20, 1993, 27–31

Mixed-Oxide Fuel Fabrication Technology and Experience at the Belgonucléaire and CFCa Plants and Further Developments for the MELOX Plant

Plutonium recycling in light water reactors (LWRs) has progressively become a fact. Over 200 t of mixed-oxide (MOX) fuel have been produced in the COGEMA and BELGONUCLEAIRE plants in the last 7 yr. Fuel loaded in European reactors -- mainly MIMAS fuel -- is presenting satisfactory in-core behavior and performance. Fuel fabrication technology and operation experience of the BELGONUCLEAIRE Po Dessel Plant (35 ton HM/yr) are reviewed. Backfitting of the Complex de Fabrication de Cadarache (CFCa) plant to MOX fabrication and recent fabrication progress are also addressed. The MELOX plant erected by COGEMA in Marcoule (South France) is a major commitment to provide the utilities with important additional plutonium recycling by the mid-1990s. This second generation plant has been designed to currently produce high plutonium-content MOX fuel and recycle degraded plutonium originating from high-burnup LWR UO[sub 2] fuels.