Present state and problems of uranium fuel fabrication businesses
International Nuclear Information System (INIS)
Yuki, Akio
1981-01-01
The businesses of uranium fuel fabrication converting uranium hexafluoride to uranium dioxide powder and forming fuel assemblies are the field of most advanced industrialization among nuclear fuel cycle industries in Japan. At present, five plants of four companies engage in this business, and their yearly sales exceeded 20 billion yen. All companies are planning the augmentation of installation capacity to meet the growth of nuclear power generation. The companies of uranium fuel fabrication make the nuclear fuel of the specifications specified by reactor manufacturers as the subcontractors. In addition to initially loaded fuel, the fuel for replacement is required, therefore the demand of uranium fuel is relatively stable. As for the safety of enriched uranium flowing through the farbicating processes, the prevention of inhaling uranium powder by workers and the precaution against criticality are necessary. Also the safeguard measures are imposed so as not to convert enriched uranium to other purposes than peacefull ones. The strict quality control and many times of inspections are carried out to insure the soundness of nuclear fuel. The growth of the business of uranium fuel fabrication and the regulation of the businesses by laws are described. As the problems for the future, the reduction of fabrication cost, the promotion of research and development and others are pointed out. (Kako, I.)
Impact of fuel fabrication and fuel management technologies on uranium management
International Nuclear Information System (INIS)
Arnsberger, P.L.; Stucker, D.L.
1994-01-01
Uranium utilization in commercial pressurized water reactors is a complex function of original NSSS design, utility energy requirements, fuel assembly design, fuel fabrication materials and fuel fabrication materials and fuel management optimization. Fuel design and fabrication technologies have reacted to the resulting market forcing functions with a combination of design and material changes. The technologies employed have included ever-increasing fuel discharge burnup, non-parasitic structural materials, burnable absorbers, and fissile material core zoning schemes (both in the axial and radial direction). The result of these technological advances has improved uranium utilization by roughly sixty percent from the infancy days of nuclear power to present fuel management. Fuel management optimization technologies have also been developed in recent years which provide fuel utilization improvements due to core loading pattern optimization. This paper describes the development and impact of technology advances upon uranium utilization in modern pressurized water reactors. 10 refs., 3 tabs., 10 figs
PHWR fuel fabrication with imported uranium - procedures and processes
International Nuclear Information System (INIS)
Rao, R.V.R.L.V.; Rameswara Rao, A.; Hemantha Rao, G.V.S.; Jayaraj, R.N.
2010-01-01
Following the 123 agreement and subsequent agreements with IAEA & NSG, Government of India has entered into bilateral agreements with different countries for nuclear trade. Department of Atomic Energy (DAE), Government of India, has entered into contract with few countries for supply of uranium material for use in the safeguarded PHWRs. Nuclear Fuel Complex (NFC), an industrial unit of DAE, established in the early seventies, is engaged in the production of Nuclear Fuel and Zircaloy items required for Nuclear Power Reactors operating in the country. NFC has placed one of its fuel fabrication facilities (NFC, Block-A, INE-) under safeguards. DAE has opted to procure uranium material in the form of ore concentrate and fuel pellets. Uranium ore concentrate was procured as per the ASTM specifications. Since no international standards are available for PHWR fuel pellets, Specifications have to be finalized based on the present fabrication and operating experience. The process steps have to be modified and fine tuned for handling the imported uranium material especially for ore concentrate. Different transportation methods are to be employed for transportation of uranium material to the facility. Cost of the uranium material imported and the recoveries at various stages of fuel fabrication have impact on the fuel pricing and in turn the unit energy costs. Similarly the operating procedures have to be modified for safeguards inspections by IAEA. NFC has successfully manufactured and supplied fuel bundles for the three 220 MWe safeguarded PHWRs. The paper describes various issues encountered while manufacturing fuel bundles with different types of nuclear material. (author)
Natural uranium metallic fuel elements: fabrication and operating experience
International Nuclear Information System (INIS)
Hammad, F.H.; Abou-Zahra, A.A.; Sharkawy, S.W.
1980-01-01
The main reactor types based on natural uranium metallic fuel element, particularly the early types, are reviewed in this report. The reactor types are: graphite moderated air cooled, graphite moderated gas cooled and heavy water moderated reactors. The design features, fabrication technology of these reactor fuel elements and the operating experience gained during reactor operation are described and discussed. The interrelation between operating experience, fuel design and fabrication was also discussed with emphasis on improving fuel performance. (author)
Babcock and Wilcox plate fabrication experience with uranium silicide spherical fuel
International Nuclear Information System (INIS)
Todd, Lawrence E.; Pace, Brett W.
1996-01-01
This report is written to present the fuel fabrication experience of Babcock and Wilcox using atomized spherical uranium silicide powder. The intent is to demonstrate the ability to fabricate fuel plates using spherical powder and to provide useful information proceeding into the next phase of work using this type of fuel. The limited quantity of resources- spherical powder and time, did not allow for much process optimizing in this work scope. However, the information contained within provides optimism for the future of spherical uranium silicide fuel plate fabrication at Babcock and Wilcox.The success of assembling fuel elements with spherical powder will enable Babcock and Wilcox to reduce overall costs to its customers while still maintaining our reputation for providing high quality research and test reactor products. (author)
The Fabrication Problem Of U3Si2-Al Fuel With Uranium High Loading
International Nuclear Information System (INIS)
Supardjo
1996-01-01
The quality of U 3 Si 2 -Al dispersion fuel product is the main aim for each fabricator. Low loading of uranium fuel element is easily fabricated, but with the increased, uranium loading, homogeneity of uranium distribution is difficult to achieve and it always formed white spots, blister, and dogboning in the fuel plates. The problem can be eliminated by the increasing treatment of the fuel/Al powder. The precise selection of fuel/Al particles diameter is needed indeed to make easier in the homogeneous process of powder and the porosities arrangement in the fuel plates. The increasing of uranium loading at constant meat thickness will increase the meat hardness, therefore to withdraw the dogboning forming, the use of harder cladding materials is necessity
Evaluation of bioassay program at uranium fuel fabrication plants
International Nuclear Information System (INIS)
Biggs, D.
1981-03-01
Results of a comprehensive study of urinalysis, lung burden and personal air sample measurements for workers at a uranium fuel fabrication plant are presented. Correlations between measurements were found and regression models used to explain the relationship between lung burden, daily intakes and urinary excretions of uranium. Assuming the ICRP lung model, the lung burden histories of ten workers were used to estimate the amounts in each of the long-term compartments of the lung. Estimates of the half lives of each compartment and of the maximum relative contributions to the urine from each compartment are given. These values were then used to predict urinary excretions from the long-term compartments for workers at another fuel fabrication plant. The standard error of estimate compared well with the daily variation in urinary excretion. (author)
Fabrication of Uranium Oxycarbide Kernels for HTR Fuel
International Nuclear Information System (INIS)
Barnes, Charles; Richardson, Clay; Nagley, Scott; Hunn, John; Shaber, Eric
2010-01-01
Babcock and Wilcox (B and W) has been producing high quality uranium oxycarbide (UCO) kernels for Advanced Gas Reactor (AGR) fuel tests at the Idaho National Laboratory. In 2005, 350-(micro)m, 19.7% 235U-enriched UCO kernels were produced for the AGR-1 test fuel. Following coating of these kernels and forming the coated-particles into compacts, this fuel was irradiated in the Advanced Test Reactor (ATR) from December 2006 until November 2009. B and W produced 425-(micro)m, 14% enriched UCO kernels in 2008, and these kernels were used to produce fuel for the AGR-2 experiment that was inserted in ATR in 2010. B and W also produced 500-(micro)m, 9.6% enriched UO2 kernels for the AGR-2 experiments. Kernels of the same size and enrichment as AGR-1 were also produced for the AGR-3/4 experiment. In addition to fabricating enriched UCO and UO2 kernels, B and W has produced more than 100 kg of natural uranium UCO kernels which are being used in coating development tests. Successive lots of kernels have demonstrated consistent high quality and also allowed for fabrication process improvements. Improvements in kernel forming were made subsequent to AGR-1 kernel production. Following fabrication of AGR-2 kernels, incremental increases in sintering furnace charge size have been demonstrated. Recently small scale sintering tests using a small development furnace equipped with a residual gas analyzer (RGA) has increased understanding of how kernel sintering parameters affect sintered kernel properties. The steps taken to increase throughput and process knowledge have reduced kernel production costs. Studies have been performed of additional modifications toward the goal of increasing capacity of the current fabrication line to use for production of first core fuel for the Next Generation Nuclear Plant (NGNP) and providing a basis for the design of a full scale fuel fabrication facility.
Development of IAEA safeguards at low enrichment uranium fuel fabrication plants
International Nuclear Information System (INIS)
Badawy, I.
1988-01-01
In this report the nuclear material at low enrichment uranium fuel fabrication plants under IAEA safeguards is studied. The current verification practices of the nuclear material and future improvements are also considered. The problems met during the implementation of the the verification measures of the nuclear material - particularly for the fuel assemblies are discussed. The additional verification activities as proposed for future improvements are also discussed including the physical inventory verification and the verification of receipts and shipments. It is concluded that the future development of the present IAEA verification practices at low enrichment uranium fuel fabrication plants would necessitate the application of quantitative measures of the nuclear material and the implementation of advanced measurement techniques and instruments. 2 fig., 4 tab
Fabrication of uranium carbide/beryllium carbide/graphite experimental-fuel-element specimens
International Nuclear Information System (INIS)
Muenzer, W.A.
1978-01-01
A method has been developed for fabricating uranium carbide/beryllium carbide/graphite fuel-element specimens for reactor-core-meltdown studies. The method involves milling and blending the raw materials and densifying the resulting blend by conventional graphite-die hot-pressing techniques. It can be used to fabricate specimens with good physical integrity and material dispersion, with densities of greater than 90% of the theoretical density, and with a uranium carbide particle size of less than 10 μm
Criticality safety studies for plutonium–uranium metal fuel pin fabrication facility
International Nuclear Information System (INIS)
Stephen, Neethu Hanna; Reddy, C.P.
2013-01-01
Highlights: ► Criticality safety limits for PUMP-F facility is identified. ► The fissile mass which can be handled safely during alloy preparation is 10.5 kg. ► The number of fuel slugs which can be handled safely during injection casting is 53. ► The number of fuel slugs which can be handled safely after fuel fabrication is 71. - Abstract: This study focuses on the criticality safety during the fabrication of fast reactor metal fuel pins comprising of the fuel type U–15Pu, U–19Pu and U–19Pu–6Zr in the Plutonium–Uranium Metal fuel Pin fabrication Facility (PUMP-F). Maximum amount of fissile mass which can be handled safely during master alloy preparation, Injection casting and fuel slug preparation following fuel pin fabrication were identified and fixed based on this study. In the induction melting furnace, the fissile mass can be limited to 10.5 kg. During fuel slug preparation and fuel pin fabrication, fuel slugs and pins were arranged in hexagonal and square lattices to identify the most reactive configuration. The number of fuel slugs which can be handled safely after injection casting can be fixed to be 53, whereas after fuel fabrication it is 71
Nondestructive assay of special nuclear material for uranium fuel-fabrication facilities
International Nuclear Information System (INIS)
Smith, H.A. Jr.; Schillebeeckx, P.
1997-01-01
A high-quality materials accounting system and effective international inspections in uranium fuel-fabrication facilities depend heavily upon accurate nondestructive assay measurements of the facility's nuclear materials. While item accounting can monitor a large portion of the facility inventory (fuel rods, assemblies, storage items), the contents of all such items and mass values for all bulk materials must be based on quantitative measurements. Weight measurements, combined with destructive analysis of process samples, can provide highly accurate quantitative information on well-characterized and uniform product materials. However, to cover the full range of process materials and to provide timely accountancy data on hard-to-measure items and rapid verification of previous measurements, radiation-based nondestructive assay (NDA) techniques play an important role. NDA for uranium fuel fabrication facilities relies on passive gamma spectroscopy for enrichment and U isotope mass values of medium-to-low-density samples and holdup deposits; it relies on active neutron techniques for U-235 mass values of high-density and heterogeneous samples. This paper will describe the basic radiation-based nondestructive assay techniques used to perform these measurements. The authors will also discuss the NDA measurement applications for international inspections of European fuel-fabrication facilities
Development of ISA procedure for uranium fuel fabrication and enrichment facilities
International Nuclear Information System (INIS)
Yamate, Kazuki; Arakawa, Tomoyuki; Yamashita, Masahiro; Sasaki, Noriaki; Hirano, Mitsumasa
2011-01-01
The integrated safety analysis (ISA) procedure has been developed to apply risk-informed regulation to uranium fuel fabrication and enrichment facilities. The major development efforts are as follows: (a) preparing the risk level matrix as an index for items-relied-on-for-safety (IROFS) identification, (b) defining requirements of IROFS, and (c) determining methods of IROFS importance based on the results of risk- and scenario-based analyses. For the risk level matrix, the consequence and likelihood categories have been defined by taking into account the Japanese regulatory laws, rules, and safety standards. The trial analyses using the developed procedure have been performed for several representative processes of the reference uranium fuel fabrication and enrichment facilities. This paper presents the results of the ISA for the sintering process of the reference fabrication facility. The results of the trial analyses have demonstrated the applicability of the procedure to the risk-informed regulation of these facilities. (author)
An overview of the regulation of uranium mining, milling, refining and fuel fabrication
International Nuclear Information System (INIS)
Smythe, W.D.
1980-07-01
The mining, milling, refining and fabrication of uranium into nuclear fuel are activities that have in common the handling of natural uranium. The occupational and environmental hazards resulting from these activities vary widely. Uranium presents a radiological hazard throughout, but the principal culprit is radium which creates an occupational hazard in the mine and mill and an environmental hazard in the waste products produced in both the mill and the refinery. The chemicals used in both these latter processes also present hazards. Fuel fabrication presents the least potential for occupational and environmental hazards. The Canadian Atomic Energy Control Board licenses eight plants, and one plant for the extraction of uranium from phosphoric acid. The licensing process is characterised by approval in stages, the placing of the burden of proof on the applicant, inspection at all stages, and joint review by all regulatory agencies involved
Fabrication and testing of uranium nitride fuel for space power reactors
Matthews, R. B.; Chidester, K. M.; Hoth, C. W.; Mason, R. E.; Petty, R. L.
1988-02-01
Uranium nitride fuel was selected for previous space power reactors because of its attractive thermal and physical properties; however, all UN fabrication and testing activities were terminated over ten years ago. An accelerated irradiation test, SP-1, was designed to demonstrate the irradiation performance of Nb-1 Zr clad UN fuel pins for the SP-100 program. A carbothermic-reduction/nitriding process was developed to synthesize UN powders. These powders were fabricated into fuel pellets by conventional cold-pressing and sintering. The pellets were loaded into Nb-1 Zr cladding tubes, irradiated in a fast-test reactor, and destructively examined after 0.8 at% burnup. Preliminary postirradiation examination (PIE) results show that the fuel pins behaved as designed. Fuel swelling, fission-gas release, and microstructural data are presented, and suggestions to enhance the reliability of UN fuel pins are discussed.
Fabrication procedures for manufacturing high uranium concentration dispersion fuel elements
International Nuclear Information System (INIS)
Souza, Jose Antonio Batista de
2011-01-01
IPEN-CNEN/SP developed the technology to produce the dispersion type fuel elements for research reactors and made it available for routine production. Today, the fuel produced in IPEN-CNEN/SP is limited to the uranium concentration of 3.0 gU/cm 3 for U 3 Si 2 -Al dispersion-based and 2.3 gU/cm 3 for U 3 O 8 -Al dispersion. The increase of uranium concentration in fuel plates enables the reactivity of the reactor core reactivity to be higher and extends the fuel life. Concerning technology, it is possible to increase the uranium concentration in the fuel meat up to the limit of 4.8 gU/cm 3 in U 3 Si 2 -Al dispersion and 3.2 gU/cm 3 U 3 O 8 -Al dispersion. These dispersions are well qualified worldwide. This work aims to develop the manufacturing process of both fuel meats with high uranium concentrations, by redefining the manufacturing procedures currently adopted in the Nuclear Fuel Center of IPEN-CNEN/SP. Based on the results, it was concluded that to achieve the desired concentration, it is necessary to make some changes in the established procedures, such as in the particle size of the fuel powder and in the feeding process inside the matrix, before briquette pressing. These studies have also shown that the fuel plates, with a high concentration of U 3 Si 2 -Al, met the used specifications. On the other hand, the appearance of the microstructure obtained from U 3 O 8 -Al dispersion fuel plates with 3.2 gU/cm 3 showed to be unsatisfactory, due to the considerably significant porosity observed. The developed fabrication procedure was applied to U 3 Si 2 production at 4.8 gU/cm 3 , with enriched uranium. The produced plates were used to assemble the fuel element IEA-228, which was irradiated in order to check its performance in the IEA-R1 reactor at IPEN-CNEN/SP. These new fuels have potential to be used in the new Brazilian Multipurpose Reactor - RMB. (author)
International Nuclear Information System (INIS)
Singh, A.K.; Hussain, M.M.; Jayachandran, N.K.; Abdulla, K.K.
2012-01-01
Natural uranium metal or its alloy is used as fuel in nuclear reactors. Usually fuel is clad with compatible material to prevent its direct contact with coolant which prevents spread of activity. One of the methods of producing fuel for nuclear reactor is by co-drawing finished uranium rods with aluminum clad tube to develop intimate contact for effective heat removal during reactor operation. Presently seam welded Aluminium tubes are used as clad for Research Reactor fuel. The paper will highlight entire fabrication process followed for the fabrication of seamless Aluminium finned tubes along with relevant characterisation results
Uranium accountability for ATR fuel fabrication. Part I. A description of the existing system
International Nuclear Information System (INIS)
Dolan, C.A.; Nieschmidt, E.B.; Vegors, S.H. Jr.; Wagner, E.P. Jr.
1977-06-01
An evaluation of the materials accountability program at the Atomics International fuel fabrication facility in Canoga Park, California, with regard to the fabrication of highly enriched uranium fuel for the Advanced Test Reactor is presented. An analysis is given of the existing standards program, the existing measurements program and the existing statistical analysis procedures. In addition a short discussion is given of our evaluation of the safeguards procedures at Atomics International together with suggestions for possible modifications and improvements. Appendices of this report contain a rather complete description of the Atomics International plant and the flow of highly enriched uranium through the plant as well as the principal documents used for material accountability records
Fabrication of high-uranium-loaded U/sub 3/O/sub 8/-Al developmental fuel plates
Energy Technology Data Exchange (ETDEWEB)
Copeland, G.L.; Martin, M.M.
1980-12-01
A common plate-type fuel for research and test reactors is U/sub 3/O/sub 8/ dispersed in aluminum and clad with an aluminum alloy. There is an impetus to reduce the /sup 235/U enrichment from above 90% to below 20% for these fuels to lessen the risk of diversion of the uranium for nonpeaceful uses. Thus, the uranium content of the fuel plates has to be increased to maintain the performance of the reactors. This paper describes work at ORNL to determine the maximal uranium loading for these fuels that can be fabricated with commercially proven materials and techniques and that can be expected to perform satisfactorily in service.
International Nuclear Information System (INIS)
Heremanns, R.H.; Vandersteene, J.J.
1983-01-01
The invention concerns a process for recovery of plutonium from fabrication residues of mixed fuels consisting of uranium oxide and plutonium oxide in the form of PuO 2 . Mixed fuels consisting of uranium oxide and plutonium oxide are being used more and more. The plants which prepare these mixed fuels have around 5% of the total mass of fuels as fabrication residue, either as waste or scrap. In view of the high cost of plutonium, it has been attempted to recover this plutonium from the fabrication residues by a process having a purchase price lower than the price of plutonium. The problem is essentially to separate the plutonium, the uranium and the impurities. The residues are fluorinated, the UF 6 and PuF 6 obtained are separated by selective absorption of the PuF 6 on NaF at a temperature of at least 400 0 C, the complex obtained by this absorption is dissolved in nitric acid solution, the plutonium is precipitated in the form of plutonium oxalate by adding oxalic acid, and the precipitated plutonium oxalate is calcined
Fabrication of uranium-plutonium mixed nitride fuel pins (88F-5A) for first irradiation test at JMTR
International Nuclear Information System (INIS)
Suzuki, Yasufumi; Iwai, Takashi; Arai, Yasuo; Sasayama, Tatsuo; Shiozawa, Ken-ichi; Ohmichi, Toshihiko; Handa, Muneo
1990-07-01
A couple of uranium-plutonium mixed nitride fuel pins was fabricated for the first irradiation tests at JMTR for the purpose of understanding the irradiation behavior and establishing the feasibility of nitride fuels as advanced FBR fuels. The one of the pins was fitted with thermocouples in order to observe the central fuel temperature. In this report, the fabrication procedure of the pins such as pin design, fuel pellet fabrication and characterizations, welding of fuel pins, and inspection of pins are described, together with the outline of the new TIG welder installed recently. (author)
International Nuclear Information System (INIS)
Viswanathan, S.; Surya Rao, B.; Lakshmanan, A.R.; Krishna Rao, T.
1991-01-01
Nuclear Fuel Complex, Hyderabad manufactures uranium dioxide fuel assemblies for PHWRs and BWRs operating in India. Starting materials are magnesium diuranate received from UCIL, Jaduguda and imported UF. Both of these are converted to UO 2 pellets by identical chemical processes and mechanical compacting. Since the uranium handled here is free of daughter product activities, external radiation is not a problem. Inhalation of airborne U compounds is one of the main source of exposure. Engineered protective measures like enclosures around U bearing powder handling equipment and local exhausts reduce worker's exposure. Installation of pre-filters, wet rotoclones and electrostatic precipitators in the ventillation system reduces the release of U into the environment. The criticality hazard in handling slightly enriched uranium is very low due to the built-in control based on geometry and inventory. Where airborne uranium is significant, workers are provided with protective respirators. The workers are regularly monitored for external exposure and also for internal exposure. The environmental releases from the NFC facility is well controlled. Soil, water and air from the NFC environment are routinely collected and analysed for all the possible pollutants. The paper describes the Health Physics experience during the last five years on occupational exposures and on environmental surveillance which reveals the high quality of safety observed in our nuclear fuel fabricating installations. (author). 4 refs., 6 tabs
Criticality Calculations for a Typical Nuclear Fuel Fabrication Plant with Low Enriched Uranium
International Nuclear Information System (INIS)
Elsayed, Hade; Nagy, Mohamed; Agamy, Said; Shaat, Mohmaed
2013-01-01
The operations with the fissile materials such as U 235 introduce the risk of a criticality accident that may be lethal to nearby personnel and can lead the facility to shutdown. Therefore, the prevention of a nuclear criticality accident should play a major role in the design of a nuclear facility. The objectives of criticality safety are to prevent a self-sustained nuclear chain reaction and to minimize the consequences. Sixty criticality accidents were occurred in the world. These are accidents divided into two categories, 22 accidents occurred in process facilities and 38 accidents occurred during critical experiments or operations with research reactor. About 21 criticality accidents including Japan Nuclear Fuel Conversion Co. (JCO) accident took place with fuel solution or slurry and only one accident occurred with metal fuel. In this study the nuclear criticality calculations have been performed for a typical nuclear fuel fabrication plant producing nuclear fuel elements for nuclear research reactors with low enriched uranium up to 20%. The calculations were performed for both normal and abnormal operation conditions. The effective multiplication factor (k eff ) during the nuclear fuel fabrication process (Uranium hexafluoride - Ammonium Diuranate conversion process) was determined. Several accident scenarios were postulated and the criticalities of these accidents were evaluated. The computer code MCNP-4B which based on Monte Carlo method was used to calculate neutron multiplication factor. The criticality calculations Monte Carlo method was used to calculate neutron multiplication factor. The criticality calculations were performed for the cases of, change of moderator to fuel ratio, solution density and concentration of the solute in order to prevent or mitigate criticality accidents during the nuclear fuel fabrication process. The calculation results are analyzed and discussed
Fabrication of high-uranium-loaded U{sub 3}O{sub 8}-Al developmental fuel plates
Energy Technology Data Exchange (ETDEWEB)
Copeland, G L; Martin, M M [Oak Ridge National Laboratory, TN (United States)
1983-08-01
A common plate-type fuel for Research and Test Reactors (RERTR) is U{sub 3}0{sub 8} dispersed in aluminum and clad with an aluminum alloy. There is an impetus to reduce the {sup 235}U enrichment from above 90% to below 20% for these fuels to lessen the risk of diversion of the uranium for non-peaceful uses. Thus, the uranium content of the fuel plates has to be increased to maintain the performance of the reactors. This paper describes work at ORNL to determine the maximal uranium loading for these fuels that can be fabricated with commercially proven materials and techniques and that can be expected to perform satisfactorily in service. We fabricated developmental fuel plates with cores containing from 60 to 100 wt U{sub 3}0{sub 8} in aluminum encapsulated in 6061 aluminum alloy and evaluated them for aspects of fabricability, nondestructive testing, and expected performance. We recommend 75 wt U{sub 3}0{sub 8}-Al 3.1 Mg U/m{sup 3}) as the highest loading in the initial irradiation test. This upper limit is based on a qualitative assessment of the mechanical integrity of the core made by using current fabrication techniques and materials. As the oxide loading is increased beyond this point, planar areas and extensive stringers of oxide and voids develop, which leave little strength in the thickness direction. Fuel plates may then blister over these areas as fission gases collect during irradiation. Current size plates are easily fabricable to the 75 wt % U{sub 3}0{sub 8}-Al core loading by current fabrication techniques. Dogboning is a potential problem at this loading for some applications; however, this can be easily solved by using tapered compact ends. Current nondestructive radiography and transmission x-ray scanning are applicable to the highly loaded plates. Ultrasonic testing for non-bonds is marginal because of the abrupt change in conductance at the cladding-core interface. Plate thickness can be increased if desired; we fabricated 75 wt % plates with
Nuclear fuel fabrication in India
International Nuclear Information System (INIS)
Kondal Rao, N.
1975-01-01
The important role of a nuclear power programme in meeting the growing needs of power in India is explained. The successful installation of Tarapur Atomic Power Station and Rajasthan Atomic Power Station as well as the work at Madras Atomic Power Station are described. The development of the Atomic Fuels Division and the Nuclear Fuel Complex, Hyderabad which is mainly concerned with the fabrication of fuel elements and the reprocessing of fuels are explained. The N.F.C. essentially has the following constituent units : Zirconium Plant (ZP) comprising of Zirconium Oxide Plant, Zirconium Sponge Plant and Zirconium Fabrication Plant; Natural Uranium Oxide Plant (UOP); Ceramic Fuel Fabrication Plant (CFFP); Enriched Uranium Oxide Plant (EUOP); Enriched Fuel Fabrication Plant (EEFP) and Quality Control Laboratory for meeting the quality control requirements of all plants. The capacities of various plants at the NFC are mentioned. The work done on mixed oxide fuels and FBTR core with blanket assemblies, nickel and steel assemblies, thermal research reactor of 100 MW capacity, etc. are briefly mentioned. (K.B.)
Nuclear fuel fabrication in India
Energy Technology Data Exchange (ETDEWEB)
Kondal Rao, N
1975-01-01
The important role of a nuclear power program in meeting the growing needs of power in India is explained. The successful installation of Tarapur Atomic Power Station and Rajasthan Atomic Power Station as well as the work at Madras Atomic Power Station are described. The development of the Atomic Fuels Division and the Nuclear Fuel Complex, Hyderabad which is mainly concerned with the fabrication of fuel elements and the reprocessing of fuels are explained. The N.F.C. essentially has the following constituent units : Zirconium Plant (ZP) comprising of Zirconium Oxide Plant, Zirconium Sponge Plant and Zirconium Fabrication Plant; Natural Uranium Oxide Plant (UOP); Ceramic Fuel Fabrication Plant (CFFP); Enriched Uranium Oxide Plant (EUOP); Enriched Fuel Fabrication Plant (EEFP) and Quality Control Laboratory for meeting the quality control requirements of all plants. The capacities of various plants at the NFC are mentioned. The work done on mixed oxide fuels and FBTR core with blanket assemblies, nickel and steel assemblies, thermal research reactor of 100 MW capacity, etc. are briefly mentioned.
Radiation protection of workers in uranium mining, ore processing and fuel fabrication in India
International Nuclear Information System (INIS)
Khan, A. H.; Jha, G.; Jha, S.; Srivastava, G. K.; Sadasivan, S.; Raj, Venkat
2002-01-01
Low grade of uranium ore mined from three underground mines is processed in a mill at Jaduguda in eastern India to recover uranium concentrate in the form of yellow cake. This concentrate is further processed at the Nuclear Fuel Complex at Hyderabad, in southern India, to produce fuel for use in nuclear power plants. Radiation protection of workers is given due importance at all stages of these operations. Dedicated Health Physics Units and Environmental Survey Laboratories established at each site regularly carry out in-plant and environmental surveillance to keep radiation exposure of workers and the members of public within the limits prescribed by the regulatory body. The limits set by the national regulatory body are based on the international standards suggested by the ICRP and the IAEA. In the uranium mines external gamma radiation, radon and airborne activity due to radioactive dust is monitored. Similarly, in the uranium mill and the fuel fabrication plant gamma radiation and airborne radioactivity due to long-lived α -emitters are monitored. Personal dosimeters are also issued to workers. The total radiation exposure of workers from external and internal sources is evaluated from the personal monitoring and area monitoring data. It has been observed that the total radiation dose to workers has been well below 20 mSv.y 1 at all stages of operations. Adequate ventilation is provided during mining, ore processing and fuel fabrication operations to keep the concentrations of airborne radioactivity well below the derived limits. Workers use personal protective appliances, where necessary, as a supplementary means of control. The monitoring methodologies, results and control measures are presented in the paper
Radiation protection of workers in uranium mining, ore processing and fuel fabrication in India
International Nuclear Information System (INIS)
Khan, A.H.; Jha, G.; Jha, S.; Srivastava, G.K.; Sadasivan, S.; Venkat Raj, V.
2002-01-01
Full text: Low grade of uranium ore mined from three underground mines is processed in a mill at Jaduguda in eastern India to recover uranium concentrate in the form of yellow cake. This concentrate is further processed at the Nuclear Fuel Complex at Hyderabad, in southern India, to produce fuel for use in nuclear power plants. Radiation protection of workers is given due importance at all stages of these operations. Dedicated Health Physics Units and Environmental Survey Laboratories established at each site regularly carry out in-plant and environmental surveillance to keep radiation exposure of workers and the members of public within the limits prescribed by the regulatory body. The limits set by the national regulatory body are based on the international standards suggested by the ICRP and the IAEA. In the uranium mines external gamma radiation, radon and airborne activity due to radioactive dust is monitored. Similarly, in the uranium mill and the fuel fabrication plant gamma radiation and airborne radioactivity due to long-lived a- emitters are monitored. Personal dosimeters are also issued to workers. The total radiation exposure of workers from external and internal sources is evaluated from the personal monitoring and area monitoring data. It has been observed that the total radiation dose to workers has been well below 20 mSvy -1 at all stages of operations. Adequate ventilation is provided during mining, ore processing and fuel fabrication operations to keep the concentrations of airborne radioactivity well below the derived limits. Workers use personal protective appliances, where necessary, as a supplementary means of control. The monitoring methodologies, results and control measures are presented in the paper
International Nuclear Information System (INIS)
Kitts, F.G.
1994-04-01
The Uranium-Atomic Vapor Laser Isotope Separation (U-AVLIS) Program has the objective of developing a cost-competitive enrichment process that will ultimately replace the gaseous diffusion process used in the United States. Current nuclear fuel fabricators are set up to process only the UF 6 product from gaseous diffusion enrichment. Enriched uranium-iron alloy from the U-AVLIS separator system must be chemically converted into an oxide form acceptable to these fabricators to make fuel pellets that meet American Society for Testing and Materials (ASTM) and utility company specifications. A critical step in this conversion is the modified direct denitration (MDD) that has been selected and presented in the AVLIS Conceptual Design for converting purified uranyl nitrate to UO 3 to be shipped to fabricators for making UO 2 pellets for power reactor fuel. This report describes the MDD process, the equipment used, and the experimental work done to demonstrate the conversion of AVLIS product to ceramic-grade UO 3 suitable for making reactor-grade fuel pellets
Study of internal exposure to uranium compounds in fuel fabrication plants in Brazil
International Nuclear Information System (INIS)
Santos, Maristela Souza
2006-01-01
The International Commission on Radiological Protection (ICRP) Publication 66 and Supporting Guidance 3) strongly recommends that specific information on lung retention parameters should be used in preference to default values wherever appropriate, for the derivation of effective doses and for bioassay interpretation of monitoring data. A group of 81 workers exposed to UO 2 at the fuel fabrication facility in Brazil was selected to evaluate the committed effective dose. The workers were monitored for determination of uranium content in the urinary and faecal excretion. The contribution of intakes by ingestion and inhalation were assessed on the basis of the ratios of urinary to fecal excretion. For the selected workers it was concluded that inhalation dominated intake. According to ICRP 66, uranium oxide is classified as insoluble Type S compound. The ICRP Supporting Guidance 3 and some recent studies have recommended specific lung retention parameters to UO 2 . The solubility parameters of the uranium oxide compound handled by the workers at the fuel fabrication facility in Brazil was evaluated on the basis of the ratios of urinary to fecal excretion. Excretion data were corrected for dietary intakes. This paper will discuss the application of lung retention parameters recommended by the ICRP models to these data and also the dependence of the effective committed dose on the lung retention parameters. It will also discuss the problems in the interpretation of monitoring results, when the worker is exposed to several uranium compounds of different solubilities. (author)
Nuclear fuel conversion and fabrication chemistry
International Nuclear Information System (INIS)
Lerch, R.E.; Norman, R.E.
1984-01-01
Following irradiation and reprocessing of nuclear fuel, two operations are performed to prepare the fuel for subsequent reuse as fuel: fuel conversion, and fuel fabrication. These operations complete the classical nuclear fuel cycle. Fuel conversion involves generating a solid form suitable for fabrication into nuclear fuel. For plutonium based fuels, either a pure PuO 2 material or a mixed PuO 2 -UO 2 fuel material is generated. Several methods are available for preparation of the pure PuO 2 including: oxalate or peroxide precipitation; or direct denitration. Once the pure PuO 2 is formed, it is fabricated into fuel by mechanically blending it with ceramic grade UO 2 . The UO 2 can be prepared by several methods which include direct denitration. ADU precipitation, AUC precipitation, and peroxide precipitation. Alternatively, UO 2 -PuO 2 can be generated directly using coprecipitation, direct co-denitration, or gel sphere processes. In coprecipitation, uranium and plutonium are either precipitated as ammonium diuranate and plutonium hydroxide or as a mixture of ammonium uranyl-plutonyl carbonate, filtered and dried. In direct thermal denitration, solutions of uranium and plutonium nitrates are heated causing concentration and, subsequently, direct denitration. In gel sphere conversion, solutions of uranium and plutonium nitrate containing additives are formed into spherical droplets, gelled, washed and dried. Refabrication of these UO 3 -PuO 2 starting materials is accomplished by calcination-reduction to UO 2 -PuO 2 followed by pellet fabrication. (orig.)
Fabrication routes for Thorium and Uranium233 based AHWR fuel
International Nuclear Information System (INIS)
Danny, K.M.; Saraswat, Anupam; Chakraborty, S.; Somayajulu, P.S.; Kumar, Arun
2011-01-01
India's economic growth is on a fast growth track. The growth in population and economy is creating huge demand for energy which has to be met with environmentally benign technologies. Nuclear Energy is best suited to meet this demand without causing undue environmental impact. Considering the large thorium reserves in India, the future nuclear power program will be based on Thorium- Uranium 233 fuel cycle. The major characteristic of thorium as the fuel of future comes from its superior fuel utilization. 233 U produced in a reactor is always contaminated with 232 U. This 232 U undergoes a decay to produce 228 Th and it is followed by decay chain including 212 Bi and 208 Tl. Both 212 Bi and 208 Tl are hard gamma emitters ranging from 0.6 MeV-1.6 MeV and 2.6 MeV respectively, which necessitates its handling in hot cell. The average concentration of 232 U is expected to exceed 1000 ppm after a burn-up of 24,000 MWD/t. Work related to developing the fuel fabrication technology including automation and remotization needed for 233 U based fuels is in progress. Various process for fuel fabrication have been developed i.e. Coated Agglomerate Pelletisation (CAP), impregnation technique (Pellet/Gel), Sol Gel Micro-sphere Pelletisation (SGMP) apart from Powder to Pellet (POP) route. This paper describes each process with respect to its advantages, disadvantages and its amenability to automation and remotisation. (author)
Fabrication procedures for manufacturing high uranium concentration dispersion fuel elements
International Nuclear Information System (INIS)
Souza, J.A.B.; Durazzo, M.
2010-01-01
IPEN developed and made available for routine production the technology for manufacturing dispersion type fuel elements for use in research reactors. However, the fuel produced at IPEN is limited to the uranium concentration of 3.0 gU/cm 3 by using the U 3 Si 2 -Al dispersion. Increasing the uranium concentration of the fuel is interesting by the possibility of increasing the reactor core reactivity and lifetime of the fuel. It is possible to increase the concentration of uranium in the fuel up to the technological limit of 4.8 gU/cm 3 for the U 3 Si 2 -Al dispersion, which is well placed around the world. This new fuel will be applicable in the new Brazilian-Multipurpose Reactor RMB. This study aimed to develop the manufacturing process of high uranium concentration fuel, redefining the procedures currently used in the manufacture of IPEN. This paper describes the main procedures adjustments that will be necessary. (author)
Fabrication procedures for manufacturing high uranium concentration dispersion fuel elements
Energy Technology Data Exchange (ETDEWEB)
Souza, Jose Antonio Batista de; Durazzo, Michelangelo, E-mail: jasouza@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)
2011-07-01
IPEN developed and made available for routine production the technology for manufacturing dispersion type fuel elements for use in research reactors. However, the fuel produced at IPEN is limited to the uranium concentration of 3.0 g U/c m3 by using the U{sub 3}Si{sub 2}-Al dispersion. Increasing the uranium concentration of the fuel is interesting by the possibility of increasing the reactor core reactivity and lifetime of the fuel. It is possible to increase the concentration of uranium in the fuel up to the technological limit of 4.8 g U/c m3 for the U{sub 3}Si{sub 2}-Al dispersion, which is well placed around the world. This new fuel will be applicable in the new Brazilian- Multipurpose Reactor RMB. This study aimed to develop the manufacturing process of high uranium concentration fuel, redefining the procedures currently used in the manufacture of IPEN. This paper describes the main procedures adjustments that will be necessary. (author)
Fabrication of particulate metal fuel for fast burner reactors
International Nuclear Information System (INIS)
Ryu, Ho Jin; Lee, Sun Yong; Kim, Jong Hwan; Woo, Yoon Myung; Ko, Young Mo; Kim, Ki Hwan; Park, Jong Man; Lee, Chan Bok
2012-01-01
U Zr metallic fuel for sodium cooled fast reactors is now being developed by KAERI as a national R and D program of Korea. In order to recycle transuranic elements (TRU) retained in spent nuclear fuel, remote fabrication capability in a shielded hot cell should be prepared. Moreover, generation of long lived radioactive wastes and loss of volatile species should be minimized during the recycled fuel fabrication step. Therefore, innovative fuel concepts should be developed to address the fabrication challenges pertaining to TRU while maintaining good performances of metallic fuel. Particulate fuel concepts have already been proposed and tested at several experimental fast reactor systems and vipac ceramic fuel of RIAR, Russia is one of the examples. However, much less work has been reported for particulate metallic fuel development. Spherical uranium alloy particles with various diameters can be easily produced by the centrifugal atomization technique developed by KAERI. Using the atomized uranium and uranium zirconium alloy particles, we fabricated various kinds of powder pack, powder compacts and sintered pellets. The microstructures and properties of the powder pack and pellets are presented
Uranium production in thorium/denatured uranium fueled PWRs
International Nuclear Information System (INIS)
Arthur, W.B.
1977-01-01
Uranium-232 buildup in a thorium/denatured uranium fueled pressurized water reactor, PWR(Th), was studied using a modified version of the spectrum-dependent zero dimensional depletion code, LEOPARD. The generic Combustion Engineering System 80 reactor design was selected as the reactor model for the calculations. Reactors fueled with either enriched natural uranium and self-generated recycled uranium or uranium from a thorium breeder and self-generated recycled uranium were considered. For enriched natural uranium, concentrations of 232 U varied from about 135 ppM ( 232 U/U weight basis) in the zeroth generation to about 260 ppM ( 232 U/U weight basis) at the end of the fifth generation. For the case in which thorium breeder fuel (with its relatively high 232 U concentration) was used as reactor makeup fuel, concentrations of 232 U varied from 441 ppM ( 232 U/U weight basis) at discharge from the first generation to about 512 ppM ( 232 U/U weight basis) at the end of the fifth generation. Concentrations in freshly fabricated fuel for this later case were 20 to 35% higher than the discharge concentration. These concentrations are low when compared to those of other thorium fueled reactor types (HTGR and MSBR) because of the relatively high 238 U concentration added to the fuel as a denaturant. Excellent agreement was found between calculated and existing experimental values. Nevertheless, caution is urged in the use of these values because experimental results are very limited, and the relevant nuclear data, especially for 231 Pa and 232 U, are not of high quality
Fabrication of HTTR first loading fuel
International Nuclear Information System (INIS)
Kato, S.; Yoshimuta, S.; Hasumi, T.; Sato, K.; Sawa, K.; Suzuki, S.; Mogi, H.; Shiozawa, S.; Tanaka, T.
2001-01-01
This paper summarizes the fabrication of the first loading fuel for HTTR, High Temperature engineering Test Reactor constructed by JAERI, Japan Atomic Energy Research Institute. The fuel fabrication started at the HTR fuel facility of NFI, Nuclear Fuel Industries, Ltd., June 1995. 4,770 fuel rods were fabricated through the fuel kernel, coated fuel particle and fuel compaction process, then 150 fuel elements were assembled in the reactor building December 1997. Fabrication technology for the fuel was established through a lot of R and D activities and fabrication experience of irradiation examination samples spread over about 30 years. Most of all, very high quality and production efficiency of fuel were achieved by the development of the fuel kernel process using the vibration dropping technology, the continuous 4-layer coating process and the automatic compaction process. As for the inspection technology, the development of the automatic measurement equipment for coated layer thickness of a coated fuel particle and uranium content of a fuel compact contributed to the higher reliability and rationalization of the inspection process. The data processing system for the fabrication and quality control, which was originally developed by NFI, made possible not only quick feedback of statistical quality data to the fabrication processes, but also automatic document preparation, such as inspection certificates and accountability control reports. The quality of the first loading fuel fully satisfied the design specifications for the fuel. In particular, average bare uranium fraction and SiC defective fraction of fuel compacts were 2x10 -6 and 8x10 -5 , respectively. According to the preceding irradiation examinations being performed at JMTR, Japan Materials Testing Reactor of JAERI, the specimen sampled from the first loading fuel shows good irradiation performance. (author)
Re-qualification of MTR-type fuel plates fabrication process
International Nuclear Information System (INIS)
Elseaidy, I.M.; Ghoneim, M.M.
2010-01-01
The fabricability issues with increased uranium loading due to use low enrichment of uranium (LEU), i.e. less than 20 % of U 235 , increase the problems which occur during compact manufacturing, roll bonding of the fuel plates, potential difficulty in forming during rolling process, mechanical integrity of the core during fabrication, potential difficulty in meat homogeneity, and the ability to fabricate plates with thicker core as a means of increasing total uranium loading. To produce MTR- type fuel plates with high uranium loading (HUL) and keep the required quality of these plates, many of qualification process must be done in the commissioning step of fuel fabrication plant. After that any changing of the fabrication parameters, for example changing of any of the raw materials, devises, operators, and etc., a re- qualification process should be done in order to keep the quality of produced plates. Objective of the present work is the general description of the activities to be accomplished for re-qualification of manufacturing MTR- type nuclear fuel plates. For each process to be re-qualified, a detailed of re-qualification process were established. (author)
Breeder reactor fuel fabrication system development
International Nuclear Information System (INIS)
Bennett, D.W.; Fritz, R.L.; McLemore, D.R.; Yatabe, J.M.
1981-01-01
Significant progress has been made in the design and development of remotely operated breeder reactor fuel fabrication and support systems (e.g., analytical chemistry). These activities are focused by the Secure Automated Fabrication (SAF) Program sponsored by the Department of Energy to provide: a reliable supply of fuel pins to support US liquid metal cooled breeder reactors and at the same time demonstrate the fabrication of mixed uranium/plutonium fuel by remotely operated and automated methods
Secure Automated Fabrication: an overview of remote breeder fuel fabrication
International Nuclear Information System (INIS)
Nyman, D.H.; Graham, R.A.
1983-10-01
The Secure Automated Fabrication (SAF) line is an automated, remotely controlled breeder fuel pin fabrication process which is to be installed in the Fuels and Materials Examination Facility (FMEF). The FMEF is presently under construction at Hanford and is scheduled for completion in 1984. The SAF line is scheduled for startup in 1987 and will produce mixed uranium-plutonium fuel pins for the Fast Flux Test Facility (FFTF) and the Clinch River Breeder Reactor Plant (CRBRP). The fabrication line and support systems are described
Energy Technology Data Exchange (ETDEWEB)
Sasajima, Hideo; Fuketa, Toyoshi; Ishijima, Kiyomi; Kuroha, Hiroshi; Ikeda, Yoshikazu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Aizawa, Keiichi
1998-08-01
An experiment plan is progressing in the Nuclear Safety Research Reactor (NSRR) to perform pulse-irradiation with uranium-zirconium hydride (U-ZrH{sub x}) fuel. This fuel is widely used in the training research and isotope production reactor of GA (TRIGA). The objectives of the experiment are to determine the fuel rod failure threshold and to investigate fuel behavior under simulated reactivity initiated accident (RIA) conditions. This report summarizes design, fabrication process and inspection data of the test fuel rods before pulse-irradiation. The experiment with U-ZrH{sub x} fuel will realize precise safety evaluation, and improve the TRIGA reactor performance. The data to be obtained in this program will also contribute development of next-generation TRIGA reactor and its safety evaluation. (author)
Automation of potentiometric titration for the determination of uranium in nuclear fuel materials
International Nuclear Information System (INIS)
Kelkar, Anoop; Pandey, Ashish; Kapoor, Y.S.; Kumar, Manish; Singh, Mamta; Fulzele, Ajeet; Prakash, Amrit; Afzal, Mohd; Panakkal, J.P.
2010-01-01
Advanced Fuel Fabrication Facility is fabricating various types of mixed oxide fuels, namely for PHWR, BWR, FBTR and PFBR. Precise determination of uranium in MOX fuel sample is important to get desired burn up in the reactor. The modified Davies and Gray method is routinely used for the potentiometric titration of uranium
International Nuclear Information System (INIS)
Sampei, T.; Sato, T.; Suzuki, N.; Kai, H.; Hirata, Y.
1993-01-01
The major portion of combustible solid wastes generated at LWR nuclear fuel fabrication plants in Japan is incinerated and stored in a warehouse. The uranium content in the incineration residue is higher compared with other categories of wastes, although only a small amount of incineration residue is generated. Hence, in the future uranium should be removed from incineration residues before they are reduced to a level appropriate for the final disposal. A system for processing the incineration residue for uranium removal has been developed and tested based on the information obtained through laboratory experiments and engineering scale tests
Process development and fabrication for sphere-pac fuel rods
International Nuclear Information System (INIS)
Welty, R.K.; Campbell, M.H.
1981-06-01
Uranium fuel rods containing sphere-pac fuel have been fabricated for in-reactor tests and demonstrations. A process for the development, qualification, and fabrication of acceptable sphere-pac fuel rods is described. Special equipment to control fuel contamination with moisture or air and the equipment layout needed for rod fabrication is described and tests for assuring the uniformity of the fuel column are discussed. Fuel retainers required for sphere-pac fuel column stability and instrumentation to measure fuel column smear density are described. Results of sphere-pac fuel rod fabrication campaigns are reviewed and recommended improvements for high throughput production are noted
International Nuclear Information System (INIS)
Perkins, B.L.
1983-01-01
At present in the United States, there are seven commercial light-water reactor uranium fuel fabrication facilities. Effluent wastes from these facilities include uranium, nitrogen, fluorine, and organic-containing compounds. These effluents may be either discharged to the ambient environment, treated and recycled internally, stored or disposed of on-site, sent off-site for treatment and/or recovery, or sent off-site for disposal (including disposal in low-level waste burial sites). Quantities of waste generated and treatment techniques vary greatly depending on the facility and circuits used internally at the facility, though in general all the fluorine entering the facility as UF 6 is discharged as waste. Further studies to determine techniques and procedures that might minimize dose (ALARA) and to give data on possible long-term effects of effluent discharge and waste disposal are needed
Features of spherical uranium-graphite HTGR fuel elements control
International Nuclear Information System (INIS)
Kreindlin, I.I.; Oleynikov, P.P.; Shtan, A.S.
1985-01-01
Control features of spherical HTGR uranium-graphite fuel elements with spherical coated fuel particles are mainly determined by their specific construction and fabrication technology. The technology is chiefly based on methods of ceramic fuel (fuel microspheres fabrication) and graphite production practice it is necessary to deal with a lot of problems from determination of raw materials properties to final fuel elements testing. These procedures are described
Features of spherical uranium-graphite HTGR fuel elements control
Energy Technology Data Exchange (ETDEWEB)
Kreindlin, I I; Oleynikov, P P; Shtan, A S
1985-07-01
Control features of spherical HTGR uranium-graphite fuel elements with spherical coated fuel particles are mainly determined by their specific construction and fabrication technology. The technology is chiefly based on methods of ceramic fuel (fuel microspheres fabrication) and graphite production practice it is necessary to deal with a lot of problems from determination of raw materials properties to final fuel elements testing. These procedures are described.
International Nuclear Information System (INIS)
Miles, T.L.; Liu, Y.
1994-06-01
The cost estimates originally developed in NUREG/CR-1266 for commissioning a reference low-enrichment uranium fuel fabrication plant are updated from 1978 to early 1993 dollars. During this time, the costs for labor and materials increased approximately at the rate of inflation, the cost of energy increased more slowly than the rate of inflation, and the cost of low-level radioactive waste disposal increased much more rapidly than the rate of inflation. The results of the analysis indicate that the estimated costs for the immediate dismantlement and decontamination for unrestricted facility release (DECON) of the reference plant have increased from the mid-1978 value of $3.57 million to $8.08 million in 1993 with in-compact low-level radioactive waste disposal at the US Ecoloay facility near Richland, Washington. The cost estimate rises to $19.62 million with out-of-compact radioactive waste disposal at the Chem-Nuclear facility near Barnwell, South Carolina. A methodology and a formula are presented for estimating the cost of decommissioning the reference uranium fuel fabrication plant at some future time, based on these early 1993 cost estimates. The formula contains essentially the same elements as the formula given in 10 CFR 50.75 for escalating the decommissioning costs for nuclear power reactors to some future time
Fabrication of metallic fuel for fast breeder reactor
International Nuclear Information System (INIS)
Saify, M.T.; Jha, S.K.; Abdulla, K.K.; Kumar, Arbind; Mittal, R.K.; Prasad, R.S.; Mahule, N.; Kumar, Arun; Prasad, G.J.
2012-01-01
Natural uranium oxide fuelled PHWRs comprises of first stage of Indian nuclear power programme. Liquid metal fast breeder reactors fuelled by Pu (from PHWR's) form the second stage. A shorter reactor doubling time is essential in order to accelerate the nuclear power growth in India. Metallic fuels are known to provide shorter doubling times, necessitating to be used as driver fuel for fast breeder reactors. One of the fabrication routes for metallic fuels having random grain orientation, is injection casting technique. The technique finds its basis in an elementary physical concept - the possibility of supporting a liquid column within a tube, by the application of a pressure difference across the liquid interface inside and outside the tube. At AFD, BARC a facility has been set-up for injection casting of uranium rods in quartz tube moulds, demoulding of cast rods, end-shearing of rods and an automated inspection system for inspection of fuel rods with respect to mass, length, diameter and diameter variation along the length and internal and external porosities/voids. All the above facilities have been set-up in glove boxes and have successfully been used for fabrication of uranium bearing fuel rods. The facility has been designed for fabrication and inspection of Pu-bearing metallic fuels also, if required
Application of robotics in remote fuel fabrication operations
International Nuclear Information System (INIS)
Nyman, D.H.; Nagamoto, T.T.
1984-01-01
The Secure Automated Fabrication (SAF) line, an automated and remotely controlled manufacturing process, is scheduled for startup in 1987 and will produce mixed uranium/plutonium oxide fuel pins for the Fast Flux Test Facility (FFTF). The application of robotics in the fuel fabrication and supporting operations is described
Mixed U/Pu oxide fuel fabrication facility co-processed feed, pelletized fuel
International Nuclear Information System (INIS)
1978-09-01
Two conceptual MOX fuel fabrication facilities are discussed in this study. The first facility in the main body of the report is for the fabrication of LWR uranium dioxide - plutonium dioxide (MOX) fuel using co-processed feed. The second facility in the addendum is for the fabrication of co-processed MOX fuel spiked with 60 Co. Both facilities produce pellet fuel. The spiked facility uses the same basic fabrication process as the conventional MOX plant but the fuel feed incorporates a high energy gamma emitter as a safeguard measure against diversion; additional shielding is added to protect personnel from radiation exposure, all operations are automated and remote, and normal maintenance is performed remotely. The report describes the fuel fabrication process and plant layout including scrap and waste processing; and maintenance, ventilation and safety measures
Typical IAEA operations at a fuel fabrication plant
International Nuclear Information System (INIS)
Morsy, S.
1984-01-01
The IAEA operations performed at a typical Fuel Fabrication Plant are explained. To make the analysis less general the case of Low Enriched Uranium (LEU) Fuel Fabrication Plants is considered. Many of the conclusions drawn from this analysis could be extended to other types of fabrication plants. The safeguards objectives and goals at LEU Fuel Fabrication Plants are defined followed by a brief description of the fabrication process. The basic philosophy behind nuclear material stratification and the concept of Material Balance Areas (MBA's) and Key Measurement Points (KMP's) is explained. The Agency operations and verification methods used during physical inventory verifications are illustrated
International light water nuclear fuel fabrication supply. Are fabrication services assured?
International Nuclear Information System (INIS)
Rothwell, Geoffrey
2010-01-01
This paper examines the cost structure of fabricating light water reactor (LWR) fuel with low-enriched uranium (LEU, with less than 5% enrichment). The LWR-LEU fuel industry is decades old, and (except for the high entry cost of designing and licensing a fuel fabrication facility and its fuel), labor and additional fabrication lines can be added at Nth-of-a-Kind cost to the maximum capacity allowed by a site license. The industry appears to be competitive: nuclear fuel fabrication capacity is assured with many competitors and reasonable prices. However, nuclear fuel assurance has become an important issue for nations now to considering new nuclear power plants. To provide this assurance many proposals equate 'nuclear fuel banks' (which would require fuel for specific reactors) with 'LEU banks' (where LEU could be blended into nuclear fuel with the proper enrichment) with local fuel fabrication. The policy issues (which are presented, but not answered in this paper) become (1) whether the construction of new nuclear fuel fabrication facilities in new nuclear power nations could lead to the proliferation of nuclear weapons, and (2) whether nuclear fuel quality can be guaranteed under current industry arrangements, given that fuel failure at one reactor can lead to forced shutdowns at many others. (author)
Chilean fuel elements fabrication progress report
International Nuclear Information System (INIS)
Baeza, J.; Contreras, H.; Chavez, J.; Klein, J.; Mansilla, R.; Marin, J.; Medina, R.
1993-01-01
Due to HEU-LEU core conversion necessity for the Chilean MTR reactors, the Fuel Elements Plant is being implemented to LEU nuclear fuel elements fabrication. A glove box line for powder-compact processing designed at CCHEN, which supposed to operate under an automatic control system, is at present under initial tests. Results of first natural uranium fuel plates manufacturing runs are shown
Minimization of waste from uranium purification, enrichment and fuel fabrication
Energy Technology Data Exchange (ETDEWEB)
NONE
1999-10-01
As any industry, nuclear industry generates a diverse range of waste which has to be managed in a safe manner to be acceptable to the public and the environment. The cost of waste management, the risks to the public and employees, and the detriment to the environment are dependent on the quantity and radioactive content of the waste generated. Waste minimization is a necessary activity needed to reduce the impact from nuclear fuel cycle operations and it is included in the national policy of some countries. In recognition of the importance of the subject, the IAEA has decided to review the current status of the work aimed at waste minimization in the nuclear fuel cycle. The waste minimization issues related to the back end of the nuclear fuel cycle are covered in Technical Reports Series No. 377 'Minimization of Radioactive Waste from Nuclear Power Plants and the Back End of the Nuclear Fuel Cycle' published in 1995. The present report deals with the front end of the nuclear fuel cycle, including existing options, approaches, developments and some specific considerations to be taken into account in decision making on waste minimization. It has been recognized that, in comparison with the back end of the nuclear fuel cycle, much less information is available, and this report should be considered as a first attempt to analyse waste minimization practices and opportunities in uranium purification, conversion, enrichment and fuel fabrication. Although mining and milling is an important part of the front end of the nuclear fuel cycle, these activities are excluded from consideration since relevant activities are covered in other IAEA publications.
Minimization of waste from uranium purification, enrichment and fuel fabrication
International Nuclear Information System (INIS)
1999-10-01
As any industry, nuclear industry generates a diverse range of waste which has to be managed in a safe manner to be acceptable to the public and the environment. The cost of waste management, the risks to the public and employees, and the detriment to the environment are dependent on the quantity and radioactive content of the waste generated. Waste minimization is a necessary activity needed to reduce the impact from nuclear fuel cycle operations and it is included in the national policy of some countries. In recognition of the importance of the subject, the IAEA has decided to review the current status of the work aimed at waste minimization in the nuclear fuel cycle. The waste minimization issues related to the back end of the nuclear fuel cycle are covered in Technical Reports Series No. 377 'Minimization of Radioactive Waste from Nuclear Power Plants and the Back End of the Nuclear Fuel Cycle' published in 1995. The present report deals with the front end of the nuclear fuel cycle, including existing options, approaches, developments and some specific considerations to be taken into account in decision making on waste minimization. It has been recognized that, in comparison with the back end of the nuclear fuel cycle, much less information is available, and this report should be considered as a first attempt to analyse waste minimization practices and opportunities in uranium purification, conversion, enrichment and fuel fabrication. Although mining and milling is an important part of the front end of the nuclear fuel cycle, these activities are excluded from consideration since relevant activities are covered in other IAEA publications
Uranium Fuel Plant. Applicants environmental report
International Nuclear Information System (INIS)
1975-05-01
The Uranium Fuel Plant, located at the Cimarron Facility, was constructed in 1964 with operations commencing in 1965 in accordance with License No. SNM-928, Docket No. 70-925. The plant has been in continuous operation since the issuance of the initial license and currently possesses contracts extending through 1978, for the production of nuclear fuels. The Uranium Plant is operated in conjunction with the Plutonium Facility, each sharing common utilities and sanitary wastes disposal systems. The operation has had little or no detrimental ecological impact on the area. For the operation of the Uranium Fuel Fabrication Plant, initial equipment provided for the production of UO 2 , UF 4 , uranium metal and recovery of scrap materials. In 1968, the plant was expanded by increasing the UO 2 and pellet facilities by the installation of another complete production line for the production of fuel pellets. In 1969, fabrication facilities were added for the production of fuel elements. Equipment initially installed for the recovery of fully enriched scrap has not been used since the last work was done in 1970. Economically, the plant has benefited the Logan County area, with approximately 104 new jobs with an annual payroll of approximately $1.3 million. In addition, $142,000 is annually paid in taxes to state, local and federal governments, and local purchases amount to approximately $1.3 million. This was all in land that was previously used for pasture land, with a maximum value of approximately 37,000 dollars. Environmental effects of plant operation have been minimal. A monitoring and measurement program is maintained in order to ensure that the ecology of the immediate area is not affected by plant operations
Status of high-density fuel plate fabrication
International Nuclear Information System (INIS)
Wiencek, T.C.; Domagala, R.F.; Thresh, H.R.
1991-01-01
Progress has continued on the fabrication of fuel plates with equivalent fuel zone loadings approaching 9 gU/cm 3 . Through hot isostatic pressing (HIP), successful diffusion bonds have been made with 1100 Al and 6061 Al alloys. Although additional study is necessary to optimize the procedure, these bonds demonstrated the most critical processing step for proof-of-concept hardware. Two types of prototype highly loaded fuel plates have been fabricated. The first is a fuel plate in which 0.030-in. (0.76-mm) uranium compound wires are bonded within an aluminum cladding; the second, a dispersion fuel plate with uniform cladding and fuel zone thickness. The successful fabrication of these fuel plates derives from the unique ability of the HIP process to produce diffusion bonds with minimal deformation. (orig.)
Nuclear fuel control in fuel fabrication plants
International Nuclear Information System (INIS)
Seki, Yoshitatsu
1976-01-01
The basic control problems of measuring uranium and of the environment inside and outside nuclear fuel fabrication plants are reviewed, excluding criticality prevention in case of submergence. The occurrence of loss scraps in fabrication and scrap-recycling, the measuring error, the uranium going cut of the system, the confirmation of the presence of lost uranium and the requirement of the measurement control for safeguard make the measurement control very complicated. The establishment of MBA (material balance area) and ICA (item control area) can make clearer the control of inventories, the control of loss scraps and the control of measuring points. Besides the above basic points, the following points are to be taken into account: 1) the method of confirmation of inventories, 2) the introduction of reliable NDT instruments for the rapid check system for enrichment and amount of uranium, 3) the introduction of real time system, and 4) the clarification of MUF analysis and its application to the reliability check of measurement control system. The environment control includes the controls of the uranium concentration in factory atmosphere, the surface contamination, the space dose rate, the uranium concentration in air and water discharged from factories, and the uranium in liquid wastes. The future problems are the practical restudy of measurement control under NPT, the definite plan of burglary protection and the realization of the disposal of solid wastes. (Iwakiri, K.)
International Nuclear Information System (INIS)
Vasudevamurthy, G.; Radecka, A.; Massey, C.
2015-01-01
High-temperature gas-cooled reactor technology is a frontrunner among generation IV nuclear reactor designs. Among the advanced nuclear fuel forms proposed for these reactors, dispersion-type fuel consisting of microencapsulated uranium di-oxide kernels, popularly known as tri-structural isotropic (TRISO) fuel, has emerged as the fuel form of choice. Generation IV gas-cooled fast reactors offer the benefit of recycling nuclear waste with increased burn-ups in addition to producing the required power and hydrogen. Uranium carbide has shown great potential to replace uranium di-oxide for use in these fast spectrum reactors. Uranium carbide microkernels for fast reactor TRISO fuel have traditionally been fabricated by long-duration carbothermic reduction and sintering of precursor uranium dioxide microkernels produced using sol-gel techniques. These long-duration conversion processes are often plagued by issues such as final product purity and process parameters that are detrimental to minor actinide retention. In this context a relatively simple, high-temperature but relatively quick-rotating electrode arc melting method to fabricate microkernels directly from a feedstock electrode was investigated. The process was demonstrated using surrogate tungsten carbide on account of its easy availability, accessibility and the similarity of its melting point relative to uranium carbide and uranium di-oxide.
Energy Technology Data Exchange (ETDEWEB)
Vasudevamurthy, G.; Radecka, A.; Massey, C. [Virginia Commonwealth Univ., Richmond, VA (United States). High Temperature Materials Lab.
2015-07-01
High-temperature gas-cooled reactor technology is a frontrunner among generation IV nuclear reactor designs. Among the advanced nuclear fuel forms proposed for these reactors, dispersion-type fuel consisting of microencapsulated uranium di-oxide kernels, popularly known as tri-structural isotropic (TRISO) fuel, has emerged as the fuel form of choice. Generation IV gas-cooled fast reactors offer the benefit of recycling nuclear waste with increased burn-ups in addition to producing the required power and hydrogen. Uranium carbide has shown great potential to replace uranium di-oxide for use in these fast spectrum reactors. Uranium carbide microkernels for fast reactor TRISO fuel have traditionally been fabricated by long-duration carbothermic reduction and sintering of precursor uranium dioxide microkernels produced using sol-gel techniques. These long-duration conversion processes are often plagued by issues such as final product purity and process parameters that are detrimental to minor actinide retention. In this context a relatively simple, high-temperature but relatively quick-rotating electrode arc melting method to fabricate microkernels directly from a feedstock electrode was investigated. The process was demonstrated using surrogate tungsten carbide on account of its easy availability, accessibility and the similarity of its melting point relative to uranium carbide and uranium di-oxide.
International Nuclear Information System (INIS)
Fishbone, L.G.; Moussalli, G.; Naegele, G.; Ikonomou, P.; Hosoya, M.; Scott, P.; Fager, J.; Sanders, C.; Colwell, D.; Joyner, C.J.
1994-01-01
An approach of short-notice random inspections (SNRIs) for inventory-change verification can enhance the effectiveness and efficiency of international safeguards at natural or low-enriched uranium (LEU) fuel fabrication plants. According to this approach, the plant operator declares the contents of nuclear material items before knowing if an inspection will occur to verify them. Additionally, items about which declarations are newly made should remain available for verification for an agreed time. This report details a six-month field test of the feasibility of such SNRIs which took place at the Westinghouse Electric Corporation Commercial Nuclear Fuel Division. Westinghouse personnel made daily declarations about both feed and product items, uranium hexafluoride cylinders and finished fuel assemblies, using a custom-designed computer ''mailbox''. Safeguards inspectors from the IAEA conducted eight SNRIs to verify these declarations. Items from both strata were verified during the SNRIs by means of nondestructive assay equipment. The field test demonstrated the feasibility and practicality of key elements of the SNRI approach for a large LEU fuel fabrication plant
Shield requirement estimation for pin storage room in fuel fabrication plant
International Nuclear Information System (INIS)
Shanthi, M.M.; Keshavamurthy, R.S.; Sivashankaran, G.
2012-01-01
Fast Reactor Fuel Cycle Facility (FRFCF) is an upcoming project in Kalpakkam. It has the facility to recycle the fuel from PFBR. It is an integrated facility, consists of fuel reprocessing plant, fuel fabrication plant (FFP), core subassembly plant, Reprocessed Uranium plant (RUP) and waste management plant. The spent fuel from PFBR would be reprocessed in fuel reprocessing plant. The reprocessed fuel material would be sent to fuel fabrication plant. The main activity of fuel fabrication plant is the production of MOX fuel pins. The fuel fabrication plant has a fuel pin storage room. The shield requirement for the pin storage room has been estimated by Monte Carlo method. (author)
Fabrication of fuel elements on the basis of increased concentration fuel composition
International Nuclear Information System (INIS)
Alexandrov, A.B.; Afanasiev, V.L.; Enin, A.A.; Suprun, V.B.
2004-01-01
As a part of Russian Program RERTR Reduced Enrichment for Research and Test Reactors), at NCCP, Inc. jointly with the State Scientific Centre VNIINM the mastering in industrial environment of design and fabrication process of fuel elements (FE) with increased concentration fuel compositions is performed. Fuel elements with fuel composition on the basis of dioxide uranium with nearly 4 g/cm 3 fuel concentration have been produced thus confirming the principal possibility of fuel enrichment reduction down to 20% for research reactors which were built up according to the projects of the former USSR, by increasing the oxide fuel concentration in fuel assemblies (FAs). The form and geometrical dimensions of FEs and FAs shall remain unchanged, only uranium mass in FA shall be increased. (author)
International Nuclear Information System (INIS)
Schneider, R.A.; Nilson, R.; Jaech, J.L.
1978-01-01
Today materials accounting in the low enriched conversion-fabrication sector of the LWR fuel cycle is of increased importance. Low enriched uranium is rapidly becoming a precious metal with current dollar values in the range of one dollar per gram comparing with gold and platinum at 7-8 dollars per gram. In fact, people argue that its dollar value exceeds its safeguards value. Along with this increased financial incentive for better material control, the nuclear industry is faced with the impending implementation of international safeguards and increased public attention over its ability to control nuclear materials. Although no quantity of low enriched uranium (LEU) constitutes a practical nuclear explosive, its control is important to international safeguards because of plutonium production or further enrichment to an explosive grade material. The purpose of the paper is to examine and discuss some factors in the area of materials safeguards and accountability as they apply to the low enriched uranium conversion-fabrication sector. The paper treats four main topics: basis for materials accounting; our assessment of the proposed new IAEA requirements; adequacy of current practices; and timing and direction of future modifications
The uranium-plutonium breeder reactor fuel cycle
International Nuclear Information System (INIS)
Salmon, A.; Allardice, R.H.
1979-01-01
All power-producing systems have an associated fuel cycle covering the history of the fuel from its source to its eventual sink. Most, if not all, of the processes of extraction, preparation, generation, reprocessing, waste treatment and transportation are involved. With thermal nuclear reactors more than one fuel cycle is possible, however it is probable that the uranium-plutonium fuel cycle will become predominant; in this cycle the fuel is mined, usually enriched, fabricated, used and then reprocessed. The useful components of the fuel, the uranium and the plutonium, are then available for further use, the waste products are treated and disposed of safely. This particular thermal reactor fuel cycle is essential if the fast breeder reactor (FBR) using plutonium as its major fuel is to be used in a power-producing system, because it provides the necessary initial plutonium to get the system started. In this paper the authors only consider the FBR using plutonium as its major fuel, at present it is the type envisaged in all, current national plans for FBR power systems. The corresponding fuel cycle, the uranium-plutonium breeder reactor fuel cycle, is basically the same as the thermal reactor fuel cycle - the fuel is used and then reprocessed to separate the useful components from the waste products, the useful uranium and plutonium are used again and the waste disposed of safely. However the details of the cycle are significantly different from those of the thermal reactor cycle. (Auth.)
Fabrication of Cerium Oxide and Uranium Oxide Microspheres for Space Nuclear Power Applications
Energy Technology Data Exchange (ETDEWEB)
Jeffrey A. Katalenich; Michael R. Hartman; Robert C. O' Brien
2013-02-01
Cerium oxide and uranium oxide microspheres are being produced via an internal gelation sol-gel method to investigate alternative fabrication routes for space nuclear fuels. Depleted uranium and non-radioactive cerium are being utilized as surrogates for plutonium-238 (Pu-238) used in radioisotope thermoelectric generators and for enriched uranium required by nuclear thermal rockets. While current methods used to produce Pu-238 fuels at Los Alamos National Laboratory (LANL) involve the generation of fine powders that pose a respiratory hazard and have a propensity to contaminate glove boxes, the sol-gel route allows for the generation of oxide microsphere fuels through an aqueous route. The sol-gel method does not generate fine powders and may require fewer processing steps than the LANL method with less operator handling. High-quality cerium dioxide microspheres have been fabricated in the desired size range and equipment is being prepared to establish a uranium dioxide microsphere production capability.
Prospect of Uranium Silicide fuel element with hypostoichiometric (Si ≤3.7%)
International Nuclear Information System (INIS)
Suripto, A.; Sardjono; Martoyo
1996-01-01
An attempt to obtain high uranium-loading in silicide dispersion fuel element using the fabrication technology applicable nowadays can reach Uranium-loading slightly above 5 gU/cm 3 . It is difficult to achieve a higher uranium-loading than that because of fabricability constraints. To overcome those difficulties, the use of uranium silicide U 3 Si based is considered. The excess of U is obtained by synthesising U 3 Si 2 in Si-hypostoichiometric stage, without applying heat treatment to the ingot as it can generate undesired U 3 Si. The U U will react with the matrix to form U al x compound, that its pressure is tolerable. This experiment is to consider possibilities of employing the U 3 Si 2 as nuclear fuel element which have been performed by synthesising U 3 Si 2 -U with the composition of 3.7 % weigh and 3 % weigh U. The ingot was obtained and converted into powder form which then was fabricated into experimental plate nuclear fuel element. The interaction between free U and Al-matrix during heat-treatment is the rolling phase of the fuel element was observed. The study of the next phase will be conducted later
Basic research on high-uranium density fuels for research and test reactors
International Nuclear Information System (INIS)
Ugajin, M.; Itoh, A.; Akabori, M.
1992-01-01
High-uranium density fuels, uranium silicides (U 3 Si 2 , U 3 Si) and U 6 Me-type uranium alloys (Me = Fe, Mn, Ni), were prepared and examined metallurgically as low-enriched uranium (LEU) fuels for research and test reactors. Miniature aluminum-dispersion plate-type fuel (miniplate) and aluminum-clad disk-type fuel specimens were fabricated and subjected to the neutron irradiation in JMTR (Japan Materials Testing Reactor). Fuel-aluminum compatibility tests were conducted to elucidate the extent of reaction and to identify reaction products. The relative stability of the fuels in an aluminum matrix was established at 350degC or above. Experiments were also performed to predict the chemical form of the solid fission-products in the uranium silicide (U 3 Si 2 ) simulating a high burnup anticipated for reactor service. (author)
International Nuclear Information System (INIS)
Kalsi, P.C.; Bhanu, A.U.; Sahoo, S.; Iyer, R.H.
1986-01-01
A passive gamma-ray spectroscopic method is employed for the assay of uranium in fused salt cake, a scrap produced at the natural uranium metal fuel fabrication plants. The method makes use of NaI(TI) detector coupled with a multichannel analyser. The 1 MeV gamma-ray of 238 U was used for the calibration. The calibration curve was made by counting synthetic mixtures made of U 3 O 8 powder, the heat treatment salt and iron in the form of fine powder. The uranium content in these synthetic mixtures was kept in the range of 1-11 per cent. 23 lots of the fused salt cake taken from three different batches of the salt cake were then analysed by this method. The uranium content of fused salt cake was found to be in the range of 1.70-11.43 per cent. To compare the gamma spectrometric results with a completely independent method, chemical analysis of all the fused salt cakes were also carried out. The NDA results were found to agree within ± 17 per cent with the chemical analysis results. (author)
Uranium and plutonium distribution in unirradiated mixed oxide fuel from industrial fabrication
International Nuclear Information System (INIS)
Hanus, D.; Kleykamp, H.
1982-01-01
Different process variants developed in the last few years by the firm ALKEM to manufacture FBR and LWR mixed oxide fuel are given. The uranium and plutonium distribution is determined on the pellets manufactured with the help of the electron beam microprobe. The stepwise improvement of the uranium-plutonium homogeneity in the short-term developed granulate variants and in the long-term developed new processes are illustrated starting with early standard processes for FBR fuel. An almost uniform uranium-plutonium distribution could be achieved for the long-term developed new processes (OKOM, AuPuC). The uranium-plutonium homogeneity are quantified in the pellets manufactured according to the considered process variants with a newly defined quality number. (orig.)
International Nuclear Information System (INIS)
Reardon, P.T.; Mullen, M.F.; Harms, N.L.
1981-02-01
As part of Task C.35 (Calculation of Parameters for Inspection Planning and Evaluation) of the US Program of Technical Assistance to IAEA Safeguards, Pacific Northwest Laboratory has performed some quantitative analyses of IAEA inspection activities at low-enriched uranium (LEU) conversion and fuel fabrication facilities. This report presents the results and conclusions of those analyses. Implementation of IAEA safeguards at LEU conversion and fuel fabrication facilities must take into account a variety of practical problems and constraints. One of the key concerns is the problem of flow verification, especially product verification. The objective of this report is to help put the problem of flow verification in perspective by presenting the results of some specific calculations of inspection effort and probability of detection for various product measurement strategies. In order to provide quantitative information about the advantages and disadvantages of the various strategies, eight specific cases were examined
Characterization of aerosols from industrial fabrication of mixed-oxide nuclear reactor fuels
International Nuclear Information System (INIS)
Hoover, M.D.; Newton, G.J.
1997-01-01
Recycling plutonium into mixed-oxide (MOX) fuel for nuclear reactors is being given serious consideration as a safe and environmentally sound method of managing plutonium from weapons programs. Planning for the proper design and safe operation of the MOX fuel fabrication facilities can take advantage of studies done in the 1970s, when recycling of plutonium from nuclear fuel was under serious consideration. At that time, it was recognized that the recycle of plutonium and uranium in irradiated fuel could provide a significant energy source and that the use of 239 Pu in light water reactor fuel would reduce the requirements for enriched 235 U as a reactor fuel. It was also recognized that the fabrication of uranium and plutonium reactor fuels would not be risk-free. Despite engineered safety precautions such as the handling of uranium and plutonium in glove-box enclosures, accidental releases of radioactive aerosols from normal containment might occur. Workers might then be exposed to the released materials by inhalation
International Nuclear Information System (INIS)
Alexandrov, A.B.; Afanasiev, V.L.; Enin, A.A.; Suprun, V.B.
1996-01-01
According to the Russian Reduced Enrichment for Research and Test Reactors (RERTR) program, that were constructed under the Russian projects, at the Novosibirsk Chemical Concentrates Plant the pilot series of different configuration (WR-M2, MR, IRT-4M) fuel elements, based on increased concentration uranium dioxide fuel, have been fabricated for reactor tests. Comprehensive fabricated fuel elements quality estimation has been carried out. (author)
Fuel Cycle Impacts of Uranium-Plutonium Co-extraction
International Nuclear Information System (INIS)
Taiwo, Temitope; Szakaly, Frank; Kim, Taek-Kyum; Hill, Robert
2008-01-01
A systematic investigation of the impacts of uranium and plutonium co-extraction during fuel separations on reactor performance and fuel cycle has been performed. Proliferation indicators, critical mass and radiation source levels of the separation products or fabricated fuel, were also evaluated. Using LWR-spent-uranium-based MOX fuel instead of natural-uranium-based fuel in a PWR MOX core requires a higher initial plutonium content (∼1%), and results in higher Np-237 content (factor of 5) in the spent fuel, and less consumption of Pu-238 (20%) and Am-241 (14%), indicating a reduction in the effective repository space utilization. Additionally, minor actinides continue to accumulate in the fuel cycle, and thus a separate solution is required for them. Differences were found to be quite smaller (∼0.4% in initial transuranics) between the equilibrium cycles of advanced fast reactor cores using spent and depleted uranium for make-up, in additional to transuranics. The critical masses of the co-extraction products were found to be higher than for weapons-grade plutonium (WG-Pu) and the decay heat and radiation sources of the materials (products) were also found to be generally higher than for WG-Pu in the transuranics content range of 10% to 100% in the heavy-metal. (authors)
Development of high uranium-density fuels for use in research reactors
International Nuclear Information System (INIS)
Ugajin, Mitsuhiro; Akabori, Mitsuo; Itoh, Akinori
1996-01-01
The uranium silicide U 3 Si 2 possesses uranium density 11.3 gU/cm 3 with a congruent melting point of 1665degC, and is now successfully in use as a research reactor fuel. Another uranium silicide U 3 Si and U 6 Me-type uranium alloys (Me=Fe,Mn,Ni) have been chosen as new fuel materials because of the higher uranium densities 14.9 and 17.0 gU/cm 3 , respectively. Experiments were carried out to fabricate miniature aluminum-dispersion plate-type and aluminum-clad disk-type fuels by using the conventional picture-frame method and a hot-pressing technique, respectively. These included the above-mentioned new fuel materials as well as U 3 Si 2 . Totally 14 miniplates with uranium densities from 4.0 to 6.3 gU/cm 3 of fuel meat were prepared together with 28 disk-type fuel containing structurally-modified U 3 Si, and subjected to the neutron irradiation in JMTR (Japan Materials Testing Reactor). Some results of postirradiation examinations are presented. (author)
Uranium dioxide and beryllium oxide enhanced thermal conductivity nuclear fuel development
International Nuclear Information System (INIS)
Andrade, Antonio Santos; Ferreira, Ricardo Alberto Neto
2007-01-01
The uranium dioxide is the most used substance as nuclear reactor fuel for presenting many advantages such as: high stability even when it is in contact with water in high temperatures, high fusion point, and high capacity to retain fission products. The conventional fuel is made with ceramic sintered pellets of uranium dioxide stacked inside fuel rods, and presents disadvantages because its low thermal conductivity causes large and dangerous temperature gradients. Besides, the thermal conductivity decreases further as the fuel burns, what limits a pellet operational lifetime. This research developed a new kind of fuel pellets fabricated with uranium dioxide kernels and beryllium oxide filling the empty spaces between them. This fuel has a great advantage because of its higher thermal conductivity in relation to the conventional fuel. Pellets of this kind were produced, and had their thermophysical properties measured by the flash laser method, to compare with the thermal conductivity of the conventional uranium dioxide nuclear fuel. (author) (author)
Fabrication technology of spherical fuel element for HTR-10
International Nuclear Information System (INIS)
He Jun; Zou Yanwen; Liang Tongxiang; Qiu Xueliang
2002-01-01
R and D on the fabrication technology of the spherical fuel elements for the 10 MW HTR Test Module (HTR-10) began from 1986. Cold quasi-isostatic molding with a silicon rubber die is used for manufacturing the spherical fuel elements.The fabrication technology and the graphite matrix materials were investigated and optimized. Twenty five batches of fuel elements, about 11000 of the fuel elements, have been produced. The cold properties of the graphite matrix materials satisfied the design specifications. The mean free uranium fraction of 25 batches was 5 x 10 -5
Radiation protection training at uranium hexafluoride and fuel fabrication plants
International Nuclear Information System (INIS)
Brodsky, A.; Soong, A.L.; Bell, J.
1985-05-01
This report provides general information and references useful for establishing or operating radiation safety training programs in plants that manufacture nuclear fuels, or process uranium compounds that are used in the manufacture of nuclear fuels. In addition to a brief summary of the principles of effective management of radiation safety training, the report also contains an appendix that provides a comprehensive checklist of scientific, safety, and management topics, from which appropriate topics may be selected in preparing training outlines for various job categories or tasks pertaining to the uranium nuclear fuels industry. The report is designed for use by radiation safety training professionals who have the experience to utilize the report to not only select the appropriate topics, but also to tailor the specific details and depth of coverage of each training session to match both employee and management needs of a particular industrial operation. 26 refs., 3 tabs
International Nuclear Information System (INIS)
Bastin, J.J.; Jump, M.J.; Lange, R.A.; Crandall, C.C.
1977-11-01
The Phase I study of the application of advanced accountability methods (DYMAC) in a uranium/plutonium mixed oxide facility was extended to cover the possible fabrication of uranium-233/thorium fuels. Revisions to Phase II of the DYMAC plan which would be necessitated by such a process are specified. These revisions include shielding requirements, measurement systems, licensing conditions, and safeguards considerations. The impact of the uranium/thorium cycle on a large-scale fuel fabrication facility was also reviewed; it was concluded that the essentially higher radioactivity of uranium/thorium feeds would lead to increased difficulties which tend to preclude early commercial application of the process. An amended schedule for Phase II is included
Development of very high-density low-enriched uranium fuels
International Nuclear Information System (INIS)
Snelgrove, J.L.; Hofman, G.L.; Trybus, C.L.; Wiencek, T.C.
1997-02-01
The RERTR program has recently begun an aggressive effort to develop dispersion fuels for research and test reactors with uranium densities of 8 to 9 g U/cm 3 , based on the use of γ-stabilized uranium alloys. Fabrication development teams and facilities are being put into place and preparations for the first irradiation test are in progress. The first screening irradiations are expected to begin in late April 1997 and first results should be available by end of 1997. Discussions with potential international partners in fabrication development and irradiation testing have begun
Zirconia based inert matrix fuel: fabrication concepts and feasibility studies
International Nuclear Information System (INIS)
Ingold, F.; Burghartz, M.; Ledergerber, G.
1999-01-01
The internal gelation process has traditionally been applied to fabricate standard fuel based on uranium, typically UO2 and MOX. To meet the recent aim to destroy plutonium in the most effective way, a uranium free fuel was evaluated. The fuel development programme at PSI has been redirected toward a fuel based on zirconium oxide or a mixture of zirconia and a conducting material to form ceramic/metal (CERMET) or ceramic/ceramic (CERCER) combinations. A feasibility study was carried out to demonstrate that microspheres based on zirconia and spinel can be fabricated with the required properties. The gelation parameters were investigated to optimise compositions of the starting solutions. Studies to fabricate a composite material (from zirconia and spinel) are ongoing. If the zirconia/spinel ratio is chosen appropriately, the low thermal conductivity of pure zirconia can be compensated by the higher thermal conductivity of spinel. Another solution to offset the low thermal conductivity of zirconia is the development of a CERMET, which consists of fine particles bearing plutonium in a cubic zirconia lattice dispersed in a metallic matrix. The fabrication of such a CERMET is also being studied. (author)
International Nuclear Information System (INIS)
Nishiyama, A.
1993-01-01
This paper contains a brief summary of activities in the field of research reactor fuel fabrication in Nuclear Fuel Industries Sumitomo and Furukawa Industries. Since 1956 2 million dollars were spent for development of nuclear fuels and plant facilities including complete manufacturing and testing capabilities. Now this company is the only fuel supplier for the research reactors in Japan. The fabrication process starts with the melting, alloying, and casting of U-Al. The uranium billets are prepared by foreign fabricators. The uranium content varies from 13 to 22 wt % according to the purchaser's specifications. In making fuel plates, the picture frame method is applied. In this case, the original procedure is sufficiently effective in avoiding dogboning. The plates are finished by hot and cold roll milling and inspected dimensionally, metallurgically, and mechanically, and at the same time the blister test and X-ray radiographic tests are performed. Fuel elements are assembled by rolling flat or curved plates into side plate grooves and end-fit welding. Finished elements are tested dimensionally and hydraulically. Nominal losses during operation are less than 1% of the uranium metal. Our present capacity licensed by the Japanese Government is approximately 950 fuel elements a year. About 35 employees including engineers are engaged in development and manufacturing of fuels. Owing to the small limited demand of the research reactor fuels in Japan during the past 20 years (mostly in last 10 years), we processed only about 350 kg of highly enriched uranium and supplied approximately 1000 fuel elements to JAERI, Kyoto University, and others, and we have been suffering red-ink balance of budget every year. Some of trials in development are briefly discussed. In case of UO 2 -Al metal fuel plates, the vibratory compacting method was very popular among many researchers about 10 years ago. A lot of time and money was spent to study the economic fabrication process of
Development of high uranium-density fuels for use in research reactors
Energy Technology Data Exchange (ETDEWEB)
Ugajin, Mitsuhiro; Akabori, Mitsuo; Itoh, Akinori [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
1996-02-01
The uranium silicide U{sub 3}Si{sub 2} possesses uranium density 11.3 gU/cm{sup 3} with a congruent melting point of 1665degC, and is now successfully in use as a research reactor fuel. Another uranium silicide U{sub 3}Si and U{sub 6}Me-type uranium alloys (Me=Fe,Mn,Ni) have been chosen as new fuel materials because of the higher uranium densities 14.9 and 17.0 gU/cm{sup 3}, respectively. Experiments were carried out to fabricate miniature aluminum-dispersion plate-type and aluminum-clad disk-type fuels by using the conventional picture-frame method and a hot-pressing technique, respectively. These included the above-mentioned new fuel materials as well as U{sub 3}Si{sub 2}. Totally 14 miniplates with uranium densities from 4.0 to 6.3 gU/cm{sup 3} of fuel meat were prepared together with 28 disk-type fuel containing structurally-modified U{sub 3}Si, and subjected to the neutron irradiation in JMTR (Japan Materials Testing Reactor). Some results of postirradiation examinations are presented. (author)
LEU fuel fabrication in Argentina
International Nuclear Information System (INIS)
Giorsetti, D.R.; Gomez, J.O.; Marajofsky, A.; Kohut, C.
1985-01-01
As an Institution, aiming to meet with its own needs, CNEA has been intensively developing reduced enriched fuel to use in its own research and test reactors. Development of the fabrication technology as well as the design, installation and operation of the manufacturing plant, have been carried out with its own funds. Irradiation and post-irradiation of test miniplates have been taking place within the framework of the RERTR program. During the last years, CNEA has developed three LEU fuel types. In the previous RERTR meetings, we presented the technological results obtained with these fuel types. This paper focuses on CNEA LEU fuel element manufacturing status and the trained personnel we can offer in design and manufacture fuel capability. CNEA has its own fuel manufacturing technology; the necessary facilities to start the fuel fabrication; qualified technicians and professionals for: fuel design and behaviour analysis; fuel manufacturing and QA; international recognition of its fuel development and manufacturing capability through its ORR miniplate irradiation; its own natural uranium and the future possibility to enrich up to 20% U 235 ; the probability to offer a competitive fuel manufacturing cost in the international market; the disposition to cooperate with all countries that wish to take part and aim to reach an self-sufficiency in their own fuel supply needs
The Role of Friction Stir Welding in Nuclear Fuel Plate Fabrication
International Nuclear Information System (INIS)
Burkes, D.; Medvedev, P.; Chapple, M.; Amritkar, A.; Wells, P.; Charit, I
2009-01-01
The friction bonding process combines desirable attributes of both friction stir welding and friction stir processing. The development of the process is spurred on by the need to fabricate thin, high density, reduced enrichment fuel plates for nuclear research reactors. The work seeks to convert research and test reactors currently operating on highly enriched uranium fuel to operate on low enriched uranium fuel without significant loss in reactor performance, safety characteristics, or significant increase in cost. In doing so, the threat of global nuclear material proliferation will be reduced. Feasibility studies performed on the process show that this is a viable option for mass production of plate-type nuclear fuel. Adapting the friction stir weld process for nuclear fuel fabrication has resulted in the development of several unique ideas and observations. Preliminary results of this adaptation and process model development are discussed
Competition still fierce in the US fuel fabrication market
International Nuclear Information System (INIS)
Schwartz, M.H.
1990-01-01
The US market for nuclear fuel fabrication services is characterized by an annual production capacity significantly in excess of both current and anticipated demand. The trends toward longer operating cycle lengths and higher burnup fuel continue in the United States. This, together with the lack of any prospects for new light water reactors coming on line in the US during the next ten years, is expected to hold the annual demand for fuel fabrication services from US LWRs at around 2000t of uranium into the next century. (author)
International Nuclear Information System (INIS)
Nolan, A.M.; Lewallen, M.A.; McNair, G.W.
1979-11-01
Environmental control aspects of alternative fuel cycles have been analyzed by evaluating fabrication, reprocessing, and waste disposal operations. Various indices have been used to assess potential environmental control requirements. For the fabrication and reprocessing operations, 50-year dose commitments were used. Waste disposal was evaluated by comparing projected nuclide concentrations in ground water at various time periods with maximum permissible concentrations (MPCs). Three different fabrication plants were analyzed: a fuel fabrication plant (FFP) to produce low-activity uranium and uranium-thorium fuel rods; a plutonium fuel refabrication plant (PFRFP) to produce plutonium-uranium and plutonium-thorium fuel rods; and a uranium fuel refabrication plant (UFRFP) to produce fuel rods containing the high-activity isotopes 232 U and 233 U. Each plant's dose commitments are discussed separately. Source terms for the analysis of effluents from the fuel reprocessing plant (FRP) were calculated using the fuel burnup codes LEOPARD, CINDER and ORIGEN. Effluent quantities are estimated for each fuel type. Bedded salt was chosen for the waste repository analysis. The repository site is modeled on the Waste Isolation Pilot Program site in New Mexico. Wastes assumed to be stored in the repository include high-level vitrified waste from the FRP, packaged fuel residue from the FRP, and transuranic (TRU) contaminated wastes from the FFP, PFRFP, and UFRFP. The potential environmental significance was determined by estimating the ground-water concentrations of the various nuclides over a time span of a million years. The MPC for each nuclide was used along with the estimated ground-water concentration to generate a biohazard index for the comparison among fuel compositions
Atomics International fuel fabrication facility and low enrichment program. Part 2
International Nuclear Information System (INIS)
Hassel, H.W.
1993-01-01
Most of you know our company from the last meeting in May in Vienna, so I won't steal your time with explaining and demonstrating the same techniques that we have heard this morning f rom the other speakers. I would just take some words to explain the order of business with highly enriched uranium. NUKEM handles around almost two tons of highly enriched uranium a year and it was necessary to satisfy all the new physical protection philosophies. That means that we have to install storage and safe fabrication sites for a lot of money, 2.5 meter thick concrete walls, and different alarm systems. So just to demonstrate how silly this business is, we have just overcome this for highly enriched uranium, and now we speak about low enriched uranium for which we don't need all of these investments to make this business safe. I would just like to concentrate my words on the status of fabrication and considerations in my company concerning the medium enriched uranium and low enriched uranium. In TABLE I are the different fuel types (see column 1) and then we have the fabrication in column 2; (The reason that I use the blackboard this morning is that I try to demonstrate all the techniques. However, all the speakers before me did this and in theory we are not so far away from each other.) the experience of my company in kg. In column 3 is the irradiation experience of these fuels types. Column 4 shows the studies and calculations made in our company for lower and medium enriched fuels. The preliminary fabrication tests and calculations are in column 5, and in column 6 we have the delivery time for a prototype core in months after UF 6 supply. Column 7 shows the time for the development of specifications including irradiation time in years for 6 and 7, and column 8 is the estimated cost of 6 and 7. There is just one fuel that is not in this summary and that is U-Zr
Evaluation of plutonium, uranium, and thorium use in power reactor fuel cycles
International Nuclear Information System (INIS)
Kasten, P.R.; Homan, F.J.
1977-01-01
The increased cost of uranium and separative work has increased the attractiveness of plutonium use in both uranium and thorium fuel cycles in thermal reactors. A technology, fuel utilization, and economic evaluation is given for uranium and thorium fuel cycles in various reactor types, along with the use of plutonium and 238 U. Reactors considered are LWRs, HWRs, LWBRs, HTGRs, and FBRs. Key technology factors are fuel irradiation performance and associated physical property values. Key economic factors are unit costs for fuel fabrication and reprocessing, and for refabrication of recycle fuels; consistent cost estimates are utilized. In thermal reactors, the irradiation performance of ceramic fuels appears to be satisfactory. At present costs for uranium ore and separative work, recycle of plutonium with thorium rather than uranium is preferable from fuel utilization and economic viewpoints. Further, the unit recovery cost of plutonium is lower from LWR fuels than from natural-uranium HWR fuels; use of LWR product permits plutonium/thorium fueling to compete with uranium cycles. Converting uranium cycles to thorium cycles increases the energy which can be extracted from a given uranium resource. Thus, additional fuel utilization improvement can be obtained by fueling all thermal reactors with thorium, but this requires use of highly enriched uranium; use of 235 U with thorium is most economic in HTGRs followed by HWRs and then LWRs. Marked improvement in long-term fuel utilization can be obtained through high thorium loadings and short fuel cycle irradiations as in the LWBR, but this imposes significant economic penalties. Similar operating modes are possible in HWRs and HTGRs. In fast reactors, use of the plutonium-uranium cycle gives advantageous fuel resource utilization in both LMFBRs and GCFRs; use of the thorium cycle provides more negative core reactivity coefficients and more flexibility relative to use of recycle fuels containing uranium of less than 20
CONCEPTUAL PROCESS DESCRIPTION FOR THE MANUFACTURE OF LOW-ENRICHED URANIUM-MOLYBDENUM FUEL
Energy Technology Data Exchange (ETDEWEB)
Daniel M. Wachs; Curtis R. Clark; Randall J. Dunavant
2008-02-01
The National Nuclear Security Agency Global Threat Reduction Initiative (GTRI) is tasked with minimizing the use of high-enriched uranium (HEU) worldwide. A key component of that effort is the conversion of research reactors from HEU to low-enriched uranium (LEU) fuels. The GTRI Convert Fuel Development program, previously known as the Reduced Enrichment for Research and Test Reactors program was initiated in 1978 by the United States Department of Energy to develop the nuclear fuels necessary to enable these conversions. The program cooperates with the research reactors’ operators to achieve this goal of HEU to LEU conversion without reduction in reactor performance. The programmatic mandate is to complete the conversion of all civilian domestic research reactors by 2014. These reactors include the five domestic high-performance research reactors (HPRR), namely: the High Flux Isotope Reactor at the Oak Ridge National Laboratory, the Advanced Test Reactor at the Idaho National Laboratory, the National Bureau of Standards Reactor at the National Institute of Standards and Technology, the Missouri University Research Reactor at the University of Missouri–Columbia, and the MIT Reactor-II at the Massachusetts Institute of Technology. Characteristics for each of the HPRRs are given in Appendix A. The GTRI Convert Fuel Development program is currently engaged in the development of a novel nuclear fuel that will enable these conversions. The fuel design is based on a monolithic fuel meat (made from a uranium-molybdenum alloy) clad in Al-6061 that has shown excellent performance in irradiation testing. The unique aspects of the fuel design, however, necessitate the development and implementation of new fabrication techniques and, thus, establishment of the infrastructure to ensure adequate fuel fabrication capability. A conceptual fabrication process description and rough estimates of the total facility throughput are described in this document as a basis for
Possibility of using metal uranium fuel in heavy water reactors
International Nuclear Information System (INIS)
Djuric, B.; Mihajlovic, A.; Drobnjak, Dj.
1965-01-01
The review of metal uranium properties including irradiation in the reactor core lead to the following conclusions. Using metal uranium in the heavy water reactors would be favourable from economic point of view for ita high density, i.e. high conversion factor and low cost of fuel elements fabrication. Most important constraint is swelling during burnup and corrosion
International Nuclear Information System (INIS)
Sears, D.F.; Berthiaume, L.C.; Herbert, L.N.
1987-01-01
The current status of Chalk River Nuclear Laboratories' (CRNL) program to develop and test low-enriched uranium (LEU), proliferation-resistant fuels for use in research reactors is reviewed. CRNL's fuel manufacturing process has been qualified by the successful demonstration irradiation of 7 full-size rods in the NRU reactor. Now industrial-scale production equipment has been commissioned, and a fuel-fabrication campaign for 30 NRU rods and a MAPLE-X core is underway. Excess capacity could be used for commercial fuel fabrication. In the irradiation testing program, mini-elements with deliberately included core surface defects performed well in-reactor, swelling by only 7 to 8 vol% at 93 atomic percent burnup of the original U-235. The additional restraint provided by the aluminium cladding which flowed into the defects during extrusion contributed to this good performance. Mini-elements containing a variety of particle size distributions were also successfully irradiated to 93 at% burnup in NRU, as part of a study to establish the optimum particle size distribution. Swelling was found to be proportional to the percentage of fines (<44μm particles) contained in the cores. The mini-elements containing the composition normally used at CRNL had swollen by 5.8 vol%, and mini-elements with a much higher percentage of fines had swollen by 6.8 vol%, at 93 at% burnup. Also, a program to develop LEU targets for Mo-99 production, via the technology developed to fabricate dispersed silicide fuel, has started, and preliminary scoping studies are underway. (Author)
International Nuclear Information System (INIS)
Costello, J.M.; Davy, D.R.; Cattell, F.C.R.; Cook, J.E.
1980-01-01
The subject is discussed under the headings: uranium mining; milling of uranium ores; manufacture of uranium hexafluoride; uranium enrichment; fuel manufacture and fabrication; environmental impact (use of natural resources; effluents from fuel cycle operations; occupational health; public health); alternative fuel cycles; additional waste treatment. (U.K.)
Development of very-high-density low-enriched uranium fuels
International Nuclear Information System (INIS)
Snegrove, J.L.; Hofmann, G.L.; Trybus, C.L.; Wiencek, T.C.
1997-01-01
The RERTR (=Reduced Enrichment for Research and Test Reactors) program has begun an aggressive effort to develop dispersion fuels for research and test reactors with uranium densities of 8 to 9 g U/cm 3 , based on the use of γ-stabilized uranium alloys. Fabrication development teams and facilities are being put into place, and preparations for the first irradiation test are in progress. The first screening irradiations are expected to begin in late April 1997 and the first results should be available by the end of 1997. Discussions with potential international partners in fabrication development and irradiation testing have begun. (author)
Occupational safety data and casualty rates for the uranium fuel cycle
International Nuclear Information System (INIS)
O'Donnell, F.R.; Hoy, H.C.
1981-10-01
Occupational casualty (injuries, illnesses, fatalities, and lost workdays) and production data are presented and used to calculate occupational casualty incidence rates for technologies that make up the uranium fuel cycle, including: mining, milling, conversion, and enrichment of uranium; fabrication of reactor fuel; transportation of uranium and fuel elements; generation of electric power; and transmission of electric power. Each technology is treated in a separate chapter. All data sources are referenced. All steps used to calculate normalized occupational casualty incidence rates from the data are presented. Rates given include fatalities, serious cases, and lost workdays per 100 man-years worked, per 10 12 Btu of energy output, and per other appropriate units of output
Summary report on fuel development and miniplate fabrication for the RERTR Program, 1978 to 1990
Energy Technology Data Exchange (ETDEWEB)
Wiencek, T.C. [Argonne National Lab., IL (United States). Energy Technology Div.
1995-08-01
This report summarizes the efforts of the Fabrication Technology Section at Argonne National Laboratory in the program of Reduced Enrichment Research and Test Reactors (RERTR). The main objective of this program was to reduce the amount of high enriched ({approx}93% {sup 235}U) uranium (HEU) used in nonpower reactors. Conversion from low-density (0.8--1.6 g U/cm{sup 3}) HEU fuel elements to highly loaded (up to 7 g U/cm{sup 3}) low-enrichment (<20% {sup 235}U) uranium (LEU) fuel elements allows the same reactor power levels, core designs and sizes to be retained while greatly reducing the possibility of illicit diversion of HEU nuclear fuel. This document is intended as an overview of the period 1978--1990, during which the Section supported this project by fabricating mainly powder metallurgy uranium-silicide dispersion fuel plates. Most of the subjects covered in detail are fabrication-related studies of uranium silicide fuels and fuel plate properties. Some data are included for out-of-pile experiments such as corrosion and compatibility tests. Also briefly covered are most other aspects of the RERTR program such as irradiation tests, full-core demonstrations, and technology transfer. References included are for further information on most aspects of the entire program. A significant portion of the report is devoted to data that were never published in their entirety. The appendices contain a list of previous RERTR reports, ANL fabrication procedures, calculations for phases present in two-phase fuels, chemical analysis of fuels, miniplate characteristics, and a summary of bonding runs made by hot isostatic pressing.
Summary report on fuel development and miniplate fabrication for the RERTR Program, 1978 to 1990
International Nuclear Information System (INIS)
Wiencek, T.C.
1995-08-01
This report summarizes the efforts of the Fabrication Technology Section at Argonne National Laboratory in the program of Reduced Enrichment Research and Test Reactors (RERTR). The main objective of this program was to reduce the amount of high enriched (∼93% 235 U) uranium (HEU) used in nonpower reactors. Conversion from low-density (0.8--1.6 g U/cm 3 ) HEU fuel elements to highly loaded (up to 7 g U/cm 3 ) low-enrichment ( 235 U) uranium (LEU) fuel elements allows the same reactor power levels, core designs and sizes to be retained while greatly reducing the possibility of illicit diversion of HEU nuclear fuel. This document is intended as an overview of the period 1978--1990, during which the Section supported this project by fabricating mainly powder metallurgy uranium-silicide dispersion fuel plates. Most of the subjects covered in detail are fabrication-related studies of uranium silicide fuels and fuel plate properties. Some data are included for out-of-pile experiments such as corrosion and compatibility tests. Also briefly covered are most other aspects of the RERTR program such as irradiation tests, full-core demonstrations, and technology transfer. References included are for further information on most aspects of the entire program. A significant portion of the report is devoted to data that were never published in their entirety. The appendices contain a list of previous RERTR reports, ANL fabrication procedures, calculations for phases present in two-phase fuels, chemical analysis of fuels, miniplate characteristics, and a summary of bonding runs made by hot isostatic pressing
Process for the fabrication of a nuclear fuel
International Nuclear Information System (INIS)
Hirose, Yasuo.
1970-01-01
Herein disclosed is a process for fabricating a nuclear fuel incorporating either uranium or plutonium. A pellet-like substrate consisting of a packed powder ceramic fuel such as uranium or plutonium is prepared with the horizontal surface of the body provided with a masking. Next, after impregnating the substrate voids with a solution consisting of a fissile material or mixture of fissile material and poison, the solvent is removed by a chemical deposition process which causes the impregnated material to migrate through capillary action toward the vicinity of the fuel body surface. Sintering and pyrolysis of the deposited material and masking are subsequently carried out to yield a fuel body having adjacent to its surface an intensely concentrated layer of either fissile material or a mixture of fissile material and poison. (Owens, K.J.)
Low-enriched uranium-molybdenum fuel plate development
International Nuclear Information System (INIS)
Wiencek, T.C.; Prokofiev, I.G.
2000-01-01
To examine the fabricability of low-enriched uranium-molybdenum powders, full-size 450 x 60 x 0.5-mm (17.7 x 2.4 x 0.020-in.) fuel zone test plates loaded to 6 g U/cm 3 were produced. U-10 wt.% Mo powders produced by two methods, centrifugal atomization and grinding, were tested. These powders were supplied at no cost to Argonne National Laboratory by the Korean Atomic Energy Research Institute and Atomic Energy of Canada Limited, respectively. Fuel homogeneity indicated that both of the powders produced acceptable fuel plates. Operator skill during loading of the powder into the compacting die and fuel powder morphology were found to be important when striving to achieve homogeneous fuel distribution. Smaller, 94 x 22 x 0.6-mm (3.7 x 0.87 x 0.025-in.) fuel zone, test plates were fabricated using U-10 wt.% Mo foil disks instead of a conventional powder metallurgy compact. Two fuel plates of this type are currently undergoing irradiation in the RERTR-4 high-density fuel experiment in the Advanced Test Reactor. (author)
MOX fuel fabrication: Technical and industrial developments
International Nuclear Information System (INIS)
Lebastard, G.; Bairiot, H.
1990-01-01
The plutonium available in the near future is generally estimated rather precisely on the basis of the reprocessing contracts and the performance of the reprocessing plants. A few years ago, decision makers were convinced that a significant share of this fissile material would be used as the feed material for fast breeder reactors (FBRs) or other advanced reactors. The facts today are that large reprocessing plants are coming into commercial operations: UP3 and soon UP2-800 and THORP, but that FBR deployment is delayed worldwide. As a consequence, large quantities of plutonium will be recycled in light water reactors as mixed oxide (MOX) fuels. MOX fuel technology has been properly demonstrated in the past 25 years. All specific problems have been addressed, efficient fabrication processes and engineering background have been implemented to a level of maturity which makes MOX fuel behaving as well as Uranium fuel. The paper concentrates on todays MOX fabrication expertise and presents the technical and industrial developments prepared by the MOX fuel fabrication industry for this last decade of the century
International Nuclear Information System (INIS)
Sifferlen, R.
1976-01-01
By establishing correlations between rejection level and fabrication parameters, quality assurance can guide corrective action for improving the consistency of fabrication and the reliability of fuel elements. The author cites examples relating to the quality of the uranium in metallic fuels, the influence of the parent metal on the welding of zirconium alloys, the behaviour of the springs inside the cladding during the welding of plugs and the behaviour of uranium oxide pellets. (author)
International Nuclear Information System (INIS)
Yamate, Kazuki; Yamada, Takashi; Takanashi, Mitsuhiro; Sasaki, Noriaki
2013-01-01
Integrated Safety Analysis (ISA) procedure for uranium fuel fabrication and enrichment facilities has been developed for aiming at applying risk-informed regulation to these uranium facilities. The development has carried out referring to the ISA (NUREG-1520) by the Nuclear Regulatory Commission (NRC). The paper presents purpose, principles and activities for the development of the ISA procedure, including Risk Level (RL) matrix and grading evaluation method of IROFS (Items Relied on for Safety), as well as general description and features of the procedure. Also described in the paper is current status in application of risk information from the ISA. Japanese four licensees of the uranium facilities have been conducting ISA for their representative processes using the developed procedure as their voluntary safety activities. They have been accumulating experiences and knowledge on the ISA procedure and risk information through the field activities. NISA (Nuclear and Industrial Safety Agency) and JNES (Japan Nuclear Energy Safety Organization) are studying how to use such risk information for the safety regulation of the uranium facilities, taking into account the licensees' experiences and knowledge. (authors)
Development of very-high-density low-enriched-uranium fuels
International Nuclear Information System (INIS)
Snelgrove, J.L.; Hofman, G.L.; Meyer, M.K.; Trybus, C.L.; Wiencek, T.C.
1997-01-01
Following a hiatus of several years and following its successful development and qualification of 4.8 g U cm -3 U 3 Si 2 -Al dispersion fuel for application with low-enriched uranium in research and test reactors, the US Reduced Enrichment for Research and Test Reactors program has embarked on the development of even-higher-density fuels. Our goal is to achieve uranium densities of 8-9 g cm -3 in aluminum-based dispersion fuels. Achieving this goal will require the use of high-density, γ-stabilized uranium alloy powders in conjunction with the most-advanced fuel fabrication techniques. Key issues being addressed are the reaction of the fuel alloys with aluminum and the irradiation behavior of the fuel alloys and any reaction products. Test irradiations of candidate fuels in very-small (micro) plates are scheduled to begin in the Advanced Test Reactor during June, 1997. Initial results are expected to be available in early 1998. We are performing out-of-reactor studies on the phase structure of the candidate alloys on diffusion of the matrix material into the aluminum. In addition, we are modifying our current dispersion fuel irradiation behavior model to accommodate the new fuels. Several international partners are participating in various phases of this work. (orig.)
High-Uranium-Loaded U3O8-Al fuel element development program. Part 1
International Nuclear Information System (INIS)
Martin, M.M.
1993-01-01
The High-Uranium-Loaded U 3 O 8 -Al Fuel Element Development Program supports Argonne National Laboratory efforts to develop high-uranium-density research and test reactor fuel to accommodate use of low-uranium enrichment. The goal is to fuel most research and test reactors with uranium of less than 20% enrichment for the purpose of lowering the potential for diversion of highly-enriched material for nonpeaceful usages. The specific objective of the program is to develop the technological and engineering data base for U 3 O 8 -Al plate-type fuel elements of maximal uranium content to the point of vendor qualification for full scale fabrication on a production basis. A program and management plan that details the organization, supporting objectives, schedule, and budget is in place and preparation for fuel and irradiation studies is under way. The current programming envisions a program of about four years duration for an estimated cost of about two million dollars. During the decades of the fifties and sixties, developments at Oak Ridge National Laboratory led to the use of U 3 O 8 -Al plate-type fuel elements in the High Flux Isotope Reactor, Oak Ridge Research Reactor, Puerto Rico Nuclear Center Reactor, and the High Flux Beam Reactor. Most of the developmental information however applies only up to a uranium concentration of about 55 wt % (about 35 vol % U 3 O 8 ). The technical issues that must be addressed to further increase the uranium loading beyond 55 wt % U involve plate fabrication phenomena of voids and dogboning, fuel behavior under long irradiation, and potential for the thermite reaction between U 3 O 8 and aluminum
International Nuclear Information System (INIS)
Coulter, C.A.; Whiteson, R.; Zardecki, A.
1991-01-01
We are developing a computer model of facility operations and nuclear materials accounting for a facility that reprocesses spent fuel and fabricates mixed oxide (MOX) fuel rods and assemblies from the recovered uranium and plutonium. The model will be used to determine the effectiveness of various materials measurement strategies for the facility and, ultimately, of other facility safeguards functions as well. This portion of the facility consists of a spent fuel storage pond, fuel shear, dissolver, clarifier, three solvent-extraction stages with uranium-plutonium separation after the first stage, and product concentrators. In this facility area mixed oxide is formed into pellets, the pellets are loaded into fuel rods, and the fuel rods are fabricated into fuel assemblies. These two facility sections are connected by a MOX conversion line in which the uranium and plutonium solutions from reprocessing are converted to mixed oxide. The model of the intermediate MOX conversion line used in the model is based on a design provided by Mike Ehinger of Oak Ridge National Laboratory (private communication). An initial version of the simulation model has been developed for the entire MOX conversion and fuel fabrication sections of the reprocessing/MOX fuel fabrication facility, and this model has been used to obtain inventory difference variance estimates for those sections of the facility. A significant fraction of the data files for the fuel reprocessing section have been developed, but these data files are not yet complete enough to permit simulation of reprocessing operations in the facility. Accordingly, the discussion in the following sections is restricted to the MOX conversion and fuel fabrication lines. 3 tabs
Decision fundamentals for emergencies in fuel fabrication plants
International Nuclear Information System (INIS)
Thomas, W.; Pfeffer, W.; Wiesemes, J.
1995-01-01
This report is a compilation of fundamental physical and chemical data for emergencies in fuel element fabrication facilities. The release of uranium and plutonium and a criticality accident constitute the main hazards to be considered. In addition information related to the chemical risk of a release of toxic uranium hexafluoride is included in the report. This fundamental information is to be applied in planning emergency measures and could be useful as advisory material for the emergency staff. (orig.) [de
Fabrication of inert matrix fuel for the incineration of plutonium - a feasibility study
International Nuclear Information System (INIS)
Burghartz, M.; Ledergerber, G.; Ingold, F.; Xie, T.; Botta, F.; Idemitsu, K.
1998-01-01
The internal gelation process has been applied to fabricate classical fuel based on uranium like UO 2 and MOX. For recent aims to destroy plutonium in the most effective way, a uranium free fuel was evaluated. The fuel development at PSI has been redirected to a fuel based on zirconium oxide or a mixture of zirconia and a conducting material leading to ceramic/metal (CERMET) or ceramic/ceramic (CERCER) combinations. A feasibility study was carried out to demonstrate that microspheres based on zirconia and spinel can be fabricated. The gelation parameters were investigated leading to optimised compositions for the starting solutions. Studies to fabricate a composite material (from zirconia and spinel) are ongoing. If the zirconia/spinel ratio is chosen appropriately, the low thermal conductivity of pure zirconia could be compensated by the higher thermal conductivity of spinel. Another solution to improve the low thermal conductivity of zirconia is the development of a CERMET, which consists of fine particles bearing plutonium in a cubic zirconia dispersed in a metallic matrix. The fabrication of such a CERMET is also being studied. (author)
Quality control in nuclear fuel fabrication
International Nuclear Information System (INIS)
Abdelhalim, A.S.; Elsayed, A.A.; Shaaban, H.I.
1988-01-01
The department of metallurgy, NRC Inchass is embarking on a programme of on a laboratory scale, fuel pins containing uranium dioxide pellets are going to be produced. The department is making use of the expertise and equipment at present available and is going to utilize the new fuel pin fabrication unit which would be shortly in operation. The fabrication and testing of uranium dioxide pellets then gradually adapt them and develop, a national know how in this field. This would also involve building up of indigenous experience through proper training of qualified personnel. That are applied to ensure quality of U o 2 pellets, the techniques implemented, the equipment used and the specifications of the equipment presently available. The following parameters are subject to quality control tests: density. O/U ration, hydrogen content, microstructure, each property will be discussed, measurements related to U o 2 powders, including flow ability, bulk density, O/U ratio, bet surface area and water content will be critically discussed. Relevant tests to ensure Q C of pellets are reviewed. These include surface integrity, density, dimensions, microstructure.4 fig., 1 tab
Coated U(Mo) Fuel: As-Fabricated Microstructures
Energy Technology Data Exchange (ETDEWEB)
Emmanuel Perez; Dennis D. Keiser, Jr.; Ann Leenaers; Sven Van den Berghe; Tom Wiencek
2014-04-01
As part of the development of low-enriched uranium fuels, fuel plates have recently been tested in the BR-2 reactor as part of the SELENIUM experiment. These fuel plates contained fuel particles with either Si or ZrN thin film coating (up to 1 µm thickness) around the U-7Mo fuel particles. In order to best understand irradiation performance, it is important to determine the starting microstructure that can be observed in as-fabricated fuel plates. To this end, detailed microstructural characterization was performed on ZrN and Si-coated U-7Mo powder in samples taken from AA6061-clad fuel plates fabricated at 500°C. Of interest was the condition of the thin film coatings after fabrication at a relatively high temperature. Both scanning electron microscopy and transmission electron microscopy were employed. The ZrN thin film coating was observed to consist of columns comprised of very fine ZrN grains. Relatively large amounts of porosity could be found in some areas of the thin film, along with an enrichment of oxygen around each of the the ZrN columns. In the case of the pure Si thin film coating sample, a (U,Mo,Al,Si) interaction layer was observed around the U-7Mo particles. Apparently, the Si reacted with the U-7Mo and Al matrix during fuel plate fabrication at 500°C to form this layer. The microstructure of the formed layer is very similar to those that form in U-7Mo versus Al-Si alloy diffusion couples annealed at higher temperatures and as-fabricated U-7Mo dispersion fuel plates with Al-Si alloy matrix fabricated at 500°C.
Analysis of fuel cycles with natural uranium; Analiza gorivnih ciklusa sa prirodnim uranom
Energy Technology Data Exchange (ETDEWEB)
Stojanovic, A [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)
1965-05-15
A method was developed and a computer code was written for analysis of fuel cycles and it was applied for heavy water and graphite moderated power reactors. Among a variety of possibilities, three methods which enable best utilization of natural uranium and plutonium production were analyzed. Analysis has shown that reprocessing of irradiated uranium and plutonium utilization in the same or similar type of reactor could increase significantly utilization of natural uranium. Increase of burnup is limited exclusively by costs of reprocessing, plutonium extraction and fabrication of new fuel elements.
recovery of enriched uranium from waste solution obtained from fuel fabrication laboratories
International Nuclear Information System (INIS)
Othman, S.H.A.
2003-01-01
reversed-phase partition chromatography is shown to be a convenient and applicable method for the quantitative recovery of uranium (19.7% enriched with 235 U) from highly impure solution . the processing of uranium compounds for atomic energy project especially in FMPP(Egyptian fuel manufacture pilot plant) gives rise to a variety of wastes in which the uranium content is of considerable importance. the recovery of uranium from concentrated mother liquors produced from ADU (ammonium diuranate ) precipitation, as well as those due to ADU washing is studied in this work. column of poly-trifluoro-monochloro-ethilene (Kel-F) supporting tri-n-butyl-phosphate (TBP) retains uranium .impurities are eluted with 6.5 M HCl, and the uranium is eluted with water and the recovery of uranium is better than 94%. A mathematical model was suggested to stimulate the sorption process of uranium ions (or any other ion ) by column of solvent impregnated resin containing organic extractant (the same as the previous column) . An excellent agreement was founded between the experimental results and the mathematical model
DUCTILE URANIUM FUEL FOR NUCLEAR REACTORS AND METHOD OF MAKING
Zegler, S.T.
1963-11-01
The fabrication process for a ductile nuclear fuel alloy consisting of uranium, fissium, and from 0.25 to 1.0 wt% of silicon or aluminum or from 0.25 to 2 wt% of titanium or yttrium is presented. (AEC)
Status of the atomized uranium silicide fuel development at KAERI
Energy Technology Data Exchange (ETDEWEB)
Kim, C.K.; Kim, K.H.; Park, H.D.; Kuk, I.H. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
1997-08-01
While developing KMRR fuel fabrication technology an atomizing technique has been applied in order to eliminate the difficulties relating to the tough property of U{sub 3}Si and to take advantage of the rapid solidification effect of atomization. The comparison between the conventionally comminuted powder dispersion fuel and the atomized powder dispersion fuel has been made. As the result, the processes, uranium silicide powdering and heat treatment for U{sub 3}Si transformation, become simplified. The workability, the thermal conductivity and the thermal compatibility of fuel meat have been investigated and found to be improved due to the spherical shape of atomized powder. In this presentation the overall developments of atomized U{sub 3}Si dispersion fuel and the planned activities for applying the atomizing technique to the real fuel fabrication are described.
Powder Metallurgy of Uranium Alloy Fuels for TRU-Burning Reactors Final Technical Report
Energy Technology Data Exchange (ETDEWEB)
McDeavitt, Sean M
2011-04-29
Overview Fast reactors were evaluated to enable the transmutation of transuranic isotopes generated by nuclear energy systems. The motivation for this was that TRU isotopes have high radiotoxicity and relatively long half-lives, making them unattractive for disposal in a long-term geologic repository. Fast reactors provide an efficient means to utilize the energy content of the TRUs while destroying them. An enabling technology that requires research and development is the fabrication metallic fuel containing TRU isotopes using powder metallurgy methods. This project focused upon developing a powder metallurgical fabrication method to produce U-Zr-transuranic (TRU) alloys at relatively low processing temperatures (500ºC to 600ºC) using either hot extrusion or alpha-phase sintering for charecterization. Researchers quantified the fundamental aspects of both processing methods using surrogate metals to simulate the TRU elements. The process produced novel solutions to some of the issues relating to metallic fuels, such as fuel-cladding chemical interactions, fuel swelling, volatility losses during casting, and casting mold material losses. Workscope There were two primary tasks associated with this project: 1. Hot working fabrication using mechanical alloying and extrusion • Design, fabricate, and assemble extrusion equipment • Extrusion database on DU metal • Extrusion database on U-10Zr alloys • Extrusion database on U-20xx-10Zr alloys • Evaluation and testing of tube sheath metals 2. Low-temperature sintering of U alloys • Design, fabricate, and assemble equipment • Sintering database on DU metal • Sintering database on U-10Zr alloys • Liquid assisted phase sintering on U-20xx-10Zr alloys Appendices Outline Appendix A contains a Fuel Cycle Research & Development (FCR&D) poster and contact presentation where TAMU made primary contributions. Appendix B contains MSNE theses and final defense presentations by David Garnetti and Grant Helmreich
Powder Metallurgy of Uranium Alloy Fuels for TRU-Burning Reactors Final Technical Report
International Nuclear Information System (INIS)
McDeavitt, Sean M.
2011-01-01
Overview Fast reactors were evaluated to enable the transmutation of transuranic isotopes generated by nuclear energy systems. The motivation for this was that TRU isotopes have high radiotoxicity and relatively long half-lives, making them unattractive for disposal in a long-term geologic repository. Fast reactors provide an efficient means to utilize the energy content of the TRUs while destroying them. An enabling technology that requires research and development is the fabrication metallic fuel containing TRU isotopes using powder metallurgy methods. This project focused upon developing a powder metallurgical fabrication method to produce U-Zr-transuranic (TRU) alloys at relatively low processing temperatures (500 C to 600 C) using either hot extrusion or alpha-phase sintering for charecterization. Researchers quantified the fundamental aspects of both processing methods using surrogate metals to simulate the TRU elements. The process produced novel solutions to some of the issues relating to metallic fuels, such as fuel-cladding chemical interactions, fuel swelling, volatility losses during casting, and casting mold material losses. Workscope There were two primary tasks associated with this project: (1) Hot working fabrication using mechanical alloying and extrusion - Design, fabricate, and assemble extrusion equipment - Extrusion database on DU metal - Extrusion database on U-10Zr alloys - Extrusion database on U-20xx-10Zr alloys - Evaluation and testing of tube sheath metals (2) Low-temperature sintering of U alloys - Design, fabricate, and assemble equipment - Sintering database on DU metal - Sintering database on U-10Zr alloys - Liquid assisted phase sintering on U-20xx-10Zr alloys Appendices Outline Appendix A contains a Fuel Cycle Research and Development (FCR and D) poster and contact presentation where TAMU made primary contributions. Appendix B contains MSNE theses and final defense presentations by David Garnetti and Grant Helmreich outlining the
Introduction to Exxon nuclear fuel fabrication plant
International Nuclear Information System (INIS)
Schneider, R.A.
1985-01-01
The Exxon Nuclear low-enriched uranium fuel fabrication plant in Richland, Washington produces fuel assemblies for both pressurized water and boiling water reactors. The Richland plant was the first US bulk-handling facility selected by the IAEA for inspection under the US-IAEA Safeguards Agreement. The plant was under IAEA inspection from March 1981 through October 1983. This text provides a written description of the plant layout, operation and process. The text also includes a one ton-a-day model (or reference) plant which was adapted from the Exxon Nuclear plant. The Model Plant provides a generic example of a low-enriched uranium (LEU) bulk-handling facility. The Model Plant is used to illustrate in a more quantitative way some of the key safeguards requirements for a bulk-handling facility
The nuclear fuel cycle, From the uranium mine to waste disposal
International Nuclear Information System (INIS)
2002-09-01
Fuel is a material that can be burnt to provide heat. The most familiar fuels are wood, coal, natural gas and oil. By analogy, the uranium used in nuclear power plants is called 'nuclear fuel', because it gives off heat too, although, in this case, the heat is obtained through fission and not combustion. After being used in the reactor, spent nuclear fuel can be reprocessed to extract recyclable energy material, which is why we speak of the nuclear fuel cycle. This cycle includes all the following industrial operations: - uranium mining, - fuel fabrication, - use in the reactor, - reprocessing the fuel unloaded from the reactor, - waste treatment and disposal. 'The nuclear fuel cycle includes an array of industrial operations, from uranium mining to the disposal of radioactive waste'. Per unit or mass (e.g. per kilo), nuclear fuel supplies far more energy than a fossil fuel (coal or oil). When used in a pressurised water reactor, a kilo of uranium generates 10,000 times more energy than a kilo of coal or oil in a conventional power station. Also, the fuel will remain in the reactor for a long time (several years), unlike conventional fuels, which are burnt up quickly. Nuclear fuel also differs from others in that uranium has to undergo many processes between the time it is mined and the time it goes into the reactor. For the sake of simplicity, the following pages will only look at nuclear fuel used in pressurised water reactors (or PWRs), because nuclear power plants consisting of one or more PWRs are the most widely used around the world. (authors)
Energy Technology Data Exchange (ETDEWEB)
Carvalho, E.F. Urano de; Lainetti, P.E.; Gomes, R.P. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)
1996-07-01
In this paper are described the main chemical process employed in the Chemical Processes Division of the Fuel Technology Department - IPEN for conversion of enriched UF{sub 6} in ammonium diuranate - DUA and uranium tetrafluoride - UF{sub 4}. These activities have assured the continuity of fuel elements production at IPEN since 1984. The uranium recovery from scraps of the fuel elements production and the purification processes are also described. Those compounds are important intermediate products in the fabrication routine and in development dispersed fuel elements with higher uranium loading for IEA{sub R}1 research reactor power increase program. (author)
Phenomenology of uranium-plutonium homogenization in nuclear fuels
International Nuclear Information System (INIS)
Marin, J.M.
1988-01-01
The uranium and plutonium cations distribution in mixed oxide fuels (U 1-y Pu y )O 2 with y ≤ 0.1 has been studied in laboratory with industrial fabrication methods. Our experiences has showed a slow cations migration. In the substoichiometry (UPu)O 2-x the diffusion is in connection with the plutonium valence which is an indicator of the oxidoreduction state of the crystal lattice. The plutonium valence is in connection with the oxygen ion deficit in order to compensate the electrical charge. The oxygen ratio of the solid depends of the oxygen partial pressure prevailing at the time of product elaboration but it can be modified by impurities. These impurities permit to increase or decrease the fuel characteristics and performances. An homogeneity analysis methodology is proposed, its objective is to classify the mixed oxide fuels according to the uranium and plutonium ions distribution [fr
Energy Technology Data Exchange (ETDEWEB)
Ro, Seung Gy; Shin, Y.J.; You, G.S.; Joo, J.S.; Min, D.K.; Chun, Y.B.; Lee, E.P.; Seo, H.S.; Ahn, S.B
1999-03-01
The simulated metallized spent fuel ingots were fabricated and evaluated the oxidation rates and the activation energies under several temperature conditions to develop an advanced spent fuel management process. It was also checked the alloying characteristics of the some elements with metal uranium. (Author). 3 refs., 1 tab., 36 figs.
Study on reprocessing of uranium-thorium fuel with solvent extraction for HTGR
International Nuclear Information System (INIS)
Jiao Rongzhou; He Peijun; Liu Bingren; Zhu Yongjun
1992-08-01
A single cycle process by solvent extraction with acid feed solution is suggested. The purpose is to reprocess uranium-thorium fuel elements which are of high burn-up and rich of 232 U from HTGR (high temperature gas cooled reactor). The extraction cascade tests have been completed. The recovery of uranium and thorium is greater than 99.6%. By this method, the requirement, under remote control to re-fabricate fuel elements, of decontamination factors for Cs, Sr, Zr-Nb and Ru has been reached
International Nuclear Information System (INIS)
Fleming Rubio, Peter Alex
2010-01-01
The current work was made in Conversion laboratory belonging to Chilean Nuclear Energy Commission, CCHEN. This is constituted by the development of three hydrometallurgical processes, belonging to the recovery of uranium from fuel plates based on uranium silicide (U_3Si_2) process, for nuclear research reactors MTR (Material Testing Reactor) type, those that come from the Fuel Elements Manufacture Plant, PEC. In the manufacturing process some of these plates are subjected to destructive tests by quality requirement or others are rejected for non-compliance with technical specifications, such as: lack of homogenization of the dispersion of uraniferous compound in the meat, as well as the appearance of the defects, such as blisters, so-called "dog bone", "fish tail", "remote islands", among others. Because the uranium used is enriched in 19.75% U_2_3_5 isotope, which explains the high value in the market, it must be recovered for reuse, returning to the production line of fuel elements. The uranium silicide, contained in the plates, is dispersed in an aluminum matrix and covered with plates and frames of ASTM 6061 Aluminum, as a sandwich coating, commonly referred to as 'meat' (sandwich meat). As aluminum is the main impurity, the process begins with this metal dissolution, present in meat and plates, by NaOH reaction, followed by a vacuum filtration, washing and drying, obtaining a powder of uranium silicide, with a small impurities percentage. Then, the crude uranium silicide reacts with a solution of hydrofluoric acid, dissolving the silicon and simultaneously precipitating UF_4 by reaction with HNO_3, obtaining an impure UO_2(NO_3)_2 solution. The experimental work was developed and implemented at laboratory scale for the three stages pertaining to the uranium recovery process, determining for each one the optimum operation conditions: temperature, molarity or concentration, reagent excess, among others (author)
Fabrication and characterization of MX-type fuels and fuel pins
International Nuclear Information System (INIS)
Richter, K.; Bartscher, W.; Benedict, U.; Gueugnon, J.F.; Kutter, H.; Sari, C.; Schmidt, H.E.
1978-01-01
This paper summarizes the most important fabrication parameters and characterization of fuel and fuel pins obtained during the investigation of uranium-plutonium carbides, oxicarbides, carbonitrides and nitrides in the past years at the European Institute for Transuranium Elements at Karlsruhe. All preparation methods discussed are based on carbothermic reduction of a mechanical blend of uranium-plutonium oxide and carbon powder. General data for carbothermic reduction processes are discussed (influence of starting material, homogeneity, control of degree of reaction, etc). A survey of different preparation methods investigated is given. Limitations with respect to temperature and atmosphere for both carbothermic reduction processes and sintering conditions for the different compounds are summarized. A special preparation process for mixed carbonitrides with low nitrogen content (U,Pu)sub(1-x)Nsub(x) in the range 0.1 0 C to 1400 0 C by means of a modulated electron beam technique. A scheme is proposed, which allows to predict the thermal properties of MX fuels on the basis of their chemical composition and porosity. Preparation, preirradiation characterization and final controls of fuel test pins for pellet and vibrocompacted type of pins are described and the most important data summarized for all advanced fuels irradiated at Dounreay (DN1) and Rapsodie Fast Reactor (DN2) within the TU irradiation programme
Energy Technology Data Exchange (ETDEWEB)
Sinha, V.P. [Metallic Fuels Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)], E-mail: vedsinha@barc.gov.in; Prasad, G.J.; Hegde, P.V.; Keswani, R.; Basak, C.B.; Pal, S.; Mishra, G.P. [Metallic Fuels Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)
2009-04-03
Most of the research and test reactors worldwide have undergone core conversion from high enriched uranium base fuel to low enriched uranium base fuel under the Reduced Enrichment for Research and Test Reactor (RERTR) program, which was launched in the late 1970s to reduce the risk of nuclear proliferation. To realize this goal, high density uranium compounds and {gamma}-stabilized uranium alloy powder were identified. In Metallic Fuels Division of BARC, R and D efforts are on to develop these high density uranium base alloys. This paper describes the preparation flow sheet for different compositions of Uranium and molybdenum alloys by an innovative powder processing route with uranium and molybdenum metal powders as starting materials. The same composition of U-Mo alloys were also fabricated by conventional method i.e. ingot metallurgy route. The U-Mo alloys prepared by both the methods were then characterized by XRD for phase analysis. The photomicrographs of alloys with different compositions prepared by powder metallurgy and ingot metallurgy routes are also included in the paper. The paper also covers the comparison of properties of the alloys prepared by powder metallurgy and ingot metallurgy routes.
International Nuclear Information System (INIS)
Sinha, V.P.; Prasad, G.J.; Hegde, P.V.; Keswani, R.; Basak, C.B.; Pal, S.; Mishra, G.P.
2009-01-01
Most of the research and test reactors worldwide have undergone core conversion from high enriched uranium base fuel to low enriched uranium base fuel under the Reduced Enrichment for Research and Test Reactor (RERTR) program, which was launched in the late 1970s to reduce the risk of nuclear proliferation. To realize this goal, high density uranium compounds and γ-stabilized uranium alloy powder were identified. In Metallic Fuels Division of BARC, R and D efforts are on to develop these high density uranium base alloys. This paper describes the preparation flow sheet for different compositions of Uranium and molybdenum alloys by an innovative powder processing route with uranium and molybdenum metal powders as starting materials. The same composition of U-Mo alloys were also fabricated by conventional method i.e. ingot metallurgy route. The U-Mo alloys prepared by both the methods were then characterized by XRD for phase analysis. The photomicrographs of alloys with different compositions prepared by powder metallurgy and ingot metallurgy routes are also included in the paper. The paper also covers the comparison of properties of the alloys prepared by powder metallurgy and ingot metallurgy routes
International Nuclear Information System (INIS)
Fishbone, L.G.; Moussalli, G.; Naegele, G.
1995-01-01
An approach of short-notice random inspections (SNRIs) for inventory-change verification can enhance the effectiveness and efficiency of international safeguards at natural or low-enriched uranium (LEU) fuel fabrication plants. According to this approach, the plant operator declares the contents of nuclear material items before knowing if an inspection will occur to verify them. Additionally, items about which declarations are newly made should remain available for verification for an agreed time. Then a statistical inference can be made from verification results for items verified during SNRIs to the entire populations, i.e. the entire strata, even if inspectors were not present when many items were received or produced. A six-month field test of the feasibility of such SNRIs took place at the Westinghouse Electric Corporation Commercial Nuclear Fuel Division during 1993. Westinghouse personnel made daily declarations about both feed and product items, uranium hexafluoride cylinders and finished fuel assemblies, using a custom-designed computer ''mailbox''. Safeguards inspectors from the IAEA conducted eight SNRIs to verify these declarations. They arrived unannounced at the plant, in most cases immediately after travel from Canada, where the IAEA maintains a regional office. Items from both strata were verified during the SNRIs by meant of nondestructive assay equipment
Accident safety analysis for 300 Area N Reactor Fuel Fabrication and Storage Facility
International Nuclear Information System (INIS)
Johnson, D.J.; Brehm, J.R.
1994-01-01
The purpose of the accident safety analysis is to identify and analyze a range of credible events, their cause and consequences, and to provide technical justification for the conclusion that uranium billets, fuel assemblies, uranium scrap, and chips and fines drums can be safely stored in the 300 Area N Reactor Fuel Fabrication and Storage Facility, the contaminated equipment, High-Efficiency Air Particulate filters, ductwork, stacks, sewers and sumps can be cleaned (decontaminated) and/or removed, the new concretion process in the 304 Building will be able to operate, without undue risk to the public, employees, or the environment, and limited fuel handling and packaging associated with removal of stored uranium is acceptable
Accident safety analysis for 300 Area N Reactor Fuel Fabrication and Storage Facility
Energy Technology Data Exchange (ETDEWEB)
Johnson, D.J.; Brehm, J.R.
1994-01-01
The purpose of the accident safety analysis is to identify and analyze a range of credible events, their cause and consequences, and to provide technical justification for the conclusion that uranium billets, fuel assemblies, uranium scrap, and chips and fines drums can be safely stored in the 300 Area N Reactor Fuel Fabrication and Storage Facility, the contaminated equipment, High-Efficiency Air Particulate filters, ductwork, stacks, sewers and sumps can be cleaned (decontaminated) and/or removed, the new concretion process in the 304 Building will be able to operate, without undue risk to the public, employees, or the environment, and limited fuel handling and packaging associated with removal of stored uranium is acceptable.
International Nuclear Information System (INIS)
Seward, Amy M.; Toomey, Christopher; Ford, Benjamin E.; Wood, Thomas W.; Perkins, Casey J.
2011-01-01
For several years, Pacific Northwest National Laboratory (PNNL) has been assessing the reliability of nuclear fuel supply in support of the U.S. Department of Energy/National Nuclear Security Administration. Three international low enriched uranium reserves, which are intended back up the existing and well-functioning nuclear fuel market, are currently moving toward implementation. These backup reserves are intended to provide countries credible assurance that of the uninterrupted supply of nuclear fuel to operate their nuclear power reactors in the event that their primary fuel supply is disrupted, whether for political or other reasons. The efficacy of these backup reserves, however, may be constrained without redundant fabrication services. This report presents the findings of a recent PNNL study that simulated outages of varying durations at specific nuclear fuel fabrication plants. The modeling specifically enabled prediction and visualization of the reactors affected and the degree of fuel delivery delay. The results thus provide insight on the extent of vulnerability to nuclear fuel supply disruption at the level of individual fabrication plants, reactors, and countries. The simulation studies demonstrate that, when a reasonable set of qualification criteria are applied, existing fabrication plants are technically qualified to provide backup fabrication services to the majority of the world's power reactors. The report concludes with an assessment of the redundancy of fuel supply in the nuclear fuel market, and a description of potential extra-market mechanisms to enhance the security of fuel supply in cases where it may be warranted. This report is an assessment of the ability of the existing market to respond to supply disruptions that occur for technical reasons. A forthcoming report will address political disruption scenarios.
Automated fuel fabrication- a vision comes true
International Nuclear Information System (INIS)
Hemantha Rao, G.V.S.; Prakash, M.S.; Setty, C.R.P.; Gupta, U.C.
1997-01-01
When New Uranium Fuel Assembly Project at Nuclear Fuel Complex (NFC) begins production, its operator will have equipment provided with intramachine handling systems working automatically by pressing a single button. Additionally simple low cost inter machine handling systems will further help in critical areas. All these inter and intra machine handling systems will result in improved reliability, productivity and quality. The fault diagnostics, mimics and real time data acquisition systems make the plant more operator friendly. The paper deals with the experience starting from layout, selection of product carriers, different handling systems, the latest technology and the integration of which made the vision on automation in fuel fabrication come true. (author)
Uranium silicide activities at Babcock and Wilcox
International Nuclear Information System (INIS)
Noel, W.W.; Freim, J.B.
1983-01-01
Babcock and Wilcox, Naval Nuclear Fuel Division (NNFD) in conjunction with Argonne National Laboratory (ANL) is actively involved in the Reduced Enrichment Research Test Reactor (RERTR) Program to produce low enriched fuel elements for research reactors. B and W and ANL have undertaken a joint effort in which NNFD will fabricate two low enriched uranium (LEU), Oak Ridge Reactor (ORR) elements with uranium silicide fuel furnished by ANL. These elements are being fabricated for irradiation testing at Oak Ridge National Laboratory (ORNL). Concurrently with this program, NNFD is developing and implementing the uranium silicide and uranium aluminide fuel fabrication technology. NNFD is fabricating the uranium silicide ORR elements in a two-phase program, Development and Production. To summarize: 1. Full size fuel plates can be made with U 3 SiAl but the fabricator must prevent oxidation of the compact prior to hot roll bonding; 2. Providing the ANL U 3 Si x irradiation results are successful, NNFD plans to provide two ORR elements during February 1983; 3. NNFD is developing and implementing U 3 Si x and UAI x fuel fabrication technology to be operational in 1983; 4. NNFD can supply U 3 O 8 high enriched uranium (HEU) or low enriched uranium (LEU) research reactor elements; 5. NNFD is capable of providing high quality, cost competitive LEU or HEU research reactor elements to meet the needs of the customer
New fabrication techniques for the nuclear fuels of tomorrow
International Nuclear Information System (INIS)
Babelot, J.F.; Bokelund, H.; Gerontopoulos, P.; Gueugnon, J.F.; Richter, K.
1995-01-01
The shift of the emphasis of the work at the Institute for Transuranium Elements (ITU) from the development of fuels based on uranium and plutonium to safety aspects concerning the use of plutonium and other of actinides, necessitates the production of targets containing appreciable amounts of minor actinides for irradiation experiments. The handling of minor actinides requires additional protective measures, combined with improved fuel fabrication techniques. The boundary conditions for a suitable process are flexibility, adaptability to remote control, and minimization of dust formation. A method based on the sol-gel fabrication technique meets these criteria, and was selected for the present developments at ITU. (author)
The uranium fuel cycle at IPEN - Energy and Nuclear Research Institute, SP, Brazil
International Nuclear Information System (INIS)
Abrao, Alcidio
1994-09-01
This paper summarizes the progress of research concerning the uranium fuel cycle set up at the IPEN, Sao Paulo, from the raw yellow-cake to the uranium hexafluoride. It covers the reconversion of the hexafluoride to ammonium uranyl tricarbonate and the manufacturing of the fuel elements for the swimming pool IEA-R1 reactor. This review extends the coverage of two pilot plants for uranium purification based upon ion exchange, one demonstration unity for the purification of uranyl nitrate by solvent extraction in pulsed columns, the unity of uranium tetrafluoride into moving bed reactors and a second one based upon the wet chemistry via uranium dioxide and aqueous hydrogen fluoride. The paper mentions the pilot plant for the preparation of uranium trioxide by the thermal decomposition of ammonium diuranate and a second unity by the thermal denitration of uranyl nitrate. The paper outlines the fluorine plant and the unity for the hexafluoride preparation, the unity for the conversion of the hexa to the ammonium uranyl tricarbonate and the fabrication of fuel elements for the IEA-R1 reactor. (author)
International Nuclear Information System (INIS)
Roddy, J.W.; Blanco, R.E.; Hill, G.S.; Moore, R.E.; Seagren, R.D.; Witherspoon, J.P.
1976-06-01
A cost/benefit study was made to determine the cost and effectiveness of various radioactive waste (radwaste) treatment systems for decreasing the release of radioactive materials from model High-Temperature Gas-Cooled (HTGR) fuel fabrication plants and to determine the radiological impact (dose commitment) of the released materials on the environment. The study is designed to assist in defining the term ''as low as reasonably achievable'' as it applies to these nuclear facilities. The base cases of the two model plants, a fresh fuel fabrication plant and a refabrication plant, are representative of current proposed commercial designs or are based on technology that is being developed to fabricate uranium, thorium, and graphite into fuel elements. The annual capacities of the fresh fuel plant and the refabrication plant are 450 and 245 metric tons of heavy metal (where heavy metal is uranium plus thorium), as charged to about fifty 1000-MW(e) HTGRs. Additional radwaste treatment systems are added to the base case plants in a series of case studies to decrease the amounts of radioactive materials released and to reduce the radiological dose commitment to the population in the surrounding area. The capital and annual costs for the added waste treatment operations and the corresponding reductions in dose commitments are calculated for each case. In the final analysis, the cost/benefit of each case, calculated as additional cost of radwaste system divided by the reduction in dose commitment, is tabulated or the dose commitment is plotted with cost as the variable. The status of each of the radwaste treatment methods is discussed. 48 figures, 74 tables
U-10Mo Baseline Fuel Fabrication Process Description
Energy Technology Data Exchange (ETDEWEB)
Hubbard, Lance R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Arendt, Christina L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Dye, Daniel F. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Clayton, Christopher K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lerchen, Megan E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lombardo, Nicholas J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lavender, Curt A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Zacher, Alan H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)
2017-09-27
This document provides a description of the U.S. High Power Research Reactor (USHPRR) low-enriched uranium (LEU) fuel fabrication process. This document is intended to be used in conjunction with the baseline process flow diagram (PFD) presented in Appendix A. The baseline PFD is used to document the fabrication process, communicate gaps in technology or manufacturing capabilities, convey alternatives under consideration, and as the basis for a dynamic simulation model of the fabrication process. The simulation model allows for the assessment of production rates, costs, and manufacturing requirements (manpower, fabrication space, numbers and types of equipment, etc.) throughout the lifecycle of the USHPRR program. This document, along with the accompanying PFD, is updated regularly
Development of Advanced High Uranium Density Fuels for Light Water Reactors
Energy Technology Data Exchange (ETDEWEB)
Blanchard, James [Univ. of Wisconsin, Madison, WI (United States); Butt, Darryl [Boise State Univ., ID (United States); Meyer, Mitchell [Idaho National Lab. (INL), Idaho Falls, ID (United States); Xu, Peng [Westinghouse Electric Corporation, Pittsburgh, PA (United States)
2016-02-15
This work conducts basic materials research (fabrication, radiation resistance, thermal conductivity, and corrosion response) on U3Si2 and UN, two high uranium density fuel forms that have a high potential for success as advanced light water reactor (LWR) fuels. The outcome of this proposed work will serve as the basis for the development of advance LWR fuels, and utilization of such fuel forms can lead to the optimization of the fuel performance related plant operating limits such as power density, power ramp rate and cycle length.
International Nuclear Information System (INIS)
Hassel, H.W.
1993-01-01
NUKEM handles around almost two tons of highly enriched uranium a year and it was necessary to satisfy all the new physical protection philosophies. That means that we have to install storage and safe fabrication sites for a lot of money, 25 meter thick concrete walls, and different alarm systems. So just to demonstrate how silly this business is, we have just overcome this for highly enriched uranium, and now we speak about low enriched uranium for which we don't need all of these investments to make this business safe. I would-just like to concentrate my words on the status of fabrication and considerations in my company concerning the medium enriched uranium and low enriched uranium. In the table are the different fuel types (see column) and then we have the fabrication in column 2 the experience of my comp any in kg. In column 3 is the irradiation experience of these fuels types. Column 4 shows the studies and calculations made in our company for lower and medium enriched fuels. The preliminary fabrication tests and calculations are in column 5, and in column 6 we have the delivery time for a prototype core in months after UF 6 supply. Column 7 shows the time for the development of specifications including irradiation time in years for 6 and 7 and column is the estimated cost of 6 and 7 There is just one fuel that is not in this summary and that is U-Zr. We now see how complex and sophisticated this business is. I have told you already that we have installed for a lot of millions of Deutsche Mark the physical protection, storage vaults and things like that. Now we have to investigate all these different types of fuels for, as you see, a lot of money. Maybe these are a lot of optimistic figures; anyway the question is, does this make all the overall nuclear situation worldwide easier or not. One cannot answer for the moment, but anyway we have a lot of problems
Energy Technology Data Exchange (ETDEWEB)
Souza, Jose Antonio Batista de
2011-07-01
IPEN-CNEN/SP developed the technology to produce the dispersion type fuel elements for research reactors and made it available for routine production. Today, the fuel produced in IPEN-CNEN/SP is limited to the uranium concentration of 3.0 gU/cm{sup 3} for U{sub 3}Si{sub 2}-Al dispersion-based and 2.3 gU/cm{sup 3} for U{sub 3}O{sub 8}-Al dispersion. The increase of uranium concentration in fuel plates enables the reactivity of the reactor core reactivity to be higher and extends the fuel life. Concerning technology, it is possible to increase the uranium concentration in the fuel meat up to the limit of 4.8 gU/cm{sup 3} in U{sub 3}Si{sub 2}-Al dispersion and 3.2 gU/cm{sup 3} U{sub 3}O{sub 8}-Al dispersion. These dispersions are well qualified worldwide. This work aims to develop the manufacturing process of both fuel meats with high uranium concentrations, by redefining the manufacturing procedures currently adopted in the Nuclear Fuel Center of IPEN-CNEN/SP. Based on the results, it was concluded that to achieve the desired concentration, it is necessary to make some changes in the established procedures, such as in the particle size of the fuel powder and in the feeding process inside the matrix, before briquette pressing. These studies have also shown that the fuel plates, with a high concentration of U{sub 3}Si{sub 2}-Al, met the used specifications. On the other hand, the appearance of the microstructure obtained from U{sub 3}O{sub 8}-Al dispersion fuel plates with 3.2 gU/cm{sup 3} showed to be unsatisfactory, due to the considerably significant porosity observed. The developed fabrication procedure was applied to U{sub 3}Si{sub 2} production at 4.8 gU/cm{sup 3}, with enriched uranium. The produced plates were used to assemble the fuel element IEA-228, which was irradiated in order to check its performance in the IEA-R1 reactor at IPEN-CNEN/SP. These new fuels have potential to be used in the new Brazilian Multipurpose Reactor - RMB. (author)
Structure, conduct, and sustainability of the international low-enriched fuel fabrication industry
International Nuclear Information System (INIS)
Rothwell, Geoffrey
2008-01-01
This paper examines the cost structures of fabricating Low-Enriched Uranium fuel (LEU, enriched to 5% enrichment) light water reactor fuels. The LEU industry is decades old, and (except for high entry cost, i.e., the cost of designing and licensing a fuel fabrication facility and its fuel), labor and additional fabrication lines can be added by industry incumbents at Nth-of-a-Kind cost to the maximum capacity allowed by the license. On the other hand, new entrants face higher First-of-a-Kind costs and high new-facility licensing costs, increasing the scale required for entry thus discouraging small scale entry by countries with only a few nuclear power plants. Therefore, the industry appears to be competitive with sustainable investment in fuel-cycle states, and structural barriers-to-entry increase its proliferation resistance. (author)
International Nuclear Information System (INIS)
Kim, Chang-Kyu; Kim, Ki-Hwan; Park, Jong-Man; Lee, Don-Bae; Sohn, Dong-Seong
1997-01-01
The RERTR program has been making an effort to develop dispersion fuels with uranium densities of 8 to 9 g U/cm3 for research and test reactors. Using atomized U-10wt%Mo powder, fuel meats have been fabricated successfully up to 55 volume % of fuel powder. The uranium density of an extruded meat with a 55 volume % of fuel powder was obtained to be 7.7 g/cm3. A relatively high porosity of 7.3% was formed due to cracking of particles, presumably induced by the impingement among agglomerated particles. Tensile test results indicated that the strength of fuel meats with 55% volume fraction decreased some and a little of ductility was maintained. Examination on the fracture surface revealed that some U-10%Mo particles appeared to be broken by the tensile force in brittle rupture mode. The increase of broken particles in high fuel fraction is considered to be induced mainly by the impingement among agglomerated particles. Uranium loading density is assumed to be improved through the development of the better homogeneous dispersion technology. (author)
Design of high density gamma-phase uranium alloys for LEU dispersion fuel applications
International Nuclear Information System (INIS)
Hofman, Gerard L.; Meyer, Mitchell K.; Ray, Allison E.
1998-01-01
Uranium alloys are candidates for the fuel phase in aluminium matrix dispersion fuels requiring high uranium loading. Certain uranium alloys have been shown to have good irradiation performance at intermediate burnup. previous studies have shown that acceptable fission gas swelling behavior and fuel-aluminium interaction is possible only if the fuel alloy can be maintained in the high temperature body-centered-cubic γ-phase during fabrication and irradiation, at temperatures at which αU is the equilibrium phase. transition metals in Groups V through VIII are known to allow metastable retention of the gamma phase below the equilibrium isotherm. These metals have varying degrees of effectiveness in stabilizing the gamma phase. Certain alloys are metastable for very long times at the relatively low fuel temperatures seen in research operation. In this paper, the existing data on the gamma stability of binary and ternary uranium alloys is analysed. The mechanism and kinetics of decomposition of the gamma phase are assessed with the help of metal alloy theory. Alloys with the highest possible uranium content, good gamma-phase stability, and good neutronic performance are identified for further metallurgical studies and irradiation tests. Results from theory will be compared with experimentally generated data. (author)
High-uranium-loaded U3O8-Al fuel element development program [contributed by N.M. Martin, ORNL
International Nuclear Information System (INIS)
Martin, M.M.
1993-01-01
The High-Uranium-Loaded U 3 O 8 -Al Fuel Element Development Program supports Argonne National Laboratory efforts to develop high-uranium-density research and test reactor fuel to accommodate use of low-uranium enrichment. The goal is to fuel most research and test reactors with uranium of less than 20% enrichment for the purpose of lowering the potential for diversion of highly-enriched material for nonpeaceful usages. The specific objective of the program is to develop the technological and engineering data base for U 3 O 8 -Al plate-type fuel elements of maximal uranium content to the point of vendor qualification for full scale fabrication on a production basis. A program and management plan that details the organization, supporting objectives, schedule, and budget is in place and preparation for fuel and irradiation studies is under way. The current programming envisions a program of about four years duration for an estimated cost of about two million dollars. During the decades of the fifties and sixties, developments at Oak Ridge National Laboratory led to the use of U 3 O 8 -Al plate-type fuel elements in the High Flux Isotope Reactor, Oak Ridge Research Reactor, Puerto Rico Nuclear Center Reactor, and the High Flux Beam Reactor. Most of the developmental information however applies only up to a uranium concentration of about 55 wt % (about 35 vol % U 3 O 8 ). The technical issues that must be addressed to further increase the uranium loading beyond 55 wt % involve plate fabrication phenomena of voids and dogboning, fuel behavior under long irradiation, and potential for the thermite reaction between U 3 O 8 and aluminum. (author)
Development of DIPRES feed for the fabrication of mixed-oxide fuels for fast breeder reactors
International Nuclear Information System (INIS)
Griffin, C.W.; Rasmussen, D.E.; Lloyd, M.H.
1983-01-01
The DIrect PREss Spheroidized feed process combines the conversion of uranium-plutonium solutions into spheres by internal gelation with conventional pellet fabrication techniques. In this manner, gel spheres could replace conventional powders as the feed material for pellet fabrication of nuclear fuels. Objective of the DIPRES feed program is to develop and qualify a process to produce mixed-oxide fuel pellets from gel spheres for fast breeder reactors. This process development includes both conversion and fabrication activities
Energy Technology Data Exchange (ETDEWEB)
Montierth, Leland M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)
2016-07-19
The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element design for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition, as part of a fuel meat thickness optimization effort for reactor performance, other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B, that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.
International Nuclear Information System (INIS)
Montierth, Leland M.
2016-01-01
The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element design for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition, as part of a fuel meat thickness optimization effort for reactor performance, other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B, that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.
Synthesis of Uranium nitride powders using metal uranium powders
International Nuclear Information System (INIS)
Yang, Jae Ho; Kim, Dong Joo; Oh, Jang Soo; Rhee, Young Woo; Kim, Jong Hun; Kim, Keon Sik
2012-01-01
Uranium nitride (UN) is a potential fuel material for advanced nuclear reactors because of their high fuel density, high thermal conductivity, high melting temperature, and considerable breeding capability in LWRs. Uranium nitride powders can be fabricated by a carbothermic reduction of the oxide powders, or the nitriding of metal uranium. The carbothermic reduction has an advantage in the production of fine powders. However it has many drawbacks such as an inevitable engagement of impurities, process burden, and difficulties in reusing of expensive N 15 gas. Manufacturing concerns issued in the carbothermic reduction process can be solved by changing the starting materials from oxide powder to metals. However, in nitriding process of metal, it is difficult to obtain fine nitride powders because metal uranium is usually fabricated in the form of bulk ingots. In this study, a simple reaction method was tested to fabricate uranium nitride powders directly from uranium metal powders. We fabricated uranium metal spherical powder and flake using a centrifugal atomization method. The nitride powders were obtained by thermal treating those metal particles under nitrogen containing gas. We investigated the phase and morphology evolutions of powders during the nitriding process. A phase analysis of nitride powders was also a part of the present work
Energy Technology Data Exchange (ETDEWEB)
Renfro, David G [ORNL; Cook, David Howard [ORNL; Freels, James D [ORNL; Griffin, Frederick P [ORNL; Ilas, Germina [ORNL; Sease, John D [ORNL; Chandler, David [ORNL
2012-03-01
This report describes progress made during FY11 in ORNL activities to support converting the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum (UMo) alloy. With both radial and axial contouring of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current levels achieved with HEU fuel. Studies are continuing to demonstrate that the fuel thermal safety margins can be preserved following conversion. Studies are also continuing to update other aspects of the reactor steady state operation and accident response for the effects of fuel conversion. Technical input has been provided to Oregon State University in support of their hydraulic testing program. The HFIR conversion schedule was revised and provided to the GTRI program. In addition to HFIR conversion activities, technical support was provided directly to the Fuel Fabrication Capability program manager.
International Nuclear Information System (INIS)
Renfro, David G.; Cook, David Howard; Freels, James D.; Griffin, Frederick P.; Ilas, Germina; Sease, John D.; Chandler, David
2012-01-01
This report describes progress made during FY11 in ORNL activities to support converting the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum (UMo) alloy. With both radial and axial contouring of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current levels achieved with HEU fuel. Studies are continuing to demonstrate that the fuel thermal safety margins can be preserved following conversion. Studies are also continuing to update other aspects of the reactor steady state operation and accident response for the effects of fuel conversion. Technical input has been provided to Oregon State University in support of their hydraulic testing program. The HFIR conversion schedule was revised and provided to the GTRI program. In addition to HFIR conversion activities, technical support was provided directly to the Fuel Fabrication Capability program manager.
Keiser, Dennis; Jue, Jan-Fong; Miller, Brandon; Gan, Jian; Robinson, Adam; Madden, James
2017-12-01
A low-enriched uranium U-10Mo monolithic nuclear fuel is being developed by the Material Management and Minimization Program, earlier known as the Reduced Enrichment for Research and Test Reactors Program, for utilization in research and test reactors around the world that currently use high-enriched uranium fuels. As part of this program, reactor experiments are being performed in the Advanced Test Reactor. It must be demonstrated that this fuel type exhibits mechanical integrity, geometric stability, and predictable behavior to high powers and high fission densities in order for it to be a viable fuel for qualification. This paper provides an overview of the microstructures observed at different regions of interest in fuel plates before and after irradiation for fuel samples that have been tested. These fuel plates were fabricated using laboratory-scale fabrication methods. Observations regarding how microstructural changes during irradiation may impact fuel performance are discussed.
International Nuclear Information System (INIS)
Raskach, A.
2001-01-01
Nowadays there are two completed design concepts of Nuclear Power Plants (NPPs) with the BN-800 type reactors developed with due regard for advanced safety requirements. One of them is the design of the fourth unit of the Beloyarsk Nuclear Power Plant; the other one is the design of three units of the South Ural Nuclear Power Plant. The both concepts are to use mixed oxide fuel (MOX fuel) based on civil plutonium. Studies on any project include economical analyses and cost of fuel is an essential parameter. In the course of the design works on the both projects such evaluations were done. For BN-800 on the Beloyarsk site nuclear fuel costs were taken from actual expenses of the BN-600 reactor and converted to rated thermal power and design capacity factor of the BN-800 and then increased by 20% in connection with turning to MOX fuel. Then this methodology was rewarding, but the ratio of uranium fuel and MOX fuel costs might change for the last years. For the project of three units of the South Ural Nuclear Power Plant nuclear fuel expenses were calculated from the data on a MOX fuel fabrication production facility (Complex-300). However, investigations performed recently shown that the methodology of economical assessments should be revised, as well as design and technology of MOX fuel fabrication at Complex-300 should be revised to meet all the existing safety requirements. Excepting there is a great bulk of civil plutonium to be reproduced, now we came up against the problem to utilize the exceeding ex-weapons plutonium that obviously can be used for MOX fuel fabrication as well. Construction of the MOX fuel fabrication facility - Complex-300 - was started in 1983. Its design output was planned to provide simultaneously 4 fast reactors of the BN-800 type with MOX fuel. By now about 50% of construction works (taking into account auxiliary buildings and arrangements) and 20% of installation works have been done at Complex-300. Along this, first works to construct
Fabrication of mixed oxide fuel using plutonium from dismantled weapons
International Nuclear Information System (INIS)
Blair, H.T.; Chidester, K.; Ramsey, K.B.
1996-01-01
A very brief summary is presented of experimental studies performed to support the use of plutonium from dismantled weapons in fabricating mixed oxide (MOX) fuel for commercial power reactors. Thermal treatment tests were performed on plutonium dioxide powder to determine if an effective dry gallium removal process could be devised. Fabrication tests were performed to determine the effects of various processing parameters on pellet quality. Thermal tests results showed that the final gallium content is highly dependent on the treatment temperature. Fabrication tests showed that the milling process, sintering parameters, and uranium feed did effect pellet properties. 1 ref., 1 tab
International Nuclear Information System (INIS)
Park, C. J.; Na, S. H.; Yang, J. H.; Kang, K. H.; Lee, J. W.
2009-01-01
During the pyroprocessing, most of the uranium is gathered in metallic form around a solid cathode during an electro-refining process, which is composed of about 94 weight percent of the spent fuel. In the previous study, a feasibility study has been done to reuse the recovered uranium for the CANDU reactor fuel following the traditional DUPIC (direct use of spent pressurized water reactor fuel into CANDU reactor) fuel fabrication process. However, the weight percent of U-235 in the recovered uranium is about 1 wt% and it is sufficiently re-utilized in a heavy water reactor which uses a natural uranium fuel. The reuse of recovered uranium will bring not only a huge economic profit and saving of uranium resources but also an alleviation of the burden on the management and the disposal of the spent fuel. The research on recycling of recovered uranium was carried out 10 years ago and most of the recovered uranium was assumed to be imported from abroad at that time. The preliminary results showed there is the sufficient possibility to recycle recovered uranium in terms of a reactor's characteristics as well as the fuel performance. However, the spent CANDU fuel is another issue in the storage and disposal problem. At present, most countries are considering that the spent CANDU fuel is disposed directly due to the low enrichment (∼0.5 wt%) of the discharge fissile content and lots of fission products. If mixing the spent CANDU fuel and the spent PWR fuel, the estimated uranium fissile enrichment will be about 0.6 wt% ∼ 1.0 wt% depending on the mixing ratio, which is sufficiently reusable in a CANDU reactor. Therefore, this paper deals with a feasibility study on the recovered uranium of the mixed spent fuel from the pyroprocessing. With the various mixing ratios between the PWR spent fuel and the CANDU spent fuel, a reactor characteristics including the safety parameters of the CANDU reactor was evaluated
Energy Technology Data Exchange (ETDEWEB)
Jang, Misuk; Jeon, Jong Seon; Kang, Hyun Sik; Kim, Seoung Rae [NESS, Daejeon (Korea, Republic of)
2016-10-15
In this paper, we would introduce and review technical standards related to sodium fire and plutonium criticality safety. This paper may be helpful to identify considerations in the development of equipment, standards, and etc., to meet the safety requirements in the design, construction and operating of TFFF, KAPF and SFR. The feasibility and conceptual designs are being examined on related facilities, for example, TRU Fuel Fabrication Facilities (TFFF), Korea Advanced Pyro-process Facility (KAPF), and Sodium Cooled Fast Reactor (SFR), in Korea. However, the safety concerns of these facilities have been controversial in part because of the Sodium fire accident and Plutonium related radiation safety caused by transport and handling accident. Thus, many researches have been performed to ensure safety and various documents including safety requirements have been developed. In separating and reducing the long-lived radioactive transuranic(TRU) in the spent nuclear fuel, reusing as the potential energy of uranium fuel resources and reducing the high level wastes, TFFF would be receiving the attention of many people. Thus, people would wonder whether compliance with technical standards that ensures safety. For new facility design, one of the important tasks is to review of technical standards, especially for sodium and Plutonium because of water related highly reactive characteristics and criticality hazard respectively. We have introduced and reviewed two important technical standards for TFFF, which are sodium fire and plutonium criticality safety, in this paper. This paper would provide a brief guidance, about how to start and what is important, to people who are responsible for the initial design to operation of TFFF.
International Nuclear Information System (INIS)
Jang, Misuk; Jeon, Jong Seon; Kang, Hyun Sik; Kim, Seoung Rae
2016-01-01
In this paper, we would introduce and review technical standards related to sodium fire and plutonium criticality safety. This paper may be helpful to identify considerations in the development of equipment, standards, and etc., to meet the safety requirements in the design, construction and operating of TFFF, KAPF and SFR. The feasibility and conceptual designs are being examined on related facilities, for example, TRU Fuel Fabrication Facilities (TFFF), Korea Advanced Pyro-process Facility (KAPF), and Sodium Cooled Fast Reactor (SFR), in Korea. However, the safety concerns of these facilities have been controversial in part because of the Sodium fire accident and Plutonium related radiation safety caused by transport and handling accident. Thus, many researches have been performed to ensure safety and various documents including safety requirements have been developed. In separating and reducing the long-lived radioactive transuranic(TRU) in the spent nuclear fuel, reusing as the potential energy of uranium fuel resources and reducing the high level wastes, TFFF would be receiving the attention of many people. Thus, people would wonder whether compliance with technical standards that ensures safety. For new facility design, one of the important tasks is to review of technical standards, especially for sodium and Plutonium because of water related highly reactive characteristics and criticality hazard respectively. We have introduced and reviewed two important technical standards for TFFF, which are sodium fire and plutonium criticality safety, in this paper. This paper would provide a brief guidance, about how to start and what is important, to people who are responsible for the initial design to operation of TFFF
Fabrication of metallic channel-containing UO2 fuels
International Nuclear Information System (INIS)
Yang, Jae Ho; Song, Kun Woo; Kim, Keon Sik; Jung, Youn Ho
2004-01-01
The uranium dioxide is widely used as a fuel material in the nuclear industry, owing to many advantages. But it has a disadvantage of having the lowest thermal conductivity of all kinds of nuclear fuels; metal, carbide, nitride. It is well known that the thermal conductivity of UO 2 fuel is enhanced by making, so called, the CERMET (ceramic-metal) composite which consists of both continuous body of highly thermal-conducting metal and UO 2 islands. The CERMET fuel fabrication technique needs metal phase of at least 30%, mostly more than 50%, of the volume of the pellet in order to keep the metal phase interconnected. This high volume fraction of metal requires such a high enrichment of U that the parasitic effect of metal should be compensated. Therefore, it is attractive to develop an innovative composite fuel that can form continuous metal phase with a small amount of metal. In this investigation, a feasibility study was made on how to make such an innovative fuel. Candidate metals (W, Mo, Cr) were selected, and fabrication process was conceptually designed from thermodynamic calculations. We have experimentally found that a metal phase envelops perfectly UO 2 grains, forming continuous channel throughout the pellet, and improving the thermal conductivity of pellet
Preliminary Evaluation of Alternate Designs for HFIR Low-Enriched Uranium Fuel
Energy Technology Data Exchange (ETDEWEB)
Renfro, David [ORNL; Chandler, David [ORNL; Cook, David [ORNL; Ilas, Germina [ORNL; Jain, Prashant [ORNL; Valentine, Jennifer [ORNL
2014-10-30
Engineering design studies of the feasibility of conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL) as part of an effort sponsored by the U.S. Department of Energy’s Global Threat Reduction Initiative (GTRI)/Reduced Enrichment for Research and Test Reactors (RERTR) program. The fuel type selected by the program for the conversion of the five high-power research reactors in the U.S. that still use HEU fuel is a new U-Mo monolithic fuel. Studies by ORNL have previously indicated that HFIR can be successfully converted using the new fuel provided (1) the reactor power can be increased from 85 MW to 100 MW and (2) the fuel can be fabricated to a specific reference design. Fabrication techniques for the new fuel are under development by the program but are still immature, especially for the “complex” aspects of the HFIR fuel design. In FY 2012, the program underwent a major shift in focus to emphasize developing and qualifying processes for the fabrication of reliable and affordable LEU fuel. In support of this new focus and in an effort to ensure that the HFIR fuel design is as suitable for reliable fabrication as possible, ORNL undertook the present study to propose and evaluate several alternative design features. These features include (1) eliminating the fuel zone axial contouring in the previous reference design by substituting a permanent neutron absorber in the lower unfueled region of all of the fuel plates, (2) relocating the burnable neutron absorber from the fuel plates of the inner fuel element to the side plates of the inner fuel element (the fuel plates of the outer fuel element do not contain a burnable absorber), (3) relocating the fuel zone inside the fuel plate to be centered on the centerline of the depth of the plate, and (4) reshaping the radial contour of the relocated fuel zone to be symmetric about this centerline. The
Preliminary Evaluation of Alternate Designs for HFIR Low-Enriched Uranium Fuel
Energy Technology Data Exchange (ETDEWEB)
Renfro, David G [ORNL; Chandler, David [ORNL; Cook, David Howard [ORNL; Ilas, Germina [ORNL; Jain, Prashant K [ORNL; Valentine, Jennifer R [ORNL
2014-11-01
Engineering design studies of the feasibility of conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL) as part of an effort sponsored by the U.S. Department of Energy s Global Threat Reduction Initiative (GTRI)/Reduced Enrichment for Research and Test Reactors (RERTR) program. The fuel type selected by the program for the conversion of the five high-power research reactors in the U.S. that still use HEU fuel is a new U-Mo monolithic fuel. Studies by ORNL have previously indicated that HFIR can be successfully converted using the new fuel provided (1) the reactor power can be increased from 85 MW to 100 MW and (2) the fuel can be fabricated to a specific reference design. Fabrication techniques for the new fuel are under development by the program but are still immature, especially for the complex aspects of the HFIR fuel design. In FY 2012, the program underwent a major shift in focus to emphasize developing and qualifying processes for the fabrication of reliable and affordable LEU fuel. In support of this new focus and in an effort to ensure that the HFIR fuel design is as suitable for reliable fabrication as possible, ORNL undertook the present study to propose and evaluate several alternative design features. These features include (1) eliminating the fuel zone axial contouring in the previous reference design by substituting a permanent neutron absorber in the lower unfueled region of all of the fuel plates, (2) relocating the burnable neutron absorber from the fuel plates of the inner fuel element to the side plates of the inner fuel element (the fuel plates of the outer fuel element do not contain a burnable absorber), (3) relocating the fuel zone inside the fuel plate to be centered on the centerline of the depth of the plate, and (4) reshaping the radial contour of the relocated fuel zone to be symmetric about this centerline. The present
Fuel reprocessing/fabrication interface
International Nuclear Information System (INIS)
Benistan, G.; Blanchon, T.; Galimberti, M.; Mignot, E.
1987-01-01
EDF has conducted a major research, development and experimental programme concerning the recycling of plutonium and reprocessed uranium in pressurized water reactors, in collaboration with its major partners in the nuclear fuel cycle industry. Studies already conducted have demonstrated the technical and economic advantages of this recycling, as also its feasibility with due observance of the safety and reliability criteria constantly applied throughout the industrial development of the nuclear power sector in France. Data feedback from actual experience will make it possible to control the specific technical characteristics of MOX and reprocessed uranium fuels to a higher degree, as also management, viewed from the economic standpoint, of irradiated fuels and materials recovered from reprocessing. The next step will be to examine the reprocessing of MOX for reprocessed uranium fuels, either for secondary recycling in the PWR units, or, looking further ahead, in the fast breeders or later generation PWR units, after a storage period of a few years
Economic analysis of thorium-uranium fuel cycle introduced into PWRs
International Nuclear Information System (INIS)
Fan Li; Sun Qian
2014-01-01
Using PWR of Daya Bay Unit l as the reference reactor, a validated computer code was used to calculate the fuel cycle costs for uranium fuel cycle and thorium-uranium fuel cycle over the following 20 0perational years respectively. The calculation results show that the thorium-uranium fuel cycle is economically competitive with the uranium fuel cycle when reprocessing mode is adopted. For thorium-uranium fuel cycle, if the price of natural uranium is higher than 120 $ /pound U_3O_8, the fuel cycle cost of the direct disposal mode is greater than that of the reprocessing mode. Therefore, when the uranium price may maintain a high level long-termly, adopting reprocessing mode will benefit the economic advantage for the thorium-uranium fuel cycle introduced into PWRs. (authors)
International Nuclear Information System (INIS)
Chandrasekaran, S.; Rajagopal, V.; Jose, M.T.; Venkatraman, B.
2012-01-01
A Fast Reactor Fuel Cycle Facility (FRFCF) being planned at Indira Gandhi Centre for Atomic Research, Kalpakkam is an integrated facility with head end and back end of fuel cycle plants co-located in a single place, to meet the refuelling needs of the prototype fast breeder reactor (PFBR). Reprocessed uranium oxide plant (RUP) is one such plant in FRFCF to built to meet annual requirements of UO 2 for fabrication of fuel sub-assemblies (FSAs) and radial blanket sub-assemblies (RSAs) for PFBR. RUP receives reprocessed uranium oxide powder (U 3 O 8 ) from fast reactor fuel reprocessing plant (FRP) of FRFCF. Unlike natural uranium oxide plant, RUP has to handle reprocessed uranium oxide which is likely to have residual fission products activity in addition to traces of plutonium. As the fuel used for PFBR is recycled within these plants, formation of higher actinides in the case of plutonium and formation of higher levels of 232 U in the uranium product would be a radiological problem to be reckoned with. The paper discussed the impact of handling of multi-recycled reprocessed uranium in RUP and the radiological considerations
Energy Technology Data Exchange (ETDEWEB)
Santos, Maristela Souza
2006-07-01
The International Commission on Radiological Protection (ICRP) Publication 66 and Supporting Guidance 3) strongly recommends that specific information on lung retention parameters should be used in preference to default values wherever appropriate, for the derivation of effective doses and for bioassay interpretation of monitoring data. A group of 81 workers exposed to UO{sub 2} at the fuel fabrication facility in Brazil was selected to evaluate the committed effective dose. The workers were monitored for determination of uranium content in the urinary and faecal excretion. The contribution of intakes by ingestion and inhalation were assessed on the basis of the ratios of urinary to fecal excretion. For the selected workers it was concluded that inhalation dominated intake. According to ICRP 66, uranium oxide is classified as insoluble Type S compound. The ICRP Supporting Guidance 3 and some recent studies have recommended specific lung retention parameters to UO{sub 2}. The solubility parameters of the uranium oxide compound handled by the workers at the fuel fabrication facility in Brazil was evaluated on the basis of the ratios of urinary to fecal excretion. Excretion data were corrected for dietary intakes. This paper will discuss the application of lung retention parameters recommended by the ICRP models to these data and also the dependence of the effective committed dose on the lung retention parameters. It will also discuss the problems in the interpretation of monitoring results, when the worker is exposed to several uranium compounds of different solubilities. (author)
Inert matrix fuel in dispersion type fuel elements
Energy Technology Data Exchange (ETDEWEB)
Savchenko, A.M. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation)]. E-mail: sav@bochvar.ru; Vatulin, A.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Morozov, A.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Sirotin, V.L. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Dobrikova, I.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Kulakov, G.V. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Ershov, S.A. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Kostomarov, V.P. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation); Stelyuk, Y.I. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (VNIINM) 123060, P.O. Box 369, Rogova Street, 5A, Moscow (Russian Federation)
2006-06-30
The advantages of using inert matrix fuel (IMF) as a dispersion fuel in an aluminium alloy matrix are considered, in particular, low temperatures in the fuel centre, achievable high burn-ups, serviceability in transients and an environmentally friendly process of fuel rod fabrication. Two main versions of IMF are under development at A.A. Bochvar Institute, i.e. heterogeneous or isolated distribution of plutonium. The out-of-pile results on IMF loaded with uranium dioxide as plutonium simulator are presented. Fuel elements with uranium dioxide composition fabricated at A.A. Bochvar Institute are currently under MIR tests (RIAR, Dimitrovgrad). The fuel elements reached a burn-up of 88 MW d kg{sup -1} (equivalent to the burn up of the standard uranium dioxide pelletized fuel) without loss of leak-tightness of the cladding. The feasibility of fabricating IMF of these particular types with plutonium dioxide is considered with a view to in-pile irradiation.
Inert matrix fuel in dispersion type fuel elements
Savchenko, A. M.; Vatulin, A. V.; Morozov, A. V.; Sirotin, V. L.; Dobrikova, I. V.; Kulakov, G. V.; Ershov, S. A.; Kostomarov, V. P.; Stelyuk, Y. I.
2006-06-01
The advantages of using inert matrix fuel (IMF) as a dispersion fuel in an aluminium alloy matrix are considered, in particular, low temperatures in the fuel centre, achievable high burn-ups, serviceability in transients and an environmentally friendly process of fuel rod fabrication. Two main versions of IMF are under development at A.A. Bochvar Institute, i.e. heterogeneous or isolated distribution of plutonium. The out-of-pile results on IMF loaded with uranium dioxide as plutonium simulator are presented. Fuel elements with uranium dioxide composition fabricated at A.A. Bochvar Institute are currently under MIR tests (RIAR, Dimitrovgrad). The fuel elements reached a burn-up of 88 MW d kg-1 (equivalent to the burn up of the standard uranium dioxide pelletized fuel) without loss of leak-tightness of the cladding. The feasibility of fabricating IMF of these particular types with plutonium dioxide is considered with a view to in-pile irradiation.
International Nuclear Information System (INIS)
Wiencek, T.C.; Domagala, R.F.; Thresh, H.R.
1990-01-01
Two embodiments of a high uranium fuel plate are disclosed which contain a meat comprising structured uranium compound confined between a pari of diffusion bonded ductile metal cladding plates uniformly covering the meat, the meat hiving a uniform high fuel loading comprising a content of uranium compound greater than about 45 Vol. % at a porosity not greater than about 10 Vol. %. In a first embodiment, the meat is a plurality of parallel wires of uranium compound. In a second embodiment, the meat is a dispersion compact containing uranium compound. The fuel plates are fabricated by a hot isostatic pressing process
International Nuclear Information System (INIS)
Ghosh, J.K.; Pandey, V.D.; Rao, T.S.; Kutty, T.R.G.; Kurup, P.K.D.; Joseph, J.K.; Ganguly, C.
1993-01-01
Ever since the inception of Radiometallurgy Laboratory (RML) in its early seventies optical metallography has played a key role in development and fabrication of plutonium, uranium and thorium bearing nuclear fuels. In this report, an album of photomicrographs depicts the different types of metallic, ceramic and dispersion fuels and welded section that have been evaluated in RML during the last two decades. (author). 14 refs., 1 tab
Development of CANFLEX fuel fabrication technology
Energy Technology Data Exchange (ETDEWEB)
Kang, M. S.; Choi, C. B.; Park, C. H.; Kwon, W. J.; Kim, C. H.; Kim, B. J.; Koo, C. H.; Cho, D. S.; So, D. Y.; Suh, S. W.; Park, C. J.; Chang, D. H.; Yun, S. H. [KEPCO Nuclear Fuel Company, Taejeon (Korea)
2000-04-01
Wolsong Unit 1 as the first heavy water reactor in Korea has been in service for 17 years since 1983. It would be about the time to prepare a plan for the solution of problems due to aging of the reactor. The aging of CANDU reactor could lead especially to the steam generator cruding and pressure tube sagging and creep and then decreases the operation margin to make some problems on reactor operations and safety. The counterplan could be made in two ways. One is to repair or modify reactor itself. The other is to develop new advanced fuel to increase of CANDU operation margin effectively, so as to compensate the reduced operation margin. Therefore, the first objectives in the present R and D is to develop the CANFLEX-NU(CANDU Flexible fuelling-Natural Uranium) fuel as a CANDU advanced fuel. One of the improvements in CANDU fuel fabrication technology, and advanced method of Zr-Be brazing was developed. For the formation of Zr-Be alloy, preheating and main heating temperature in the furnace is 700 deg C, 1200 deg C respectively. In order to find an appropriate material for the brazing joints in the CANDU fuel, the composition of Zr based amorphous metals were designed. And, the effect of hydrogen on the mechanical properties of cladding sheath and feasibility of the eddy current test to evaluate quality of end cap weld were also studied for the fundamental research purpose. As a preliminary study to suggest optimal way for the mass production of CANFLEX-NU fuel at KNFC the existing CANDU fuel facilities and fabrication/inspection processes were reviewed. The best way is that the current CANDU facility shall be modified to produce small diametrial CANFLEX elements and a new facility shall be constructed to produce large diametrial CANFLEX fuel elements. 46 refs., 99 figs., 10 tabs. (Author)
Energy Technology Data Exchange (ETDEWEB)
Kulakov, Mykola, E-mail: mykola.kulakov@cnl.ca [Fuel Development Branch, Canadian Nuclear Laboratories, Chalk River, ON K0J 1J0 Canada (Canada); Saoudi, Mouna [Fuel Development Branch, Canadian Nuclear Laboratories, Chalk River, ON K0J 1J0 Canada (Canada); Piro, Markus H.A. [Fuel and Fuel Channel Safety Branch, Canadian Nuclear Laboratories, Chalk River, ON K0J 1J0 Canada (Canada); Donaberger, Ronald L. [Canadian Neutron Beam Centre, Chalk River, ON K0J 1J0 Canada (Canada)
2017-02-15
Aluminum-clad U-7Mo/Mg and U-10Mo/Mg pin-type mini-elements (with a core uranium loading of 4.5 gU/cm{sup 3}) have been fabricated at the Canadian Nuclear Laboratories for experimental tests and ultimately for use in research and test reactors. In this study, the microstructure and phase composition of unirradiated U-7Mo/Mg and U-10Mo/Mg fuel cores were analyzed using optical and scanning electron microscopy, and neutron powder diffraction. Thermal properties were characterized using a combination of experimental measurements and thermodynamic calculations. The thermal diffusivity was measured using the laser flash method. The temperature-dependent specific heat capacities were calculated based on the linear rule of mixture using the weight fraction of different crystalline phases and their specific heat capacity values taken from the literature. The thermal conductivity was then calculated using the measured thermal diffusivity, the measured density and the calculated specific heat capacity. The resulting thermal conductivity is practically identical for both types of fuel. The in-reactor temperatures were predicted using conjugate heat transfer simulations. - Highlights: • Neutron diffraction analysis shows that most of the γ-U(Mo) phase was retained in as-fabricated U-7Mo/Mg and U-10Mo/Mg fuel cores. • The experimental thermal conductivity of both types of fuel is practically identical. • Based on conjugate heat transfer simulations, under normal operating conditions, the in-reactor fuel centreline temperature is about 510 K.
Metallic Reactor Fuel Fabrication for SFR
Energy Technology Data Exchange (ETDEWEB)
Song, Hoon; Kim, Jong-Hwan; Ko, Young-Mo; Woo, Yoon-Myung; Kim, Ki-Hwan; Lee, Chan-Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2015-05-15
The metal fuel for an SFR has such advantages such as simple fabrication procedures, good neutron economy, high thermal conductivity, excellent compatibility with a Na coolant, and inherent passive safety 1. U-Zr metal fuel for SFR is now being developed by KAERI as a national R and D program of Korea. The fabrication technology of metal fuel for SFR has been under development in Korea as a national nuclear R and D program since 2007. The fabrication process for SFR fuel is composed of (1) fuel slug casting, (2) loading and fabrication of the fuel rods, and (3) fabrication of the final fuel assemblies. Fuel slug casting is the dominant source of fuel losses and recycled streams in this fabrication process. Fabrication on the rod type metallic fuel was carried out for the purpose of establishing a practical fabrication method. Rod-type fuel slugs were fabricated by injection casting. Metallic fuel slugs fabricated showed a general appearance was smooth.
Fabrication of Fast Reactor Fuel Pins for Test Irradiations
Energy Technology Data Exchange (ETDEWEB)
Karsten, G. [Institute for Applied Reactor Physics, Kernforschungszentrum Karlsruhe, Karlsruhe, Federal Republic of Germany (Germany); Dippel, T. [Institute for Radiochemistry, Kernforschungszentrum Karlsruhe, Karlsruhe, Federal Republic of Germany (Germany); Laue, H. J. [Institute for Applied Reactor Physics, Kernforschungszentrum Karlsruhe, Karlsruhe, Federal Republic of Germany (Germany)
1967-09-15
An extended irradiation programme is being carried out for the fuel element development of the Karlsruhe fast breeder project. A very important task within the programme is the testing of plutonium-containing fuel pins in a fast-reactor environment. This paper deals with fabrication of such pins by our laboratories at Karlsruhe. For the fast reactor test positions at present envisaged a fuel with 15% plutonium and the uranium fully enriched is appropriate. Hie mixed oxide is both pelletized and vibro-compacted with smeared densities between 80 and 88% theoretical. The pin design is, for example, such that there are two gas plena at the top and bottom, and one blanket above the fuel with the fuel zone fitting to the test reactor core length. The specifications both for fuel and cladding have been adapted to the special purpose of a fast-breeder reactor - the outer dimensions, the choice of cladding and fuel types, the data used and the kind of tests outline the targets of the development. The fuel fabrication is described in detail, and also the powder line used for vibro-compaction. The source materials for the fuel are oxalate PuO{sub 2} and UO{sub 2} from the UF{sub 6} process. The special problems of mechanical mixing and of plutonium homogeneity have been studied. The development of the sintering technique and grain characteristics for vibratory compactive fuel had to overcome serious problems in order to reach 82-83% theoretical. The performance of the pin fabrication needed a major effort in welding, manufacturing of fits and decontamination of the pin surfaces. This was a stimulation for the development of some very subtle control techniques, for example taking clear X-ray photographs and the tube testing. In general the selection of tests was a special task of the production routine. In conclusion the fabrication of the pins resulted in valuable experiences for the further development of fast reactor fuel elements. (author)
Hydrothermal synthesis for fabrication and reprocessing of MOX nuclear fuel
International Nuclear Information System (INIS)
Ohta, Suguru; Yamamura, Tomoo; Shirasaki, Kenji; Satoh, Isamu; Shikama, Tatsuo
2011-01-01
To improve the nuclear proliferation resistance and to minimize use of chemicals, a new reprocessing and fabrication process of 'mixed oxide' (MOX) fuel was proposed and studied by using simulated spent fuel solutions. The process is consisting of the two steps, i.e. the removal of fission product (FP) from dissolved spent fuel by using carbonate solutions (Step-1), and hydrothermal synthesis of uranium dioxides (Step-2). In Step-1, rare earth (the precipitation ratio: 90%) and alkaline earth (10-50% for Sr) as FP were removed based on their low solubility of hydroxides and carbonate salts, with uranium kept dissolved for the certain carbonate solutions of weak base (Type 2) or mixtures of relatively strong base and weak base (Type 3). In Step-2, the features of uranium dioxides UO 2+x particles, i.e. stoichiometry (x=0.05-0.2), size (0.2-3 μm) and shape (cubic, spherical, rectangular parallelpiped, etc.), were controlled, and the cesium was removed down to 40 ppm by an addition of organic additives. The decontamination factors (DF) for cesium exceeds 10 5 , whereas the total DF of all the simulated FP were as low as the order of 10 which requires future studies for removal of alkaline earth, Re and Tc etc. (author)
International Nuclear Information System (INIS)
Gough, J.; Wredberg, L.; Zobor, E.; Zuccaro-Labellarte, G.
1988-01-01
IAEA has implemented safeguards at three Low Enriched Uranium (LEU) fuel fabrication plants in the USA during the period 1982 to 1987, and it is in the process of safeguarding a fourth plant from 01 January 1988. In Canada IAEA safeguards inspections were implemented at all Natural Uranium (NU) fuel fabrication plants form 1972 onwards, and there are, at present, three plants under safeguards. The direct responsibility for the implementation of safeguards inspections in the USA and Canada lies with the Division of Operations B (SGOB) within the IAEA Department of Safeguards. The senior staff that is at present directly engaged in the implementation activities has accumulated supervising inspection experience at about 50 Physical Inventory Verification (PIV) inspections at the Canadian and US fabrication plants during the period 1978 to 1987. This experience has been gained in close cooperation with the facility operators and with the support of the state authorities. The paper describes the latest PIV inspections at the Westinghouse Columbia plant and the Zircatec Precision Industries Inc. Port Hope plant. Furthermore, the paper describes the initial activities for the 1988 PIV inspection at the General Electric Wilmington plant including computerized book audit activities
Quality assurance in the manufacture of metallic uranium fuel for research reactors
International Nuclear Information System (INIS)
Shah, B.K.; Kumar, Arbind; Nanekar, P.P.; Vaidya, P.R.
2009-01-01
Two Research Reactors viz. CIRUS and DHRUVA are operating at Trombay since 1960 and 1985 respectively. Cirus is a 40 MWth reactor using heavy water as moderator and light water as coolant. Dhruva is a 100 MWth reactor using heavy water as moderator and coolant. The maximum neutron flux of these reactors are 6.7 x 10 13 n/cm 2 /s (Cirus) and 1.8 x 10 14 n/cm 2 /s (Dhruva). Both these reactors are used for basic research, R and D in reactor technology, isotope production and operator training. Fuel material for these reactors is natural uranium metallic rods claded in finned aluminium (99.5%) tubes. This presentation will discuss various issues related to fabrication quality assurance and reactor behavior of metallic uranium fuel used in research reactors
Quality control in nuclear fuel fabrication on the inspection basis
International Nuclear Information System (INIS)
Fuentes S, A.
1997-01-01
Every plant productive of electric power requires the use of energetics for the transformation to electricity. In the nucleo electric plant the energetic is the uranium, in which it makes ensembles and is used as fuel in the reactor. To assure that the fuel ensembles fulfill the specifications and requirements of design stipulated in the nucleo electric plant is that under a quality control through inspections during the fabrication process. The purpose of this work is to study and verify that the lineaments of the standard 10 CFR 50 appendix B 'Quality assurement for nuclear plants' specially in the criteria 'Inspections' that is used to guarantee the quality of the ensembles. This standard is the one that rules every activity and operation inside the pilot plant and its established in the quality program in the production of nuclear fuel for the Laguna Verde plant. The quality of the assemble is verified through each one of the tests or inspections due to the importance of it in the fabrication of fuel. (Author)
International Nuclear Information System (INIS)
Carter, W.L.; Olsen, A.R.
1979-06-01
The classification of water-cooled nuclear reactors offers a number of fuel cycles that present inherently low risk of weapons proliferation while making power available to the international community. Eight fuel cycles in light water reactor (LWR), heavy water reactor (HWR), and the spectral shift controlled reactor (SSCR) systems have been proposed to promote these objectives in the International Fuel Cycle Evaluation (INFCE) program. Each was examined in an effort to provide technical and economic data to INFCE on fuel fabrication, refabrication, and reprocessing for an initial comparison of alternate cycles. The fuel cycles include three once-through cycles that require only fresh fuel fabrication, shipping, and spent fuel storage; four cycles that utilize denatured uranium--thorium and require all recycle operations; and one cycle that considers the LWR--HWR tandem operation requiring refabrication but no reprocessing
Energy Technology Data Exchange (ETDEWEB)
Carter, W.L.; Olsen, A.R.
1979-06-01
The classification of water-cooled nuclear reactors offers a number of fuel cycles that present inherently low risk of weapons proliferation while making power available to the international community. Eight fuel cycles in light water reactor (LWR), heavy water reactor (HWR), and the spectral shift controlled reactor (SSCR) systems have been proposed to promote these objectives in the International Fuel Cycle Evaluation (INFCE) program. Each was examined in an effort to provide technical and economic data to INFCE on fuel fabrication, refabrication, and reprocessing for an initial comparison of alternate cycles. The fuel cycles include three once-through cycles that require only fresh fuel fabrication, shipping, and spent fuel storage; four cycles that utilize denatured uranium--thorium and require all recycle operations; and one cycle that considers the LWR--HWR tandem operation requiring refabrication but no reprocessing.
A Study on the Fabrication of Uranium-Cadmium Alloy and its Distillation Behavior
International Nuclear Information System (INIS)
Kim, Ji Yong; Ahn, Do Hee; Kim, Kwang Rag; Paek, Seung Woo; Kim, Si Hyung
2010-01-01
The pyrometallurgical nuclear fuel recycle process, called pyroprocessing, has been known as a promising nuclear fuel recycling technology. Pyroprocessing technology is crucial to advanced nuclear systems due to increased nuclear proliferation resistance and economic efficiency. The basic concept of pyroprocessing is group actinide recovery, which enhances the nuclear proliferation resistance significantly. One of the key steps in pyroprocessing is 'electrowinning' which recovers group actinides with lanthanide from the spent nuclear fuels. In this study, a vertical cadmium distiller was manufactured. The evaporation rate of pure cadmium in vertical cadmium distiller varied from 12.3 to 40.8 g/cm 2 /h within a temperature range of 773 ∼ 923 K and pressure below 0.01 torr. Uranium - cadmium alloy was fabricated by electrolysis using liquid cadmium cathode in a high purity argon atmosphere glove box. The distillation behavior of pure cadmium and cadmium in uranium - cadmium alloy was investigated. The distillation behavior of cadmium from this study could be used to develop an actinide recovery process from a liquid cadmium cathode in a cadmium distiller
A Study on the Fabrication of Uranium-Cadmium Alloy and its Distillation Behavior
Energy Technology Data Exchange (ETDEWEB)
Kim, Ji Yong [University of Science and Technology, Daejeon (Korea, Republic of); Ahn, Do Hee; Kim, Kwang Rag; Paek, Seung Woo; Kim, Si Hyung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2010-12-15
The pyrometallurgical nuclear fuel recycle process, called pyroprocessing, has been known as a promising nuclear fuel recycling technology. Pyroprocessing technology is crucial to advanced nuclear systems due to increased nuclear proliferation resistance and economic efficiency. The basic concept of pyroprocessing is group actinide recovery, which enhances the nuclear proliferation resistance significantly. One of the key steps in pyroprocessing is 'electrowinning' which recovers group actinides with lanthanide from the spent nuclear fuels. In this study, a vertical cadmium distiller was manufactured. The evaporation rate of pure cadmium in vertical cadmium distiller varied from 12.3 to 40.8 g/cm{sup 2}/h within a temperature range of 773 {approx} 923 K and pressure below 0.01 torr. Uranium - cadmium alloy was fabricated by electrolysis using liquid cadmium cathode in a high purity argon atmosphere glove box. The distillation behavior of pure cadmium and cadmium in uranium - cadmium alloy was investigated. The distillation behavior of cadmium from this study could be used to develop an actinide recovery process from a liquid cadmium cathode in a cadmium distiller.
Chemical aspects of nuclear fuel fabrication processes
Energy Technology Data Exchange (ETDEWEB)
Naylor, A; Ellis, J F; Watson, R H
1986-04-01
Processes used by British Nuclear Fuels plc for the conversion of uranium ore concentrates to uranium metal and uranium hexafluoride, are reviewed. Means of converting the latter compound, after enrichment, to sintered UO/sub 2/ fuel bodies are also described. An overview is given of the associated chemical engineering technology.
Slightly enriched uranium fuel for a PHWR
International Nuclear Information System (INIS)
Notari, C.; Marajofsky, A.
1997-01-01
An improved fuel element design for a PHWR using slightly enriched uranium fuel is presented. It maintains the general geometric disposition of the currently used in the argentine NPP's reactors, replacing the outer ring of rods by rods containing annular pellets. Power density reduction is achieved with modest burnup losses and the void volume in the pellets can be used to balance these two opposite effects. The results show that with this new design, the fuel can be operated at higher powers without violating thermohydraulic limits and this means an improvement in fuel management flexibility, particularly in the transition from natural uranium to slightly enriched uranium cycle. (author)
Performance evaluation of WDXRF as a process control technique for MOX fuel fabrication
International Nuclear Information System (INIS)
Pandey, A.; Khan, F.A.; Das, D.K.; Behere, P.G.; Afzal, Mohd
2015-01-01
This paper presents studies on Wavelength Dispersive X-Ray Fluorescence (WDXRF), as a powerful non destructive technique (NDT) for the compositional analysis of various types of MOX fuels. The sample has come after mixing and milling of UO 2 and PuO 2 powder for the estimation of plutonium, as a process control step of fabrication of (U, Pu)O 2 mixed oxide (MOX) fuel. For the characterization for heavy metal in various MOX fuel, a WDXRF method was established as a process control technique. The attractiveness of our system is that it can analyze the samples in solid form as well as in liquid form. The system is adapted in a glove box for handling of plutonium based fuels. The glove box adapted system was optimized with Uranium and Thorium based MOX sample before introduction of Pu. Uranium oxide and thorium oxide have been estimated in uranium thorium MOX samples. Standard deviation for the analysis of U 3 O 8 and ThO 2 were found to be 0.14 and 0.15 respectively. The results are validated against the conventional wet chemical methods of analysis. (author)
International Nuclear Information System (INIS)
Patzer, G.; Geier, F.
1988-01-01
The fuel assembly NY-205 has been irradiated during the first and the second core of KNK II with a total residence time of 832 equivalent full-power days. A maximum burnup of 175.000 MWd/tHM or 18.6 % was reached with a maximum steel damage of 66 dpa-NRT. For the cladding the materials 1.4970 and 1.4981 have been used in different metallurgical conditions, and for the Uranium/Plutonium mixed- oxide fuel the most important variants of the major fabrication parameters had been realized. The assembly will be brought to the Hot Cells of the KfK Karlsruhe for post-irradiation examination in February 1988, so that the knowledge of the fabrication data is of interest for the selection of fuel pins and for the evaluation of the examination results. Therefore this report compiles the fuel and fuel pin fabrication data from the INTERATOM data bank system BESEX and additionally, an overview of the irradiation history of the assembly is given [de
Fabrication of chamfered uranium-plutonium mixed carbide pellets
International Nuclear Information System (INIS)
Arai, Yasuo; Iwai, Takashi; Shiozawa, Kenichi; Handa, Muneo
1985-10-01
Chamfered uranium-plutonium mixed carbide pellets for high burnup irradiation test in JMTR were fabricated in glove boxes with purified argon gas. The size of die and punch in a press was decided from pellet densities and dimensions including the angle of chamfered parts. No chip or crack caused by adopting chamfered pellets was found in both pressing and sintering stages. In addition to mixed carbide pellets, uranium carbide pellets used as insulators were also successfully fabricated. (author)
International Nuclear Information System (INIS)
Choi, Hangbok; Ko, Won Il; Yang, Myung Seung
2001-01-01
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
Refining of crude uranium by solvent extraction for production of nuclear pure uranium metal
International Nuclear Information System (INIS)
Gupta, S.K.; Manna, S.; Singha, M.; Hareendran, K.N.; Chowdhury, S.; Satpati, S.K.; Kumar, K.
2007-01-01
Uranium is the primary fuel material for any nuclear fission energy program. Natural uranium contains only 0.712% of 235 U as fissile constituent. This low concentration of fissile isotope in natural uranium calls for a very high level of purity, especially with respect to neutron poisons like B, Cd, Gd etc. before it can be used as nuclear fuel. Solvent extraction is a widely used technique by which crude uranium is purified for reactor use. Uranium metal plant (UMP), BARC, Trombay is engaged in refining of uranium concentrate for production of nuclear pure uranium metal for fabrication of fuel for research reactors. This paper reviews some of the fundamental aspects of this refining process with some special references to UMP, BARC. (author)
Measurement of enriched uranium and uranium-aluminum fuel materials with the AWCC
International Nuclear Information System (INIS)
Krick, M.S.; Menlove, H.O.; Zick, J.; Ikonomou, P.
1985-05-01
The active well coincidence counter (AWCC) was calibrated at the Chalk River Nuclear Laboratories (CRNL) for the assay of 93%-enriched fuel materials in three categories: (1) uranium-aluminum billets, (2) uranium-aluminum fuel elements, and (3) uranium metal pieces. The AWCC was a standard instrument supplied to the International Atomic Energy Agency under the International Safeguards Project Office Task A.51. Excellent agreement was obtained between the CRNL measurements and previous Los Alamos National Laboratory measurements on similar mockup fuel material. Calibration curves were obtained for each sample category. 2 refs., 8 figs., 15 tabs
International Nuclear Information System (INIS)
Demarthon, F.; Donnars, O.; Dupuy-Maury, F.
2002-01-01
This dossier gives a broad overview of the present day status of the nuclear fuel cycle in France: 1 - the revival of nuclear power as a solution to the global warming and to the increase of worldwide energy needs; 2 - the security of uranium supplies thanks to the reuse of weapon grade highly enriched uranium; 3 - the fabrication of nuclear fuels from the mining extraction to the enrichment processes, the fabrication of fuel pellets and the assembly of fuel rods; 4 - the new composition of present day fuels (UO x and chromium-doped pellets); 5 - the consumption of plutonium stocks and the Corail and Apa fuel assemblies for the reduction of plutonium stocks and the preservation of uranium resources. (J.S.)
Potential health hazard of nuclear fuel waste and uranium ore
International Nuclear Information System (INIS)
Mehta, K.; Sherman, G.R.; King, S.G.
1991-06-01
The variation of the radioactivity of nuclear fuel waste (used fuel and fuel reprocessing waste) with time, and the potential health hazard (or inherent radiotoxicity) resulting from its ingestion are estimated for CANDU (Canada Deuterium Uranium) natural-uranium reactors. Four groups of radionuclides in the nuclear fuel waste are considered: actinides, fission products, activation products of zircaloy, and activation products of fuel impurities. Contributions from each of these groups to the radioactivity and to the potential health hazard are compared and discussed. The potential health hazard resulting from used fuel is then compared with that of uranium ore, mine tailings and refined uranium (fresh fuel) on the basis of equivalent amounts of uranium. The computer code HAZARD, specifically developed for these computations, is described
International Nuclear Information System (INIS)
1998-05-01
After a short introduction about nuclear power in the world, fission physics and the French nuclear power plants, this brochure describes in a digest way the different steps of the nuclear fuel cycle: uranium prospecting, mining activity, processing of uranium ores and production of uranium concentrates (yellow cake), uranium chemistry (conversion of the yellow cake into uranium hexafluoride), fabrication of nuclear fuels, use of fuels, reprocessing of spent fuels (uranium, plutonium and fission products), recycling of energetic materials, and storage of radioactive wastes. (J.S.)
International Nuclear Information System (INIS)
Jones, R.J.
1984-09-01
Some States have expressed a need for more detailed guidance with regard to the technical elements in the design and operation of SSACs for both the national and the international objectives. To meet this need the present document has been prepared, describing the technical elements of an SSAC in considerable detail. The purpose of this document is therefore, to provide a detailed description of a system for the accounting for and control of nuclear material in a model low enriched uranium conversion and fuel fabrication facility which can be used by a facility operator to establish his own system in a way which will provide the necessary information for compliance with a national system for nuclear material accounting and control and for the IAEA to carry out its safeguards responsibilities
One year of operation of the Belgonucleaire (Dessel) plutonium fuel fabrication plant
International Nuclear Information System (INIS)
Leblanc, J.M.
1975-01-01
Based on experience with plutonium since 1958, Belgonucleaire has successively launched a pilot plant and then a fuel fabrication plant for mixed uranium and plutonium oxides in 1968 and 1973 respectively. After describing briefly the plants and the most important stages in the planning, construction and operation of the Dessel plant, the present document describes the principal problems which were met during the course of operation of the plant and their direct incidence on the capacity and quality of the production of fuel elements
International Nuclear Information System (INIS)
Dimayuga, F.C.; Jeffs, A.T.
1995-01-01
Atomic Energy of Canada Limited's mixed-oxide (MOX) fuel fabrication activities are conducted in the Recycle Fuel Fabrication Laboratories (RFFL) at the Chalk River Laboratories. The RFFL facility is designed to produce experimental quantities of CANDU MOX fuel for reactor physics tests or demonstration irradiations. From 1979 to 1987, several MOX fuel fabrication campaigns were run in the RFFL, producing various quantities of fuel with different compositions. About 150 bundles, containing over three tonnes of MOX, were fabricated in the RFFL before operations in the facility were suspended. In late 1987, the RFFL was placed in a state of active standby, a condition where no fuel fabrication activities are conducted, but the monitoring and ventilation systems in the facility are maintained. Currently, a project to rehabilitate the RFFL and resume MOX fuel fabrication is nearing completion. This project is funded by the CANDU Owners' Group (COG). The initial fabrication campaign will consist of the production of thirty-eight 37-element (U,Pu)O 2 bundles containing 0.2 wt% Pu in Heavy Element (H.E.) destined for physics tests in the zero-power ZED-2 reactor. An overview of the Rehabilitation Project will be given. (author)
Product Conversion: The Link between Separations and Fuel Fabrication
International Nuclear Information System (INIS)
Felker, L.K.; Vedder, R.J.; Walker, E.A.; Collins, E.D.
2008-01-01
Several chemical processing flowsheets are under development for the separation and isolation of the actinide, lanthanide, and fission product streams in spent nuclear fuel. The conversion of these product streams to solid forms, typically oxides, is desired for waste disposition and recycle of product fractions back into transmutation fuels or targets. The modified direct denitration (MDD) process developed at Oak Ridge National Laboratory (ORNL) in the 1980's offers significant advantages for the conversion of the spent fuel products to powder form suitable for direct fabrication into recycle fuels. A glove-box-contained MDD system and a fume-hood-contained system have been assembled at ORNL for the purposes of testing the co-conversion of uranium and mixed-actinide products. The current activities are focused on the conversion of the first products from the processing of spent nuclear fuel in the Coupled End-to-End Demonstration currently being conducted at ORNL. (authors)
Nuclear power and its fuel cycle
International Nuclear Information System (INIS)
Wymer, R.G.
1986-01-01
A series of viewgraphs describes the nuclear fuel cycle and nuclear power, covering reactor types, sources of uranium, enrichment of uranium, fuel fabrication, transportation, fuel reprocessing, and radioactive wastes
Fuel powder production from ductile uranium alloys
International Nuclear Information System (INIS)
Clark, C.R.; Meyer, M.K.
1998-01-01
Metallic uranium alloys are candidate materials for use as the fuel phase in very-high-density LEU dispersion fuels. These ductile alloys cannot be converted to powder form by the processes routinely used for oxides or intermetallics. Three methods of powder production from uranium alloys have been investigated within the US-RERTR program. These processes are grinding, cryogenic milling, and hydride-dehydride. In addition, a gas atomization process was investigated using gold as a surrogate for uranium. (author)
Method of fabricating a uranium-bearing foil
Gooch, Jackie G [Seymour, TN; DeMint, Amy L [Kingston, TN
2012-04-24
Methods of fabricating a uranium-bearing foil are described. The foil may be substantially pure uranium, or may be a uranium alloy such as a uranium-molybdenum alloy. The method typically includes a series of hot rolling operations on a cast plate material to form a thin sheet. These hot rolling operations are typically performed using a process where each pass reduces the thickness of the plate by a substantially constant percentage. The sheet is typically then annealed and then cooled. The process typically concludes with a series of cold rolling passes where each pass reduces the thickness of the plate by a substantially constant thickness amount to form the foil.
Simple process to fabricate nitride alloy powders
International Nuclear Information System (INIS)
Yang, Jae Ho; Kim, Dong-Joo; Kim, Keon Sik; Rhee, Young Woo; Oh, Jang-Soo; Kim, Jong Hun; Koo, Yang Hyun
2013-01-01
Uranium mono-nitride (UN) is considered as a fuel material [1] for accident-tolerant fuel to compensate for the loss of fissile fuel material caused by adopting a thickened cladding such as SiC composites. Uranium nitride powders can be fabricated by a carbothermic reduction of the oxide powders, or the nitriding of metal uranium. Among them, a direct nitriding process of metal is more attractive because it has advantages in the mass production of high-purity powders and the reusing of expensive 15 N 2 gas. However, since metal uranium is usually fabricated in the form of bulk ingots, it has a drawback in the fabrication of fine powders. The Korea Atomic Energy Research Institute (KAERI) has a centrifugal atomisation technique to fabricate uranium and uranium alloy powders. In this study, a simple reaction method was tested to fabricate nitride fuel powders directly from uranium metal alloy powders. Spherical powder and flake of uranium metal alloys were fabricated using a centrifugal atomisation method. The nitride powders were obtained by thermal treating the metal particles under nitrogen containing gas. The phase and morphology evolutions of powders were investigated during the nitriding process. A phase analysis of nitride powders was also part of the present work. KAERI has developed the centrifugal rotating disk atomisation process to fabricate spherical uranium metal alloy powders which are used as advanced fuel materials for research reactors. The rotating disk atomisation system involves the tasks of melting, atomising, and collecting. A nozzle in the bottom of melting crucible introduces melt at the center of a spinning disk. The centrifugal force carries the melt to the edge of the disk and throws the melt off the edge. Size and shape of droplets can be controlled by changing the nozzle size, the disk diameter and disk speed independently or simultaneously. By adjusting the processing parameters of the centrifugal atomiser, a spherical and flake shape
Ault, Timothy M.
The environment, health, and safety properties of thorium-uranium-based (''thorium'') fuel cycles are estimated and compared to those of analogous uranium-plutonium-based (''uranium'') fuel cycle options. A structured assessment methodology for assessing and comparing fuel cycle is refined and applied to several reference fuel cycle options. Resource recovery as a measure of environmental sustainability for thorium is explored in depth in terms of resource availability, chemical processing requirements, and radiological impacts. A review of available experience and recent practices indicates that near-term thorium recovery will occur as a by-product of mining for other commodities, particularly titanium. The characterization of actively-mined global titanium, uranium, rare earth element, and iron deposits reveals that by-product thorium recovery would be sufficient to satisfy even the most intensive nuclear demand for thorium at least six times over. Chemical flowsheet analysis indicates that the consumption of strong acids and bases associated with thorium resource recovery is 3-4 times larger than for uranium recovery, with the comparison of other chemical types being less distinct. Radiologically, thorium recovery imparts about one order of magnitude larger of a collective occupational dose than uranium recovery. Moving to the entire fuel cycle, four fuel cycle options are compared: a limited-recycle (''modified-open'') uranium fuel cycle, a modified-open thorium fuel cycle, a full-recycle (''closed'') uranium fuel cycle, and a closed thorium fuel cycle. A combination of existing data and calculations using SCALE are used to develop material balances for the four fuel cycle options. The fuel cycle options are compared on the bases of resource sustainability, waste management (both low- and high-level waste, including used nuclear fuel), and occupational radiological impacts. At steady-state, occupational doses somewhat favor the closed thorium option while low
Special equipment for the fabrication and quality control of nuclear fuel elements
International Nuclear Information System (INIS)
Guse, K.; Herbert, W.; Jaeger, K.
1989-01-01
For the fabrication of LWR fuel elements, columns are packed of up to 4 m in length, consisting of fuel pellets with different uranium enrichment, their weight and total length to be measured prior to further processing to fuel rods. An automated column packing device has been developed for this purpose. The packing jobs and other tasks are computer-controlled, measured data are stored and are printed out for quality documentation. The forces in the springs of fuel spacers of LWR fuel elements are to be measured and compared with the standard requirements, deviations to be documented. For this task, another computer-controlled, automated device has been developed for measuring the spring forces at all required positions after positioning and fixation of the spacers. (orig./DG) [de
Interpretation of bioassay data from nuclear fuel fabrication workers
International Nuclear Information System (INIS)
Melo, D.; Xavier, M.
2005-01-01
Full text: In nuclear fuel fabrication facilities, workers are exposed to different compounds of enriched uranium. Although in this kind of facility the main route of intake is inhalation, ingestion may occur in some situations. The interpretation of the bioassay data is very complex, since it is necessary taking into account all the different parameters, which is a big challenge. Due to the high cost of the individual monitoring programme for internal dose assessment in the routine monitoring programmes, usually only one type of measurement is assigned. In complex situations like the one described in this paper, where several parameters can compromise the accuracy of the bioassay interpretation it is need to have a combination of techniques to evaluate the internal dose. According to ICRP 78 (1997), the general order of preference in terms of accuracy of interpretation is: body activity measurement, excreta analysis and personal air sampling. Results of monitoring of working environment may provide information that assists in interpretation on particle size, chemical form and solubility, time of intake. A group of seventeen workers from controlled area of the fuel fabrication facility was selected to evaluate the internal dose using all different available techniques during a certain period. The workers were monitored for determination of uranium content in the daily urinary and faecal excretion (collected over a period of 3 consecutive days), chest counting and personal air sampling. The results have shown that at least two types of sensitivity techniques must be used, since there are some sources of uncertainties on the bioassay interpretation, like mixture of uranium compounds intake and different routes of intake. The combination of urine and faeces analysis has shown to be the more appropriate methodology for assessing internal dose in this situation. (author)
Mechanism of 232U production in MTR fuel evolution of activity in reprocessed uranium
International Nuclear Information System (INIS)
Harbonnier, G.; Lelievre, B.; Fanjas, Y.; Naccache, S.J.P.
1993-01-01
The use of reprocessed uranium for research reactor fuel fabrication implies to keep operators safe from the hard gamma rays emitted by 232 U daughter products. CERCA has carried out, with the help of French CEA and COGEMA, a detailed study to determine the evolution of the radiation dose rate associated with the use of this material. (author)
PWR core design, neutronics evaluation and fuel cycle analysis for thorium-uranium breeding recycle
International Nuclear Information System (INIS)
Bi, G.; Liu, C.; Si, S.
2012-01-01
This paper was focused on core design, neutronics evaluation and fuel cycle analysis for Thorium-Uranium Breeding Recycle in current PWRs, without any major change to the fuel lattice and the core internals, but substituting the UOX pellet with Thorium-based pellet. The fuel cycle analysis indicates that Thorium-Uranium Breeding Recycle is technically feasible in current PWRs. A 4-loop, 193-assembly PWR core utilizing 17 x 17 fuel assemblies (FAs) was taken as the model core. Two mixed cores were investigated respectively loaded with mixed reactor grade Plutonium-Thorium (PuThOX) FAs and mixed reactor grade 233 U-Thorium (U 3 ThOX) FAs on the basis of reference full Uranium oxide (UOX) equilibrium-cycle core. The UOX/PuThOX mixed core consists of 121 UOX FAs and 72 PuThOX FAs. The reactor grade 233 U extracted from burnt PuThOX fuel was used to fabrication of U 3 ThOX for starting Thorium-. Uranium breeding recycle. In UOX/U 3 ThOX mixed core, the well designed U 3 ThOX FAs with 1.94 w/o fissile uranium (mainly 233 U) were located on the periphery of core as a blanket region. U 3 ThOX FAs remained in-core for 6 cycles with the discharged burnup achieving 28 GWD/tHM. Compared with initially loading, the fissile material inventory in U 3 ThOX fuel has increased by 7% via 1-year cooling after discharge. 157 UOX fuel assemblies were located in the inner of UOX/U 3 ThOX mixed core refueling with 64 FAs at each cycle. The designed UOX/PuThOX and UOX/U 3 ThOX mixed core satisfied related nuclear design criteria. The full core performance analyses have shown that mixed core with PuThOX loading has similar impacts as MOX on several neutronic characteristic parameters, such as reduced differential boron worth, higher critical boron concentration, more negative moderator temperature coefficient, reduced control rod worth, reduced shutdown margin, etc.; while mixed core with U 3 ThOX loading on the periphery of core has no visible impacts on neutronic characteristics compared
Dissolution of mixed oxide fuel as a function of fabrication variables
International Nuclear Information System (INIS)
Lerch, R.E.
1979-08-01
Dissolution properties of mechanically blended mixed oxide fuel were very dependent on the six fuel fabrication variables studied. Fuel sintering temperature, source of PuO 2 and PuO 2 content of the fuel had major effects: (1) as the sintering temperature was increased from 1400 to 1700 0 C, pellet dissolution was more complete; (2) pellets made from burned metal derived PuO 2 were more completely dissolved than pellets made from calcined nitrate derived PuO 2 which in turn were more completely dissolved than pellets made from calcined nitrate derived PuO 2 ; (3) as the PuO 2 content decreased from 25 to 15 wt % PuO 2 , pellet dissolution was more complete. Preferential dissolution of uranium occurred in all the mechanically blended mixed oxide. Unirradiated mixed oxide fuel pellets made by the Sol Gel process were generally quite soluble in nitric acid. Unirradiated mixed oxide fuel pellets made by the coprecipitation process dissolved completely and rapidly in nitric acid. Fuel made by the coprecipitation process was more completely dissolved than fuel made by the Sol Gel process which, in turn, was more completely dissolved than fuel made by mechanically blending UO 2 and PuO 2 as shown below. Addition of uncomplexed fluoride to nitric acid during fuel dissolution generally rendered all fuel samples completely dissolvable. In boiling 12M nitric acid, 95 to 99% of the plutonium which was going to dissolve did so in the first hour. Irradiated mechanically blended mixed oxide fuel with known fuel fabrication conditions was also subjected to fuel dissolution tests. While irradiation was shown to increase completeness of plutonium dissolution, poor dissolubility due to adverse fabrication conditions (e.g., low sintering temperature) remained after irradiation
Advanced high throughput MOX fuel fabrication technology and sustainable development
International Nuclear Information System (INIS)
Krellmann, Juergen
2005-01-01
The MELOX plant in the south of France together with the La Hague reprocessing plant, are part of the two industrial facilities in charge of closing the nuclear fuel cycle in France. Started up in 1995, MELOX has since accumulated a solid know-how in recycling plutonium recovered from spent uranium fuel into MOX: a fuel blend comprised of both uranium and plutonium oxides. Converting recovered Pu into a proliferation-resistant material that can readily be used to power a civil nuclear reactor, MOX fabrication offers a sustainable solution to safely take advantage of the plutonium's high energy content. Being the first large-capacity industrial facility dedicated to MOX fuel fabrication, MELOX distinguishes itself from the first generation MOX plants with high capacity (around 200 tHM versus around 40 tHM) and several unique operational features designed to improve productivity, reliability and flexibility while maintaining high safety standards. Providing an exemplary reference for high throughput MOX fabrication with 1,000 tHM produced since start-up, the unique process and technologies implemented at MELOX are currently inspiring other MOX plant construction projects (in Japan with the J-MOX plant, in the US and in Russia as part of the weapon-grade plutonium inventory reduction). Spurred by the growing international demand, MELOX has embarked upon an ambitious production development and diversification plan. Starting from an annual level of 100 tons of heavy metal (tHM), MELOX demonstrated production capacity is continuously increasing: MELOX is now aiming for a minimum of 140 tHM by the end of 2005, with the ultimate ambition of reaching the full capacity of the plant (around 200 tHM) in the near future. With regards to its activity, MELOX also remains deeply committed to sustainable development in a consolidated involvement within AREVA group. The French minister of Industry, on August 26th 2005, acknowledged the benefits of MOX fuel production at MELOX: 'In
Nuclear fuel, with emphasis on its utilization in pressurized water reactor
International Nuclear Information System (INIS)
Khazaneh, R.; Roshanzamir, M.
1997-01-01
Production processes of nuclear fuel on one hand and using nuclear fuels in reactors, particularly PWR Type reactors on the other hand is investigated. The first chapter reviews the relationship between fuel and reactors; The principals of reactor physics in relation with fuel are described shortly. The second chapter reviews uranium exploration and extraction as well as production of uranium concentrate and uranium dioxides. The third chapter is specified to the different procedures of uranium enrichment. In the fourth chapter, processing of uranium dioxide powder and fuel pellet is described. In the fifth chapter fabrication of fuel rod and fuel assemblies is explained thoroughly. The sixth chapter devoted to the different phenomena which occur ed in fuel structure and can during operational time of reactor; damage to fuel rods and developing theoretical models to describe these phenomena and analysis of fuel structure. The seventh chapter discusses how fuel rods are to be experimented during fabrication, operation and development of technology. The eighth chapter explains different fuels such as uranium compounds and mixed oxide fuel of uranium Gadolinium and uranium plutonium and the process of fabrication of zircaloy. In the tenth chapter, fuel reprocessing is investigated and the difficulties of developing this technology is referred
Energy Technology Data Exchange (ETDEWEB)
Kessler, G.
2017-07-01
A brief outline of the historical development of the proliferation problem is followed by a description of the uranium-plutonium nuclear fuel cycle with uranium enrichment, fuel fabrication, the light-water reactors mainly in operation, and the breeder reactors still under development. The next item discussed is reprocessing of spent fuel with plutonium recycling and the future possibility to incinerate plutonium and the minor actinides: neptunium, americium, and curium. Much attention is devoted to the technical and scientific treatment of the IAEA surveillance concept of the uranium-plutonium fuel cycle. In this context, especially the physically possible accuracy of measuring U/Pu flow in the fuel cycle, and the criticism expressed of the accuracy in measuring the plutonium balance in large reprocessing plants of non-nuclear weapon states are analyzed. The second part of the book initially examines the assertion that reactor-grade plutonium could be used to build nuclear weapons whose explosive yield cannot be predicted accurately, but whose minimum explosive yield is still far above that of chemical explosive charges. Methods employed in reactor physics are used to show that such hypothetical nuclear explosive devices (HNEDs) would attain too high temperatures in the required implosion lenses as a result of the heat generated by the Pu-238 isotope always present in reactor plutonium of current light-water reactors. These lenses would either melt or tend to undergo chemical auto-explosion. Limits to the content of the Pu-238 isotope are determined above which such hypothetical nuclear weapons are not feasible on technical grounds. This situation is analyzed for various possibilities of the technical state of the art of making implosion lenses and various ways of cooling up to the use of liquid helium. The outcome is that, depending on the existing state of the art, reactor-grade plutonium from spent fuel elements of light-water reactors with a burnup of 35 to 58
Fuels for Canadian research reactors
International Nuclear Information System (INIS)
Feraday, M.A.
1993-01-01
This paper includes some statements and remarks concerning the uranium silicide fuels for which there is significant fabrication in AECL, irradiation and defect performance experience; description of two Canadian high flux research reactors which use high enrichment uranium (HEU) and the fuels currently used in these reactors; limited fabrication work done on Al-U alloys to uranium contents as high as 40 wt%. The latter concerns work aimed at AECL fast neutron program. This experience in general terms is applied to the NRX and NRU designs of fuel
Individual monitoring of internal exposure to uranium oxides in two fuel fabrication plants
International Nuclear Information System (INIS)
Bourdeix, F.; Achiary, J.; Berard, P.
1989-01-01
Individual monitoring of personal exposure to inhalation of uranium oxides throughout the manufacture of fuel for pressurized water reactor (PWR) includes lung gamma-spectrometry, fecal analysis and urine analysis. Examination of the results shows the following: internal exposure is the consequence of repeated intake incidents as revealed by early peaks of urinary and particularly fecal elimination; a shift is often observed with the results of aerosol concentration measured through air collectors; the measured variations of uranium lung incorporations are relatively fast (apparent mean period 165 d). Correct evaluation of the effective dose equivalent from inhalation requires further information concerning the aerosol size distribution at work stations, the physico-chemical characteristics of the product leading to an estimate of its actual biological solubility, and the measurement of the fraction of aerosol liable to intake with an individual portable collector [fr
Overview of chemical characterization of FBTR fuel
International Nuclear Information System (INIS)
Venkatesan, V.; Nandi, C.; Patil, A.B.; Prakash, Amrit; Khan, K.B.; Arun Kumar
2015-01-01
Uranium Plutonium mixed carbide fuel is the driver fuel for Fast Breeder Test Reactor (FBTR) at IGCAR. The fuel is being fabricated at Radiometallurgy Division, BARC by conventional powder metallurgy route. During the fabrication of fuel, chemical quality control of process intermediates is very important to reach stringent specification of the final fuel product. Different steps are involved in the fabrication of uranium-plutonium carbide (MC) for FBTR. The main steps in the fabrication of MC fuel pellets are carbothermic reduction (CR) of mixture of uranium oxide, plutonium oxide and graphite powder to prepare MC clinkers, crushing and milling of MC clinkers and consolidation of MC powders into fuel pellets and sintering. As a part of process control, analysis of uranium (U), plutonium (Pu), carbon in oxide graphite mixture and U, Pu, carbon, oxygen, nitrogen, MC, M 2 C 3 contents in mixed carbide powder (MC clinkers) are carried out at our laboratory. Analysis of U, Pu, carbon, oxygen, nitrogen, MC and M 2 C 3 contents in mixed carbide sintered pellets are carried out as a part of quality control. This paper describes an overview of analytical instruments used during chemical quality control of mixed carbide fuel
Fabrication of PWR fuel assembly and CANDU fuel bundle
International Nuclear Information System (INIS)
Lee, G.S.; Suh, K.S.; Chang, H.I.; Chung, S.H.
1980-01-01
For the project of localization of nuclear fuel fabrication, the R and D to establish the fabrication technology of CANDU fuel bundle as well as PWR fuel assembly was carried out. The suitable boss height and the prober Beryllium coating thickness to get good brazing condition of appendage were studied in the fabrication process of CANDU fuel rod. Basic Studies on CANLUB coating method also were performed. Problems in each fabrication process step and process flow between steps were reviewed and modified. The welding conditions for top and bottom nozzles, guide tube, seal and thimble screw pin were established in the fabrication processes of PWR fuel assembly. Additionally, some researches for a part of PWR grid brazing problems are also carried out
New Concept of Designing Composite Fuel for Fast Reactors with Closing Fuel Cycle
International Nuclear Information System (INIS)
Savchenko, A.; Vatulin, A.; Uferov, O.; Kulakov, G.; Sorokin, V.
2013-01-01
For fast reactors a novel type of promising composite U-PuO2 fuel is proposed which is based on dispersion fuel elements. 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. Novel fuel features higher characteristics in comparison to metallic or MOX fuel its fabrication technology is readily accomplished and is environmentally clean. A possibility is demonstrated of fabricating coated steel claddings to protect from interaction with fuel and fission products when use standard rod type MOX or metallic U-Pu-Zr fuel. Novel approach to reprocessing of composite fuel is demonstrated, which allows to separate uranium from burnt plutonium as well as the newly generated fissile plutonium from burnt one without chemical processes, which simplifies the closing of the nuclear fuel cycle. Novel composite fuel combines the advantages of metallic and ceramic types of fuel and has high uranium density that allows also to implicate it in BREST types reactor with conversion ratio more than 1. Peculiarities of closing nuclear cycle with composite fuel are demonstrated that allows more effective re-usage of generated Pu as well as, minimizing r/a wastes by incineration of MA in specially developed IMF design
Recycling of reprocessed uranium
International Nuclear Information System (INIS)
Randl, R.P.
1987-01-01
Since nuclear power was first exploited in the Federal Republic of Germany, the philosophy underlying the strategy of the nuclear fuel cycle has been to make optimum use of the resource potential of recovered uranium and plutonium within a closed fuel cycle. Apart from the weighty argument of reprocessing being an important step in the treatment and disposal of radioactive wastes, permitting their optimum ecological conditioning after the reprocessing step and subsequent storage underground, another argument that, no doubt, carried weight was the possibility of reducing the demand of power plants for natural uranium. In recent years, strategies of recycling have emerged for reprocessed uranium. If that energy potential, too, is to be exploited by thermal recycling, it is appropriate to choose a slightly different method of recycling from the one for plutonium. While the first generation of reprocessed uranium fuel recycled in the reactor cuts down natural uranium requirement by some 15%, the recycling of a second generation of reprocessed, once more enriched uranium fuel helps only to save a further three per cent of natural uranium. Uranium of the second generation already carries uranium-232 isotope, causing production disturbances, and uranium-236 isotope, causing disturbances of the neutron balance in the reactor, in such amounts as to make further fabrication of uranium fuel elements inexpedient, even after mixing with natural uranium feed. (orig./UA) [de
Overview of MOX fuel fabrication achievements
International Nuclear Information System (INIS)
Bairiot, H.; Vliet, J. van; Chiarelli, G.; Edwards, J.; Nagai, S.H.; Reshetnikov, F.
2000-01-01
Such overview having been adequately covered in an OECD/NEA publication providing the situation as of end 1994, this paper is mainly devoted to an update as of end 1998. The Belgian plant, Belgonucleaire/Dessel, is now dedicated exclusively to the fabrication of MOX fuel and has operated consistently around its nameplate capacity (35tHM/a) through the 1990s involving a large variety of PWR and BWR fuels. The two French plants have also achieved routine operation during the 1990s. CFCa, historically the largest FBR MOX fuel manufacturer, is utilizing the genuine COCA process for that type of fuel and the MIMAS process for LWR fuel: a nominal capacity (40 tHM/a) has been gradually approached. MELOX has operated at 100 tHM/a, as defined in the operating licence granted originally. The British plant, MDF/Sellafield with 8tHM/a nameplate capacity is devoted to fuel and has manufactured several small fabrication campaigns. In Japan, JNC operates three facilities located at Tokai: PFDF, devoted to basic research and fabrication of test fuels, PFFF/ATR line, for the fabrication of Fugen fuel and of corresponding fuel for the critical facility DCA, and PFPF for the fabrication of FBR fuel. In Russia, fabrication techniques have been developed to fuel four BN-800 FBRs contemplated to be constructed and be fuelled with the civilian Pu stockpile. Two demonstration facilities Paket (Mayak) and RIAR (Dimitrovgrad) fabricated respectively pellet and vipac type FBR MOX fuel for BR-5, BOR-60, BN-350 and BN-600. The paper includes a brief description of each of the fabrication routes mentioned, as well as the production of respectively LWR and FBR MOX fuel in each fabrication facility, since the start-up of the plant, since 1 January 1993 and since 1 January 1998 up to 31 December 1998. (author)
Energy Technology Data Exchange (ETDEWEB)
Souza, J.A.B.; Durazzo, M., E-mail: jasouza@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)
2010-07-01
IPEN developed and made available for routine production the technology for manufacturing dispersion type fuel elements for use in research reactors. However, the fuel produced at IPEN is limited to the uranium concentration of 3.0 gU/cm{sup 3} by using the U{sub 3}Si{sub 2}-Al dispersion. Increasing the uranium concentration of the fuel is interesting by the possibility of increasing the reactor core reactivity and lifetime of the fuel. It is possible to increase the concentration of uranium in the fuel up to the technological limit of 4.8 gU/cm{sup 3} for the U{sub 3}Si{sub 2}-Al dispersion, which is well placed around the world. This new fuel will be applicable in the new Brazilian-Multipurpose Reactor RMB. This study aimed to develop the manufacturing process of high uranium concentration fuel, redefining the procedures currently used in the manufacture of IPEN. This paper describes the main procedures adjustments that will be necessary. (author)
FY16 Status Report for the Uranium-Molybdenum Fuel Concept
International Nuclear Information System (INIS)
Bennett, Wendy D.; Doherty, Ann L.; Henager, Charles H.; Lavender, Curt A.; Montgomery, Robert O.; Omberg, Ronald P.; Smith, Mark T.; Webster, Ryan A.
2016-01-01
The Fuel Cycle Research and Development program of the Office of Nuclear Energy has implemented a program to develop a Uranium-Molybdenum metal fuel for light water reactors. Uranium-Molybdenum fuel has the potential to provide superior performance based on its thermo-physical properties. With sufficient development, it may be able to provide the Light Water Reactor industry with a melt-resistant, accident-tolerant fuel with improved safety response. The Pacific Northwest National Laboratory has been tasked with extrusion development and performing ex-reactor corrosion testing to characterize the performance of Uranium-Molybdenum fuel in both these areas. This report documents the results of the fiscal year 2016 effort to develop the Uranium-Molybdenum metal fuel concept for light water reactors.
FY16 Status Report for the Uranium-Molybdenum Fuel Concept
Energy Technology Data Exchange (ETDEWEB)
Bennett, Wendy D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Doherty, Ann L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Henager, Charles H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lavender, Curt A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Montgomery, Robert O. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Omberg, Ronald P. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Smith, Mark T. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Webster, Ryan A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)
2016-09-22
The Fuel Cycle Research and Development program of the Office of Nuclear Energy has implemented a program to develop a Uranium-Molybdenum metal fuel for light water reactors. Uranium-Molybdenum fuel has the potential to provide superior performance based on its thermo-physical properties. With sufficient development, it may be able to provide the Light Water Reactor industry with a melt-resistant, accident-tolerant fuel with improved safety response. The Pacific Northwest National Laboratory has been tasked with extrusion development and performing ex-reactor corrosion testing to characterize the performance of Uranium-Molybdenum fuel in both these areas. This report documents the results of the fiscal year 2016 effort to develop the Uranium-Molybdenum metal fuel concept for light water reactors.
Progress in irradiation performance of experimental uranium - Molybdenum dispersion fuel
International Nuclear Information System (INIS)
Hofman, Gerard L.; Meyer, Mitchell K.
2002-01-01
High-density dispersion fuel experiment, RERTR-4, was removed from the Advanced Test Reactor (ATR) after reaching a peak U-235 burnup of ∼80% and is presently undergoing postirradiation examination at the ANL alpha-gamma hot cells. This test consists of 32 mini fuel plates of which 27 were fabricated with nominally 6 and 8 g cm -3 atomized and machined uranium alloy powders containing 7 wt% and 10 wt% molybdenum. In addition, two miniplates containing solid U-10 wt% Mo foils and three containing 6 g cm -3 U 3 Si 2 are part of the test. The results of the postirradiation examination and analysis of RERTR-4 in conjunction with data from previous tests performed to lower burnup will be presented. (author)
Yalina booster subcritical assembly performance with low enriched uranium fuel
International Nuclear Information System (INIS)
Talamo, Alberto; Gohar, Yousry
2011-01-01
The YALINA Booster facility is a subcritical assembly located in Minsk, Belarus. The facility has special features that result in fast and thermal neutron spectra in different zones. The fast zone of the assembly uses a lead matrix and uranium fuels with different enrichments: 90% and 36%, 36%, or 21%. The thermal zone of the assembly contains 10% enriched uranium fuel in a polyethylene matrix. This study discusses the performance of the three YALINA Booster configurations with the different fuel enrichments. In order to maintain the same subcriticality level in the three configurations, the number of fuel rods in the thermal zone is increased as the uranium fuel enrichment in the fast zone is decreased. The maximum number of fuel rods that can be loaded in the thermal zone is about 1185. Consequently, the neutron multiplication of the configuration with 21% enriched uranium fuel in the fast zone is enhanced by changing the position of the boron carbide and the natural uranium absorber rods, located between the fast and the thermal zones, to form an annular rather than a square arrangement. (author)
Yalina booster subcritical assembly performance with low enriched uranium fuel
Energy Technology Data Exchange (ETDEWEB)
Talamo, Alberto; Gohar, Yousry, E-mail: alby@anl.gov [Argonne National Laboratory, Lemont, IL (United States)
2011-07-01
The YALINA Booster facility is a subcritical assembly located in Minsk, Belarus. The facility has special features that result in fast and thermal neutron spectra in different zones. The fast zone of the assembly uses a lead matrix and uranium fuels with different enrichments: 90% and 36%, 36%, or 21%. The thermal zone of the assembly contains 10% enriched uranium fuel in a polyethylene matrix. This study discusses the performance of the three YALINA Booster configurations with the different fuel enrichments. In order to maintain the same subcriticality level in the three configurations, the number of fuel rods in the thermal zone is increased as the uranium fuel enrichment in the fast zone is decreased. The maximum number of fuel rods that can be loaded in the thermal zone is about 1185. Consequently, the neutron multiplication of the configuration with 21% enriched uranium fuel in the fast zone is enhanced by changing the position of the boron carbide and the natural uranium absorber rods, located between the fast and the thermal zones, to form an annular rather than a square arrangement. (author)
Uranium recovering from slags generated in the metallic uranium by magnesiothermic reduction
International Nuclear Information System (INIS)
Fornarolo, F.; Carvalho, E.F. Urano de; Durazzo, M.; Riella, H.G.
2008-01-01
The Nuclear Fuel Center of IPEN/CNEN-SP has recent/y concluded a program for developing the fabrication technology of the nuclear fuel based on the U 3 Si 2 -Al dispersion, which is being used in the IEA-R1 research reactor. The uranium silicide (U 3 Si 2 ) fuel production starts with the uranium hexafluoride (UF 6 ) processing and uranium tetrafluoride (UF 4 ) precipitation. Then, the UF 4 is converted to metallic uranium by magnesiothermic reduction. The UF 4 reduction by magnesium generates MgF 2 slag containing considerable concentrations of uranium, which could reach 20 wt%. The uranium contained in that slag should be recovered and this work presents the results obtained in recovering the uranium from that slag. The uranium recovery is accomplished by acidic leaching of the calcined slag. The calcination transforms the metallic uranium in U 3 O 8 , promoting the pulverization of the pieces of metallic uranium and facilitating the leaching operation. As process variables, have been considered the nitric molar concentration, the acid excess regarding the stoichiometry and the leaching temperature. As result, the uranium recovery reached a 96% yield. (author)
International Nuclear Information System (INIS)
Primm, Trent; Guida, Tracey
2010-01-01
Under the auspices of the Global Threat Reduction Initiative Reduced Enrichment for Research and Test Reactors Program, the National Nuclear Security Administration/Department of Energy (NNSA/DOE) has, as a goal, to convert research reactors worldwide from weapons grade to non-weapons grade uranium. The High Flux Isotope Reactor (HFIR) at Oak Ridge National Lab (ORNL) is one of the candidates for conversion of fuel from high enriched uranium (HEU) to low enriched uranium (LEU). A well documented business model, including tasks, costs, and schedules was developed to plan the conversion of HFIR. Using Microsoft Project, a detailed outline of the conversion program was established and consists of LEU fuel design activities, a fresh fuel shipping cask, improvements to the HFIR reactor building, and spent fuel operations. Current-value costs total $76 million dollars, include over 100 subtasks, and will take over 10 years to complete. The model and schedule follows the path of the fuel from receipt from fuel fabricator to delivery to spent fuel storage and illustrates the duration, start, and completion dates of each subtask to be completed. Assumptions that form the basis of the cost estimate have significant impact on cost and schedule.
Waste arisings from a high-temperature reactor with a uranium-thorium fuel cycle
International Nuclear Information System (INIS)
1979-09-01
This paper presents an equilibrium-recycle condition flow sheet for a high-temperature gas-cooled reactor (HTR) fuel cycle which uses thorium and high-enriched uranium (93% U-235) as makeup fuel. INFCE Working Group 7 defined percentage losses to various waste streams are used to adjust the heavy-element mass flows per gigawatt-year of electricity generated. Thorium and bred U-233 are recycled following Thorex reprocessing. Fissile U-235 is recycled one time following Purex reprocessing and then is discarded to waste. Plutonium and other transuranics are discarded to waste. Included are estimates of volume, radioactivity, and heavy-element content of wastes arising from HTR fuel element fabrication; HTR operation, maintenance, and decommissioning; and reprocessing spent fuel where the waste is unique to the HTR fuel cycle
Swiss R and D on uranium-free LWR fuels for plutonium incineration
International Nuclear Information System (INIS)
Stanculescu, A.; Chawla, R.; Degueldre, C.; Kasemeyer, U.; Ledergerber, G.; Paratte, J.M.
1999-01-01
The most efficient way to enhance the plutonium consumption in LWRs is to eliminate plutonium production altogether. This requirement leads to fuel concepts in which the uranium is replaced by an inert matrix. The inert matrix material studied at PSI is zirconium oxide. For reactivity control reasons, adding a burnable poison to this fuel proves to be necessary. The studies performed at PSI have identified erbium oxide as the most suitable candidate for this purpose. With regard to material technology aspects, efforts have concentrated on the evaluation of fabrication feasibility and on the determination of the physicochemical properties of the chosen single phase zirconium/ erbium/plutonium oxide material stabilised as a cubic solution by yttrium. The results to-date, obtained for inert matrix samples containing thorium or cerium as plutonium substitute, confirm the robustness and stability of this material. With regard to reactor physics aspects, our studies indicate the feasibility of uranium-free, plutonium-fuelled cores having operational characteristics quite similar to those of conventional UO 2 -fuelled ones, and much higher plutonium consumption rates, as compared to 100% MOX loadings. The safety features of such cores, based on results obtained from static neutronics calculations, show no cliff edges. However, the need for further detailed transient analyses is clearly recognised. Summarising, PSI's studies indicate the feasibility of a uranium-free plutonium fuel to be considered in 'maximum plutonium consumption LWRs' operating in a 'once-through' mode. With regard to reactor physics, future efforts will concentrate on strengthening the safety case of uranium-free cores, as well as on improving the integral data base for validation of the neutronics calculations. Material technology studies will be continued to investigate the physico-chemical properties of the inert matrix fuel containing plutonium and will focus on the planning and evaluation of
International Nuclear Information System (INIS)
Hrovat, M.F.; Hassel, H.-W.
1983-01-01
According to the present state of development full size test fuel elements with UAl x , U 3 O 8 , and U 3 Si 2 fuel were fabricated at Nukem in production scale. The maximum uranium densities amount to 1.8 g/cc for UAI x , 2.9 g/cc for U 3 O 8 , and 4.76 g/cc for U 3 Si 2 . The irradiation performance of these fuel elements is good: Up to the end of September 1982 the following burnups were achieved: 73% with UA1 x , 60% with U 3 O 8 , 39% with U 3 Si 2 ; no defects could be detected. For an economical fuel element production with reduced 235-U enrichment chemical uranium recycling methods were developed allowing immediate scrap recovery at minimum waste generation. In addition test plates with UAl x and U 3 O 8 fuel were successfully irradiated in the ORR up to a burnup of 75 %. The relatively high uranium meat densities of these test plates amount to 2.2 g/cc for UAI x , and 3.14 g/cc for U 3 O 8 fuel. Apart from plates with standard geometry also plates with increased meat thickness were inserted. (author)
Fabrication of simulated plate fuel elements: Defining role of out-of-plane residual shear stress
Energy Technology Data Exchange (ETDEWEB)
Rakesh, R., E-mail: rakesh.rad87@gmail.com [DAE Graduate Fellows, IIT Bombay, Powai, Mumbai 400076 (India); Metallic Fuels Division, BARC, Trombay, Mumbai 400085 (India); Kohli, D. [DAE Graduate Fellows, IIT Bombay, Powai, Mumbai 400076 (India); Metallic Fuels Division, BARC, Trombay, Mumbai 400085 (India); Sinha, V.P.; Prasad, G.J. [Metallic Fuels Division, BARC, Trombay, Mumbai 400085 (India); Samajdar, I. [Department of Metallurgical Engineering and Materials Science, IIT Bombay, Powai, Mumbai 400076 (India)
2014-02-01
Bond strength and microstructural developments were investigated during fabrication of simulated plate fuel elements. The study involved roll bonding of aluminum–aluminum (case A) and aluminum–aluminum + yttria (Y{sub 2}O{sub 3}) dispersion (case B). Case B approximated aluminum–uranium silicide (U{sub 3}Si{sub 2}) ‘fuel-meat’ in an actual plate fuel. Samples after different stages of fabrication, hot and cold rolling, were investigated through peel and pull tests, micro-hardness, residual stresses, electron and micro-focus X-ray diffraction. Measurements revealed a clear drop in bond strength during cold rolling: an observation unique to case B. This was related to significant increase in ‘out-of-plane’ residual shear stresses near the clad/dispersion interface, and not from visible signatures of microstructural heterogeneities.
Research Establishment progress report 1978 - uranium fuel cycle
International Nuclear Information System (INIS)
1978-12-01
A report of research programs continuing in the following areas is presented: mining and treatment of uranium ores, uranium enrichment, waste treatment, reprocessing and the uranium fuel cycle. Staff responsible for each project are indicated
Energy Technology Data Exchange (ETDEWEB)
Molecke, M.A.; Gregson, M.W.; Sorenson, K.B. [Sandia National Labs. (United States); Billone, M.C.; Tsai, H. [Argonne National Lab. (United States); Koch, W.; Nolte, O. [Fraunhofer Inst. fuer Toxikologie und Experimentelle Medizin (Germany); Pretzsch, G.; Lange, F. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (Germany); Autrusson, B.; Loiseau, O. [Inst. de Radioprotection et de Surete Nucleaire (France); Thompson, N.S.; Hibbs, R.S. [U.S. Dept. of Energy (United States); Young, F.I.; Mo, T. [U.S. Nuclear Regulatory Commission (United States)
2004-07-01
We provide a detailed overview of an ongoing, multinational test program that is developing aerosol data for some spent fuel sabotage scenarios on spent fuel transport and storage casks. Experiments are being performed to quantify the aerosolized materials plus volatilized fission products generated from actual spent fuel and surrogate material test rods, due to impact by a high energy density device, HEDD. The program participants in the U.S. plus Germany, France, and the U.K., part of the international Working Group for Sabotage Concerns of Transport and Storage Casks, WGSTSC have strongly supported and coordinated this research program. Sandia National Laboratories, SNL, has the lead role for conducting this research program; test program support is provided by both the U.S. Department of Energy and Nuclear Regulatory Commission. WGSTSC partners need this research to better understand potential radiological impacts from sabotage of nuclear material shipments and storage casks, and to support subsequent risk assessments, modeling, and preventative measures. We provide a summary of the overall, multi-phase test design and a description of all explosive containment and aerosol collection test components used. We focus on the recently initiated tests on ''surrogate'' spent fuel, unirradiated depleted uranium oxide, and forthcoming actual spent fuel tests. The depleted uranium oxide test rodlets were prepared by the Institut de Radioprotection et de Surete Nucleaire, in France. These surrogate test rodlets closely match the diameter of the test rodlets of actual spent fuel from the H.B. Robinson reactor (high burnup PWR fuel) and the Surry reactor (lower, medium burnup PWR fuel), generated from U.S. reactors. The characterization of the spent fuels and fabrication into short, pressurized rodlets has been performed by Argonne National Laboratory, for testing at SNL. The ratio of the aerosol and respirable particles released from HEDD-impacted spent
The industrial production of fuel elements; La fabrication en france des elements combustibles
Energy Technology Data Exchange (ETDEWEB)
Boussard, R [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires; Nadal, J [Societe Industrielle de Combustible Nucleaire (SICN), 75 - Paris (France); Pellen, A [Compagnie pour l' Etude et la Realisation de Combustibles Atomiques (CERCA), 75 - Paris (France)
1964-07-01
The authors deal successively with the industrial production of fuel elements for power reactors of the natural uranium-graphite-gas type, and more particularly for the EDF power stations, and with the industrial production of fuel elements containing enriched uranium designed for swimming-pool type reactors. 1. part: advanced fuel elements for the EDF reactors. After recalling the characteristics of the fuel elements now being produced industrially for the Marcoule and Chinon reactors, the authors give the various steps leading to the industrial production of a new type of fuel element both as concerns the can, and in certain cases the graphite sleeve, and the fuel itself. As for as the production of the fuel is concerned, they describe the various operations, stressing the original aspects of the production and of the equipment such as: - casting in hot moulds, - thermal treatments, of Uranium containing 1% in weight molybdenum, - welding of the pellets for closing the tubes of uranium, - canning, - controls in the various steps. As far as can production is concerned they show why the extruded can was replaced by a machined can and give a few characteristics of the equipment used as well as the controls effected. They give also some details concerning the production and machining of the sleeves. After recalling the state of the nuclear fuel industry in France in mid-1964 the authors stress the economic aspects of the production of fuel elements. They show the relative importance of capital costs on the cost price of the fuel itself and examine the various items involved. They analyse the cost price of a completed fuel element using present date knowledge. In conclusion the authors show the particular points which should be the subject of future efforts in order to decrease the cost of a production which is perhaps delicate but now will define, and review the development of this new industrial branch. 2. part: industrial production of fuel elements for swimming
Feasibility study on AFR-100 fuel conversion from uranium-based fuel to thorium-based fuel
Energy Technology Data Exchange (ETDEWEB)
Heidet, F.; Kim, T.; Grandy, C. (Nuclear Engineering Division)
2012-07-30
Although thorium has long been considered as an alternative to uranium-based fuels, most of the reactors built to-date have been fueled with uranium-based fuel with the exception of a few reactors. The decision to use uranium-based fuels was initially made based on the technology maturity compared to thorium-based fuels. As a result of this experience, lot of knowledge and data have been accumulated for uranium-based fuels that made it the predominant nuclear fuel type for extant nuclear power. However, following the recent concerns about the extent and availability of uranium resources, thorium-based fuels have regained significant interest worldwide. Thorium is more abundant than uranium and can be readily exploited in many countries and thus is now seen as a possible alternative. As thorium-based fuel technologies mature, fuel conversion from uranium to thorium is expected to become a major interest in both thermal and fast reactors. In this study the feasibility of fuel conversion in a fast reactor is assessed and several possible approaches are proposed. The analyses are performed using the Advanced Fast Reactor (AFR-100) design, a fast reactor core concept recently developed by ANL. The AFR-100 is a small 100 MW{sub e} reactor developed under the US-DOE program relying on innovative fast reactor technologies and advanced structural and cladding materials. It was designed to be inherently safe and offers sufficient margins with respect to the fuel melting temperature and the fuel-cladding eutectic temperature when using U-10Zr binary metal fuel. Thorium-based metal fuel was preferred to other thorium fuel forms because of its higher heavy metal density and it does not need to be alloyed with zirconium to reduce its radiation swelling. The various approaches explored cover the use of pure thorium fuel as well as the use of thorium mixed with transuranics (TRU). Sensitivity studies were performed for the different scenarios envisioned in order to determine the
Energy Technology Data Exchange (ETDEWEB)
Troyanov, V.; Molchanov, V.; Tuzov, A. [TVEL Corporation, 49 Kashirskoe shosse, Moscow 115409 (Russian Federation); Semchenkov, Yu.; Lizorkin, M. [RRC ' Kurchatov Institute' (Russian Federation); Vasilchenko, I.; Lushin, V. [OKB ' Gidropress' (Russian Federation)
2009-06-15
Current nuclear fuel cycle of Russian nuclear power involves reprocessed low-enriched uranium in nuclear fuel production for some NPP units with VVER-type LWR. This paper discusses design and performance characteristics of commercial nuclear fuel based on natural and reprocessed uranium. It presents the review of results of commercial operation of nuclear fuel based on reprocessed uranium on Russian NPPs-unit No.2 of Kola NPP and unit No.2 of Kalinin NPP. The results of calculation and experimental validation of safe fuel operation including necessary isotope composition conformed to regulation requirements and results of pilot fuel operation are also considered. Meeting the customer requirements the possibility of high burn-up achieving was demonstrated. In addition the paper compares the characteristics of nuclear fuel cycles with maximum length based on reprocessed and natural uranium considering relevant 5% enrichment limitation and necessity of {sup 236}U compensation. The expedience of uranium-235 enrichment increasing over 5% is discussed with the aim to implement longer fuel cycles. (authors)
Research on calculation of mixing fraction for natural uranium equivalent fuel
International Nuclear Information System (INIS)
Huang Shien; Wang Lianjie; Wei Yanqin; Li Qing; Zheng Jiye
2013-01-01
Based on the first-order perturbation theory and reasonable approximations, the calculation method of recycled uranium (RU) and depleted uranium (DU) mixing fraction for natural uranium equivalent (NUE) fuel was studied, so the equivalence between NUE fuel and natural uranium (NU) fuel was assured. The adopted calculation method accurately takes the variation of micro cross sections alone with fuel depletion into account. A computer code named ALPHA was programmed to execute the calculation procedure. Then the ALPHA code and the WIMS-AECL code compose a processing system, which is applicable to the mixing fraction calculation for heavy water reactor NUE fuel. The validation shows that the processing system can accurately calculate the mixing fraction for NUE fuel. (authors)
Security of supply of uranium as nuclear fuel
International Nuclear Information System (INIS)
Guzman Gomez-Selles, L.
2011-01-01
When we talk about Sustainability related to nuclear fuel, the first concern that comes to our mind is about the possibility of having guarantees on the uranium supply for a sufficient period of time. In this paper we are going to analyze the last Reserves data published by the OCD's Red Book and also how the Reserve concept in fully linked to the uranium price. Additionally, it is demonstrated how the uranium Security of supply is guaranteed for, at least, the next 100 years. finally, some comments are made regarding other sources of nuclear fuel as it is the uranium coming from the phosphates or the thorium. (Author)
Property-process relationships in nuclear fuel fabrication
International Nuclear Information System (INIS)
Tikare, V.
2015-01-01
Nuclear fuels are fabricated using many different techniques as they come in a large variety of shapes and compositions. The design and composition of nuclear fuels are predominantly dictated by the engineering requirements necessary for their function in reactors of various designs. Other engineering properties requirements originate from safety and security concerns, and the easy of handling, storing, transporting and disposing of the radioactive materials. In this chapter, the more common of these fuels will be briefly reviewed and the methods used to fabricate them will be presented. The fuels considered in this paper are oxide fuels used in LWRs and FRs, metal fuels in FRs and particulate fuels used in HTGRs. Fabrication of alternative fuel forms and use of standard fuels in alternative reactors will be discussed briefly. The primary motivation to advance fuel fabrication is to improve performance, reduce cost, reduce waste or enhance safety and security of the fuels. To achieve optimal performance, developing models to advance fuel fabrication has to be done in concert with developing fuel performance models. The specific properties and microstructures necessary for improved fuel performance must be identified using fuel performance models, while fuel fabrication models that can determine processing variables to give the desired microstructure and materials properties must be developed. (author)
International Nuclear Information System (INIS)
Biermann, W.K.; Heuvel, H.J.; Pilate, S.; Vanderborck, Y.; Pelckmans, E.; Vanhellemont, G.; Roepenack, H.; Stoll, W.
1987-01-01
The core and fuel were designed for the SNR-300 first core by Interatom GmbH and Belgonucleaire. The fuel was fabricated by Alkem/RBU and Belgonucleaire. Based on the preparation of drawings and specifications and on the results of the prerun fabrication, an extensive interplay took place between design requirements, specifications, and fabrication processes at both fuel plants. During start-up of pellet and pin fabrication, this solved such technical questions as /sup 239/Pu equivalent linear weight, pellet density, stoichiometry of the pellets, and impurity content. Close cooperation of designers and manufacturers has allowed manufacture of 205 fuel assemblies without major problems
Nuclear fuel cycle head-end enriched uranium purification and conversion into metal
International Nuclear Information System (INIS)
Bonini, A.; Cabrejas, J.; Lio, L. de; Dell'Occhio, L.; Devida, C.; Dupetit, G.; Falcon, M.; Gauna, A.; Gil, D.; Guzman, G.; Neuringer, P.; Pascale, A.; Stankevicius, A.
1998-01-01
The CNEA (Comision Nacional de Energia Atomica - Argentina) operated two facilities at the Ezeiza Atomic Center which supply purified enriched uranium employed in the production of nuclear fuels. At one of those facilities, the Triple Height Laboratory scraps from the production of MTR type fuel elements (mainly out of specification U 3 O 8 plates or powder) are purified to nuclear grade. The purification is accomplished by a solvent extraction process. The other facility, the Enriched Uranium Laboratory produces 90% enriched uranium metal to be used in Mo 99 production (originally the uranium was used for the manufacture of MTR fuel elements made of aluminium-uranium alloy). This laboratory also provided metallic uranium with a lower enrichment (20%) for a first uranium-silicon testing fuel element, and in the near future it is going to recommence 20% enriched uranium related activities in order to provide the metal for the silicon-based fuel elements production (according to the policy of enrichment reduction for MTR reactors). (author)
Development of nuclear fuel cycle technologies
International Nuclear Information System (INIS)
Suzuoki, Akira; Matsumoto, Takashi; Suzuki, Kazumichi; Kawamura, Fumio
1995-01-01
In the long term plan for atomic energy that the Atomic Energy Commission decided the other day, the necessity of the technical development for establishing full scale fuel cycle for future was emphasized. Hitachi Ltd. has engaged in technical development and facility construction in the fields of uranium enrichment, MOX fuel fabrication, spent fuel reprocessing and so on. In uranium enrichment, it took part in the development of centrifuge process centering around Power Reactor and Nuclear Fuel Development Corporation (PNC), and took its share in the construction of the Rokkasho uranium enrichment plant of Japan Nuclear Fuel Service Co., Ltd. Also it cooperates with Laser Enrichment Technology Research Association. In Mox fuel fabrication, it took part in the construction of the facilities for Monju plutonium fuel production of PNC, for pellet production, fabrication and assembling processes. In spent fuel reprocessing, it cooperated with the technical development of maintenance and repair of Tokai reprocessing plant of PNC, and the construction of spent fuel stores in Rokkasho reprocessing plant is advanced. The centrifuge process and the atomic laser process of uranium enrichment are explained. The high reliability of spent fuel reprocessing plants and the advancement of spent fuel reprocessing process are reported. Hitachi Ltd. Intends to exert efforts for the technical development to establish nuclear fuel cycle which increases the importance hereafter. (K.I.)
Radioactive decay properties of CANDU fuel. Volume 1: the natural uranium fuel cycle
International Nuclear Information System (INIS)
Clegg, L.J.; Coady, J.R.
1977-01-01
The computer code CANIGEN was used to obtain the mass, activity, decay heat and toxicity of CANDU fuel and its component isotopes. Data are also presented on gamma spectra and neutron emissions. Part 1 presents these data for unirradiated fuel, uranium ore and uranium mill tailings. In Part 2 they have been computed for fuel irradiated to levels of burnup ranging from 140 GJ/kg U to 1150 GJ/kg U. (author)
Metallic uranium as fuel for fast reactors
International Nuclear Information System (INIS)
Moura Neto, C. de
1988-01-01
This paper presents a first overview of the use of metallic uranium and its alloys as an option for fuel for rapid reactors. Aspects are discussed concerning uranium alloys which present high solubility in the gamma phase. (author)
Research on using depleted uranium as nuclear fuel for HWR
International Nuclear Information System (INIS)
Zhang Jiahua; Chen Zhicheng; Bao Borong
1999-01-01
The purpose of our work is to find a way for application of depleted uranium in CANDU reactor by using MOX nuclear fuel of depleted U and Pu instead of natural uranium. From preliminary evaluation and calculation, it was shown that MOX nuclear fuel consisting of depleted uranium enrichment tailings (0.25% 235 U) and plutonium (their ratio 99.5%:0.5%) could replace natural uranium in CANDU reactor to sustain chain reaction. The prospects of application of depleted uranium in nuclear energy field are also discussed
The relationship between natural uranium and advanced fuel cycles in CANDU reactors
International Nuclear Information System (INIS)
Lane, A.D.; McDonnell, F.N.; Griffiths, J.
1988-11-01
CANDU is the most uranium-economic type of thermal power reactor, and is the only type used in Canada. CANDU reactors consume approximately 15% of Canadian uranium production and support a fuel service industry valued at ∼$250 M/a. In addition to their once-through, natural-uranium fuel cycle, CANDU reactors are capable of operating with slightly-enriched uranium (SEU), uranium-plutonium and thorium cycles, more efficiently than other reactors. Only SEU is economically attractive in Canada now, but the other cycles are of interest to countries without indigenous fuel resources. A program is underway to establish the fuel technologies necessary for the use of SEU and the other fuel cycles in CANDU reactors. 22 refs
Energy Technology Data Exchange (ETDEWEB)
Galahom, Ahmed Abdelghafar [Higher Technological Institute, Ramadan (Egypt)
2016-06-15
The present work discusses two different models of boiling water reactor (BWR) bundle to compare the neutronic characteristics of uranium dioxide (UO{sub 2}) and uranium zirconium hydride (UZrH{sub 1.6}) fuel. Each bundle consists of four assemblies. The BWR assembly fueled with UO{sub 2} contains 8 × 8 fuel rods while that fueled with UZrH{sub 1.6} contains 9 × 9 fuel rods. The Monte Carlo N-Particle Transport code, based on the Mont Carlo method, is used to design three dimensional models for BWR fuel bundles at typical operating temperatures and pressure conditions. These models are used to determine the multiplication factor, pin-by-pin power distribution, axial power distribution, thermal neutron flux distribution, and axial thermal neutron flux. The moderator and coolant (water) are permitted to boil within the BWR core forming steam bubbles, so it is important to calculate the reactivity effect of voiding at different values. It is found that the hydride fuel bundle design can be simplified by eliminating water rods and replacing the control blade with control rods. UZrH{sub 1.6} fuel improves the performance of the BWR in different ways such as increasing the energy extracted per fuel assembly, reducing the uranium ore, and reducing the plutonium accumulated in the BWR through burnup.
International Nuclear Information System (INIS)
Suripto, A; Yuwono
1998-01-01
Heat treatment experiments upon U 3 Si 2 - U ss ingot have been cam e d out to obtain free uranium particle size improvement which is required to enhance the U-Al inter-diffusion reaction in the fuel plate meat. . Heat treatment experiments upon fuel plates containing dispersion of U 3 Si 2 - U ss in Al matrix have also been carried out to study the effect of temperature and treatment duration on the extent of inter-diffusion reaction between free uranium particle and aluminium matrix in the fuel plate meat. Both the experiments indicate that a drastic size improvement has occurred with the U 3 Si 2 as well as free uranium particles upon heat treatment at controlled temperature between the U 3 Si 2 peritectic and peritectoid temperatures and that the inter-diffusion reaction between free uranium and Al matrix occurs quite significantly at temperatures higher than that ordinarily used in the fabrication procedure
High-uranium-loaded U3O8--Al fuel element development program
International Nuclear Information System (INIS)
Martin, M.M.
1978-01-01
The High-Uranium-Loaded U 3 O 8 --Al Fuel Development Program supports Argonne National Laboratory efforts to develop high-uranium-density research and test reactor fuel to accommodate use of low-uranium enrichment. The goal is to fuel most research and test reactors with uranium of less than 20% enrichment for the purpose of lowering the potential for diversion of highly-enriched material for nonpeaceful usages
Licensed fuel facility status report
International Nuclear Information System (INIS)
1990-04-01
NRC is committed to the periodic publication of licensed fuel facilities inventory difference data, following agency review of the information and completion of any related NRC investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, plutonium, or uranium-233
Licensed fuel facility status report
International Nuclear Information System (INIS)
Joy, D.; Brown, C.
1993-04-01
NRC is committed to the periodic publication of licensed fuel facilities inventory difference data, following agency review of the information and completion of any related NRC investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, plutonium, or uranium-233
Depleted uranium management alternatives
Energy Technology Data Exchange (ETDEWEB)
Hertzler, T.J.; Nishimoto, D.D.
1994-08-01
This report evaluates two management alternatives for Department of Energy depleted uranium: continued storage as uranium hexafluoride, and conversion to uranium metal and fabrication to shielding for spent nuclear fuel containers. The results will be used to compare the costs with other alternatives, such as disposal. Cost estimates for the continued storage alternative are based on a life-cycle of 27 years through the year 2020. Cost estimates for the recycle alternative are based on existing conversion process costs and Capital costs for fabricating the containers. Additionally, the recycle alternative accounts for costs associated with intermediate product resale and secondary waste disposal for materials generated during the conversion process.
Depleted uranium management alternatives
International Nuclear Information System (INIS)
Hertzler, T.J.; Nishimoto, D.D.
1994-08-01
This report evaluates two management alternatives for Department of Energy depleted uranium: continued storage as uranium hexafluoride, and conversion to uranium metal and fabrication to shielding for spent nuclear fuel containers. The results will be used to compare the costs with other alternatives, such as disposal. Cost estimates for the continued storage alternative are based on a life-cycle of 27 years through the year 2020. Cost estimates for the recycle alternative are based on existing conversion process costs and Capital costs for fabricating the containers. Additionally, the recycle alternative accounts for costs associated with intermediate product resale and secondary waste disposal for materials generated during the conversion process
Uranium for Nuclear Power: Resources, Mining and Transformation to Fuel
International Nuclear Information System (INIS)
Hore-Lacy, Ian
2016-01-01
Uranium for Nuclear Power: Resources, Mining and Transformation to Fuel discusses the nuclear industry and its dependence on a steady supply of competitively priced uranium as a key factor in its long-term sustainability. A better understanding of uranium ore geology and advances in exploration and mining methods will facilitate the discovery and exploitation of new uranium deposits. The practice of efficient, safe, environmentally-benign exploration, mining and milling technologies, and effective site decommissioning and remediation are also fundamental to the public image of nuclear power. This book provides a comprehensive review of developments in these areas: • Provides researchers in academia and industry with an authoritative overview of the front end of the nuclear fuel cycle • Presents a comprehensive and systematic coverage of geology, mining, and conversion to fuel, alternative fuel sources, and the environmental and social aspects • Written by leading experts in the field of nuclear power, uranium mining, milling, and geological exploration who highlight the best practices needed to ensure environmental safety
Fabrication of uranium alloy fuel slug for sodium-cooled fast reactor by injection casting
International Nuclear Information System (INIS)
Jong Hwan Kim; Hoon Song; Ki Hwan Kim; Chan Bock Lee
2014-01-01
Metal fuel slugs of U-Zr alloys for a sodium-cooled fast reactor (SFR) have been fabricated using an injection casting method. However, casting alloys containing volatile radioactive constituents such as Am can cause problems in a conventional injection casting method. Therefore, in this study, several injection-casting methods were applied to evaluate the volatility of the metal-fuel elements and control the transport of volatile elements. Mn was selected as a volatile surrogate alloy since it possesses a total vapor pressure equivalent to that of minor actinide-bearing fuels for SFRs. U-10 wt% Zr and U-10 wt% Zr-5 wt% Mn metal fuels were prepared, and the casting processes were evaluated. The casting soundness of the fuel slugs was characterized by gamma-ray radiography and immersion density measurements. Inductively coupled plasma atomic emission spectroscopy was used to determine the chemical composition of fuel slugs. Fuel losses after casting were also evaluated according to the casting conditions. (author)
The low enriched uranium fuel cycle in Ontario
International Nuclear Information System (INIS)
Archinoff, G.H.
1979-02-01
Six fuel-cycle strategies for use in CANDU reactors are examined in terms of their uranium-conserving properties and their ease of commercialization for three assumed growth rates of installed nuclear capacity in Ontario. The fuel cycle strategies considered assume the continued use of the natural uranium cycle up to the mid-1990's. At that time, the low-enriched uranium (LEU) cycle is gradually introduced into the existing power generation grid. In the mid-2020's one of four advanced cycles is introduced. The advanced cycles considered are: mixed oxide, intermediate burn-up thorium (Pu topping), intermediate burn-up thorium (U topping), and LMFBR. For comparison purposes an all natural uranium strategy and a natural uranium-LEU strategy (with no advanced cycle) are also included. None of the strategies emerges as a clear, overall best choice. (LL)
Irradiation behavior of miniature experimental uranium silicide fuel plates
International Nuclear Information System (INIS)
Hofman, G.L.; Neimark, L.A.; Mattas, R.F.
1983-01-01
Uranium silicides, because of their relatively high uranium density, were selected as candidate dispersion fuels for the higher fuel densities required in the Reduced Enrichment Research and Test Reactor (RERTR) Program. Irradiation experience with this type of fuel, however, was limited to relatively modest fission densities in the bulk from, on the order of 7 x 10 20 cm -3 , far short of the approximately 20 x 10 20 cm -3 goal established for the RERTR program. The purpose of the irradiation experiments on silicide fuels on the ORR, therefore, was to investigate the intrinsic irradiation behavior of uranium silicide as a dispersion fuel. Of particular interest was the interaction between the silicide particles and the aluminum matrix, the swelling behavior of the silicide particles, and the maximum volume fraction of silicide particles that could be contained in the aluminum matrix
Energy Technology Data Exchange (ETDEWEB)
Nakano, H [Power Reactor and Nuclear Fuel Development Corp., Tokyo (Japan)
1976-10-01
It is expected that nuclear power generation will reach 49 million kW in 1985 and 129 million kW in 1995, and the nuclear fuel having to be supplied and processed will increase in proportion to these values. The technical problems concerning nuclear fuel are presented on the basis of the balance between the benefit for human beings and the burden on the human beings. Recently, especially the downstream of nuclear fuel attracts public attention. Enriched uranium as the raw material for light water reactor fuel is almost monopolized by the U.S., and the technical information has not been published for fear of the diversion to nuclear weapons. In this paper, the present situations of uranium enrichment, fuel fabrication, transportation, reprocessing and waste disposal and the future problems are described according to the path of nuclear fuel cycle. The demand and supply of enriched uranium in Japan will be balanced up to about 1988, but afterwards, the supply must rely upon the early establishment of the domestic technology by centrifugal separation method. No problem remains in the fabrication of light water reactor fuel, but for the fabrication of mixed oxide fuel, the mechanization of the production facility and labor saving are necessary. The solution of the capital risk for the construction of the second reprocessing plant is the main problem. Japan must develop waste disposal techniques with all-out efforts.
Process for the fabrication of nuclear fuel oxide pellets
International Nuclear Information System (INIS)
Francois, Bernard; Paradis, Yves.
1977-01-01
Process for the fabrication of nuclear fuel oxide pellets of the type for which particles charged with an organic binder -selected from the group that includes polyvinyl alcohol, carboxymethyl cellulose, polyvinyl compounds and methyl cellulose- are prepared from a powder of such an oxide, for instance uranium dioxide. These particles are then compressed into pellets which are then sintered. Under this process the binder charged particles are prepared by stirring the powder with a gas, spraying on to the stirred powder a solution or a suspension in a liquid of this organic binder in order to obtain these particles and then drying the particles so obtained with this gas [fr
Transition from uranium to denatured uranium/thorium fuel in an existing PWR
International Nuclear Information System (INIS)
Walters, M.A.
1982-01-01
The purpose of this research was to determine whether it is possible to make a gradual transition from uranium to denatured uranium/thorium (DUTH) fuel in an existing PWR by adding DUTH assemblies during each scheduled refueling and, if the transition is possible, to develop a general procedure for making it. The feasibility of the transition was established by identifying acceptable refueling schemes for a series of transition cores, and in the process, a method for identifying acceptable schemes evolved. The utility of the method was then demonstrated by applying it to a standard reactor operating under normal conditions. The vehicle used to examine proposed fuel mixtures and to select acceptable ones was a set of one-dimensional computer codes. The core was modeled as a set of five concentric fuel zones with a reflector. Fuel mixtures were proposed and the computer codes were used to determine whether a mixture was acceptable, i.e., whether it had the desired k-effective and flux and power distributions. The parameters allowed to vary in selection of proposed fuel mixtures were enrichment of fresh fuel assemblies, number of uranium and DUTH assemblies added during each refueling, and distribution of fuel in the core. Results of the research showed that a gradual transition is possible. Furthermore, there is a method that allows the identification of fuel mixtures that are likely to be acceptable. It requires the calculation of K-infinity for the entire proposed core and for some of its regions. These values of K-infinity and relationships developed in this research can be used to predict the flux distribution and the final k-effective for the proposed fuel mixture
Role of thermal analysis in uranium oxide fuel fabrication process
International Nuclear Information System (INIS)
Balaji Rao, Y.; Yadav, R.B.
2006-01-01
The present paper discusses the application of thermal analysis, particularly, differential thermal analysis (Dta) at various stages of fuel fabrication process. The useful role of Dta in knowing the decomposition pattern and calcination temperature of Adu along with de-nitration temperature is explained. The decomposition pattern depends upon the type of drying process adopted for wet ADU cake (ADU C). Also, the paper highlights the utility of DTA in determining the APS and SSA of UO 2+x and U 3 O 8 powders as an alternate technique. Further, the temperature difference (ΔT max ) between the two exothermic peaks obtained in UO 2+x powder oxidation is related to sintered density of UO 2 pellets. (author)
The nuclear fuel cycle; Le cycle du combustible nucleaire
Energy Technology Data Exchange (ETDEWEB)
NONE
1998-05-01
After a short introduction about nuclear power in the world, fission physics and the French nuclear power plants, this brochure describes in a digest way the different steps of the nuclear fuel cycle: uranium prospecting, mining activity, processing of uranium ores and production of uranium concentrates (yellow cake), uranium chemistry (conversion of the yellow cake into uranium hexafluoride), fabrication of nuclear fuels, use of fuels, reprocessing of spent fuels (uranium, plutonium and fission products), recycling of energetic materials, and storage of radioactive wastes. (J.S.)
Fabrication of preliminary fuel rods for SFR
International Nuclear Information System (INIS)
Kim, Sun Ki; Oh, Seok Jin; Ko, Young Mo; Woo, Youn Myung; Kim, Ki Hwan
2012-01-01
Metal fuels was selected for fueling many of the first reactors in the US, including the Experimental Breeder Reactor-I (EBR-I) and the Experimental Breeder Reactor-II (EBR-II) in Idaho, the FERMI-I reactor, and the Dounreay Fast Reactor (DFR) in the UK. Metallic U.Pu.Zr alloys were the reference fuel for the US Integral Fast Reactor (IFR) program. Metallic fuel has advantages such as simple fabrication procedures, good neutron economy, high thermal conductivity, excellent compatibility with a Na coolant and inherent passive safety. U-Zr-Pu alloy fuels have been used for SFR (sodium-cooled fast reactor) related to the closed fuel cycle for managing minor actinides and reducing a high radioactivity levels since the 1980s. Fabrication technology of metallic fuel for SFR has been in development in Korea as a national nuclear R and D program since 2007. For the final goal of SFR fuel rod fabrication with good performance, recently, three preliminary fuel rods were fabricated. In this paper, the preliminary fuel rods were fabricated, and then the inspection for QC(quality control) of the fuel rods was performed
International Nuclear Information System (INIS)
Borgonovi, G.M.; McCartin, T.J.; Miller, C.L.
1978-12-01
The report presents information on foreign nuclear fuel fabrication facilities. Fuel descriptions and fuel fabrication information for three basic reactor types are presented: The information presented for LWRs assumes that Pu--U Mixed Oxide Fuel (MOX) will be used as fuel
Fuel elements for LWR power plants
International Nuclear Information System (INIS)
Roepenack, H.
1977-01-01
About five times more expensive than the fabrication of a fuel element is the enriched uranium contained therein; soon the monthly interest charges for the uranium value of a fuel element reload will account for five percent of the fabrication costs, and much more expensive than all this together can it be if reactor operation has to be interrupted because of damaged elements. Thus, quality assurance comes first. (orig.) [de
Choice and utilization of slightly enriched uranium fuel for high performance research reactors
International Nuclear Information System (INIS)
Cerles, J.M.; Schwartz, J.P.
1978-01-01
Problems relating to the replacement of highly enriched (90% or 93% U 235 ) uranium fuel: by moderately enriched (20% or 40% in U 235 ) metallic uranium fuel and slightly enriched (3% or 8% in U 235 ) uranium oxide fuel are discussed
Powder fabrication of U-Mo alloys for nuclear dispersion fuels
Energy Technology Data Exchange (ETDEWEB)
Durazzo, Michelangelo; Rocha, Claudio Jose da; Mestnik Filho, Jose; Leal Neto, Ricardo Mendes, E-mail: mdurazzo@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)
2011-07-01
For the last 30 years high uranium density dispersion fuels have been developed in order to accomplish the low enrichment goals of the Reduced Enrichment for Research and Test Reactors (RERTR) Program. Gamma U-Mo alloys, particularly with 7 to 10 wt% Mo, as a fuel phase dispersed in aluminum matrix, have shown good results concerning its performance under irradiation tests. That's why this fissile phase is considered to be used in the nuclear fuel of the Brazilian Multipurpose Research Reactor (RMB), currently being designed. Powder production from these ductile alloys has been attained by atomization, mechanical (machining, grinding, cryogenic milling) and chemical (hydriding-de hydriding) methods. This work is a part of the efforts presently under way at IPEN to investigate the feasibility of these methods. Results on alloy fabrication by induction melting and gamma-stabilization of U-10Mo alloys are presented. Some results on powder production and characterization are also discussed. (author)
Powder fabrication of U-Mo alloys for nuclear dispersion fuels
International Nuclear Information System (INIS)
Durazzo, Michelangelo; Rocha, Claudio Jose da; Mestnik Filho, Jose; Leal Neto, Ricardo Mendes
2011-01-01
For the last 30 years high uranium density dispersion fuels have been developed in order to accomplish the low enrichment goals of the Reduced Enrichment for Research and Test Reactors (RERTR) Program. Gamma U-Mo alloys, particularly with 7 to 10 wt% Mo, as a fuel phase dispersed in aluminum matrix, have shown good results concerning its performance under irradiation tests. That's why this fissile phase is considered to be used in the nuclear fuel of the Brazilian Multipurpose Research Reactor (RMB), currently being designed. Powder production from these ductile alloys has been attained by atomization, mechanical (machining, grinding, cryogenic milling) and chemical (hydriding-de hydriding) methods. This work is a part of the efforts presently under way at IPEN to investigate the feasibility of these methods. Results on alloy fabrication by induction melting and gamma-stabilization of U-10Mo alloys are presented. Some results on powder production and characterization are also discussed. (author)
Linking fuel design features ampersand plant management to uranium, SWU savings
International Nuclear Information System (INIS)
Anon.
1992-01-01
This article, contributed by Scott Garrett, Manager of Planning and Uranium Operations for Siemens Power Corporation in Bellevue, Washington, explores the impact of advances in fuel design and fuel management strategies on uranium utilization in the United States. Nuclear plant operators are deriving substantial benefits from these changes, including longer fuel cycle lengths, increased burnup, and added capacity - and experiencing cost savings in both uranium and enrichment services at the same time
Experiments of JRR-4 low-enriched-uranium-silicied fuel core
International Nuclear Information System (INIS)
Hirane, Nobuhiko; Ishikuro, Yasuhiro; Nagadomi, Hideki; Yokoo, Kenji; Horiguchi, Hironori; Nemoto, Takumi; Yamamoto, Kazuyoshi; Yagi, Masahiro; Arai, Nobuyoshi; Watanabe, Shukichi; Kashima, Yoichi
2006-03-01
JRR-4, a light-water-moderated and cooled, swimming pool type research reactor using high-enriched uranium plate-type fuels had been operated from 1965 to 1996. In order to convert to low-enriched-uranium-silicied fuels, modification work had been carried out for 2 years, from 1996 to 1998. After the modification, start-up experiments were carried out to obtain characteristics of the low-enriched-uranium-silicied fuel core. The measured excess reactivity, reactor shutdown margin and the maximum reactivity addition rate satisfied the nuclear limitation of the safety report for licensing. It was confirmed that conversion to low-enriched-uranium-silicied fuels was carried out properly. Besides, the necessary data for reactor operation were obtained, such as nuclear, thermal hydraulic and reactor control characteristics. This report describes the results of start-up experiments and burnup experiments. The first criticality of low-enriched-uranium-silicied core was achieved on 14th July 1998, and the operation for joint-use has been carried out since 6th October 1998. (author)
International Nuclear Information System (INIS)
1982-05-01
Producing and using uranium as a fuel for the generation of electricity involves many activities and industries. Some 100,000 Canadian jobs are dependent upon the nuclear power industry. Of these, 38,000 people are employed directly by more than 100 corporations. This publication describes the components, relationships and some special aspects of 'the nuclear fuel cycle,' which includes exploration, mining, milling, refining, conversion, enrichment, fuel fabrication, reactors, and waste management
Energy Technology Data Exchange (ETDEWEB)
Kelly, J.E.
1995-02-08
Uranium fuel assemblies and other uranium associated with the shutdown N Reactor are stored in the 300 Area N Reactor Fuel Fabrication and Storage Facility (Facility). The 3712 Building, where the majority of the fuel assemblies and other uranium is stored, is where there could be a potential for a criticality bounding case. The purpose of this study is to evaluate the probability of potential fires in the Facility 3712 Building that could lead to criticality. This study has been done to support the criticality update. For criticality to occur, the wooden fuel assembly containers would have to burn such that the fuel inside would slump into a critical geometry configuration, a sufficient number of containers burn to form an infinite wide configuration, and sufficient water (about a 17 inch depth) be placed onto the slump. To obtain the appropriate geometric configuration, enough fuel assembly containers to form an infinite array on the floor would have to be stacked at least three high. Administrative controls require the stacks to be limited to two high for 1.25 wt% enriched finished fuel. This is not sufficient to allow for a critical mass regardless of the fire and accompanying water moderation. It should be noted that 0.95 wt% enriched fuel and billets are stacked higher than only two high. In this analysis, two initiating events will be considered. The first is a random fire that is hot enough and sufficiently long enough to burn away the containers and fuel separators such that the fuel can establish a critical mass. The second is a seismically induced fire with the same results.
Comparison of the radiological impacts of thorium and uranium nuclear fuel cycles
International Nuclear Information System (INIS)
Meyer, H.R.; Witherspoon, J.P.; McBride, J.P.; Frederick, E.J.
1982-03-01
This report compares the radiological impacts of a fuel cycle in which only uranium is recycled, as presented in the Final Generic Environmental Statement on the Use of Recycle Plutonium in Mixed Oxide Fuel in Light Water Cooled Reactors (GESMO), with those of the light-water breeder reactor (LWBR) thorium/uranium fuel cycle in the Final Environmental Statement, Light Water Breeder Reactor Program. The significant offsite radiological impacts from routine operation of the fuel cycles result from the mining and milling of thorium and uranium ores, reprocessing spent fuel, and reactor operations. The major difference between the impacts from the two fuel cycles is the larger dose commitments associated with current uranium mining and milling operations as compared to thorium mining and milling. Estimated dose commitments from the reprocessing of either fuel type are small and show only moderate variations for specific doses. No significant differences in environmental radiological impact are anticipated for reactors using either of the fuel cycles. Radiological impacts associated with routine releases from the operation of either the thorium or uranium fuel cycles can be held to acceptably low levels by existing regulations
Once-through uranium thorium fuel cycle in CANDU reactors
International Nuclear Information System (INIS)
Ozdemir, S.; Cubukcu, E.
2000-01-01
In this study, the performance of the once-through uranium-thorium fuel cycle in CANDU reactors is investigated. (Th-U)O 2 is used as fuel in all fuel rod clusters where Th and U are mixed homogeneously. CANDU reactors have the advantage of being capable of employing various fuel cycle options because of its good neutron economy, continuous on line refueling ability and axial fuel replacement possibility. For lattice cell calculations transport code WIMS is used. WIMS cross-section library is modified to achieve precise lattice cell calculations. For various enrichments and Th-U mixtures, criticality, heavy element composition changes, diffusion coefficients and cross-sections are calculate. Reactor core is modeled by using the diffusion code CITATION. We conclude that an overall saving of 22% in natural uranium demand can be achieved with the use of Th cycle. However, slightly enriched U cycle still consumes less natural Uranium and is a lot less complicated. (author)
Uranium-thorium fuel cycle in a very high temperature hybrid system
International Nuclear Information System (INIS)
Hernandez, C.R.G.; Oliva, A.M.; Fajardo, L.G.; Garcia, J.A.R.; Curbelo, J.P.; Abadanes, A.
2011-01-01
Thorium is a potentially valuable energy source since it is about three to four times as abundant as Uranium. It is also a widely distributed natural resource readily accessible in many countries. Therefore, Thorium fuels can complement Uranium fuels and ensure long term sustainability of nuclear power. The main advantages of the use of a hybrid system formed by a Pebble Bed critical nuclear reactor and two Pebble Bed Accelerator Driven Systems (ADSs) using a Uranium-Thorium (U + Th) fuel cycle are shown in this paper. Once-through and two step U + Th fuel cycle was evaluated. With this goal, a preliminary conceptual design of a hybrid system formed by a Graphite Moderated Gas-Cooled Very High Temperature Reactor and two ADSs is proposed. The main parameters related to the neutronic behavior of the system in a deep burn scheme are optimized. The parameters that describe the nuclear fuel breeding and Minor Actinide stockpile are compared with those of a simple Uranium fuel cycle. (author)
International Nuclear Information System (INIS)
Nardi, A.J.; Lamantia, L.
2009-01-01
Recent soil excavation activities on a site identified the presence of buried uranium contaminated building construction debris. The site previously was the location of a low enriched uranium fuel fabrication facility. This resulted in the collection of excavated materials from the two locations where contaminated subsurface debris was identified. The excavated material was temporarily stored in two piles on the site until a determination could be made as to the appropriate disposition of the material. Characterization of the excavated material was undertaken in a manner that involved the collection of large scale samples of the excavated material in 1 cubic meter Super Sacks. Twenty bags were filled with excavated material that consisted of the mixture of both the construction debris and the associated soil. In order to obtain information on the level of activity associated with the construction debris, ten additional bags were filled with construction debris that had been separated, to the extent possible, from the associated soil. Radiological surveys were conducted of the resulting bags of collected materials and the soil associated with the waste mixture. The 30 large samples, collected as bags, were counted using an In-Situ Object Counting System (ISOCS) unit to determine the average concentration of U-235 present in each bag. The soil fraction was sampled by the collection of 40 samples of soil for analysis in an on-site laboratory. A fraction of these samples were also sent to an off-site laboratory for additional analysis. This project provided the necessary soil characterization information to allow consideration of alternate options for disposition of the material. The identified contaminant was verified to be low enriched uranium. Concentrations of uranium in the waste were found to be lower than the calculated site specific derived concentration guideline levels (DCGLs) but higher than the NRC's screening values. The methods and results are presented
Airborne effluent control at fuel enrichment, conversion, and fabrication plants
International Nuclear Information System (INIS)
Mitchell, M.E.
1976-01-01
Uranium conversion, enrichment, and fuel fabrication facilities generate gaseous wastes that must be treated prior to being discharged to the atmosphere. Since all three process and/or handle similar compounds, they also encounter similar gaseous waste disposal problems, the majority of which are treated in a similar manner. Ventilation exhausts from personnel areas and equipment off-gases that do not contain corrosive gases (such as HF) are usually passed through roughening and/or HEPA filters prior to release. Ventilation exhausts that contain larger quantities of particles, such as the conversion facilities' U 3 O 8 sampling operation, are passed through bag filters or cyclone separators, while process off-gases containing corrosive materials are normally treated by sintered metal filters or scrubbers. The effectiveness of particle removal varies from about 90 percent for a scrubber alone to more than 99.9 percent for HEPA filters or a combination of the various filters and scrubbers. The removal of nitrogen compounds (N 2 , HNO 3 , NO/sub x/, and NH 3 ) is accomplished by scrubbers in the enrichment and fuel fabrication facilities. The conversion facility utilizes a nitric acid recovery facility for both pollution control and economic recovery of raw materials. Hydrogen removal from gaseous waste streams is generally achieved with burners. Three different systems are currently utilized by the conversion, enrichment, and fuel fabrication plants to remove gaseous fluorides from airborne effluents. The HF-rich streams, such as those emanating from the hydrofluorination and fluorine production operations of the conversion plant, are passed through condensers to recover aqueous hydrofluoric acid
Tuning up and fabrication of U3Si2 nuclear material
International Nuclear Information System (INIS)
Pasqualini, Enrique E.; Echenique, Patricia N.; Rossi, Gustavo S.; Canil, Eduardo E.; Esteban, Adolfo; Lopez, Marisol; Adelfang, Pablo
2000-01-01
This work describes the tuning up and fabrication of uranium-silicide powder for its utilization as nuclear fuel in material testing reactors taking in account NUREG-1313 recommendations, the experience of several suppliers and the one acquired in this work.Several alloy compositions were melted with natural uranium at temperatures of about 1800 degree C for adjusting composition and ingot homogeneity. Alumina, magnesia and zirconia-5% stabilized yttria crucibles were tested to evaluate the degree of contamination introduced by chemical attack of molten uranium and silicon. The fabrication procedure of 20% enriched uranium-silicide powder was established for building up the P-06 fuel element that actually is being irradiated at the RA-3 reactor facility. The selected procedures of fabrication and the critical analysis for the interpretation of several specifications are discussed. Results are shown of the obtained ingots and powder produced with the enriched uranium-silicide. (author)
The role of electricity utilities in ensuring environmental compliance of uranium suppliers
International Nuclear Information System (INIS)
Lindholm, I.
2001-01-01
The Swedish Utilities, Vattenfall Fuel, OKG and BKAB (purchasing uranium and fuel cycle services for all Swedish reactors), have started 'Nuclear Fuel and Environment Project, NFE'. The purpose is to make environmental audits of companies with uranium mines, conversion, isotope enrichment and fuel fabrication facilities. Up to now five environmental audits have been carried out and another three are being processed. (author)
Uranium Oxide Rate Summary for the Spent Nuclear Fuel (SNF) Project (OCRWM)
Energy Technology Data Exchange (ETDEWEB)
PAJUNEN, A.L.
2000-09-20
The purpose of this document is to summarize the uranium oxidation reaction rate information developed by the Hanford Spent Nuclear Fuel (SNF) Project and describe the basis for selecting reaction rate correlations used in system design. The selection basis considers the conditions of practical interest to the fuel removal processes and the reaction rate application during design studies. Since the reaction rate correlations are potentially used over a range of conditions, depending of the type of evaluation being performed, a method for transitioning between oxidation reactions is also documented. The document scope is limited to uranium oxidation reactions of primary interest to the SNF Project processes. The reactions influencing fuel removal processes, and supporting accident analyses, are: uranium-water vapor, uranium-liquid water, uranium-moist air, and uranium-dry air. The correlation selection basis will consider input from all available sources that indicate the oxidation rate of uranium fuel, including the literature data, confirmatory experimental studies, and fuel element observations. Trimble (2000) summarizes literature data and the results of laboratory scale experimental studies. This document combines the information in Trimble (2000) with larger scale reaction observations to describe uranium oxidation rate correlations applicable to conditions of interest to the SNF Project.
Uranium Oxide Rate Summary for the Spent Nuclear Fuel (SNF) Project (OCRWM)
International Nuclear Information System (INIS)
PAJUNEN, A.L.
2000-01-01
The purpose of this document is to summarize the uranium oxidation reaction rate information developed by the Hanford Spent Nuclear Fuel (SNF) Project and describe the basis for selecting reaction rate correlations used in system design. The selection basis considers the conditions of practical interest to the fuel removal processes and the reaction rate application during design studies. Since the reaction rate correlations are potentially used over a range of conditions, depending of the type of evaluation being performed, a method for transitioning between oxidation reactions is also documented. The document scope is limited to uranium oxidation reactions of primary interest to the SNF Project processes. The reactions influencing fuel removal processes, and supporting accident analyses, are: uranium-water vapor, uranium-liquid water, uranium-moist air, and uranium-dry air. The correlation selection basis will consider input from all available sources that indicate the oxidation rate of uranium fuel, including the literature data, confirmatory experimental studies, and fuel element observations. Trimble (2000) summarizes literature data and the results of laboratory scale experimental studies. This document combines the information in Trimble (2000) with larger scale reaction observations to describe uranium oxidation rate correlations applicable to conditions of interest to the SNF Project
Energy Technology Data Exchange (ETDEWEB)
Brim, Cornelia P.
2013-04-01
An important requirement for the international safeguards community is the ability to determine the enrichment level of uranium in gas centrifuge enrichment plants and nuclear fuel fabrication facilities. This is essential to ensure that countries with nuclear nonproliferation commitments, such as States Party to the Nuclear Nonproliferation Treaty, are adhering to their obligations. However, current technologies to verify the uranium enrichment level in gas centrifuge enrichment plants or nuclear fuel fabrication facilities are technically challenging and resource-intensive. NNSA’s Office of Nonproliferation and International Security (NIS) supports the development, testing, and evaluation of future systems that will strengthen and sustain U.S. safeguards and security capabilities—in this case, by automating the monitoring of uranium enrichment in the entire inventory of a fuel fabrication facility. One such system is HEVA—hybrid enrichment verification array. This prototype was developed to provide an automated, nondestructive assay verification technology for uranium hexafluoride (UF6) cylinders at enrichment plants.
International Nuclear Information System (INIS)
Meyer, M.K.; Clark, C.R.; Hayes, S.L.; Strain, R.V.; Hofman, G.L.; Snelgrove, J.L.; Park, J.M.; Kim, K.H.
1999-01-01
This paper presents quantitative data on the irradiation behavior of uranium-molybdenum fuels from the low temperature RERTR-1 and -2 experiments. Fuel swelling measurements of U-Mo fuels at ∼40% and ∼70% burnup are presented. The rate of fuel-matrix interaction layer growth is estimated. Microstructures of fuel in the pre- and postirradiation condition were compared. Based on these data, a qualitative picture of the evolution of the U-Mo fuel microstructure during irradiation has been developed. Estimates of uranium-molybdenum fuel swelling and fuel-matrix interaction under high-power research reactor operating conditions are presented. (author)
International Nuclear Information System (INIS)
Cox, C.M.; Leggett, R.D.; Weber, E.T.
1981-01-01
Uranium plutonium oxide is the principal fuel material for liquid metal fast breeder reactors (LMFBR's) throughout the world. Development of this material has been a reasonably straightforward evolution from the UO 2 used routinely in the light water reactor (LWR's); but, because of the lower neutron capture cross sections and much lower coolant pressures in the sodium cooled LMFBR's, the fuel is operated to much higher discharge exposures than that of a LWR. A typical LMFBR fuel assembly is shown. Depending on the required power output and the configuration of the reactor, some 70 to 400 such fuel assemblies are clustered to form the core. There is a wide variation in cross section and length of the assemblies where the increasing size reflects a chronological increase in plant size and power output as well as considerations of decreasing the net fuel cycle cost. Design and performance characteristics are described
Technical study report on fuel fabrication system
International Nuclear Information System (INIS)
Kono, Shusaku; Tanaka, Kenya; Ono, Kiyoshi; Iwasa, Katsuyoshi; Hoshino, Yasushi; Shinkai, Yasuo
2000-07-01
The feasibility study of FBR and related fuel cycle is performed for developing the FBR recycle system which ensures safety, economic competitiveness, efficient utilization of resources, reduction of environmental burden and enhancement of nuclear non-proliferation under consistency of FBR reactor and fuel cycle systems. In this study, a conceptual design study and system characteristics evaluation are conducted for fuel fabrication systems of pellet process, vibropack process for oxide and nitride fuel and casting process for metal fuel. Technical issues in each process are also extracted. In 1999 fiscal year, a conceptual design study were conducted for the fuel fabrication plants adopting (1) the short pellet process which simplifies the conventional MOX pellet fabrication processes, (2) vibropack processes of aqueous gelation process, improved RIAR process, improved ANL process and fluoride volatility process, (3) casting processes of injection process, centrifuging process. As a result, attainable perspective was obtained for each fuel fabrication system through the evaluation of apparatuses, layout and facility volume, etc. In each fuel fabrication system, technical issues for practical use were made clear. Hereafter, more detailed study will be performed for each system, and research programs for phase II study will be planned. (author)
The quality challenge for fuel fabrication
International Nuclear Information System (INIS)
Lannegrace, J.-P.
1990-01-01
Fuel fabrication is a key segment of the nuclear fuel cycle, since safe and economic operation of reactors is highly dependent on the quality of the fuel. Achieving and controlling quality is, therefore, of paramount importance to fuel fabricators dominating nearly every aspect of the business. The quality policy, concepts and assurance system at three French plants are outlined. The need for integrated inspection, process optimization and good employee motivation is stressed. (author)
Irradiation performance of uranium-molybdenum alloy dispersion fuels
International Nuclear Information System (INIS)
Almeida, Cirila Tacconi de
2005-01-01
The U-Mo-Al dispersion fuels of Material Test Reactors (MTR) are analyzed in terms of their irradiation performance. The irradiation performance aspects are associated to the neutronic and thermal hydraulics aspects to propose a new core configuration to the IEA-R1 reactor of IPEN-CNEN/SP using U-Mo-Al fuels. Core configurations using U-10Mo-Al fuels with uranium densities variable from 3 to 8 gU/cm 3 were analyzed with the computational programs Citation and MTRCR-IEA R1. Core configurations for fuels with uranium densities variable from 3 to 5 gU/cm 3 showed to be adequate to use in IEA-R1 reactor e should present a stable in reactor performance even at high burn-up. (author)
International Nuclear Information System (INIS)
Hofman, G.L.
1996-01-01
A fuel development campaign that results in an aluminum plate-type fuel of unlimited LEU burnup capability with an uranium loading of 9 grams per cm 3 of meat should be considered an unqualified success. The current worldwide approved and accepted highest loading is 4.8 g cm -3 with U 3 Si 2 as fuel. High-density uranium compounds offer no real density advantage over U 3 Si 2 and have less desirable fabrication and performance characteristics as well. Of the higher-density compounds, U 3 Si has approximately a 30% higher uranium density but the density of the U 6 X compounds would yield the factor 1.5 needed to achieve 9 g cm -3 uranium loading. Unfortunately, irradiation tests proved these peritectic compounds have poor swelling behavior. It is for this reason that the authors are turning to uranium alloys. The reason pure uranium was not seriously considered as a dispersion fuel is mainly due to its high rate of growth and swelling at low temperatures. This problem was solved at least for relatively low burnup application in non-dispersion fuel elements with small additions of Si, Fe, and Al. This so called adjusted uranium has nearly the same density as pure α-uranium and it seems prudent to reconsider this alloy as a dispersant. Further modifications of uranium metal to achieve higher burnup swelling stability involve stabilization of the cubic γ phase at low temperatures where normally α phase exists. Several low neutron capture cross section elements such as Zr, Nb, Ti and Mo accomplish this in various degrees. The challenge is to produce a suitable form of fuel powder and develop a plate fabrication procedure, as well as obtain high burnup capability through irradiation testing
Nuclear fuel elements design, fabrication and performance
Frost, Brian R T
1982-01-01
Nuclear Fuel Elements: Design, Fabrication and Performance is concerned with the design, fabrication, and performance of nuclear fuel elements, with emphasis on fast reactor fuel elements. Topics range from fuel types and the irradiation behavior of fuels to cladding and duct materials, fuel element design and modeling, fuel element performance testing and qualification, and the performance of water reactor fuels. Fast reactor fuel elements, research and test reactor fuel elements, and unconventional fuel elements are also covered. This volume consists of 12 chapters and begins with an overvie
Informal presentations by fuel fabricators and others [contributed by W. Ross, U.S. NRC
International Nuclear Information System (INIS)
Ross, W.
1993-01-01
Al-28 wt.% U alloy and its ductility certainly will not increase with irradiation. The effect of this brittleness on behavior of the long NRX/NRU rods during thermal cycling in reactor would have to be investigated. Therefore from a fabrication point of view, it may be possible to make the rods from Al-40 wt.% U but it is obvious that an extensive fabrication, safety, and irradiation study would be required before a definitive answer could be given. Development of Al-50 wt.% U alloy for such rods would be even more difficult, the probability of success smaller, and the development program somewhat larger. The development of a completely new design driver fuel for the reactors using Zircaloy clad powder packed UO 2 or dispersion type fuels would require an even larger and more expensive program. At 20% enrichment level, the current Al-U designs of driver fuel could not be used even by increasing the number of fuel rods and/or by removing experimental facilities. A new fuel design would have to be tested to burnups of 60-70% for NRX and over 80% for NRU before it could be considered to be acceptable. The defect performance of the higher uranium alloys would also have to be checked out. In summary, the NRX and NRU are high performance, high flux research reactors having a very heavy experimental and radioisotope load. With the present designs of driver fuel and reactor loadings the use of 50% enriched uranium would be possible only if the brittle Al-40 to 50 wt.% U alloys can be successfully developed for high burnups. Extensive fabrication, safety, and physics work and irradiation of intact and defective elements would have to be done for each of the fuel designs considered before a definitive answer could be given. If these alloys could not be successfully developed, new designs of driver fuel would be required. The fuel developments would be in three phases: fuel alloy development, pin development, and irradiation testing and would probably take 3 to years. A change
Quantitative determination of uranium distribution homogeneity in MTR fuel type plates
International Nuclear Information System (INIS)
Ferrufino, Felipe Bonito Jaldin
2011-01-01
IPEN/CNEN-SP produces the fuel to supply its nuclear research reactor IEA-R1. The fuel is assembled with fuel plates containing an U 3 Si 2 -Al composite meat. A good homogeneity in the uranium distribution inside the fuel plate meat is important from the standpoint of irradiation performance. Considering the lower power of reactor IEA-R1, the uranium distribution in the fuel plate has been evaluated only by visual inspection of radiographs. However, with the possibility of IPEN to manufacture the fuel for the new Brazilian Multipurpose Reactor (RMB), with higher power, it urges to develop a methodology to determine quantitatively the uranium distribution into the fuel. This paper presents a methodology based on X-ray attenuation, in order to quantify the uranium concentration distribution in the meat of the fuel plate by using optical densities in radiographs and comparison with standards. The results demonstrated the inapplicability of the method, considering the current specification for the fuel plates due to the high intrinsic error to the method. However, the study of the errors involved in the methodology, seeking to increase their accuracy and precision, can enable the application of the method to qualify the final product. (author)
Status of fuel element technology for plate type dispersion fuels with high uranium density
International Nuclear Information System (INIS)
Hrovat, M.; Huschka, H.; Koch, K.H.; Nazare, S.; Ondracek, G.
1983-01-01
A number of about 20 Material Test and Research Reactors in Germany and abroad is supplied with fuel elements by the company NUKEM. The power of these reactors differs widely ranging from up to about 100 MW. Consequently, the uranium density of the fuel elements in the meat varies considerably depending on the reactor type and is usually within the range from 0.4 to 1.3 g U/cm 3 if HEU is used. In order to convert these reactors to lower uranium enrichment (19.75% 235-U) extensive work is carried out at NUKEM since about two years with the goal to develop fuel elements with high U-density. This work is sponsored by the German Ministry for Research and Technology in the frame of the AF-program. This paper reports on the present state of development for fuel elements with high U-density fuels at NUKEM is reported. The development works were so far concentrated on UAl x , U 3 O 8 and UO 2 fuels which will be described in more detail. In addition fuel plates with new fuels like e.g. U-Si or U-Fe compounds are developed in collaboration with KfK. The required uranium densities for some typical reactors with low, medium, and high power are listed allowing a comparison of HEU and LEU uranium density requirements. The 235-U-content in the case of LEU is raised by 18%. Two different meat thicknesses are considered: Standard thickness of 0.5 mm; and increased thickness of 0.76 mm. From this data compilation the objective follows: in the case of conversion to LEU (19.75% 235-U-enrichment), uranium densities have to be made available up to 24 gU/cm 3 meat for low power level reactors, up to 33 gU/cm 3 meat for medium power level reactors, and between 5.75 and 7.03 g/cm 3 meat for high power level reactors according to this consideration
Alternative Fabrication of Recycling Fast Reactor Metal Fuel
International Nuclear Information System (INIS)
Kim, Ki-Hwan; Kim, Jong Hwan; Song, Hoon; Kim, Hyung-Tae; Lee, Chan-Bock
2015-01-01
Metal fuels such as U-Zr/U-Pu-Zr alloys have been considered as a nuclear fuel for a sodium-cooled fast reactor (SFR) related to the closed fuel cycle for managing minor actinides and reducing a high radioactivity levels since the 1980s. In order to develop innovative fabrication method of metal fuel for preventing the evaporation of volatile elements such as Am, modified casting under inert atmosphere has been applied for metal fuel slugs for SFR. Alternative fabrication method of fuel slugs has been introduced to develop an improved fabrication process of metal fuel for preventing the evaporation of volatile elements. In this study, metal fuel slugs for SFR have been fabricated by modified casting method, and characterized to evaluate the feasibility of the alternative fabrication method. In order to prevent evaporation of volatile elements such as Am and improve quality of fuel slugs, alternative fabrication methods of metal fuel slugs have been studied in KAERI. U-10Zr-5Mn fuel slug containing volatile surrogate element Mn was soundly cast by modified injection casting under modest pressure. Evaporation of Mn during alternative casting could not be detected by chemical analysis. Mn element was most recovered with prevention of evaporation by alternative casting. Modified injection casting has been selected as an alternative fabrication method in KAERI, considering evaporation prevention, and proven benefits of high productivity, high yield, and good remote control
Bharadwaj, S R
2013-01-01
This book presents the state of the art on thermophysical and thermochemical properties, fabrication methodologies, irradiation behaviours, fuel reprocessing procedures, and aspects of waste management for oxide fuels in general and for thoria-based fuels in particular. The book covers all the essential features involved in the development of and working with nuclear technology. With the help of key databases, many of which were created by the authors, information is presented in the form of tables, figures, schematic diagrams and flow sheets, and photographs. This information will be useful for scientists and engineers working in the nuclear field, particularly for design and simulation, and for establishing the technology. One special feature is the inclusion of the latest information on thoria-based fuels, especially on the use of thorium in power generation, as it has less proliferation potential for nuclear weapons. Given its natural abundance, thorium offers a future alternative to uranium fuels in nuc...
Uranium isotopic signatures measured in samples of dirt collected at two former uranium facilities
International Nuclear Information System (INIS)
Meyers, L.A.; Stalcup, A.M.; LaMont, S.P.; Spitz, H.B.
2014-01-01
Nuclear forensics is a multidisciplinary science that uses a variety of analytical methods and tools to explore the physical, chemical, and isotopic characteristics of nuclear and radiological materials. These characteristics, when evaluated alone or in combination, become signatures that may reveal how and when the material was fabricated. The signatures contained in samples of dirt collected at two different uranium metal processing facilities in the United States were evaluated to determine uranium isotopic composition and compare results with processes that were conducted at these sites. One site refined uranium and fabricated uranium metal ingots for fuel and targets and the other site rolled hot forged uranium and other metals into dimensional rods. Unique signatures were found that are consistent with the activities and processes conducted at each facility and establish confidence in using these characteristics to reveal the provenance of other materials that exhibit similar signatures. (author)
Pilot production of 325 kg of uranium carbide
International Nuclear Information System (INIS)
Clozet, C.; Dessus, J.; Devillard, J.; Guibert, M.; Morlot, G.
1969-01-01
This report describes the pilot fabrication of uranium carbide rods to be mounted in bundles and assayed in two channels of the EL 4 reactor. The fabrication process includes: - elaboration of uranium carbide granules by carbothermic reduction of uranium dioxide; - electron bombardment melting and continuous casting of the granules; - machining of the raw ingots into rods of the required dimensions; finally, the rods will be piled-up to make the fuel elements. Both qualitative and quantitative results of this pilot production chain are presented and discussed. (authors) [fr
Process for preparing sintered uranium dioxide nuclear fuel
International Nuclear Information System (INIS)
Carter, R.E.
1975-01-01
Uranium dioxide is prepared for use as fuel in nuclear reactors by sintering it to the desired density at a temperature less than 1300 0 C in a chemically controlled gas atmosphere comprised of at least two gases which in equilibrium provide an oxygen partial pressure sufficient to maintain the uranium dioxide composition at an oxygen/uranium ratio of at least 2.005 at the sintering temperature. 7 Claims, No Drawings
The recycling of reprocessed uranium
International Nuclear Information System (INIS)
Lannegrace, J.-P.
1991-01-01
The 1990 update to the Uranium Institute's report ''Uranium Market Issues'', presented to this Symposium last year (1990) stated that the impact of recycled reprocessing products on uranium demand would be limited in the near future to that due to MOX fuel fabrication. The report stated that the recycling of reprocessed uranium was still at an early discussion stage, rather than being a short-term prospect. This paper will set out to challenge this assertion, on the basis both of facts and of economic and environmental incentives. (author)
Remote technologies for handling spent fuel
International Nuclear Information System (INIS)
Ramakumar, M.S.
1999-01-01
The nuclear programme in India involves building and operating power and research reactors, production and use of isotopes, fabrication of reactor fuel, reprocessing of irradiated fuel, recovery of plutonium and uranium-233, fabrication of fuel containing plutonium-239, uranium-233, post-irradiation examination of fuel and hardware and handling solid and liquid radioactive wastes. Fuel that could be termed 'spent' in thermal reactors is a source for second generation fuel (plutonium and uranium-233). Therefore, it is only logical to extend remote techniques beyond handling fuel from thermal reactors to fuel from fast reactors, post-irradiation examination etc. Fabrication of fuel containing plutonium and uranium-233 poses challenges in view of restriction on human exposure to radiation. Hence, automation will serve as a step towards remotisation. Automated systems, both rigid and flexible (using robots) need to be developed and implemented. Accounting of fissile material handled by robots in local area networks with appropriate access codes will be possible. While dealing with all these activities, it is essential to pay attention to maintenance and repair of the facilities. Remote techniques are essential here. There are a number of commonalities in these requirements and so development of modularized subsystems, and integration of different configurations should receive attention. On a long-term basis, activities like decontamination, decommissioning of facilities and handling of waste generated have to be addressed. While robotized remote systems have to be designed for existing facilities, future designs of facilities should take into account total operation with robotic remote systems. (author)
Estimates of future demand for uranium and nuclear fuel cycle services
Energy Technology Data Exchange (ETDEWEB)
Krymm, R; Woite, G [International Atomic Energy Agency, Division of Nuclear Power and Reactors, Economic Studies Section, Vienna (Austria)
1976-07-01
As a review of forecasts made over the last few years amply demonstrates, projections of nuclear power capacity on a country, regional or world basis are subject to uncertainties. It summarizes the evolution of estimates made in the recent past, should provide a sobering reminder of the advisability of relying on ranges rather than on single figures. Although they are derived from a relatively narrow range of assumptions for nuclear power capacity, the alternative estimates of demands for uranium and nuclear-fuel-cycle services differ by about 50%. If plausible variations in breeder penetration, load factors, tails assays and fuel performance were taken into account, a ratio of 2 between maximum and minimum possible demands for the 2000 could easily be approached. Thus, for instance, a 15% (instead of 5%) breeder penetration by the year 2000 would decrease annual natural uranium demand by about 10%, a drop of load factor from 0.7 to 0.6 would drop the demand by another 10%, a decrease in tail assay from 0.25% to 0.2% would drop the demand by 8%. These momentous uncertainties, characteristic of medium- and long-term demand projections, offer a sharp contrast to the inflexibility of short-term requirements. Once a nuclear plant is ordered, the demand for the fuel services required for its core and for its replacement loadings is practically fixed (subject to minor trade-offs) and it can only be delayed in time by accepting exceedingly heavy additional costs. The demand for uranium can be characterized as being uncertain in the future and inelastic in the present. It faces sources of supply which, with the exception of fabrication and conversion facilities, are characterized by long planning times, lengthy prospecting and construction times, and above all by heavy capital investments. This combination offers an almost ideal framework for instability and wild price fluctuations if consumers and suppliers operate independently seeking temporary guidance in their
Estimates of future demand for uranium and nuclear fuel cycle services
International Nuclear Information System (INIS)
Krymm, R.; Woite, G.
1976-01-01
As a review of forecasts made over the last few years amply demonstrates, projections of nuclear power capacity on a country, regional or world basis are subject to uncertainties. It summarizes the evolution of estimates made in the recent past, should provide a sobering reminder of the advisability of relying on ranges rather than on single figures. Although they are derived from a relatively narrow range of assumptions for nuclear power capacity, the alternative estimates of demands for uranium and nuclear-fuel-cycle services differ by about 50%. If plausible variations in breeder penetration, load factors, tails assays and fuel performance were taken into account, a ratio of 2 between maximum and minimum possible demands for the 2000 could easily be approached. Thus, for instance, a 15% (instead of 5%) breeder penetration by the year 2000 would decrease annual natural uranium demand by about 10%, a drop of load factor from 0.7 to 0.6 would drop the demand by another 10%, a decrease in tail assay from 0.25% to 0.2% would drop the demand by 8%. These momentous uncertainties, characteristic of medium- and long-term demand projections, offer a sharp contrast to the inflexibility of short-term requirements. Once a nuclear plant is ordered, the demand for the fuel services required for its core and for its replacement loadings is practically fixed (subject to minor trade-offs) and it can only be delayed in time by accepting exceedingly heavy additional costs. The demand for uranium can be characterized as being uncertain in the future and inelastic in the present. It faces sources of supply which, with the exception of fabrication and conversion facilities, are characterized by long planning times, lengthy prospecting and construction times, and above all by heavy capital investments. This combination offers an almost ideal framework for instability and wild price fluctuations if consumers and suppliers operate independently seeking temporary guidance in their
Interfacing robotics with plutonium fuel fabrication
International Nuclear Information System (INIS)
Bowen, W.W.; Moore, F.W.
1986-01-01
Interfacing robotic systems with nuclear fuel fabrication processes resulted in a number of interfacing challenges. The system not only interfaces with the fuel process, but must also interface with nuclear containment, radiation control boundaries, criticality control restrictions, and numerous other safety systems required in a fuel fabrication plant. The robotic system must be designed to allow operator interface during maintenance and recovery from an upset as well as normal operations
Radiation shielding calculation for the MOX fuel fabrication plant Melox
International Nuclear Information System (INIS)
Lee, Y.K.; Nimal, J.C.; Chiron, M.
1994-01-01
Radiation shielding calculation is an important engineering work in the design of the MOX fuel fabrication plant MELOX. Due to the recycle of plutonium and uranium from UO2 spent fuel reprocessing and the large capacity of production (120t HM/yr.), the shielding design requires more attention in this LWR fuel plant. In MELOX, besides several temporary storage facilities of massive fissile material, about one thousand radioactive sources with different geometries, forms, densities, quantities and Pu concentrations, are distributed through different workshops from the PuO 2 powder reception unit to the fuel assembly packing room. These sources, with or without close shield, stay temporarily in different locations, containers and glove boxes. In order to optimize the dimensions, the material and the cost of shield as well as to limit the calculation work in a reasonable engineer-hours, a calculation scheme for shielding design of MELOX is developed. This calculation scheme has been proved to be useful in consideration of the feedback from the evolutionary design and construction. The validated shielding calculations give a predictive but reliable radiation doses information. (authors). 2 figs., 10 refs
Fabrication development for the Advanced Neutron Source Reactor
International Nuclear Information System (INIS)
Pace, B.W.; Copeland, G.L.
1995-08-01
This report presents the fuel fabrication development for the Advanced Neutron Source (ANS) reactor. The fuel element is similar to that successfully fabricated and used in the High Flux Isotope Reactor (HFIR) for many years, but there are two significant differences that require some development. The fuel compound is U 3 Si 2 rather than U 3 O 8 , and the fuel is graded in the axial as well as the radial direction. Both of these changes can be accomplished with a straightforward extension of the HFIR technology. The ANS also requires some improvements in inspection technology and somewhat more stringent acceptance criteria. Early indications were that the fuel fabrication and inspection technology would produce a reactor core meeting the requirements of the ANS for the low volume fraction loadings needed for the highly enriched uranium design (up to 1.7 Mg U/m 3 ). Near the end of the development work, higher volume fractions were fabricated that would be required for a lower- enrichment uranium core. Again, results look encouraging for loadings up to ∼3.5 Mg U/m 3 ; however, much less evaluation was done for the higher loadings
Porous nuclear fuel element for high-temperature gas-cooled nuclear reactors
Youchison, Dennis L [Albuquerque, NM; Williams, Brian E [Pacoima, CA; Benander, Robert E [Pacoima, CA
2011-03-01
Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.
Standard specification for sintered gadolinium oxide-uranium dioxide pellets
American Society for Testing and Materials. Philadelphia
2008-01-01
1.1 This specification is for finished sintered gadolinium oxide-uranium dioxide pellets for use in light-water reactors. It applies to gadolinium oxide-uranium dioxide pellets containing uranium of any 235U concentration and any concentration of gadolinium oxide. 1.2 This specification recognizes the presence of reprocessed uranium in the fuel cycle and consequently defines isotopic limits for gadolinium oxide-uranium dioxide pellets made from commercial grade UO2. Such commercial grade UO2 is defined so that, regarding fuel design and manufacture, the product is essentially equivalent to that made from unirradiated uranium. UO2 falling outside these limits cannot necessarily be regarded as equivalent and may thus need special provisions at the fuel fabrication plant or in the fuel design. 1.3 This specification does not include (1) provisions for preventing criticality accidents or (2) requirements for health and safety. Observance of this specification does not relieve the user of the obligation to be aw...
International Nuclear Information System (INIS)
Sorenson, R.J.; McSweeney, T.I.; Hartman, M.G.; Brouns, R.J.; Stewart, K.B.; Granquist, D.P.
1977-11-01
This report describes the application of methodology for planning an inspection according to the procedures of the International Atomic Energy Agency (IAEA), and an example evaluation of data representative of low-enriched uranium fuel fabrication facilities. Included are the inspection plan test criteria, the inspection sampling plans, the sample data collected during the inspection, acceptance testing of physical inventories with test equipment, material unaccounted for (MUF) evaluation, and quantitative statements of the results and conclusions that could be derived from the inspection. The analysis in this report demonstrates the application of inspection strategies which produce quantitative results. A facility model was used that is representative of large low-enriched uranium fuel fabrication plants with material flows, inventory sizes, and compositions of material representative of operating commercial facilities. The principal objective was to determine and illustrate the degree of assurance against a diversion of special nuclear materials (SNM) that can be achieved by an inspection and the verification of material flows and inventories. This work was performed as part of the USA program for technical assistance to the IAEA. 10 figs, 14 tables
Integrated planning for a fuel industry with emphasis on minimum size to fabricate own fuel
International Nuclear Information System (INIS)
Kondal Rao, N.; Katiyar, H.C.; Rajendran, R.; Sinha, K.K.; Swaminathan, N.; Subramanyam, R.B.; Pande, B.P.; Krishnan, T.S.; Agarwala, G.C.; Chandramouli, V.A.
1977-01-01
The Indian nuclear energy programme is based on the utilization of indigenous resources for the economic generation of power, developing its own know-how. In order to gain time, the first nuclear power station at Tarapur is a turn-key job based on enriched uranium fuel. Taking into consideration the established resources of uranium and thorium in the country, a strategy for nuclear power programme has been drawn up. The first phase is based on natural uranium fuel, the second phase on the recycle of plutonium and conversion of thorium and the third phase is the breeder system based on utilization of U 233 and conversion of thorium. This programme is specially significant for India in view of its vast resources of thorium. After the experience and confidence gained with the manufacture of metallic uranium fuel for the research reactors and about 40 tonnes of fuel for the initial loading of the Rajasthan Reactor, the fuel manufacturing programme within the country has been implemented to meet the entire initial and reload fuel requirements. The plant capacities are small compared to similar activities in developed countries. Further, by planning for an integrated fuel and component manufacturing complex, any draw-back in smaller scale of some of the operations is off-set. At the Nuclear Fuel Complex, set up on the above principles, production plants are in operation for the manufacture of reload fuel for the 400 MW Tarapur station, natural uranium oxide fuel, various zircaloy components such as fuel sheaths, pressure tubes, calandria tubes, channels and various other zircaloy components. Provisions have been made to expand the production facilities as the demand for reload fuel grows. With the facilities provided, the production programme can be diversified to take up the production of fast breeder reactor components of stainless steel and also the blanket thorium elements. The unitary control of all aspects of the manufacture and quality control of different types
International Nuclear Information System (INIS)
1997-05-01
The Technical Committee Meeting on Recycling of Plutonium and Uranium in Water Reactor Fuel was recommended by the International Working Group on Fuel Performance and Technology (IWGFPT). Its aim was to obtain an overall picture of MOX fabrication capacity and technology, actual performance of this kind of fuel, and ways explored to dispose of the weapons grade plutonium. The subject of this meeting had been reviewed by the International Atomic Energy Agency every 5 to 6 years and for the first time the problem of weapons grade plutonium disposal was included. The papers presented provide a summary of experience on MOX fuel and ongoing research in this field in the participating countries. The meeting was hosted by British Nuclear Fuels plc, at Newby Bridge, United Kingdom, from 3 to 7 July 1995. Fifty-six participants from twelve countries or international organizations took part. Refs, figs, tabs
Quality assurance for breeder reactor fuel fabrication
International Nuclear Information System (INIS)
Marx, E.R.
1978-01-01
Fuel performance in the Fast Flux Test Facility (FFTF) depends on fabrication of fuel to rigorous quality standards. The quality program including Management, Procurement, Fabrication, Inspection, Records, and Audits is discussed as well as unique mixed oxide fuel inspections such as homogeneity inspection, analytical chemistry, and nondestructive fissile assay
Remote Handling Devices for Disposition of Enriched Uranium Reactor Fuel Using Melt-Dilute Process
International Nuclear Information System (INIS)
Heckendorn, F.M.
2001-01-01
Remote handling equipment is required to achieve the processing of highly radioactive, post reactor, fuel for the melt-dilute process, which will convert high enrichment uranium fuel elements into lower enrichment forms for subsequent disposal. The melt-dilute process combines highly radioactive enriched uranium fuel elements with deleted uranium and aluminum for inductive melting and inductive stirring steps that produce a stable aluminum/uranium ingot of low enrichment
Behavior of metallic uranium-fissium fuel in TREAT transient overpower tests
International Nuclear Information System (INIS)
Bauer, T.H.; Klickman, A.E.; Lo, R.K.; Rhodes, E.A.; Robinson, W.R.; Stanford, G.S.; Wright, A.E.
1986-01-01
TREAT tests M2, M3, and M4 were performed to obtain information on two key behavior characteristics of fuel under transient overpower accident conditions in metal-fueled fast reactors: the prefailure axial self-extrusion (elongation beyond thermal expansion) of fuel within intact cladding and the margin to cladding breach. Uranium-5 wt% fissium Experimental Breeder Reactor-II driver fuel pins were used for the tests since they were available as suitable stand-ins for the uranium-plutonium-zirconium ternary fuel, which is the reference fuel of the integral fast reactor (IFR) concept. The ternary fuel will be used in subsequent TREAT tests. Preliminary results from tests M2 and M3 were presented earlier. The present report includes significant advances in analysis as well as additional data from test M4. Test results and analysis have led to the development and validation of pin cladding failure and fuel extrusion models for metallic fuel, within reasonable uncertainties for the uranium-fissium alloy. Concepts involved are straightforward and readily extendable to ternary alloys and behavior in full-size reactors
Fuel cycle and waste management: A perspective from British nuclear fuels plc
International Nuclear Information System (INIS)
Holmes, R.G.G.; Fairhall, G.A.; Robbins, R.A.
1996-01-01
The phrase fuel cycle and waste management implies two separate and distinct activities. British Nuclear Fuels plc (BNFL) has adopted a holistic approach to the fuel cycle that integrates the traditional fuel cycle activities of conversion to uranium hexafluoride, fuel fabrication, power generation, and reprocessing with waste arisings, its subsequent treatment, and disposal
Thermal-hydraulic calculations for KUHFR with reduced enrichment uranium fuel
International Nuclear Information System (INIS)
Mishima, Kaichiro; Shibata, Toshikazu.
1982-01-01
This report provides the preliminary results of the thermal-hydraulic calculations to study the safety aspects in fueling the KUHFR with reduced enrichment uranium. The calculations were based on what was outlined in the Safety Analysis Report for the KUHFR and the guidebook for research reactor core conversion, IAEA-TECDOC-233, published by the International Atomic Energy Agency. No significant differences in the thermal-hydraulic operating conditions have been found between HEU and MEU fuels. However, in LEU cases, the combination of three factors - larger power peaking with LEU fuel, smaller thermal conductivity of U 3 O 8 -Al fuel with high uranium densities, and thicker fuel meat - resulted in higher maximum fuel and surface temperatures with the LEU oxide fuel. (author)
76 FR 72984 - Revised Application for a License To Export High-Enriched Uranium
2011-11-28
... NUCLEAR REGULATORY COMMISSION Revised Application for a License To Export High-Enriched Uranium The application for a license to export high-enriched Uranium has been revised as noted below. Notice... fabricate fuel France. Security Complex; October 18, Uranium (93.35%). uranium (174.0 elements in France...
Back-end fuel cycle efficiencies with respect to improved uranium utilization
International Nuclear Information System (INIS)
Kuczera, B.; Hennies, H.H.
1983-01-01
The world-wide nuclear power plant (NPP) capacity is at present 160 GW(e). If one adds the power stations under construction and ordered, a plant capacity of approximately 480 GW(e) is obtained for 1990, with the share of LWRs making up more than 80%. A modern LWR consumes in the open fuel cycle about 4400 metric tonnes of natural uranium per GW(e), assuming a lifetime of 30 years and a load factor of 70%. Considering the natural uranium reserves known at present and exploitable under economic conditions, it can be conveniently estimated that, with the present NPP capacity extension perspective, the natural uranium resources may be exhausted in a few decades. This trend can be counteracted in a flexible manner by various approaches in fuel cycle technology and strategy: (i) by steady further development of the established LWR technology the uranium consumption can be reduced by about 15%; (ii) closing the nuclear fuel cycle on the basis of LWRs (i.e. thermal uranium and plutonium recycling) implies up to 40% savings in natural uranium consumption; (iii) more recent considerations include the advanced pressurized water reactor (APWR). The APWR combines the proven PWR technology with a newly developed tight lattice core with greatly improved conversion characteristics (conversion ratio = 0.90 to 0.95). In terms of uranium utilization, the APWR has an efficiency three to five times higher than a PWR; (iv) Commercial introduction of FBR systems results in an optimal utilization of uranium which, at the same time, guarantees the supply of nuclear fuel well beyond the present century. For a corresponding transition period an energy supply system can be conceived which relies essentially on extended back-end fuel cycle capacities. These would facilitate a symbiosis of PWR, APWR and FBR, characterized by high flexibility with respect to long-term developments on the energy market. (author)
International Nuclear Information System (INIS)
Bernard, H.
1989-01-01
This paper deals with the fabrication of advanced fuels, such as mixed oxides for Pressurized Water Reactors or mixed nitrides for Fast Breeder Reactors. Although an extensive production experience exists for the mixed oxides used in the FBR, important work is still needed to improve the theoretical and technical knowledge of the production route which will be introduced in the future European facility, named Melox, at Marcoule. Recently, the feasibility of nitride fuel fabrication in existing commercial oxide facilities was demonstrated in France. The process, based on carbothermic reduction of oxides with subsequent comminution of the reaction product, cold pressing and sintering provides (U, Pu)N pellets with characteristics suitable for irradiation testing. Two experiments named NIMPHE 1 and 2 fabricated in collaboration with ITU, Karlsruhe, involve 16 nitride and 2 carbide pins, operating at a linear power of 45 and 73 kW/m with a smear density of 75-80% TD and a high burn-up target of 15 at%. These experiments are currently being irradiated in Phenix, at Marcoule. (orig.)
The use of medium enriched uranium fuel for research reactors
International Nuclear Information System (INIS)
1979-01-01
The evaluation described in the present paper concerns the use of medium enriched uranium fuel for our research reactors. The underlying assumptions set up for the evaluation are as follows: (1) At first, the use of alternative fuel should not affect, even to a small extent, research and development programs in nuclear energy utilization, which were described in the previous paper. Hence the use of lower enrichment fuel should not cause any reduction in reactor performances. (2) The fuel cycle cost for operating research reactors with alternative fuel, excepting R and D cost for such fuel, should not increase beyond an acceptable limit. (3) The use of alternative fuel should be satisfactory with respect to non-proliferation purposes, to the almost same degree as the use of 20% enriched uranium fuel
Natural uranium equivalent fuel an innovative design for proven CANDU technology
Energy Technology Data Exchange (ETDEWEB)
Pineiro, F.; Ho, K.; Khaial, A.; Boubcher, M.; Cottrell, C.; Kuran, S., E-mail: fabricia.pineiro@candu.com [Candu Energy Inc., Mississauga, ON (Canada); Zhenhua, Z.; Zhiliang, M. [Third Qinshan Nuclear Power Company, Haiyan, Zhejiang (China)
2015-07-01
The high neutron economy, on-power refuelling capability and fuel bundle design simplicity in CANDU reactors allow for the efficient utilization of alternative fuels. Candu Energy Inc. (Candu), in collaboration with the Third Qinshan Nuclear Power Company (TQNPC), the China North Nuclear Fuel Corporation (CNNFC), and the Nuclear Power Institute of China (NPIC), has successfully developed an advanced fuel called Natural Uranium Equivalent (NUE). This innovative design consists of a mixture of recycled and depleted uranium, which can be implemented in existing CANDU stations thereby bringing waste products back into the energy stream, increasing fuel resources diversity and reducing fuel costs. (author)
Natural uranium equivalent fuel. An innovative design for proven CANDU technology
Energy Technology Data Exchange (ETDEWEB)
Pineiro, F.; Ho, K.; Khaial, A.; Boubcher, M.; Cottrell, C.; Kuran, S. [Candu Energy Inc., Mississauga, Ontario (Canada); Zhenhua, Z.; Zhiliang, M. [Third Qinshan Nuclear Power Co., Haiyan, Zhejiang (China)
2015-09-15
The high neutron economy, on-power refuelling capability and fuel bundle design simplicity in CANDU® reactors allow for the efficient utilization of alternative fuels. Candu Energy Inc. (Candu), in collaboration with the Third Qinshan Nuclear Power Company (TQNPC), the China North Nuclear Fuel Corporation (CNNFC), and the Nuclear Power Institute of China (NPIC), has successfully developed an advanced fuel called Natural Uranium Equivalent (NUE). This innovative design consists of a mixture of recycled and depleted uranium, which can be implemented in existing CANDU stations thereby bringing waste products back into the energy stream, increasing fuel resources diversity and reducing fuel costs. (author)
Role of the DIAMO corporation in the nuclear fuel cycle
International Nuclear Information System (INIS)
Jansky, J.
1992-01-01
The Czechoslovak Uranium Industry corporation changed its name to DIAMO after 1989 and started entrepreneurial activities. The principal field of interest is the uranium mining and milling business, now extended to include the fuel cycle (both the front and back ends) and environmental issues. Uranium mining has been decreasing and, in the future, will be concentrated in the Straz pod Ralskem and Dolni Rozinka regions. A project has been developed aimed at the survey, assessment and reclamation of land affected by uranium mining. Engaged in the introduction of fuel fabrication for the Czech and Slovak power industries, DIAMO has been negotiating with foreign partners on technology transfer. The company intends to build its own fuel fabrication and assembling plant. It participates in studies concerned with the construction of an underground spent fuel storage facility and possibly a spent fuel disposal facility. (M.D.). 1 fig
Nuclear power and the nuclear fuel cycle
International Nuclear Information System (INIS)
Hardy, C.J.; Silver, J.M.
1985-09-01
The report provides data and assessments of the status and prospects of nuclear power and the nuclear fuel cycle. The report discusses the economic competitiveness of nuclear electricity generation, the extent of world uranium resources, production and requirements, uranium conversion and enrichment, fuel fabrication, spent fuel treatment and radioactive waste management. A review is given of the status of nuclear fusion research
Milling uranium silicide powder for dispersion nuclear fuels
Energy Technology Data Exchange (ETDEWEB)
Vieira, E.; Silva, D.G.; Souza, J.A.B.; Durazzo, M. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Riella, H.G. [Universidade Federal de Santa Catarina (UFSC), Florianopolis, SC (Brazil)
2009-07-01
Full text: Uranium silicide (U3Si2) is presently considered the best fuel qualified so far in terms of uranium loading and performance. Stability of the U3Si2 fuel with uranium density of 4.8 g/cm3 was confirmed by burnup stability tests performed during the Reduced Enrichment for Research and Test Reactors (RERTR) program. This fuel was chosen to compose the first core of the new Brazilian Multipurpose Research Reactor (RMB), planned to be constructed in the next years. This new reactor will consume bigger quantities of U3Si2 powder, when compared with the small consumption of the IEA-R1 research reactor of IPEN-CNEN/SP, the unique MTR type research reactor operating in the country. At the present time, the milling operation of U3Si2 ingots is made manually. In order to increase the powder production capacity, the manual milling must be replaced by an automated procedure. This paper describes a new milling machine and procedure developed to produce U3Si2 powder with higher efficiency. (author)
International Nuclear Information System (INIS)
Hakkila, E.A.; Barnes, J.W.; Dayem, H.A.; Dietz, R.J.; Shipley, J.P.
1978-10-01
This report addresses preliminary concepts for coordinated safeguards materials management in a typical generic thorium--uranium-fueled light-water reactor (LWR) fuels reprocessing plant. The reference facility is designed to recover thorium and uranium from first-generation (denatured 235 U) startup fuels, first-recycle and equilibrium (denatured 233 U) thorium--uranium LWR fuels, and to recover the plutonium generated in the 238 U denaturant as well. 12 figures, 3 tables
Quantifying Tc-99 contamination in a fuel fabrication plant - 59024
International Nuclear Information System (INIS)
Darbyshire, Carol; Burgess, Pete
2012-01-01
The Springfields facility manufactures nuclear fuel products for the UK's nuclear power stations and for international customers. Fuel manufacture is scheduled to continue into the future. In addition to fuel manufacture, Springfields is also undertaking decommissioning activities. Today it is run and operated by Springfields Fuels Limited, under the management of Westinghouse Electric UK Limited. The site has been operating since 1946 manufacturing nuclear fuel. As part of the decommissioning activities, there was a need was to quantify contamination in a large redundant building. This building had been used to process uranium derived from uranium ore concentrate but had also processed a limited quantity of recycled uranium. The major non-uranic contaminant was Tc-99. The aim was to be able to identify any areas where the bulk activity exceeded 0.4 Bq/g Tc-99 as this would preclude the demolition rubble being sent to the local disposal facility. The problems associated with this project were the presence of significant uranium contamination, the realisation that both the Tc-99 and the uranium had diffused into the brickwork to a significant depth and the relatively low beta energy of Tc-99. The uranium was accompanied by Pa-234m, an energetic beta emitter. The concentration/depth profile was determined for several areas on the plant for Tc-99 and for uranium. The radiochemical analysis was performed locally but the performance of the local laboratory was checked during the initial investigation by splitting samples three ways and having confirmation analyses performed by 2 other laboratories. The results showed surprisingly consistent concentration gradients for Tc-99 and for uranium across the samples. Using that information, the instrument response was calculated for Tc-99 using the observed diffusion gradient and averaged through the full 225 mm of brick wall, as agreed by the regulator. The Tc-99 and uranium contributions to the detector signal were separated
Fuel-cycle facilities: preliminary safety and environmental information document. Volume VII
Energy Technology Data Exchange (ETDEWEB)
1980-01-01
Information is presented concerning the mining and milling of uranium and thorium; uranium hexafluoride conversion; enrichment; fuel fabrication; reprocessing; storage options; waste disposal options; transportation; heavy-water-production facilities; and international fuel service centers.
Fuel-cycle facilities: preliminary safety and environmental information document. Volume VII
International Nuclear Information System (INIS)
1980-01-01
Information is presented concerning the mining and milling of uranium and thorium; uranium hexafluoride conversion; enrichment; fuel fabrication; reprocessing; storage options; waste disposal options; transportation; heavy-water-production facilities; and international fuel service centers
Recent Progress on the DUPIC Fuel Fabrication Technology at KAERI
International Nuclear Information System (INIS)
Jung-Won Lee; Ho-Jin Ryu; Geun-Il Park; Kee-Chan Song
2008-01-01
Since 1991, KAERI has been developing the DUPIC fuel cycle technology. The concept of a direct use of spent PWR fuel in Candu reactors (DUPIC) is based on a dry processing method to re-fabricate Candu fuel from spent PWR fuel without any intentional separation of the fissile materials and fission products. A DUPIC fuel pellet was successfully fabricated and the DUPIC fuel element fabrication processes were qualified on the basis of a Quality Assurance program. Consequently, the DUPIC fuel fabrication technology was verified and demonstrated on a laboratory-scale. Recently, the fuel discharge burn-up of PWRs has been extended to reduce the amount of spent fuel and the fuel cycle costs. Considering this trend of extending the fuel burn-up in PWRs, the DUPIC fuel fabrication technology should be improved to process high burn-up spent fuels. Particularly the release behavior of cesium from the pellet prepared with a high burn-up spent fuel was assessed. an improved DUPIC fuel fabrication technology was experimentally established with a fuel burn-up of 65,000 MWd/tU. (authors)
International Nuclear Information System (INIS)
Pigford, T.H.; Ang, K.P.
1975-01-01
The quantities of plutonium and other fuel actinides have been calculated for equilibrium fuel cycles for 1000-MW water reactors fueled with slightly enriched uranium, water reactors fueled with plutonium and natural uranium, fast-breder reactors, gas-cooled reactors fueled with thorium and highly enriched uranium, and gas-cooled reactors fueled with thorium, plutonium and recycled uranium. The radioactivity quantities of plutonium, americium and curium processed yearly in these fuel cycles are greatest for the water reactors fueled with natural uranium and recycled plutonium. The total amount of actinides processed is calculated for the predicted future growth of the U.S. nuclear power industry. For the same total installed nuclear power capacity, the introduction of the plutonium breeder has little effect upon the total amount of plutonium in this century. The estimated amount of plutonium in the low-level process wastes in the plutonium fuel cycles is comparable to the amount of plutonium in the high-level fission product wastes. The amount of plutonium processed in the nuclear fuel cycles can be considerably reduced by using gas-cooled reactors to consume plutonium produced in uranium-fueled water reactors. These, and other reactors dedicated for plutonium utilization, could be co-located with facilities for fuel reprocessing ad fuel fabrication to eliminate the off-site transport of separated plutonium. (author)
International Nuclear Information System (INIS)
Muhammad, Farhan; Majid, Asad
2009-01-01
The reactivity feedbacks of a material test research reactor using various low enriched uranium fuels, having same uranium density were calculated. For this purpose, the original aluminide fuel (UAl x -Al) containing 4.40 gU/cm 3 of an MTR was replaced with silicide (U 3 Si-Al and U 3 Si 2 -Al) and oxide (U 3 O 8 -Al) dispersion fuels having the same uranium density as of the original fuel. Calculations were carried out to find the fuel temperature reactivity feedback, moderator temperature reactivity feedback, moderator density reactivity feedback and moderator void reactivity feedback. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It was observed that the magnitudes all the respective reactivity feedbacks from 38 deg. C to 50 deg. C and 100 deg. C, at the beginning of life, of all the fuels were very close to each other. The fuel temperature reactivity feedback of the U 3 O 8 -Al was about 2% more than the original UAl x -Al fuel. The magnitudes of the moderator temperature, moderator density and moderator void reactivity feedbacks of all the fuels, showed very minor variations from the original aluminide fuel.
Fuel Fabrication and Nuclear Reactors
Energy Technology Data Exchange (ETDEWEB)
Karpius, Peter Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2017-02-02
The uranium from the enrichment plant is still in the form of UF6. UF6 is not suitable for use in a reactor due to its highly corrosive chemistry as well as its phase diagram. UF6 is converted into UO2 fuel pellets, which are in turn placed in fuel rods and assemblies. Reactor designs are variable in moderators, coolants, fuel, performance etc.The dream of energy ‘too-cheap to meter’ is no more, and now the nuclear power industry is pushing ahead with advanced reactor designs.
Separation and recovery method for depleted uranium from spent fuel
International Nuclear Information System (INIS)
Imoto, Yoshie; Fujita, Reiko.
1993-01-01
Spent oxide fuels are reduced in a molten salt of CaCl 2 -CaF 2 to convert them into metals, then melted in an Fe-U bath disposed in an electrolytic refining vessel and brought into contact with molten Mg, to extract transuranium elements and rare earth elements contained in the Fe-U bath as metals in the molten Mg. Then molten Mg is removed and the residue is brought into contact with KCl-LiCl molten salt and electrolyzed using the Fe-U as an anode. Then, uranium is recovered by deposition on an iron cathode disposed in chloride electrolytes of the electrolytic refining vessel. Uranium and transuranium elements can be thus separated and, for example, depleted uranium for use in blanket fuels can be recovered easily. This can greatly reduce the temporary storage amount of depleted uranium, to eliminate requirement for a large-scaled facility used exclusively for storing uranium and long time management for uranium. (T.M.)
Assessment of Nuclear Fuels using Radiographic Thickness Measurement Method
Energy Technology Data Exchange (ETDEWEB)
Muhammad Abir; Fahima Islam; Hyoung Koo Lee; Daniel Wachs
2014-11-01
The Convert branch of the National Nuclear Security Administration (NNSA) Global Threat Reduction Initiative (GTRI) focuses on the development of high uranium density fuels for research and test reactors for nonproliferation. This fuel is aimed to convert low density high enriched uranium (HEU) based fuel to high density low enriched uranium (LEU) based fuel for high performance research reactors (HPRR). There are five U.S. reactors that fall under the HPRR category, including: the Massachusetts Institute of Technology Reactor (MITR), the National Bureau of Standards Reactor (NBSR), the Missouri University Research Reactor (UMRR), the Advanced Test Reactor (ATR), and the High Flux Isotope Reactor (HFIR). U-Mo alloy fuel phase in the form of either monolithic or dispersion foil type fuels, such as ATR Full-size In center flux trap Position (AFIP) and Reduced Enrichment for Research and Test Reactor (RERTR), are being designed for this purpose. The fabrication process1 of RERTR is susceptible to introducing a variety of fuel defects. A dependable quality control method is required during fabrication of RERTR miniplates to maintain the allowable design tolerances, therefore evaluating and analytically verifying the fabricated miniplates for maintaining quality standards as well as safety. The purpose of this work is to analyze the thickness of the fabricated RERTR-12 miniplates using non-destructive technique to meet the fuel plate specification for RERTR fuel to be used in the ATR.
Recent irradiation tests of uranium-plutonium-zirconium metal fuel elements
International Nuclear Information System (INIS)
Pahl, R.G.; Lahm, C.E.; Villarreal, R.; Hofman, G.L.; Beck, W.N.
1986-09-01
Uranium-Plutonium-Zirconium metal fuel irradiation tests to support the ANL Integral Fast Reactor concept are discussed. Satisfactory performance has been demonstrated to 2.9 at.% peak burnup in three alloys having 0, 8, and 19 wt % plutonium. Fuel swelling measurements at low burnup in alloys to 26 wt % plutonium show that fuel deformation is primarily radial in direction. Increasing the plutonium content in the fuel diminishes the rate of fuel-cladding gap closure and axial fuel column growth. Chemical redistribution occurs by 2.1 at.% peak burnup and generally involves the inward migration of zirconium and outward migration of uranium. Fission gas release to the plenum ranges from 46% to 56% in the alloys irradiated to 2.9 at.% peak burnup. No evidence of deleterious fuel-cladding chemical or mechanical interaction was observed
International Nuclear Information System (INIS)
Asakura, K.; Aono, S.; Yamaguchi, T.; Deguchi, M.
2000-01-01
The Japan Nuclear Cycle Development Institute, JNC, designed, constructed and has operated the Plutonium Fuel Production Facility, PFPF, at the JNC Tokai Works to supply MOX fuels to the proto-type Fast Breeder Reactor, FBR, 'MONJU' and the experimental FBR 'JOYO' with 5 tonMOX/year of fabrication capability. Reduction of personal radiation exposure to a large amount of plutonium is one of the most important subjects in the development of MOX fabrication facility on a large scale. As the solution of this issue, the PFPF has introduced automated and/or remote controlled equipment in conjunction with computer controlled operation scheme. The PFPF started its operation in 1988 with JOYO reload fuel fabrication and has demonstrated MOX fuel fabrication on a large scale through JOYO and MONJU fuel fabrication for this decade. Through these operations, it has become obvious that several numbers of equipment initially installed in the PFPF need improvements in their performance and maintenance for commercial utilization of plutonium in the future. Furthermore, fuel fabrication of low density MOX pellets adopted in the MONJU fuel required a complete inspection because of difficulties in pellet fabrication compared with high density pellet for JOYO. This paper describes new pressing equipment with a powder recovery system, and pellet finishing and inspection equipment which has multiple functions, such as grinding measurements of outer diameter and density, and inspection of appearance to improve efficiency in the pellet finishing and inspection steps. Another development of technology concerning an annular pellet and an innovative process for MOX fuel fabrication are also described in this paper. (author)
Automated breeder fuel fabrication
International Nuclear Information System (INIS)
Goldmann, L.H.; Frederickson, J.R.
1983-01-01
The objective of the Secure Automated Fabrication (SAF) Project is to develop remotely operated equipment for the processing and manufacturing of breeder reactor fuel pins. The SAF line will be installed in the Fuels and Materials Examination Facility (FMEF). The FMEF is presently under construction at the Department of Energy's (DOE) Hanford site near Richland, Washington, and is operated by the Westinghouse Hanford Company (WHC). The fabrication and support systems of the SAF line are designed for computer-controlled operation from a centralized control room. Remote and automated fuel fabriction operations will result in: reduced radiation exposure to workers; enhanced safeguards; improved product quality; near real-time accountability, and increased productivity. The present schedule calls for installation of SAF line equipment in the FMEF beginning in 1984, with qualifying runs starting in 1986 and production commencing in 1987. 5 figures
Extending the world's uranium resources through advanced CANDU fuel cycles
Energy Technology Data Exchange (ETDEWEB)
De Vuono, Tony; Yee, Frank; Aleyaseen, Val; Kuran, Sermet; Cottrell, Catherine
2010-09-15
The growing demand for nuclear power will encourage many countries to undertake initiatives to ensure a self-reliant fuel source supply. Uranium is currently the only fuel utilized in nuclear reactors. There are increasing concerns that primary uranium sources will not be enough to meet future needs. AECL has developed a fuel cycle vision that incorporates other sources of advanced fuels to be adaptable to its CANDU technology.
International Nuclear Information System (INIS)
Guimaraes, Leonam dos Santos; Degrange, Jean Pierre
1998-01-01
The study introduced by this paper compose the first step to the implementation of ALARA (As Low As Reasonable Achievable) for a nuclear fuel assembly factory which one of its two production lines will be designed to work with Enriched Reprocessing Uranium (ERU). This step includes the reference situation analysis is based on previsional dosimetric evaluations for individual and collective exposures of each factory operator (117 in total) working on 7 work stations, considering 6 annual production scenarios (10, 50 75, 100 and 150 ERU tons), which corresponds to an annual production of 600 tons (ERU plus enriched natural uranium ENU). The exposure indicators evolution, expressed in terms of collective dose, annual individual dose and radiological detrimental cost for workers, is also used in a complimentary way to guide the analysis. (author)
Mixed Uranium/Refractory Metal Carbide Fuels for High Performance Nuclear Reactors
International Nuclear Information System (INIS)
Knight, Travis; Anghaie, Samim
2002-01-01
Single phase, solid-solution mixed uranium/refractory metal carbides have been proposed as an advanced nuclear fuel for advanced, high-performance reactors. Earlier studies of mixed carbides focused on uranium and either thorium or plutonium as a fuel for fast breeder reactors enabling shorter doubling owing to the greater fissile atom density. However, the mixed uranium/refractory carbides such as (U, Zr, Nb)C have a lower uranium densities but hold significant promise because of their ultra-high melting points (typically greater than 3700 K), improved material compatibility, and high thermal conductivity approaching that of the metal. Various compositions of (U, Zr, Nb)C were processed with 5% and 10% metal mole fraction of uranium. Stoichiometric samples were processed from the constituent carbide powders, while hypo-stoichiometric samples with carbon-to-metal (C/M) ratios of 0.92 were processed from uranium hydride, graphite, and constituent refractory carbide powders. Processing techniques of cold uniaxial pressing, dynamic magnetic compaction, sintering, and hot pressing were investigated to optimize the processing parameters necessary to produce high density (low porosity), single phase, solid-solution mixed carbide nuclear fuels for testing. This investigation was undertaken to evaluate and characterize the performance of these mixed uranium/refractory metal carbides for high performance, ultra-safe nuclear reactor applications. (authors)
Use of enriched uranium as a fuel in CANDU reactors
International Nuclear Information System (INIS)
Zech, H.J.
1976-08-01
The use of slightly enriched uranium as a fuel in CANDU-reactors is studied in a simple parametric way. The results show the possibility of 1) about 30% savings in natural uranium consumption 2) about 35% increase in the utilization of the natural uranium 3) a decrease in fuelling costs to about 70 - 80% of the normal case of natural uranium fuelling. (orig.) [de
Loading ion exchange resins with uranium for HTGR fuel kernels
International Nuclear Information System (INIS)
Notz, K.J.; Greene, C.W.
1976-12-01
Uranium-loaded ion exchange beads provide an excellent starting material in the production of uranium carbide microspheres for nuclear fuel applications. Both strong-acid (sulfonate) and weak-acid (carboxylate) resins can be fully loaded with uranium from a uranyl nitrate solution utilizing either a batch method or a continuous column technique
Design and fabrication of fuel for the prototype heavy water reactor Fugen
International Nuclear Information System (INIS)
Hasumi, Takashi; Yamanaka, Ryozi; Osawa, Masahide; Asami, Tomohiro; Kaziyama, Takashi
1983-01-01
For the advanced thermal reactor Fugen, 224 fuel assemblies were charged as the initial charge fuel, of which 96 were uranium-plutonium mixed oxide fuel, and 128 were uranium dioxide fuel. Since the full scale operation was started in March, 1979, fuel exchange was carried out five times, and 240 fuel assemblies were taken out, but fuel breaking was never found, and the fuel for Fugen has shown good result. For 16 mixed oxide fuel assemblies for the third exchange and thereafter, the domestically produced plutonium extracted in the Tokai reprocessing plant has been used, and for 15 UO 2 fuel assemblies for the fifth exchange, the enriched uranium produced in the Ningyo Pass plant was used. These fuels are burning in the core without causing trouble. The course of the development of the fuel is described as follows: trial manufacture, evaluation test outside the core, heat transferring flow characteristic test, irradiation test, design of fuel elements and fuel assemblies, production of fuel and quality assurance, and results of production and use. (Kako, I.)
LOFT fuel modules design, characterization, and fabrication program
International Nuclear Information System (INIS)
Russell, M.L.
1977-06-01
The loss-of-fluid test [LOFT) fuel modules have evolved from a comprehensive five-year design, characterization, and fabrication program which has resulted in the accomplishment of many technical activities of interest in pressurized water reactor fuel design development and safety research. Information is presented concerning: determination of fundamental high-temperature reactor material properties; design invention related to in-core instrumentation attachment; implementation of advanced and/or unique fuel bundle characterization techniques; implementation of improved fuel bundle fabrication techniques; and planning and execution of a multimillion dollar design, characterization, and fabrication program for pressurized water reactor fuel
Composition and Distribution of Tramp Uranium Contamination on BWR and PWR Fuel Rods
International Nuclear Information System (INIS)
Schienbein, Marcel; Zeh, Peter; Hurtado, Antonio; Rosskamp, Matthias; Mailand, Irene; Bolz, Michael
2012-09-01
In a joint research project of VGB and AREVA NP GmbH the behaviour of alpha nuclides in nuclear power plants with light water reactors has been investigated. Understanding the source and the behaviour of alpha nuclides is of big importance for planning radiation protection measures for outages and upcoming dismantling projects. Previous publications have shown the correlation between plant specific alpha contamination of the core and the so called 'tramp fuel' or 'tramp uranium' level which is linked to the defect history of fuel assemblies and accordingly the amount of previously washed out fuel from defective fuel rods. The methodology of tramp fuel estimation is based on fission product concentrations in reactor coolant but also needs a good knowledge of tramp fuel composition and in-core distribution on the outer surface of fuel rods itself. Sampling campaigns of CRUD deposits of irradiated fuel assemblies in different NPPs were performed. CRUD analyses including nuclide specific alpha analysis have shown systematic differences between BWR and PWR plants. Those data combined with literature results of fuel pellet investigations led to model improvements showing that a main part of fission products is caused by fission of Pu-239 an activation product of U-238. CRUD investigations also gave a better picture of the in-core composition and distribution of the tramp uranium contamination. It was shown that the tramp uranium distribution in PWR plants is time dependent. Even new fuel assemblies will be notably contaminated after only one cycle of operation. For PWR applies the following logic: the higher the local power the higher the contamination. With increasing burnup the local rod power usually decreases leading to decreasing tramp uranium contamination on the fuel rod surface. This is not applicable for tramp uranium contamination in BWR. CRUD contamination (including the tramp fuel deposits) is much more fixed and is constantly increasing
International Nuclear Information System (INIS)
Wang Jiangong
2010-01-01
In order to meet the dosage test the operational needs of the laser fluorescence determination of trace radioactive waste cotton fabric uranium research and exploration, to determine the sample ashing time, measured dosage of acidity and digestion and other technical parameters, gives the laser fluorescence determination of radioactive abandoned cotton fabric of trace uranium method. Method of high sensitivity, strong anti-interference, the detection limit of 0.025μg/g(Burning down dregs), relative standard deviation was 3.96%, the mean recovery 93.3%-103% for masks, gloves and other radioactive waste to the determination of trace uranium. (authors)
Fabrication of simulated DUPIC fuel
Kang, Kweon Ho; Song, Ki Chan; Park, Hee Sung; Moon, Je Sun; Yang, Myung Seung
2000-12-01
Simulated DUPIC fuel provides a convenient way to investigate the DUPIC fuel properties and behavior such as thermal conductivity, thermal expansion, fission gas release, leaching, and so on without the complications of handling radioactive materials. Several pellets simulating the composition and microstructure of DUPIC fuel are fabricated by resintering the powder, which was treated through OREOX process of simulated spent PWR fuel pellets, which had been prepared from a mixture of UO2 and stable forms of constituent nuclides. The key issues for producing simulated pellets that replicate the phases and microstructure of irradiated fuel are to achieve a submicrometre dispersion during mixing and diffusional homogeneity during sintering. This study describes the powder treatment, OREOX, compaction and sintering to fabricate simulated DUPIC fuel using the simulated spent PWR fuel. The homogeneity of additives in the powder was observed after attrition milling. The microstructure of the simulated spent PWR fuel agrees well with the other studies. The leading structural features observed are as follows: rare earth and other oxides dissolved in the UO2 matrix, small metallic precipitates distributed throughout the matrix, and a perovskite phase finely dispersed on grain boundaries.
Fuel balance in nuclear power with fast reactors without a uranium blanket
International Nuclear Information System (INIS)
Naumov, V.V.; Orlov, V.V.; Smirnov, V.S.
1994-01-01
General aspects related to replacing the uranium blanket of a lead-cooled fast reactor burning uranium-plutonium nitride fuel with a more efficient lead reflector are briefly discussed in the article. A study is very briefly summarized, which showed that a breeding ratio of about 1 and electric power of about 300 MW were achievable. A nuclear fuel balance is performed to estimate the increased consumption of uranium to produce power and the gains achievable by eliminating the uranium blanket. Elimination of the uranium blanket has the advantages of simplifying and improving the fast reactor and eliminating the production of weapons quality plutonium. 3 figs
75 FR 15743 - Application for a License To Export High-Enriched Uranium
2010-03-30
... NUCLEAR REGULATORY COMMISSION Application for a License To Export High-Enriched Uranium Pursuant to 10 CFR 110.70(c) ``Public notice of receipt of an application,'' please take notice that the...-Enriched 160.0 kilograms To fabricate fuel France. Complex, March 3, 2010. Uranium (93.35%). uranium (149...
Reproduction of the RA reactor fuel element fabrication
International Nuclear Information System (INIS)
Novakovic, M.
1961-12-01
This document includes the following nine reports: Final report on task 08/12 - testing the Ra reactor fuel element; design concept for fabrication of RA reactor fuel element; investigation of the microstructure of the Ra reactor fuel element; Final report on task 08/13 producing binary alloys with Al, Mo, Zr, Nb and B additions; fabrication of U-Al alloy; final report on tasks 08/14 and 08/16; final report on task 08/32 diffusion bond between the fuel and the cladding of the Ra reactor fuel element; Final report on task 08/33, fabrication of the RA reactor fuel element cladding; and final report on task 08/36, diffusion of solid state metals [sr
Fuel buyers guide: company data
International Nuclear Information System (INIS)
Anon.
1991-01-01
Four major listings relating to nuclear fuel services are provided. 1. A fuel buyer's guide listing companies under alphabetical order of country and giving addresses and an indication of the services offered. 2. A fuel buyers guide classifying companies in alphabetical order of the services offered. 3. A fuel and front end facility listing subdivided into companies involved in: uranium ore processing; uranium refining and conversion; enrichment; fuel fabrication; heavy water production; zirconium metal production; and zirconium tube production. 4. A fuel and front end facilities listing giving operators' addresses under alphabetical order of country. (UK)
Irradiation behavior of experimental miniature uranium silicide fuel plates
International Nuclear Information System (INIS)
Hofman, Gerard L.; Neimark, L.A.; Mattas, R.F.
1983-01-01
Uranium silicides, because of their relatively high uranium density, were selected as candidate dispersion fuels for the higher fuel densities required in the Reduced Enrichment Research and Test Reactor (RERTR) Program. Irradiation experience with this type of fuel, however, was limited to relatively modest fission densities in the bulk form, on the order of 7 x 10 20 cm -3 , far short of he approximately 20 x 10 20 cm -3 goal established for the RERTR Program. The purpose of the irradiation experiments on silicide fuels in the ORR, therefore, was to investigate the intrinsic irradiation behavior of uranium silicide as a dispersion fuel. Of particular interest was the interaction between the silicide particles and the aluminum matrix, the swelling behavior of the silicide particles, and the maximum volume fraction of silicide particles that could be contained in the aluminum matrix. The first group of experimental 'mini' fuel plates have recently reached the program's goal burnup and are in various stages of examination. Although the results to date indicate some limitations, it appears that within the range of parameters examined thus far the uranium silicide dispersion holds promise for satisfying most of the needs of the RERTR Program. The twelve experimental silicide dispersion fuel plates that were irradiated to approximately their goal exposure show the 30-vol % U 3 Si-Al plates to be in a stage of relatively rapid fission-gas-driven swelling at a fission density of 2 x 10 20 cm -3 . This fuel swelling will likely result in unacceptably large plate-thickness increases. The U 3 Si plates appear to be superior in this respect; however, they, too, are starting to move into the rapid fuel-swelling stage. Analysis of the currently available post irradiation data indicates that a 40-vol % dispersed fuel may offer an acceptable margin to the onset of unstable thickness changes at exposures of 2 x 10 21 fission/cm 3 . The interdiffusion between fuel and matrix
International Nuclear Information System (INIS)
Bastin, J.J.; Jump, M.J.; Lange, R.A.; Randall, C.C.
1977-11-01
The Phase I study of the application of advanced accountability methods (DYMAC) in a uranium/plutonium mixed oxide facility was extended to assess the effect of coprocessed UO 2 --PuO 2 feed on the observations made in the original Phase I effort and on the proposed Phase II program. The retention of plutonium mixed with uranium throughout the process was also considered. This addendum reports that coprocessed feed would have minimal effect on the DYMAC program, except in the areas of material specifications, starting material delivery schedule, and labor requirements. Each of these areas is addressed, as are the impact of coprocessed feed at a large fuel fabrication facility and the changes needed in the dirty scrap recovery process to maintain the lower plutonium levels which may be required by future nonproliferation philosophy. An amended schedule for Phase II is included
Supply and demand estimates for the nuclear fuel cycle
International Nuclear Information System (INIS)
Haussermann, W.; Hogroian, P.; Krymm, R.; Cameron, J.
1977-01-01
Based on the nuclear power growth forecasts described in the papers for Session I.B., estimates of requirements in the nuclear fuel cycle are given, notably concerning: - natural uranium, - enriched uranium, - fuel fabrication services, and - reprocessing services. The influence of realistic scenarios of uranium and plutonium recycling on fuel cycle requirements is discussed. Furthermore, the known plans for uranium and related fuel cycle production capacities are compared with the foreseeable demand. These estimates cover the period between now and the year 2000. However, in order to determine the influence of possible variations in reactor strategies on uranium demand, notably the introduction of breeder reactors, power growth projections and resulting fuel cycle requirements beyond the year 2000 are also briefly considered [fr
Foreign research reactor uranium supply program: The Y-12 national security complex process
International Nuclear Information System (INIS)
Nelson, T.; Eddy, B.G.
2010-01-01
The Foreign Research Reactor (FRR) Uranium Supply Program at the Y-12 National Security Complex supports the nonproliferation objectives of the HEU Disposition Program, the Reduced Enrichment Research and Test Reactors (RERTR) Program, and the United States FRR Spent Nuclear Fuel (SNF) Acceptance Program. The Y-12 National Nuclear Security Administration (NNSA) Y-12 Site Office maintains the prime contracts with foreign governments for the supply of Low-Enriched Uranium (LEU) for their research reactors. The LEU is produced by down blending Highly Enriched Uranium (HEU) that has been declared surplus to the U.S. national defense needs. The down blending and sale of the LEU supports the Surplus HEU Disposition Program Record of Decision to make the HEU non-weapons usable and to recover the economic value of the uranium to the extent feasible. This program supports the important U.S. government and nuclear nonproliferation commitment to serve as a reliable and cost-effective uranium supplier for those foreign research reactors that are converting or have converted to LEU fuel under the guidance of the NNSA RERTR Program. In conjunction with the FRR SNF Acceptance Program which supports the global nonproliferation efforts to disposition U.S.-origin HEU, the Y-12 FRR Uranium Supply Program can provide the LEU for the replacement fuel fabrication. In addition to feedstock for fuel fabrication, Y-12 supplies LEU for target fabrication for medical isotope production. The Y-12 process uses supply forecasting tools, production improvements and efficient delivery preparations to successfully support the global research reactor community
Sodium-cooled Fast Reactor Cores using Uranium-Free Metallic Fuels for Maximizing TRU Support Ratio
International Nuclear Information System (INIS)
You, WuSeung; Hong, Ser Gi
2014-01-01
The depleted uranium plays important roles in the SFR burner cores because it substantially contributes to the inherent safety of the core through the negative Doppler coefficient and large delayed neutron. However, the use of depleted uranium as a diluent nuclide leads to a limited value of TRU support ratio due to the generation of TRUs through the breeding. In this paper, we designed sodium cooled fast reactor (SFR) cores having uranium-free fuels 3,4 for maximization of TRU consumption rate. However, the uranium-free fuelled burner cores can be penalized by unacceptably small values of the Doppler coefficient and small delayed neutron fraction. In this work, metallic fuels of TRU-(W or Ni)-Zr are considered to improve the performances of the uranium-free cores. The objective of this work is to consistently compare the neutronic performances of uranium-free sodium cooled fast reactor cores having TRU-Zr metallic fuels added with Ni or W and also to clarify what are the problematic features to be resolved. In this paper, a consistent comparative study of 400MWe sodium cooled burner cores having uranium-based fuels and uranium-free fuels was done to analyze the relative core neutronic features. Also, we proposed a uranium-free metallic fuel based on Nickel. From the results, it is found that tungsten-based uranium-free metallic fuel gives large negative Doppler coefficient due to high resonance of tungsten isotopes but this core has large sodium void worth and small effective delayed neutron fraction while the nickel-based uranium-free metallic fuelled core has less negative Doppler coefficient but smaller sodium void worth and larger effective delayed neutron fraction than the tungsten-based one. On the other hand, the core having TRU-Zr has very high burnup reactivity swing which may be problematic in compensating it using control rods and the least negative Doppler coefficient
Porous nuclear fuel element with internal skeleton for high-temperature gas-cooled nuclear reactors
Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.
2013-09-03
Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.
Isotopic composition and radiological properties of uranium in selected fuel cycles
International Nuclear Information System (INIS)
Fleischman, R.M.; Liikala, R.C.
1975-04-01
Three major topic areas are discussed: First, the properties of the uranium isotopes are defined relative to their respective roles in the nuclear fuel cycle. Secondly, the most predominant fuel cycles expected in the U. S. are described. These are the Light Water Reactor (LWR), High Temperature Gas Cooled Reactor (HTGR), and Liquid Metal Fast Breeder Reactor (LMFBR) fuel cycles. The isotopic compositions of uranium and plutonium fuels expected for these fuel cycles are given in some detail. Finally the various waste streams from these fuel cycles are discussed in terms of their relative toxicity. Emphasis is given to the high level waste streams from reprocessing of spent fuel. Wastes from the various fuel cycles are compared based on projected growth patterns for nuclear power and its various components. (U.S.)
Nuclear Fuel Complex - a landmark of indigenous nuclear technology
Energy Technology Data Exchange (ETDEWEB)
Katiyar, H C [Nuclear Fuel Complex, Hyderabad (India)
1977-01-01
The Nuclear Fuel Complex (NFC) set up in India for manufacturing fuel and related hardware has proved to be a significant step towards self-sufficiency and saving of foreign exchange. The complex is involved in the entire operations starting from processing of raw material concentrates to finishing of fuel assemblies and other zircaloy reactor components. The complex consists of the following units : (1) Zirconium Oxide Plant, (2) Zirconium Sponge Plant, (3) Zircaloy Fabrication Plant, (4) Uranium Oxide Plant, (5) Ceramic Fabrication Plant, (6) Enriched Uranium Oxide Plant, (7) Enriched Fuel Fabrication Plant, (8) Special Materials Plant and (9) Titanium Plant. A brief description of the activities of the various units of the complex are given. The effluent management scheme is outlined. The requirements and cost of fuel and zircaloy components for the power stations at Tarapur, Kota and Kalpakkam are mentioned.
Analysis of the Reuse of Uranium Recovered from the Reprocessing of Commercial LWR Spent Fuel
International Nuclear Information System (INIS)
DelCul, Guillermo D.; Trowbridge, Lee D.; Renier, John-Paul; Ellis, Ronald James; Williams, Kent Alan; Spencer, Barry B.; Collins, Emory D.
2009-01-01
This report provides an analysis of the factors involved in the reuse of uranium recovered from commercial light-water-reactor (LWR) spent fuels (1) by reenrichment and recycling as fuel to LWRs and/or (2) by recycling directly as fuel to heavy-water-reactors (HWRs), such as the CANDU (registered trade name for the Canadian Deuterium Uranium Reactor). Reuse is an attractive alternative to the current Advanced Fuel Cycle Initiative (AFCI) Global Nuclear Energy Partnership (GNEP) baseline plan, which stores the reprocessed uranium (RU) for an uncertain future or attempts to dispose of it as 'greater-than-Class C' waste. Considering that the open fuel cycle currently deployed in the United States already creates a huge excess quantity of depleted uranium, the closed fuel cycle should enable the recycle of the major components of spent fuel, such as the uranium and the hazardous, long-lived transuranic (TRU) actinides, as well as the managed disposal of fission product wastes. Compared with the GNEP baseline scenario, the reuse of RU in the uranium fuel cycle has a number of potential advantages: (1) avoidance of purchase costs of 11-20% of the natural uranium feed; (2) avoidance of disposal costs for a large majority of the volume of spent fuel that is reprocessed; (3) avoidance of disposal costs for a portion of the depleted uranium from the enrichment step; (4) depending on the 235 U assay of the RU, possible avoidance of separative work costs; and (5) a significant increase in the production of 238 Pu due to the presence of 236 U, which benefits somewhat the transmutation value of the plutonium and also provides some proliferation resistance
The development and localization of nuclear fuel technology for KMRR
International Nuclear Information System (INIS)
Kim, Seong Yun; Lee, Ji Bok; Suk, Ho Chun; Kuk, Il Hyun; Hwang, Woan; Kim, Bong Goo; Park, Joo Hwan; Kim, Young Jin; Kang, Thae Khapp; Lee, Jae Choon
1988-05-01
This project was implemented aiming at localizing the fabrication of the KMRR fuel by october 1993. The contents of this project were divided into three parts: fuel design, fuel fabrication and process criticality analysis. In the fuel design, the radial power distribution in the fuel core was modeled and formulated taking account of the neutron flux depression in the radial direction. It was also performed to model and formulate the thermal characteristics such as the thermal conductivity and specific heat of the fuel core, U3Si-Al, the swelling and the film coefficient of heat transfer between the aluminum clad and light water coolant. The two dimensional heat transfer in the finned fuel element was equated based on the general equation governing the heat transfer in materials in order to develope a computer code, TEMP2D. TEMP2D solves finite differenced equations to calculate a two dimensional fuel temperature distribution under the steady and transient states. In the fuel fabrication, the technologies of fabricating uranium silicide fuel meat were tried by using depleted uranium as a raw material. These were extended to find the problems in technologies and to establish the ways of approach. The end product, so called fuel meat, was a metallic powder compound, U3Six(1≤x≤2), dispersed in Al matrix. The fuel meat was fabricated by the horizontal extrusion technique, and powder extrusion technique. Fabrication technologies comprise five different continuous processes: melting and casting of metallic uranium with silicon and aluminum, heat treatment, chipping and crushing, pulverizing, and extrusion. In the process criticality analysis, AMPX-KENO benchmark calculation was performed and calculational error of AMPX-KENO system was established. (Author)
The Y-12 National Security Complex Foreign Research Reactor Uranium Supply Production
Energy Technology Data Exchange (ETDEWEB)
Nelson, T. [Nuclear Technology and Nonproliferation Programs, B and W Y-12, L.L.C., Y-12 National Security Complex, Oak Ridge, Tennessee (United States); Keller, A.P. [Disposition and Supply Programs, B and W Y-12, L.L.C., Y-12 National Security Complex, Oak Ridge, Tennessee (United States)
2011-07-01
The Foreign Research Reactor (FRR) Uranium Supply Program at the Y-12 National Security Complex supports the nonproliferation objectives of the National Nuclear Security Administration (NNSA) HEU Disposition, the Reduced Enrichment Research and Test Reactors (RERTR), and the United States (U.S.) FRR Spent Nuclear Fuel (SNF) Acceptance Programs. The FRR Supply Program supports the important U.S. government nuclear nonproliferation commitment to serve as a reliable and cost-effective uranium supplier for those foreign research reactors that are converting or have converted to Low-Enriched Uranium (LEU) fuel under the RERTR Program. The NNSA Y-12 Site Office maintains the prime contracts with foreign government agencies for the supply of LEU for their research reactors. The LEU is produced by down blending Highly Enriched Uranium (HEU) that has been declared surplus to the U.S. national defense needs. The down blending and sale of the LEU supports the Surplus HEU Disposition Program Record of Decision to make the HEU non-weapons usable and to recover the economic value of the uranium to the extent feasible. In addition to uranium metal feedstock for fuel fabrication, Y-12 can produce LEU in different forms to support new fuel development or target fabrication for medical isotope production. With production improvements and efficient delivery preparations, Y-12 continues to successfully support the global research reactor community. (author)
Gel-sphere-pac reactor fuel fabrication and its application to a variety of fuels
International Nuclear Information System (INIS)
Olsen, A.R.; Judkins, R.R.
1979-12-01
The gel-sphere-pac fuel fabrication option was evaluated for its possible application to commercial scale fuel fabrication for 19 fuel element designs that use oxide fuel in metal clad rods. The dry gel spheres are prepared at the reprocessing plant and are then calcined, sintered, inspected, and loaded into fuel rods and packed by low-energy vibration. A fuel smear density of 83 to 88% theoretical can be obtained. All fuel fabrication process steps were defined and evaluated from fuel receiving to finished fuel element shipping. The evaluation also covers the feasibility of the process, the current status of technology, estimates of the required time and cost to develop the technology to commercial status, and the safety and licensability of commercial scale plants. The primary evaluation was for a Light-Water Reactor fuel element containing (U,Pu)O 2 fuel. The other 18 fuel element types - 3 for Light-Water Reactors, 1 for a Heavy-Water Reactor, 1 for a Gas-Cooled Fast Reactor, 7 for Liquid-Metal-Cooled Fast Breeder Reactors, and 3 pairs for Light-Water Prebreeder and Breeder Reactors - were compared with the Light-Water Reactor. The gel-sphere-pac option was found applicable to 17 of the 19 element types; the characteristics of a commercial scale plant were defined for these for making cost estimates for such plants. The evaluation clearly shows the gel-sphere-pac process to be a viable fuel fabrication option. Estimates indicate a significant potential fabrication cost advantage for the gel-sphere-pac process if a remotely operated and remotely maintained fuel fabrication plant is required
Sustainable and safe energy supply with seawater uranium fueled HTGR and its economy
International Nuclear Information System (INIS)
Fukaya, Y.; Goto, M.
2017-01-01
Highlights: • We discussed uranium resources with an energy security perspective. • We concluded seawater uranium is preferable for sustainability and energy security. • We evaluated electricity generation cost of seawater uranium fueled HTGR. • We concluded electricity generation with seawater uranium is reasonable. - Abstract: Sustainable and safe energy supply with High Temperature Gas-cooled Reactor (HTGR) fueled by uranium from seawater have been investigated and discussed. From the view point of safety feature of self-regulation with thermal reactor of HTGR, the uranium resources should be inexhaustible. The seawater uranium is expected to be alternative resources to conventional resources because it exists so much in seawater as a solute. It is said that 4.5 billion tons of uranium is dissolved in the seawater, which corresponds to a consumption of approximately 72 thousand years. Moreover, a thousand times of the amount of 4.5 trillion tU of uranium, which corresponds to the consumption of 72 million years, also is included in the rock on the surface of the sea floor, and that is also recoverable as seawater uranium because uranium in seawater is in an equilibrium state with that. In other words, the uranium from seawater is almost inexhaustible natural resource. However, the recovery cost with current technology is still expensive compared with that of conventional uranium. Then, we assessed the effect of increase in uranium purchase cost on the entire electricity generation cost. In this study, the economy of electricity generation of cost of a commercial HTGR was evaluated with conventional uranium and seawater uranium. Compared with ordinary LWR using conventional uranium, HTGR can generate electricity cheaply because of small volume of simple direct gas turbine system compared with water and steam systems of LWR, rationalization by modularizing, and high thermal efficiency, even if fueled by seawater uranium. It is concluded that the HTGR
Survey of Worldwide Light Water Reactor Experience with Mixed Uranium-Plutonium Oxide Fuel
Energy Technology Data Exchange (ETDEWEB)
Cowell, B.S.; Fisher, S.E.
1999-02-01
The US and the Former Soviet Union (FSU) have recently declared quantities of weapons materials, including weapons-grade (WG) plutonium, excess to strategic requirements. One of the leading candidates for the disposition of excess WG plutonium is irradiation in light water reactors (LWRs) as mixed uranium-plutonium oxide (MOX) fuel. A description of the MOX fuel fabrication techniques in worldwide use is presented. A comprehensive examination of the domestic MOX experience in US reactors obtained during the 1960s, 1970s, and early 1980s is also presented. This experience is described by manufacturer and is also categorized by the reactor facility that irradiated the MOX fuel. A limited summary of the international experience with MOX fuels is also presented. A review of MOX fuel and its performance is conducted in view of the special considerations associated with the disposition of WG plutonium. Based on the available information, it appears that adoption of foreign commercial MOX technology from one of the successful MOX fuel vendors will minimize the technical risks to the overall mission. The conclusion is made that the existing MOX fuel experience base suggests that disposition of excess weapons plutonium through irradiation in LWRs is a technically attractive option.
Survey of Worldwide Light Water Reactor Experience with Mixed Uranium-Plutonium Oxide Fuel
International Nuclear Information System (INIS)
Cowell, B.S.; Fisher, S.E.
1999-01-01
The US and the Former Soviet Union (FSU) have recently declared quantities of weapons materials, including weapons-grade (WG) plutonium, excess to strategic requirements. One of the leading candidates for the disposition of excess WG plutonium is irradiation in light water reactors (LWRs) as mixed uranium-plutonium oxide (MOX) fuel. A description of the MOX fuel fabrication techniques in worldwide use is presented. A comprehensive examination of the domestic MOX experience in US reactors obtained during the 1960s, 1970s, and early 1980s is also presented. This experience is described by manufacturer and is also categorized by the reactor facility that irradiated the MOX fuel. A limited summary of the international experience with MOX fuels is also presented. A review of MOX fuel and its performance is conducted in view of the special considerations associated with the disposition of WG plutonium. Based on the available information, it appears that adoption of foreign commercial MOX technology from one of the successful MOX fuel vendors will minimize the technical risks to the overall mission. The conclusion is made that the existing MOX fuel experience base suggests that disposition of excess weapons plutonium through irradiation in LWRs is a technically attractive option
Shock and vibration tests of uranium mononitride fuel pellets for a space power nuclear reactor
Adams, D. W.
1972-01-01
Shock and vibration tests were conducted on cylindrically shaped, depleted, uranium mononitride (UN) fuel pellets. The structural capabilities of the pellets were determined under exposure to shock and vibration loading which a nuclear reactor may encounter during launching into space. Various combinations of diametral and axial clearances between the pellets and their enclosing structures were tested. The results of these tests indicate that for present fabrication of UN pellets, a diametral clearance of 0.254 millimeter and an axial clearance of 0.025 millimeter are tolerable when subjected to launch-induced loads.
Licensed fuel facility status report. Inventory difference data, January-June 1982
International Nuclear Information System (INIS)
1983-02-01
NRC is committed to the periodic publication of licensed fuel facilities inventory difference data, following agency review of the information and completion of any related investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, plutonium, or uranium-233
Studies involving direct heating of uranium and plutonium oxides by microwaves
Energy Technology Data Exchange (ETDEWEB)
Mallik, G K; Malav, R K; Karande, A P; Bhargava, V K; Kamath, H S [Bhabha Atomic Research Centre, Tarapur (India). Advanced Fuel Fabrication Facility
1997-08-01
Nuclear fuel fabrication and recovery of nuclear materials from scraps generated during fabrication involve heating steps like dewaxing, sintering, roasting of scraps, calcination, etc. The dielectric properties of uranium and plutonium oxides place them in the category of materials which are susceptible to absorption of microwaves. The studies were carried out to explore the microwave heating technique for these steps required in nuclear fuel fabrication and scrap recovery laboratories. (author). 1 ref.
Estimates of Canadian fuel fabrication costs for alternative fuel cycles and systems
International Nuclear Information System (INIS)
Blahnik, C.
1979-04-01
Unit fuel fabrication costs are estimated for alternate fuel cycles and systems that may be of interest in Ontario Hydro's strategy analyses. A method is proposed for deriving the unit fuel fabrication price to be paid by a Canadian utility as a function of time (i.e. the price that reflects the changing demand/supply situation in the particular scenario considered). (auth)
The cycle of the nuclear fuel used in EDF power plants
International Nuclear Information System (INIS)
2011-11-01
This document briefly indicates the different stages of the nuclear fuel cycle, from the purchase of natural uranium to waste storage. It also indicates the main responsibilities of EDF regarding this fuel cycle (to secure supplies, to organise material transportation, to process and store used fuels and associated wastes). It presents the different associated processes: uranium extraction, purification and concentration, conversion or fluoridation, enrichment. It briefly describes the fuel assembly fabrication, and indicates the main uranium producers in the world. Other addressed steps are: the transportation of fuel assembly, fuel loading, and spent fuel management, the processing of spent fuel and radioactive wastes
Experience in the development of metal uranium-base nuclear fuel for heavy-water gas-cooled reactors
International Nuclear Information System (INIS)
Ashikhmin, V.P.; Vorob'ev, M.A.; Gusarov, M.S.; Davidenko, A.S.; Zelenskij, V.F.; Ivanov, V.E.; Krasnorutskij, V.S.; Petel'guzov, I.A.; Stukalov, A.I.
1978-01-01
Investigations were carried out to solve the problem of making the development of radiation-resistant uranium fuel for power reactors including the heavy-water gas-cooled KS-150 reactor. Factors are considered that limit the lifetime of uranium fuel elements, and the ways of suppressing them are discussed. Possible reasons of the insufficient radiation resistance of uranium rod fuel element and the progress attained are analyzed. Some general problems on the fuel manufacture processes are discussed. The main results are presented on the operation of the developed fuel in research reactor loops and the commercial heavy-water KS-150 reactor. The results confirm an exceptionally high radiation resistance of fuel to burn-ups of 1.5-2%. The successful solution of a large number of problems associated with the development of metal uranium fuel provides for new possibilities of using metal uranium in power reactors
International Nuclear Information System (INIS)
Boczar, P.G.; Van Dyk, M.T.
1987-02-01
A new configuration of reactivity devices is proposed for future CANDU reactors which improves the core characteristics with enriched fuels, while still allowing the use of natural uranium fuel. Physics calculations for this new configuration are presented for four fuel types: natural uranium, mixed plutonium - uranium oxide (MOX) having a burnup of 21 MWd/kg, and slightly enriched uranium (SEU) having burnups of either 21 or 31 MWd/kg
Automated manufacturing of breeder reactor fuels
International Nuclear Information System (INIS)
Nyman, D.H.; Benson, E.M.; Bennett, D.W.
1983-09-01
The Secure Automated Fabrication (SAF) line is an automated, remotely controlled breeder fuel pin fabrication process which is to be installed in the Fuels and Materials Examination Facility (FMEF). The FMEF is presently under construction at Hanford and is scheduled for completion in 1984. The SAF line is scheduled for startup in 1987 and will produce mixed uranium-plutonium oxide fuel pins for the Fast Flux Test Facility (FFTF). Radiological protection requirements, computer control equipment, use of robotics, and the fabrication process is described
Analysis of the Reuse of Uranium Recovered from the Reprocessing of Commercial LWR Spent Fuel
Energy Technology Data Exchange (ETDEWEB)
DelCul, Guillermo Daniel [ORNL; Trowbridge, Lee D [ORNL; Renier, John-Paul [ORNL; Ellis, Ronald James [ORNL; Williams, Kent Alan [ORNL; Spencer, Barry B [ORNL; Collins, Emory D [ORNL
2009-02-01
This report provides an analysis of the factors involved in the reuse of uranium recovered from commercial light-water-reactor (LWR) spent fuels (1) by reenrichment and recycling as fuel to LWRs and/or (2) by recycling directly as fuel to heavy-water-reactors (HWRs), such as the CANDU (registered trade name for the Canadian Deuterium Uranium Reactor). Reuse is an attractive alternative to the current Advanced Fuel Cycle Initiative (AFCI) Global Nuclear Energy Partnership (GNEP) baseline plan, which stores the reprocessed uranium (RU) for an uncertain future or attempts to dispose of it as 'greater-than-Class C' waste. Considering that the open fuel cycle currently deployed in the United States already creates a huge excess quantity of depleted uranium, the closed fuel cycle should enable the recycle of the major components of spent fuel, such as the uranium and the hazardous, long-lived transuranic (TRU) actinides, as well as the managed disposal of fission product wastes. Compared with the GNEP baseline scenario, the reuse of RU in the uranium fuel cycle has a number of potential advantages: (1) avoidance of purchase costs of 11-20% of the natural uranium feed; (2) avoidance of disposal costs for a large majority of the volume of spent fuel that is reprocessed; (3) avoidance of disposal costs for a portion of the depleted uranium from the enrichment step; (4) depending on the {sup 235}U assay of the RU, possible avoidance of separative work costs; and (5) a significant increase in the production of {sup 238}Pu due to the presence of {sup 236}U, which benefits somewhat the transmutation value of the plutonium and also provides some proliferation resistance.
International Nuclear Information System (INIS)
Muhammad, Farhan; Majid, Asad
2009-01-01
The effects of using different low enriched uranium fuels, having same uranium density, on the kinetic parameters of a material test research reactor were studied. For this purpose, the original aluminide fuel (UAl x -Al) containing 4.40 gU/cm 3 of an MTR was replaced with silicide (U 3 Si-Al and U 3 Si 2 -Al) and oxide (U 3 O 8 -Al) dispersion fuels having the same uranium density as of the original fuel. Simulations were carried out to calculate prompt neutron generation time, effective delayed-neutron fraction, core excess reactivity and neutron flux spectrum. Nuclear reactor analysis codes including WIMS-D4 and CITATION were used to carry out these calculations. It was observed that both the silicide fuels had the same prompt neutron generation time 0.02% more than that of the original aluminide fuel, while the oxide fuel had a prompt neutron generation time 0.05% less than that of the original aluminide fuel. The effective delayed-neutron fraction decreased for all the fuels; the decrease was maximum at 0.06% for U 3 Si 2 -Al followed by 0.03% for U 3 Si-Al, and 0.01% for U 3 O 8 -Al fuel. The U 3 O 8 -Al fueled reactor gave the maximum ρ excess at BOL which was 21.67% more than the original fuel followed by U 3 Si-Al which was 2.55% more, while that of U 3 Si 2 -Al was 2.50% more than the original UAl x -Al fuel. The neutron flux of all the fuels was more thermalized, than in the original fuel, in the active fuel region of the core. The thermalization was maximum for U 3 O 8 -Al followed by U 3 Si-Al and then U 3 Si 2 -Al fuel.
Status of SFR Metal Fuel Development
International Nuclear Information System (INIS)
Lee, Chan Bock; Lee, Byoung Oon; Kim, Ki Hwan; Kim, Sung Ho
2013-01-01
Conclusion: • Metal fuel recycling in SFR: - Enhanced utilization of uranium resource; - Efficient transmutation of minor actinides; - Inherent passive reactor safety; - Proliferation resistance with pyro-electrochemical fuel recycling. • Demonstration of technical feasibility of recycling TRU metal fuel by 2020: - Remote fuel fabrication; - Irradiation performance up to high burnup
Fuel Fabrication Capability Research and Development Plan
Energy Technology Data Exchange (ETDEWEB)
Senor, David J.; Burkes, Douglas
2014-04-17
The purpose of this document is to provide a comprehensive review of the mission of the Fuel Fabrication Capability (FFC) within the Global Threat Reduction Initiative Convert Program, along with research and development (R&D) needs that have been identified as necessary to ensuring mission success. The design and fabrication of successful nuclear fuels must be closely linked endeavors. Therefore, the overriding motivation behind the FFC R&D program described in this plan is to foster closer integration between fuel design and fabrication to reduce programmatic risk. These motivating factors are all interrelated, and progress addressing one will aid understanding of the others. The FFC R&D needs fall into two principal categories, 1) baseline process optimization, to refine the existing fabrication technologies, and 2) manufacturing process alternatives, to evaluate new fabrication technologies that could provide improvements in quality, repeatability, material utilization, or cost. The FFC R&D Plan examines efforts currently under way in regard to coupon, foil, plate, and fuel element manufacturing, and provides recommendations for a number of R&D topics that are of high priority but not currently funded (i.e., knowledge gaps). The plan ties all FFC R&D efforts into a unified vision that supports the overall Convert Program schedule in general, and the fabrication schedule leading up to the MP-1 and FSP-1 irradiation experiments specifically. The fabrication technology decision gates and down-selection logic and schedules are tied to the schedule for fabricating the MP-1 fuel plates, which will provide the necessary data to make a final fuel fabrication process down-selection. Because of the short turnaround between MP-1 and the follow-on FSP-1 and MP-2 experiments, the suite of specimen types that will be available for MP-1 will be the same as those available for FSP-1 and MP-2. Therefore, the only opportunity to explore parameter space and alternative processing
Development of a recovery process of scraps resulting from the manufacture of metallic uranium fuels
International Nuclear Information System (INIS)
Camilo, Ruth L.; Kuada, Terezinha A.; Forbicini, Christina A.L.G.O.; Cohen, Victor H.; Araujo, Bertha F.; Lobao, Afonso S.T.
1996-01-01
The study of the dissolution of natural metallic uranium fuel samples with aluminium cladding is presented, in order to obtain optimized conditions for the system. The aluminium cladding was dissolved in an alkaline solution of Na OH/Na NO 3 and the metallic uranium with HNO 3 . A fumeless dissolution with total recovery of nitrous gases was achieved. The main purpose of this project was the recovery of uranium from scraps resulting from the manufacture of the metallic uranium fuel or other non specified fuels. (author)
International Nuclear Information System (INIS)
Ito, Mitsuo; Iso, Shuichi; Hoshino, Akira; Suzuki, Shuichi.
1992-03-01
Potassium fluoride-gravimetric method has been developed for the determination of uranium in TRISO type-coated fuel particle compact. Graphite matrix in the fuel compact is burned off by heating it in a platinum crucible at 850degC. The coated fuel particles thus obtained are decomposed by fusion with potassium fluoride at 900degC. The melt was dissolved with sulfuric acid. Uranium is precipitated as ammonium diuranate, by passing ammonia gas through the solution. The resulting precipitate is heated in a muffle furnace at 850degC, to convert uranium into triuranium octoxide. Uranium in the triuranium octoxide was determined gravimetrically. Ten grams of caoted fuel particles were completely decomposed by fusion with 50 g of potassium fluoride at 900degC for 3 hrs. Analytical result for uranium in the fuel compact by the proposed method was 21.04 ± 0.05 g (n = 3), and was in good agreement with that obtained by non-destructive γ-ray measurement method : 21.01 ± 0.07 g (n = 3). (author)
International Nuclear Information System (INIS)
1988-06-01
With the transition to high density uranium LEU fuel, fabrication costs of research reactor fuel elements have a tendency to increase because of two reasons. First, the amount of the powder of the uranium compound required increases by more than a factor of five. Second, fabrication requirements are in many cases nearer the fabrication limits. Therefore, it is important that measures be undertaken to eliminate or reduce unnecessary requirements in the specification or inspection procedures of research reactor fuel elements utilizing LEU. An additional stimulus for standardizing specifications and inspection procedures at this time is provided by the fact that most LEU conversions will occur within a short time span, and that nearly all of them will require preparation of new specifications and inspection procedures. In this sense, the LEU conversions offer an opportunity for improving the rationality and efficiency of the fuel fabrication and inspection processes. This report focuses on the standardization of specifications and inspection processes of high uranium density LEU fuels for research reactors. However, in many cases the results can also be extended directly to other research reactor fuels. 15 refs, 1 fig., 3 tabs
Kinetic and thermodynamic bases to resolve issues regarding conditioning of uranium metal fuels
International Nuclear Information System (INIS)
Johnson, A.B.; Ballinger, R.G.; Simpson, K.A.
1994-12-01
Numerous uranium - bearing fuels are corroding in fuel storage pools in several countries. At facilities where reprocessing is no longer available, dry storage is being evaluated to preclude aqueous corrosion that is ongoing. It is essential that thermodynamic and kinetic factors are accounted for in transitions of corroding uranium-bearing fuels to dry storage. This paper addresses a process that has been proposed to move Hanford N-Reactor fuel from wet storage to dry storage
Uranium-236 in light water reactor spent fuel recycled to an enriching plant
International Nuclear Information System (INIS)
de la Garza, A.
1977-01-01
The introduction of 236 U to an enriching plant by recycling spent fuel uranium results in enriched products containing 236 U, a parasitic neutron absorber in reactor fuel. Convenient approximate methodology determines 235 236 U, and total uranium flowsheets with associated separative work requirements in enriching plant operations for use by investigators of the light water reactor fuel cycle not having recourse to specialized multicomponent cascade technology. Application of the methodology has been made to compensation of an enriching plant product for 236 U content and to the value at an enriching plant of spent fuel uranium. The approximate methodology was also confirmed with more exact calculations and with some experience with 236 U in an enriching plant
Occupational exposures to uranium: processes, hazards, and regulations
International Nuclear Information System (INIS)
Stoetzel, G.A.; Fisher, D.R.; McCormack, W.D.; Hoenes, G.R.; Marks, S.; Moore, R.H.; Quilici, D.G.; Breitenstein, B.D.
1981-04-01
The United States Uranium Registry (USUR) was formed in 1978 to investigate potential hazards from occupational exposure to uranium and to assess the need for special health-related studies of uranium workers. This report provides a summary of Registry work done to date. The history of the uranium industry is outlined first, and the current commercial uranium industry (mining, milling, conversion, enrichment, and fuel fabrication) is described. This description includes information on basic processes and areas of greatest potential radiological exposure. In addition, inactive commercial facilities and other uranium operations are discussed. Regulation of the commercial production industry for uranium fuel is reported, including the historic development of regulations and the current regulatory agencies and procedures for each phase of the industry. A review of radiological health practices in the industry - facility monitoring, exposure control, exposure evaluation, and record-keeping - is presented. A discussion of the nonradiological hazards of the industry is provided, and the final section describes the tissue program developed as part of the Registry
International Nuclear Information System (INIS)
Marin, J.-F.; Pillet, Claude; Francois, Bernard; Morize, Pierre; Petitgrand, Sylvie; Atabek, R.-M.; Houdaille, Brigitte.
1981-06-01
Fabrication of fuel pins made of uranium oxide pellets and of a zircaloy 4 cladding is described. Irradiation experiment results are given. Thermomechanical behavior of the fuel pin in a power reactor is examined [fr
MOX fuel fabrication, in reactor performance and improvement
International Nuclear Information System (INIS)
Vliet, J. van; Deramaix, P.; Nigon, J.L.; Fournier, W.
1998-01-01
In Europe, MOX fuel for light water reactors (LWRs) has first been manufactured in Belgium and Germany. Belgonucleaire (BN) loaded the first MOX assembly in the BR3 Pressurised Water Reactor (PWR) in 1963. In June 1998, more than 750 tHM LWR MOX fuel assemblies were manufactured on a industrial scale in Europe without any particular difficulty relating to fuel fabrication, reactor operation or fuel behaviour. So, today plutonium recycling through MOX fuel is a mature industry, with successful operational experience and large-scale fabrication plants. In this field, COGEMA and BELGONUCLEAIRE are the main actors by operating simultaneously three complete multidesign fuel production plants: MELOX plant (in Marcoule), CADARACHE plant and P0 plant (in Dessel, Belgium). Present MOX production capacity available to COGEMA and BN fits 175 tHM per year and is to be extended to reach about 325 tHM in the year 2000. This will represent 75% of the total MOX fabrication capacity in Europe. The industrial mastery and the high production level in MOX fabrication assured by high technology processes confer to these companies a large expertise for Pu recycling. This allows COGEMA and BN to be major actors in Pu-based fuels in the coming second nuclear era with advanced fuel cycles. (author)
Possibilities of using metal uranium fuel in heavy water reactors
International Nuclear Information System (INIS)
Djuric, B.; Mihajlovic, A.; Drobnjak, Dj.
1965-11-01
There are serious economic reasons for using metal uranium in heavy water reactors, because of its high density, i.e. high conversion factor, and low cost of fuel elements production. Most important disadvantages are swelling at high burnup and corrosion risk. Some design concepts and application of improved uranium obtained by alloying are promising for achievement of satisfactory stability of metal uranium under reactor operation conditions [sr
Equipment specifications for an electrochemical fuel reprocessing plant
International Nuclear Information System (INIS)
Hemphill, Kevin P.
2010-01-01
Electrochemical reprocessing is a technique used to chemically separate and dissolve the components of spent nuclear fuel, in order to produce new metal fuel. There are several different variations to electrochemical reprocessing. These variations are accounted for by both the production of different types of spent nuclear fuel, as well as different states and organizations doing research in the field. For this electrochemical reprocessing plant, the spent fuel will be in the metallurgical form, a product of fast breeder reactors, which are used in many nuclear power plants. The equipment line for this process is divided into two main categories, the fuel refining equipment and the fuel fabrication equipment. The fuel refining equipment is responsible for separating out the plutonium and uranium together, while getting rid of the minor transuranic elements and fission products. The fuel fabrication equipment will then convert this plutonium and uranium mixture into readily usable metal fuel.
Corrosion testing of uranium silicide fuel specimens
International Nuclear Information System (INIS)
Bourns, W.T.
1968-09-01
U 3 Si is the most promising high density natural uranium fuel for water-cooled power reactors. Power reactors fuelled with this material are expected to produce cheaper electricity than those fuelled with uranium dioxide. Corrosion tests in 300 o C water preceded extensive in-reactor performance tests of fuel elements and bundles. Proper heat-treatment of U-3.9 wt% Si gives a U 3 5i specimen which corrodes at less than 2 mg/cm 2 h in 300 o C water. This is an order of magnitude lower than the maximum corrosion rate tolerable in a water-cooled reactor. U 3 Si in a defected unbonded Zircaloy-2 sheath showed only a slow uniform sheath expansion in 300 o C water. All tests were done under isothermal conditions in an out-reactor loop. (author)
Licensed-fuel-facility status report: inventory difference data, July 1981-December 1981
International Nuclear Information System (INIS)
1982-10-01
NRC is committed to the periodic publication of licensed fuel facilities inventory difference data, following agency review of the information and completion of any related investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, plutonium, or uranium-233
Licensed fuel facility status report: Inventory difference data, January 1986-June 1986
International Nuclear Information System (INIS)
1987-02-01
NRC is committed to the periodic publication of licensed fuel facilities' inventory difference data, following agency review of the information and completion of any related investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, plutonium, or uranium-233
Licensed fuel facility status report: Inventory difference data, January 1987-June 1987
International Nuclear Information System (INIS)
1988-03-01
NRC is committed to the periodic publication of licensed fuel facilities' inventory difference data, following agency review of the information and completion of any related investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, plutonium, or uranium-233
Licensed fuel facility status report: Inventory difference data, July 1987-December 1987
International Nuclear Information System (INIS)
1988-09-01
NRC is committed to the periodic publication of licensed fuel facilities' inventory difference data, following agency review of the information and completion of any related investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, plutonium, or uranium-233
International Nuclear Information System (INIS)
Bricout, J.
2012-01-01
Powder Injection Moulding (PIM) is identified as an innovative process for the nuclear fuel fabrication. Technological breakthrough compared to the current process of powder metallurgy, the impact of actinide powder's specificities on the different steps of PIM is performed. Alumina powders simulating actinide powder have been implemented with a reference binders system. Thermal and rheological studies show the injectability and the de-binding of feedstocks with adequate solid loading (≥50 %vol), thanks to the de-agglomeration during the mixing step, which allow to obtain net shape fuel pellet. Specific surface area of powders, acting as a key role in behaviour's feedstocks, has been integrated in analysis models of viscosity prediction according to the shear rate. Also conducted studies on uranium oxide powder show that the selected binders systems, which have a compatible rheological behaviour with PIM process, impact the de-agglomeration of powder and final microstructure of the fuel pellet, consistent with the results obtained on alumina powders. Independent behaviour of binders and uranium oxide powder, showing no adverse chemical reaction against the PIM process, show a residual mass of carbon of about 150 ppm after sintering. Binders system using polystyrene, resistant to radiolysis phenomena and loadable more than 50 %(vol) of actinide powder, shows the promising potential of PIM process for the fuel fabrication. (author) [fr
Detailed analysis of uranium silicide dispersion fuel swelling
International Nuclear Information System (INIS)
Hofmann, G.L.; Ryu, Woo-Seog
1991-01-01
Swelling of U 3 Si and U 3 Si 2 is analyzed. The growth of fission gas bubbles appears to be affected by fission rate, fuel loading, and micro structural change taking place in the fuel compounds during irradiation. Several mechanisms are explored to explain the observations. The present work is aimed at a better understanding of the basic swelling phenomenon in order to accurately model irradiation behavior of uranium silicide dispersion fuel. (orig.)
Feasibility of Low Enriched Uranium Fuel for Space Nuclear Propulsion
Energy Technology Data Exchange (ETDEWEB)
Venneri, Paolo; Kim, Yonghee [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)
2013-05-15
The purpose of this initial study is to create a baseline with which to perform further analysis and to build a solid understanding of the neutronic characteristics of a solid core for the nuclear thermal rocket. Once consistency with work done at Idaho National Laboratory (INL) is established, this paper will provide a study of other fuel types, such as low and medium-enriched uranium fuels. This paper will examine how the implementation of each fuel type affects the multiplication factor of the reactor, and will then explore different possibilities for alterations needed to accommodate their successful usage. The reactor core analysis was done using the MCNP5 code. While this study has not shown that the SNRE can be easily retrofitted for low-enriched U fuel, it has made a detailed study of the SNRE, and identified the difficulties of the implementation of low-enriched fuels in small nuclear rockets. These difficulties are the need for additional moderation and fuel mass in order to achieve a critical mass. Neither of these is insurmountable. Future work includes finding the best method by which to increase the internal moderation of the reactor balanced with appropriate sizing to prevent neutron leakage. Both of these are currently being studied. This paper will present a study of the Small Nuclear Rocket Engine (SNRE) and the feasibility of using low enriched Uranium (LEU) instead of the traditional high enriched Uranium (HEU) fuels.
Development of PHWR fuel fabrication in Korea
International Nuclear Information System (INIS)
Suh, K.S.; Yang, M.S.; Kim, D.H.; Rim, C.S.
1988-01-01
Korea Advanced Energy Research Institute (KAERI) started a research project to develop the PHWR (CANDU) nuclear fuel fabrication technology in 1981. Based on the results of the intensive developmental work, several prototype fuel bundles were fabricated and tested in the Hot Test Loop at KAERI continuously in 1983 and 1984. After that, irradiation test and post-irradiation examination were carried out for two KAERI-made fuel bundles at Chalk River Nuclear Laboratories in Canada in 1984. Since the results of in-pile and out-of-pile tests with prototype fuel bundles proved to be satisfactory, 48 additional fuel bundles were loaded in Wolsung reactor (CANDU) in 1984 and 1985, and all of them were discharged without a defect after excellent performance in the power reactor. In 1985, the Korean government decided that KAERI supplies all the fuel necessary for the Wolsung reactor. For the mass production of nuclear fuel bundle, several process equipment, facilities and automation methods have been improved making use of experience accumulated during research. A quality assurance program was also established, and quality inspection technology was reviewed and improved to fit the mass production. This paper deals with the development experience so far obtained with the design and fabrication of the Korean PHWR fuel
Hot pressing of uranium nitride and mixed uranium plutonium nitride
International Nuclear Information System (INIS)
Chang, J.Y.
1975-01-01
The hot pressing characteristics of uranium nitride and mixed uranium plutonium nitride were studied. The utilization of computer programs together with the experimental technique developed in the present study may serve as a useful purpose of prediction and fabrication of advanced reactor fuel and other high temperature ceramic materials for the future. The densification of nitrides follow closely with a plastic flow theory expressed as: d rho/ dt = A/T(t) (1-rho) [1/1-(1-rho)/sup 2/3/ + B1n (1-rho)] The coefficients, A and B, were obtained from experiment and computer curve fitting. (8 figures) (U.S.)
International Nuclear Information System (INIS)
Anon.
2010-01-01
In december 2010 IAEA gave its agreement for the creation of a nuclear fuel bank. This bank will allow IAEA to help member countries that renounce to their own uranium enrichment capacities. This bank located on one or several member countries will belong to IAEA and will be managed by IAEA and its reserve of low enriched uranium will be sufficient to fabricate the fuel for the first load of a 1000 MW PWR. Fund raising has been successful and the running of the bank will have no financial impact on the regular budget of the IAEA. Russia has announced the creation of the first nuclear fuel bank. This bank will be located on the Angarsk site (Siberia) and will be managed by IAEA and will own 120 tonnes of low-enriched uranium fuel (between 2 and 4.95%), this kind of fuel is used in most Russian nuclear power plants. (A.C.)
Behavior of silicon in nitric media. Application to uranium silicides fuels reprocessing
International Nuclear Information System (INIS)
Cheroux, L.
2001-01-01
Uranium silicides are used in some research reactors. Reprocessing them is a solution for their cycle end. A list of reprocessing scenarios has been set the most realistic being a nitric dissolution close to the classic spent fuel reprocessing. This uranium silicide fuel contains a lot of silicon and few things are known about polymerization of silicic acid in concentrated nitric acid. The study of this polymerization allows to point out the main parameters: acidity, temperature, silicon concentration. The presence of aluminum seems to speed up heavily the polymerization. It has been impossible to find an analytical technique smart and fast enough to characterize the first steps of silicic acid polymerization. However the action of silicic species on emulsions stabilization formed by mixing them with an organic phase containing TBP has been studied, Silicon slows down the phase separation by means of oligomeric species forming complex with TBP. The existence of these intermediate species is short and heating can avoid any stabilization. When non irradiated uranium silicide fuel is attacked by a nitric solution, aluminum and uranium are quickly dissolved whereas silicon mainly stands in solid state. That builds a gangue of hydrated silica around the uranium silicide particulates without preventing uranium dissolution. A small part of silicon passes into the solution and polymerize towards the highly poly-condensed forms, just 2% of initial silicon is still in molecular form at the end of the dissolution. A thermal treatment of the fuel element, by forming inter-metallic phases U-Al-Si, allows the whole silicon to pass into the solution and next to precipitate. The behavior of silicon in spent fuels should be between these two situations. (author)
Development of nuclear fuel. Development of CANDU advanced fuel bundle
International Nuclear Information System (INIS)
Suk, Ho Chun; Hwang, Woan; Jeong, Young Hwan; Jung, Sung Hoon
1991-07-01
In order to develop CANDU advanced fuel, the agreement of the joint research between KAERI and AECL was made on February 19, 1991. AECL conceptual design of CANFLEX bundle for Bruce reactors was analyzed and then the reference design and design drawing of the advanced fuel bundle with natural uranium fuel for CANDU-6 reactor were completed. The CANFLEX fuel cladding was preliminarily investigated. The fabricability of the advanced fuel bundle was investigated. The design and purchase of the machinery tools for the bundle fabrication for hydraulic scoping tests were performed. As a result of CANFLEX tube examination, the tubes were found to be meet the criteria proposed in the technical specification. The dummy bundles for hydraulic scoping tests have been fabricated by using the process and tools, where the process parameters and tools have been newly established. (Author)
International Nuclear Information System (INIS)
Kurillovich, A.N.; Pimonov, Yu.I.; Biryukov, A.S.
1988-01-01
A technique for mass-spectroimetric determination of moisture content in fuel elements and fuek claddings in the 2x10 -4 -1.5x10 -2 g range is developed. The relative standard deviation is 0.13. A character of moisture extraction from oxide uranium fuels in the 20-700 deg C temperature range is studied. Approximately 80% of moisture is extracted from the fuels at 300 deg C. The moisture content in fuel elements with granular uranium oxide fuels is measured. Dependence of fuel element moisture content on conditions of hot vacuum drying is shown. The technique permits to optimize the fuel element fabrication process to decrease the moisture content in them. 4 refs.; 3 figs.; 2 tabs
Y-12 product improvements expected to reduce metal production costs and decrease fabrication losses
International Nuclear Information System (INIS)
Parker, Elaine; Hassler, Morris
2004-01-01
Full text: The Y-12 National Security Complex supplies uranium metal and uranium oxide feed material that is then fabricated into fuel for research reactors around the world. Over the past two to three years, Y-12 has learned a great deal about its Low Enriched Uranium (LEU) product. The LEU is produced by taking U.S. surplus Highly Enriched Uranium (HEU) and blending it with depleted or natural uranium. The surplus HEU comes from dismantled U.S. weapons parts that have been declared as surplus. Those research reactors that use LEU from Y-12 are making important contributions to international nuclear non-proliferation by using LEU rather than HEU, and by helping to disposition former weapons material. We clearly understand that our customers want to keep fuel costs as low as possible. We at Y-12 are making every effort to improve efficiencies in producing the uranium through standardizing the chemical specifications as well as the product mass and dimensional qualities. This paper will discuss the new standard specification that we have proposed to existing LEU metal customers and fuel fabricators. It will also cover Y-12's progress on a new mold-design that will result in a more uniform, higher quality product that is less expensive to produce. This new product is expected to decrease overall fabrication losses by 5-10%, depending on the fabricator's process. The paper will include planned activities and the schedule associated with implementation of the new specification and product form. (author)
Y-12 product improvements expected to reduce metal production costs and decrease fabrication losses
International Nuclear Information System (INIS)
Hassler, Morris E.
2005-01-01
The Y-12 National Security Complex (Y-12) supplies uranium metal and uranium oxide feed material for fabrication into fuel for research reactors around the world. Over the past few years, Y-12 has continued to improve its Low Enriched Uranium (LEU) product. The LEU is produced by taking U.S. surplus Highly Enriched Uranium (HEU) and blending it with depleted or natural uranium. The surplus HEU comes from dismantled U.S. weapons parts. Those research reactors that use LEU from Y-12 are making important contributions to international nuclear nonproliferation by using LEU rather than HEU, and helping to disposition former U.S. weapons material. It is clearly understood that the research reactor community must keep fuel costs as low as possible and Y-12 is making every effort to improve efficiencies in producing the uranium through standardizing the chemical specifications as well as the product mass and dimensional qualities. These production cost reductions allows for the U.S. to keep the LEU product price low even with the dramatic increase in the uranium enrichment and feed component market prices in the last few years. This paper will discuss a new standard specification that has been proposed to existing LEU metal customers and fuel fabricators. It will also cover Y-12's progress on a new mold-design that will result in a more uniform, higher quality product and eliminates two steps of the production process. This new product is expected to decrease fabrication losses by 5-10%, depending on the fabricator's process. The paper will include planned activities and the schedule associated with implementation of the new specification and product form. (author)
International Nuclear Information System (INIS)
Canck, H. de; Ingels, R.; Lefevre, R.
1991-01-01
In a modern MOX Fuel Fabrication Plant, with a large throughput of nuclear materials, computerized real-time accountancy systems are applied. Following regulations and prescriptions imposed by the Inspectorates EURATOM-IAEA, the State and also by internal plant safety rules, the accountancy is kept in plutonium element, uranium element and 235 U for enriched uranium. In practice, Safeguards Authorities are concerned with quantities of the element (U tot , Pu tot ) and to some extent with its fissile content. Custom Authorities are for historical reasons, interested in fissile quantities (U fiss , Pu fiss ) whereas owners wish to recover the energetic value of their material (Pu equivalent). Balancing the accountancy simultaneously in all these related but not proportional units is a new problem in a MOX-plant where pool accountancy is applied. This paper indicates possible ways to solve the balancing problem created by these different units used for expressing nuclear material quantities
Fuel Fabrication Capability Research and Development Plan
Energy Technology Data Exchange (ETDEWEB)
Senor, David J.; Burkes, Douglas
2013-06-28
The purpose of this document is to provide a comprehensive review of the mission of the Fuel Fabrication Capability (FFC) within the Global Threat Reduction Initiative (GTRI) Convert Program, along with research and development (R&D) needs that have been identified as necessary to ensuring mission success. The design and fabrication of successful nuclear fuels must be closely linked endeavors.
Design of the MOX fuel fabrication facility
International Nuclear Information System (INIS)
Johnson, J.V.; Brabazon, E.J.
2001-01-01
A consortium of Duke Engineering and Services, Inc., COGEMA, Inc. and Stone and Webster (DCS) are designing a mixed oxide fuel fabrication facility (MFFF) for the U.S. Department of Energy (DOE) to convert surplus plutonium to mixed oxide (MOX) fuel to be irradiated in commercial nuclear power plants based on the proven European technology of COGEMA and BELGONUCLEAIRE. This paper describes the MFFF processes, and how the proven MOX fuel fabrication technology is being adapted as required to comply with U.S. requirements. (author)
Design of the MOX fuel fabrication facility
Energy Technology Data Exchange (ETDEWEB)
Johnson, J.V. [MFFF Technical Manager, U.S. dept. of Energy, Washington, DC (United States); Brabazon, E.J. [MFFF Engineering Manager, Duke Cogema Stone and Webster, Charlotte, NC (United States)
2001-07-01
A consortium of Duke Engineering and Services, Inc., COGEMA, Inc. and Stone and Webster (DCS) are designing a mixed oxide fuel fabrication facility (MFFF) for the U.S. Department of Energy (DOE) to convert surplus plutonium to mixed oxide (MOX) fuel to be irradiated in commercial nuclear power plants based on the proven European technology of COGEMA and BELGONUCLEAIRE. This paper describes the MFFF processes, and how the proven MOX fuel fabrication technology is being adapted as required to comply with U.S. requirements. (author)
Detailed analysis of uranium silicide dispersion fuel swelling
International Nuclear Information System (INIS)
Hofman, G.L.; Ryu, Woo-Seog.
1989-01-01
Swelling of U 3 Si and U 3 Si 2 is analyzed. The growth of fission gas bubbles appears to be affected by fission rate, fuel loading, and microstructural change taking place in the fuel compounds during irradiation. Several mechanisms are explored to explain the observations. The present work is aimed at a better understanding of the basic swelling phenomenon in order to accurately model irradiation behavior of uranium silicide disperson fuel. 5 refs., 10 figs
TRIGA International, a new TRIGA fuel fabrication facility at CERCA
International Nuclear Information System (INIS)
Harbonnier, G.
1997-01-01
At the time when General Atomics expressed its intention to cease fuel fabrication on its site of San Diego, CERCA has been chosen to carry on the fabrication of TRIGA fuel. After negotiations in 1994 and 1995, a partnership 50%/50% was decided and on July 1995, a new company was founded, with the name TRIGA INTERNATIONAL SAS, head office in Paris and fuel fabrication facility at CERCA in Romans. The intent of this presentation is, after a short reminder about TRIGA fuel design and fabrication to describe the new facility with special emphasis on the safety features associated with the modification of existing fabrication buildings. (author)
Highlights of 50 years of nuclear fuels developments
International Nuclear Information System (INIS)
Simnad, M.T.
1989-01-01
The development of nuclear fuels since the discovery of nuclear fission is briefly surveyed in this paper. The fabrication of the uranium fuel for the first nuclear pile, CP-1, is described. The research and development studies and fabrication of the different types of nuclear fuels for the variety of research and power reactors are reviewed. The important factors involved to achieve low fuel cycle costs and reliable performance in the fuel elements are discussed in the historical context
Basic research and industrialization of CANDU advanced fuel
Energy Technology Data Exchange (ETDEWEB)
Chun, Suk Ho; Park, Joo Hwan; Jun, Ji Su [and others
2000-04-01
Wolsong Unit 1 as the first heavy water reactor in Korea has been in service for 17 years since 1983. It would be about the time to prepare a plan for the solution of problems due to aging of the reactor. The aging of CANDU reactor could lead especially to the steam generator cruding and pressure tube sagging and creep and then decreases the operation margin to make some problems on reactor operations and safety. The counterplan could be made in two ways. One is to repair or modify reactor itself. The other is to develop new advanced fuel to increase of CANDU operation margin effectively, so as to compensate the reduced operation margin. Therefore, the first objectives in the present R and D is to develop the CANFLEX-NU (CANDU Flexible fuelling-Natural Uranium) fuel as a CANDU advanced fuel. The second objectives is to develop CANDU advanced fuel bundle to utilize advanced fuel cycles such as recovered uranium, slightly enriched uranium, etc. and so to raise adaptability for change in situation of uranium market. Also, it is to develop CANDU advanced fuel technology which improve uranium utilization to cope with a world-wide imbalance between uranium supply and demand, without significant modification of nuclear reactor design and refuelling strategies. As the implementations to achieve the above R and D goal, the work contents and scope of technology development of CANDU advanced fuel using natural uranium (CANFLEX-NU) are the fuel element/bundle designs, the nuclear design and fuel management analysis, the thermalhydraulic analysis, the safety analysis, fuel fabrication technologies, the out-pile thermalhydraulic test and in-pile irradiation tests performed. At the next, the work scopes and contents of feasibility study of CANDU advanced fuel using recycled uranium (CANFLEX-RU) are the fuel element/bundle designs, the reactor physics analysis, the thermalhydraulic analysis, the basic safety analysis of a CANDU-6 reactor with CANFLEX-RU fuel, the fabrication and
Basic research and industrialization of CANDU advanced fuel
International Nuclear Information System (INIS)
Chun, Suk Ho; Park, Joo Hwan; Jun, Ji Su
2000-04-01
Wolsong Unit 1 as the first heavy water reactor in Korea has been in service for 17 years since 1983. It would be about the time to prepare a plan for the solution of problems due to aging of the reactor. The aging of CANDU reactor could lead especially to the steam generator cruding and pressure tube sagging and creep and then decreases the operation margin to make some problems on reactor operations and safety. The counterplan could be made in two ways. One is to repair or modify reactor itself. The other is to develop new advanced fuel to increase of CANDU operation margin effectively, so as to compensate the reduced operation margin. Therefore, the first objectives in the present R and D is to develop the CANFLEX-NU (CANDU Flexible fuelling-Natural Uranium) fuel as a CANDU advanced fuel. The second objectives is to develop CANDU advanced fuel bundle to utilize advanced fuel cycles such as recovered uranium, slightly enriched uranium, etc. and so to raise adaptability for change in situation of uranium market. Also, it is to develop CANDU advanced fuel technology which improve uranium utilization to cope with a world-wide imbalance between uranium supply and demand, without significant modification of nuclear reactor design and refuelling strategies. As the implementations to achieve the above R and D goal, the work contents and scope of technology development of CANDU advanced fuel using natural uranium (CANFLEX-NU) are the fuel element/bundle designs, the nuclear design and fuel management analysis, the thermalhydraulic analysis, the safety analysis, fuel fabrication technologies, the out-pile thermalhydraulic test and in-pile irradiation tests performed. At the next, the work scopes and contents of feasibility study of CANDU advanced fuel using recycled uranium (CANFLEX-RU) are the fuel element/bundle designs, the reactor physics analysis, the thermalhydraulic analysis, the basic safety analysis of a CANDU-6 reactor with CANFLEX-RU fuel, the fabrication and
Fuel loading and homogeneity analysis of HFIR design fuel plates loaded with uranium silicide fuel
International Nuclear Information System (INIS)
Blumenfeld, P.E.
1995-08-01
Twelve nuclear reactor fuel plates were analyzed for fuel loading and fuel loading homogeneity by measuring the attenuation of a collimated X-ray beam as it passed through the plates. The plates were identical to those used by the High Flux Isotope Reactor (HFIR) but were loaded with uranium silicide rather than with HFIR's uranium oxide fuel. Systematic deviations from nominal fuel loading were observed as higher loading near the center of the plates and underloading near the radial edges. These deviations were within those allowed by HFIR specifications. The report begins with a brief background on the thermal-hydraulic uncertainty analysis for the Advanced Neutron Source (ANS) Reactor that motivated a statistical description of fuel loading and homogeneity. The body of the report addresses the homogeneity measurement techniques employed, the numerical correction required to account for a difference in fuel types, and the statistical analysis of the resulting data. This statistical analysis pertains to local variation in fuel loading, as well as to ''hot segment'' analysis of narrow axial regions along the plate and ''hot streak'' analysis, the cumulative effect of hot segment loading variation. The data for all twelve plates were compiled and divided into 20 regions for analysis, with each region represented by a mean and a standard deviation to report percent deviation from nominal fuel loading. The central regions of the plates showed mean values of about +3% deviation, while the edge regions showed mean values of about -7% deviation. The data within these regions roughly approximated random samplings from normal distributions, although the chi-square (χ 2 ) test for goodness of fit to normal distributions was not satisfied
WWER-1000 nuclear fuel manufacturing process at PJSC MSZ
International Nuclear Information System (INIS)
Morylev, A.; Bagdatyeva, E.; Aksenov, P.
2015-01-01
In this report a brief description of WWER-1000 fuel manufacturing process steps at PJSC MSZ as: uranium dioxide powder fabrication; fuel pellet manufacture fuel rod manufacture working assembly and fuel assembly manufacture is given. The implemented innovations are also presented
Corrosion testing of uranium silicide fuel specimens
Energy Technology Data Exchange (ETDEWEB)
Bourns, W T
1968-09-15
U{sub 3}Si is the most promising high density natural uranium fuel for water-cooled power reactors. Power reactors fuelled with this material are expected to produce cheaper electricity than those fuelled with uranium dioxide. Corrosion tests in 300{sup o}C water preceded extensive in-reactor performance tests of fuel elements and bundles. Proper heat-treatment of U-3.9 wt% Si gives a U{sub 3}5i specimen which corrodes at less than 2 mg/cm{sup 2} h in 300{sup o}C water. This is an order of magnitude lower than the maximum corrosion rate tolerable in a water-cooled reactor. U{sub 3}Si in a defected unbonded Zircaloy-2 sheath showed only a slow uniform sheath expansion in 300{sup o}C water. All tests were done under isothermal conditions in an out-reactor loop. (author)
International Nuclear Information System (INIS)
Supardjo; Rojak, A.; Boybul; Suyoto; Datam, A. S.
2000-01-01
The calibration of X-Ray intensity of the U 3 Si 2 -AI fuel plates with the uranium loading between 3.60 up to 5.20 g/cm 3 and varied thickness of AIMgSi1 reference block have been performed. The measurement with changing variable slit diameter and energy of X-Ray attenuation, are produced enough representative X-Ray intensity at 18 mm slit diameter and energy of 43 kV. From the correlation of X-ray intensities vs variation of uranium loading in the fuel plates and thickness of the AIMgSi1 materials, the equivalence of thickness of the AIMgSi1 block to the uranium loading of fuel plates are determined. By assuming that the tolerance of the homogeneity measurement is + 20 % from normal thickness staircase of the AIMgSi1 standard could be determined and than together with fuel plate were scanned to determine the uranium homogeneity. The test result on the U 3 Si 2 -AI fuel plates with uranium loading of 4.80 and 5.20 g/cm 3 (each 4 fuel plates) indicated that uranium distribution in the fuel plates is relatively homogeneous, with each maximum deviation being 6.30 % and 6.90%. It is showed that measurement method is relatively good, easy, and fast so that this method is suitable to control the uranium homogeneity in the fuel plate. (author)
UO2 fuel pellets fabrication via Spark Plasma Sintering using non-standard molybdenum die
Papynov, E. K.; Shichalin, O. O.; Mironenko, A. Yu; Tananaev, I. G.; Avramenko, V. A.; Sergienko, V. I.
2018-02-01
The article investigates spark plasma sintering (SPS) of commercial uranium dioxide (UO2) powder of ceramic origin into highly dense fuel pellets using non-standard die instead of usual graphite die. An alternative and formerly unknown method has been suggested to fabricate UO2 fuel pellets by SPS for excluding of typical problems related to undesirable carbon diffusion. Influence of SPS parameters on chemical composition and quality of UO2 pellets has been studied. Also main advantages and drawbacks have been revealed for SPS consolidation of UO2 in non-standard molybdenum die. The method is very promising due to high quality of the final product (density 97.5-98.4% from theoretical, absence of carbon traces, mean grain size below 3 μm) and mild sintering conditions (temperature 1100 ºC, pressure 141.5 MPa, sintering time 25 min). The results are interesting for development and probable application of SPS in large-scale production of nuclear ceramic fuel.
The choice of the fuel assembly for VVER-1000 in a closed fuel cycle based on REMIX-technology
International Nuclear Information System (INIS)
Bobrov, E.; Alekseev, P.; Chibinyaev, A.; Teplov, P.; Dudnikov, A.
2016-01-01
REMIX (Regenerated Mixture) fuel is produced directly from a non-separated mix of recycled uranium and plutonium from reprocessed used fuel and the fabrication technology of such fuel is called REMIX-technology. This paper shows basic features of different fuel assembly (FA) application for VVER-1000 in a closed fuel cycle based on REMIX-technology. This investigation shows how the change in the water-fuel ratio in the VVER FA affects the fuel characteristics produced by REMIX technology during multiple recycling. It is shown that for for the traditional REMIX-fuel it does not make sense to change anything in the design of VVER FA, because there are no advantages in the fuel feed consumption. The natural uranium economy by the fifth cycle reached about 29%. In the case of the REMIX fuel based on uranium-plutonium from SNF MOX fuel, it would be appropriate to use fuel assemblies with a water-fuel ratio of 1.5
Study on microstructure change of Uranium nitride coated U-7wt%Mo powder by heat treatment
Energy Technology Data Exchange (ETDEWEB)
Cho, Woo Hyoung; Park, Jae Soon; Lee, Hae In; Kim, Woo Jeong; Yang, Jae Ho; Park, Jong Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2011-10-15
Uranium-molybdenum alloy particle dispersion fuel in an aluminum matrix with a high uranium density has been developed for a high performance research reactor in the RERTR program. In order to retard the fuel-matrix interaction in U-Mo/Al dispersion fuel in which the U-Mo fuel particles were dispersed in Al matrix, nitride layer coated U-Mo fuel particle has been designed and techniques to fabricate nitride-layer coated U-7wt%Mo particles have been developed in our lab. In this study, uranium nitride coated U-Mo particle has heat treatment for several times and degree. And we suggested for interaction layer remedy in U-Mo dispersion fuel. We investigate effect of heat treatment interaction layer evolution on uranium nitride coated U-Mo powder. The EDS and XRD analysis to investigate the phase evolution in uranium nitride coated layer is also a part of the present work
Natural uranium fueled light water moderated breeding hybrid power reactors
International Nuclear Information System (INIS)
Greenspan, E.; Schneider, A.; Misolovin, A.; Gilai, D.; Levin, P.
The feasibility of fission-fusion hybrid reactors based on breeding light water thermal fission systems is investigated. The emphasis is on fuel-self-sufficient (FSS) hybrid power reactors that are fueled with natural uranium. Other LWHRs considered include FSS-LWHRs that are fueled with spent fuel from LWRs, and LWHRs which are to supplement LWRs to provide a tandem LWR-LWHR power economy that is fuel-self-sufficient
Development of Nuclear Fuel Remote Fabrication Technology
International Nuclear Information System (INIS)
Lee, Jung Won; Yang, M. S.; Kim, S. S. and others
2005-04-01
The aim of this study is to develop the essential technology of dry refabrication using spent fuel materials in a laboratory scale on the basis of proliferation resistance policy. The emphasis is placed on the assessment and the development of the essential technology of dry refabrication using spent fuel materials. In this study, the remote fuel fabrication technology to make a dry refabricated fuel with an enhanced quality was established. And the instrumented fuel pellets and mini-elements were manufactured for the irradiation testing in HANARO. The design and development technology of the remote fabrication equipment and the remote operating and maintenance technology of the equipment in hot cell were also achieved. These achievements will be used in and applied to the future back-end fuel cycle and GEN-IV fuel cycle and be a milestone for Korea to be an advanced nuclear country in the world
Fuel Cycle of Reactor SVBR-100
Energy Technology Data Exchange (ETDEWEB)
Zrodnikov, A.V.; Toshinsky, G.I.; Komlev, O.G. [FSUE State Scientific Center Institute for Physics and Power Engineering, 1, Bondarenko sq., Obninsk, Kaluga rg., 249033 (Russian Federation)
2009-06-15
Modular fast reactor with lead-bismuth heavy liquid-metal coolant in 100 MWe class (SVBR 100) is referred to the IV Generation reactors and shall operate in a closed nuclear fuel cycle (NFC) without consumption of natural uranium. Usually it is considered that launch of fast reactors (FR) is realized using mixed uranium-plutonium fuel. However, such launch of FRs is not economically effective because of the current costs of natural uranium and uranium enrichment servicing. This is conditioned by the fact that the quantity of reprocessing the spent nuclear fuel (SNF) of thermal reactors (TR) calculated for a ton of plutonium that determines the expenditures for construction and operation of the corresponding enterprise is very large due to low content of plutonium in the TR SNF. The economical effectiveness of FRs will be reduced as the enterprises on reprocessing the TR SNF have to be built prior to FRs have been implemented in the nuclear power (NP). Moreover, the pace of putting the FRs in the NP will be constrained by the quantity of the TR SNF. The report grounds an alternative strategy of FRs implementation into the NP, which is considered to be more economically effective. That is conditioned by the fact that in the nearest future use of the mastered uranium oxide fuel for FRs and operation in the open fuel cycle with postponed reprocessing will be most economically expedient. Changeover to the mixed uranium-plutonium fuel and closed NFC will be economically effective when the cost of natural uranium is increased and the expenditures for construction of enterprises on SNF reprocessing, re-fabrication of new fuel with plutonium and their operating becomes lower than the corresponding costs of natural uranium, uranium enrichment servicing, expenditures for fabrication of fresh uranium fuel and long temporary storage of the SNF. As when operating in the open NFC, FRs use much more natural uranium as compared with TRs, and at a planned high pace of NP development
Modified ADS molten salt processes for back-end fuel cycle of PWR spent fuel
International Nuclear Information System (INIS)
Choi, In-Kyu; Yeon, Jei-Won; Kim, Won-Ho
2002-01-01
The back-end fuel cycle concept for PWR spent fuel is explained. This concept is adequate for Korea, which has operated both PWR and CANDU reactors. Molten salt processes for accelerator driven system (ADS) were modified both for the transmutation of long-lived radioisotopes and for the utilisation of the remained fissile uranium in PWR spent fuels. Prior to applying molten salt processes to PWR fuel, hydrofluorination and fluorination processes are applied to obtain uranium hexafluoride from the spent fuel pellet. It is converted to uranium dioxide and fabricated into CANDU fuel. From the remained fluoride compounds, transuranium elements can be separated by the molten salt technology such as electrowinning and reductive extraction processes for transmutation purpose without weakening the proliferation resistance of molten salt technology. The proposed fuel cycle concept using fluorination processes is thought to be adequate for our nuclear program and can replace DUPIC (Direct Use of spent PWR fuel in CANDU reactor) fuel cycle. Each process for the proposed fuel cycle concept was evaluated in detail
New options to fuel plate for MTR reactor
International Nuclear Information System (INIS)
Macedo, C.R.
1988-01-01
The main datas of fuel elements and the new materials for good performance of the MTR reactor are described. A study to verify the possibility of introduction a new element on the alloy is presented. After verification the stages of nucleus fabrication with dispersion cermets of uranium oxide is gave a special emphasis to cermet fabrication of uranium-aluminium alloys. (C.G.C.) [pt
Assuaging Nuclear Energy Risks: The Angarsk International Uranium Enrichment Center
International Nuclear Information System (INIS)
Myers, Astasia
2011-01-01
The recent nuclear renaissance has motivated many countries, especially developing nations, to plan and build nuclear power reactors. However, domestic low enriched uranium demands may trigger nations to construct indigenous enrichment facilities, which could be redirected to fabricate high enriched uranium for nuclear weapons. The potential advantages of establishing multinational uranium enrichment sites are numerous including increased low enrichment uranium access with decreased nuclear proliferation risks. While multinational nuclear initiatives have been discussed, Russia is the first nation to actualize this concept with their Angarsk International Uranium Enrichment Center (IUEC). This paper provides an overview of the historical and modern context of the multinational nuclear fuel cycle as well as the evolution of Russia's IUEC, which exemplifies how international fuel cycle cooperation is an alternative to domestic facilities.
Risk-analysis of the fuel cycle in the Netherlands
International Nuclear Information System (INIS)
1975-06-01
The introduction of nuclear power production in the Netherlands in 1985 up to an installed power of 3500MWe, requires a certain capacity for the fabrication of fuel elements. In view of the risk analysis, a study group has originated a plan to develop a Dutch fuel fabrication plant with developing economic prospectives. Until 1986, only light enriched uranium will be manufactured; after that period, the recycling of plutonium. The location of the projected factory is yet indefinite. The possibilities of waste disposal could influence the selection of the location. The threat of critical accidents remains low according to the permissible levels of polution to the environment. The fabrication of fuel elements would not give any significant contribution in the radiation burden to the environment. The working conditions are strongly supervised and follow the standing procedures and disciplines. The manufacturing processes in the uranium fuel cycle for light water reactors have been described with particular reference to the chemical conversion of UF 6 , sintering of the fuel pellets, the fabrication of uranium oxide-vibrasol 6 fuel and the steel assembling of fuel elements. The safeguarding of the fuel cycle has been submitted to strictly enforced administrative control. The recycling of plutonium in light water reactors on an industrial scale would not be foreseeable for some time in Holland. Because of the much higher specific radioactivity of the material (6 x 10 -2 Ci/g), the processing of plutonium requires additional provisions. For the present, the Dutch factory should process only Pu bearing fuel with 5% Pu and in the form of high density sintered globules with 1mm diameter. The specific radioactivity of this material is 3 x 10 -3 Ci/g, about 2000 times that of light enriched uranium. Experiences in the safe handling of this material has taught the processing of radioactivated fuel in gloveboxes which are connected to ventilation systems with extra filters
Reduction of uranium in disposal conditions of spent nuclear fuel
International Nuclear Information System (INIS)
Myllykylae, E.
2008-02-01
This literature study is a summary of publications, in which the reduction of uranium by iron has been investigated in anaerobic groundwater conditions or in aqueous solution in general. The basics of the reduction phenomena and the oxidation states, complexes and solubilities of uranium and iron in groundwaters are discussed as an introduction to the subject, as well as, the Finnish disposal concept of spent nuclear fuel. The spent fuel itself mainly (∼96 %) consists of a sparingly soluble uranium(IV) dioxide, UO 2 (s), which is stable phase in the anticipated reducing disposal conditions. If spent fuel gets in contact with groundwater, oxidizing conditions might be induced by the radiolysis of water, or by the intrusion of oxidizing glacial melting water. Under these conditions, the oxidation and dissolution of uranium dioxide to more soluble U(VI) species could occur. This could lead to the mobilization of uranium and other components of spent fuel matrix including fission products and transuranium elements. The reduction of uranium back to oxidation state U(IV) can be considered as a favourable immobilization mechanism in a long-term, leading to precipitation due to the low solubility of U(IV) species. The cast iron insert of the disposal canister and its anaerobic corrosion products are the most important reductants under disposal conditions, but dissolved ferrous iron may also function as reductant. Other iron sources in the buffer or near-field rock, are also considered as possible reductants. The reduction of uranium is a very challenging phenomenon to investigate. The experimental studies need e.g. well-controlled anoxic conditions and measurements of oxidation states. Reduction and other simultaneous phenomena are difficult to distinghuish. The groundwater conditions (pH, Eh and ions) influence on the prevailing complexes of U and Fe and on forming corrosion products of iron and, thus they determine also the redox chemistry. The partial reduction of
Highlights of 50 years of nuclear fuel development
International Nuclear Information System (INIS)
Simnad, M.T.
1989-01-01
The development of nuclear fuels since the discovery of nuclear fission is briefly surveyed in this paper. The fabrication of the uranium fuel for the first nuclear pile, CP-1, is described. The research and development studies and fabrication of the different types of nuclear fuels for the variety of research and power reactors are reviewed. The important factors involved to achieve low fuel-cycle costs and reliable performance in the fuel elements are discussed in the historical context. 10 refs
Improvements in fabrication of metallic fuels
International Nuclear Information System (INIS)
Tracy, D.B.; Henslee, S.P.; Dodds, N.E.; Longua, K.J.
1989-12-01
Argonne National Laboratory is currently developing a new liquid- metal cooled breeder reactor known as the Integral Fast Reactor (IFR). IFR fuels represent the state-of-the-art in metal-fueled reactor technology. Improvements in the fabrication of metal fuel, to be discussed below, will support the fully remote fuel cycle facility that as an integral part of the IFR concept will be demonstrated at the EBR-II site. 3 refs
High U-density nuclear fuel development with application of centrifugal atomization technology
International Nuclear Information System (INIS)
Kim, Chang Kyu; Kim, Ki Hwan; Lee, Don Bae
1997-01-01
In order to simplify the preparation process and improve the properties of uranium silicide fuels prepared by mechanical comminution, a fuel fabrication process applying rotating-disk centrifugal atomization technology was invented in KAERI in 1989. The major characteristic of atomized U 3 Si and U 3 Si 2 powders have been examined. The out-pile properties, including the thermal compatibility between atomized particle and aluminum matrix in uranium silicide dispersion fuels, have generally showed a superiority to the comminuted fuels. Moreover, the RERTR (reduced enrichment for research and test reactors) program, which recently begins to develop very-high-density uranium alloy fuels, including U-Mo fuels, requires the centrifugal atomization process to overcome the contaminations of impurities and the difficulties of the comminution process. In addition, a cooperation with ANL in the U.S. has been performed to develop high-density fuels with an application of atomization technology since December 1996. If the microplate and miniplate irradiation tests of atomized fuels, which have been performed with ANL, demonstrated the stability and improvement of in-reactor behaviors, nuclear fuel fabrication technology by centrifugal atomization could be most-promising to the production method of very-high-uranium-loading fuels. (author). 22 refs., 2 tabs., 12 figs
Licensed fuel facility status report: Inventory difference data, July 1, 1990--June 30, 1991
International Nuclear Information System (INIS)
1992-03-01
NRC is committed to the periodic publication of licensed fuel facilities inventory difference data, following agency review of the information and completion of any related NRC investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, plutonium, or uranium-233
Process for producing uranium oxide rich compositions from uranium hexafluoride
International Nuclear Information System (INIS)
DeHollander, W.R.; Fenimore, C.P.
1978-01-01
Conversion of gaseous uranium hexafluoride to a uranium dioxide rich composition in the presence of an active flame in a reactor defining a reaction zone is achieved by separately introducing a first gaseous reactant comprising a mixture of uranium hexafluoride and a reducing carrier gas, and a second gaseous reactant comprising an oxygen-containing gas. The reactants are separated by a shielding gas as they are introduced to the reaction zone. The shielding gas temporarily separates the gaseous reactants and temporarily prevents substantial mixing and reacting of the gaseous reactants. The flame occurring in the reaction zone is maintained away from contact with the inlet introducing the mixture to the reaction zone. After suitable treatment, the uranium dioxide rich composition is capable of being fabricated into bodies of desired configuration for loading into nuclear fuel rods. Alternatively, an oxygen-containing gas as a third gaseous reactant is introduced when the uranium hexafluoride conversion to the uranium dioxide rich composition is substantially complete. This results in oxidizing the uranium dioxide rich composition to a higher oxide of uranium with conversion of any residual reducing gas to its oxidized form
Progress in developing very-high-density low-enriched-uranium fuels
International Nuclear Information System (INIS)
Snelgrove, J.L.; Hofman, G.L.; Meyer, M.K.; Hayes, S.L.; Wiencek, T.C.; Strain, R.V.
1999-01-01
Preliminary results from the postirradiation examinations of microplates irradiated in the RERTR-1 and -2 experiments in the ATR have shown several binary and ternary U-Mo alloys to be promising candidates for use in aluminum-based dispersion fuels with uranium densities up to 8 to 9 g/cm 3 . Ternary alloys of uranium, niobium, and zirconium performed poorly, however, both in terms of fuel/matrix reaction and fission-gas-bubble behavior, and have been dropped from further study. Since irradiation temperatures achieved in the present experiments (approximately 70 deg. C) are considerably lower than might be experienced in a high-performance reactor, a new experiment is being planned with beginning-of-cycle temperatures greater than 200 deg. C in 8-g U/cm 3 fuel. (author)
International Nuclear Information System (INIS)
Bolon, A.E.; Straka, M.; Freeman, D.W.
1997-01-01
The objectives of this project were to convert the UMR Reactor fuel from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel and to ship the HEU fuel back to the Department of Energy Savannah River Site. The actual core conversion was completed in the summer of 1992. The HEU fuel was offloaded to an onsite storage pit where it remained until July, 1996. In July, 1996, the HEU fuel was shipped to the DOE Savannah River Site. The objectives of the project have been achieved. DOE provided the following funding for the project. Several papers were published regarding the conversion project and are listed in the Attachment. In retrospect, the conversion project required much more time and effort than originally thought. Several difficulties were encountered including the unavailability of a shipping cask for several years. The authors are grateful for the generous funding provided by DOE for this project but wish to point out that much of their efforts on the conversion project went unfunded
Uranium transport to solid electrodes in pyrochemical reprocessing of nuclear fuel
International Nuclear Information System (INIS)
Tomczuk, Z.; Ackerman, J.P.; Wolson, R.D.; Miller, W.E.
1992-01-01
A unique pyrochemical process developed for the separation of metallic nuclear fuel from fission products by electrotransport through molten LiCl-KCl eutectic salt to solid and liquid metal cathodes. The process allow for recovery and reuse of essentially all of the actinides in spent fuel from the integral fast reactor (IFR) and disposal of wastes in satisfactory forms. Electrotransport is used to minimize reagent consumption and, consequently, waste volume. In particular, electrotransport to solid cathodes is used for recovery of an essentially pure uranium product in the presence of other actinides; removal of pure uranium is used to adjust the electrolyte composition in preparation for recovery of a plutonium-rich mixture with uranium in liquid cadmium cathodes. This paper presents experiments that delineate the behavior of key actinide and rare-earth elements during electrotransport to a solid electrode over a useful range of PuCl 3 /UCl 3 ratios in the electrolyte, a thermodynamic basis for that behavior, and a comparison of the observed behavior with that calculated from a thermodynamic model. This work clearly established that recovery of nearly pure uranium can be a key step in the overall pyrochemical-fuel-processing strategy for the IFR
Nuclear fuels and development of nuclear fuel elements
International Nuclear Information System (INIS)
Sundaram, C.V.; Mannan, S.L.
1989-01-01
Safe, reliable and economic operation of nuclear fission reactors, the source of nuclear power at present, requires judicious choice, careful preparation and specialised fabrication procedures for fuels and fuel element structural materials. These aspects of nuclear fuels (uranium, plutonium and their oxides and carbides), fuel element technology and structural materials (aluminium, zircaloy, stainless steel etc.) are discussed with particular reference to research and power reactors in India, e.g. the DHRUVA research reactor at BARC, Trombay, the pressurised heavy water reactors (PHWR) at Rajasthan and Kalpakkam, and the Fast Breeder Test Reactor (FBTR) at Kalpakkam. Other reactors like the gas-cooled reactors operating in UK are also mentioned. Because of the limited uranium resources, India has opted for a three-stage nuclear power programme aimed at the ultimate utilization of her abundant thorium resources. The first phase consists of natural uranium dioxide-fuelled, heavy water-moderated and cooled PHWR. The second phase was initiated with the attainment of criticality in the FBTR at Kalpakkam. Fast Breeder Reactors (FBR) utilize the plutonium and uranium by-products of phase 1. Moreover, FBR can convert thorium into fissile 233 U. They produce more fuel than is consumed - hence, the name breeders. The fuel parameters of some of the operating or proposed fast reactors in the world are compared. FBTR is unique in the choice of mixed carbides of plutonium and uranium as fuel. Factors affecting the fuel element performance and life in various reactors e.g. hydriding of zircaloys, fuel pellet-cladding interaction etc. in PHWR and void swelling; irradiation creep and helium embrittlement of fuel element structural materials in FBR are discussed along with measures to overcome some of these problems. (author). 15 refs., 9 tabs., 23 figs
Licensed fuel facility status report. Inventory difference data, January-June 1983. Volume 4, No. 1
International Nuclear Information System (INIS)
1984-03-01
NRC is committed to the periodic publication of licensed fuel facilities inventory difference data, following agency review of the information and completion of any related investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, or uranium-233
Various problems in establishment of fuel cycle business in Japan
International Nuclear Information System (INIS)
Murata, Hiroshi
1985-01-01
Since Japan instituted the Atomic Energy Act in 1956, and organized the Atomic Energy Commission, as the fundamental policy of the peaceful use of atomic energy, the industrialization and establishment of fuel cycle technology have been advanced as well as the development of power reactors. The consistent and harmonious industrialization of uranium enrichment, fuel fabrication, reprocessing, the utilization of recovered plutonium and uranium, and the storage, treatment and disposal of wastes has been the target. As the nuclear power generation in Japan grew, the enhancement of the various factors of nuclear fuel cycle as the base of supporting nuclear power generation has become necessary. The effort of technical development has been continued in the fields of uranium enrichment, fuel reprocessing, plutonium fuel and waste treatment by the Power Reactor and Nuclear Fuel Development Corp., Japan Atomic Energy Research Institute and related industries. The plan and present status of nuclear fuel cycle business in Japan, the problems such as the roles of the government and private enterprises, technology transfer, the economy of nuclear fuel cycle business, the industrialization of mixed oxide fuel fabrication, nuclear nonproliferation policy and location are discussed. (Kako, I.)
Fuel fabrication and reprocessing for nuclear fuel cycle with inherent safety demands
Energy Technology Data Exchange (ETDEWEB)
Shadrin, Andrey Yurevich; Dvoeglazov, Konstantin Nikolaevich; Ivanov, Valentine Borisovich; Volk, Vladimir Ivanovich; Skupov, Mikhail Vladimirovich; Glushenkov, Alexey Evgenevich [Joint Stock Company ' ' The High Technological Research Institute of Inorganic Materials' ' , Moscow (Russian Federation); Troyanov, Vladimir Mihaylovich; Zherebtsov, Alexander Anatolievich [Innovation and Technology Center of Project ' ' PRORYV' ' , State Atomic Energy Corporation ' ' Rosatom' ' , Moscow (Russian Federation)
2015-06-01
The strategies adopted in Russia for a closed nuclear fuel cycle with fast reactors (FR), selection of fuel type and recycling technologies of spent nuclear fuel (SNF) are discussed. It is shown that one of the possible technological solutions for the closing of a fuel cycle could be the combination of pyroelectrochemical and hydrometallurgical methods of recycling of SNF. This combined scheme allows: recycling of SNF from FR with high burn-up and short cooling time; decreasing the volume of stored SNF and the amount of plutonium in a closed fuel cycle in FR; recycling of any type of SNF from FR; obtaining the high pure end uranium-plutonium-neptunium end-product for fuel refabrication using pellet technology.
NEUTRONICS STUDIES OF URANIUM-BASED FULLY CERAMIC MICRO-ENCAPSULATED FUEL FOR PWRs
Energy Technology Data Exchange (ETDEWEB)
George, Nathan M [ORNL; Maldonado, G Ivan [ORNL; Terrani, Kurt A [ORNL; Gehin, Jess C [ORNL; Godfrey, Andrew T [ORNL
2012-01-01
This study evaluates the core neutronics and fuel cycle characteristics that result from employing uranium-based fully ceramic micro-encapsulated (FCM) fuel in a pressurized water reactor (PWR). Specific PWR bundle designs with FCM fuel have been developed, which by virtue of their TRISO particle based elements, are expected to safely reach higher fuel burnups while also increasing the tolerance to fuel failures. The SCALE 6.1 code package, developed and maintained at ORNL, was the primary software employed to model these designs. Analysis was performed using the SCALE double-heterogeneous (DH) fuel modeling capabilities. For cases evaluated with the NESTLE full-core three-dimensional nodal simulator, because the feature to perform DH lattice physics branches with the SCALE/TRITON sequence is not yet available, the Reactivity-Equivalent Physical Transformation (RPT) method was used as workaround to support the full core analyses. As part of the fuel assembly design evaluations, fresh feed lattices were modeled to analyze the within-assembly pin power peaking. Also, a color-set array of assemblies was constructed to evaluate power peaking and power sharing between a once-burned and a fresh feed assembly. In addition, a parametric study was performed by varying the various TRISO particle design features; such as kernel diameter, coating layer thicknesses, and packing fractions. Also, other features such as the selection of matrix material (SiC, Zirconium) and fuel rod dimensions were perturbed. After evaluating different uranium-based fuels, the higher physical density of uranium mononitride (UN) proved to be favorable, as the parametric studies showed that the FCM particle fuel design will need roughly 12% additional fissile material in comparison to that of a standard UO2 rod in order to match the lifetime of an 18-month PWR cycle. Neutronically, the FCM fuel designs evaluated maintain acceptable design features in the areas of fuel lifetime, temperature
Fabrication and characterization of fully ceramic microencapsulated fuels
Energy Technology Data Exchange (ETDEWEB)
Terrani, K.A., E-mail: kurt.terrani@gmail.com [Fuel Cycle and Isotopes Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Kiggans, J.O.; Katoh, Y. [Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Shimoda, K. [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Montgomery, F.C.; Armstrong, B.L.; Parish, C.M. [Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Hinoki, T. [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Hunn, J.D. [Fuel Cycle and Isotopes Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Snead, L.L. [Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)
2012-07-15
The current generation of fully ceramic microencapsulated fuels, consisting of Tristructural Isotropic fuel particles embedded in a silicon carbide matrix, is fabricated by hot pressing. Matrix powder feedstock is comprised of alumina-yttria additives thoroughly mixed with silicon carbide nanopowder using polyethyleneimine as a dispersing agent. Fuel compacts are fabricated by hot pressing the powder-fuel particle mixture at a temperature of 1800-1900 Degree-Sign C using compaction pressures of 10-20 MPa. Detailed microstructural characterization of the final fuel compacts shows that oxide additives are limited in extent and are distributed uniformly at silicon carbide grain boundaries, at triple joints between silicon carbide grains, and at the fuel particle-matrix interface.
Description of ECRI (CNEA'S MTR fuel fabrication plant)
International Nuclear Information System (INIS)
Echenique, P.; Fabro, J.; Podesta, D.; Restelli, M.; Rossi, G.; Alvarez, L.; Adelfang, P.
2002-01-01
The ECRI Plant is dedicated to the development and fabrication of high-density fuel elements and targets for 99 Mo. In this sector had been done the start up Fuel Elements for the Reactors of Peru, Iran, Algeria and Egypt. All of them were made with U 3 O 8 . The targets for 99 Mo using HEU were fabricated too in the last years. The new material of high-density for Fuel Elements as U 3 Si 2 were done in this sector, three prototypes were fabricated, two are still under irradiation. (P06 and P07). As new developments we are working with U-Mo (7%) Fuel Plates with both material Korean and HMD. This work is under the RERTR Program and two fuel elements, manufactured by us, with both powders, will be irradiated in Petten. For 99 Mo targets, we are fabricating miniplates of LEU with an AlUx powder by pulvi-metallurgy technique. And it is under development the foils targets under the RERTR Program. A general view of the fabrication facilities and control sector will be shown. The different operations that are done in each sector will be explained. All our activities will be certified under the ISO 9000 and we are working hard to get it in the middle of 2003. (author)
Energy Technology Data Exchange (ETDEWEB)
Stohr, J A; Englander, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires
1958-07-01
Description of some theoretical and experimental data concerning the design and most economic preparation of metallic uranium based fuel elements, which are intended to produce an energy of 3 kW days/g of uranium in a thermal reactor, at a sufficiently high mean temperature. Experimental results obtained by testing by analogy or by actually trying out fuel elements obtained by alloying uranium with other metals in proportions such that the resistance to deformation of the alloy produced is much higher than that of pure metallic uranium and that the thermal utilisation factor is only slightly different from that of the uranium. (author) [French] Description de quelques donnees theoriques et experimentales concernant la conception et la preparation la plus economique d'elements combustibles a base d'uranium metallique naturel, destines a degager dans un reacteur thermique une energie de l'ordre de 3 kWj/g d'uranium a une temperature moyenne suffisamment elevee. Resultats experimentaux acquis par tests analogiques ou reels sur combustibles obtenus par alliage de l'uranium avec des elements metalliques en proportions telles que la resistance a la deformation soit bien superieure a celle de l'uranium metal pur et que le facteur propre d'utilisation thermique n ne soit que peu affecte. (auteur)
MOX fuel fabrication technology in J-MOX
International Nuclear Information System (INIS)
Osaka, Shuichi; Yoshida, Ryouichi; Yamazaki, Yukiko; Ikeda, Hiroyuki
2014-01-01
Japan Nuclear Fuel Ltd. (JNFL) has constructed JNFL MOX Fuel Fabrication Plant (J-MOX) since 2010. The MIMAS process has been introduced in the powder mixing process from AREVA NC considering a lot of MOX fuel fabrication experiences at MELOX plant in France. The feed material of Pu for J-MOX is MH-MOX powder from Rokkasho Reprocessing Plant (RRP) in Japan. The compatibility of the MH-MOX powder with the MIMAS process was positively evaluated and confirmed in our previous study. This paper describes the influences of the UO2 powder and the recycled scrap powder on the MOX pellet density. (author)
Fabrication of MOX fuel element clusters for irradiation in PWL, CIRUS
International Nuclear Information System (INIS)
Roy, P.R.; Purushotham, D.S.C.; Majumdar, S.
1983-01-01
Three clusters, each containing 6 zircaloy-2 clad short length fuel elements of either MOX or UO 2 fuel pellets were fabricated for irradiation in pressurized water loop of CIRUS. The major objectives of the programme were: (a) to optimize the various fabrication parameters for developing a flow sheet for MOX fuel element fabrication; (b) to study the performance of the MOX fuel elements at a peak heat flux of 110 W/cm 2 ; and (c) to study the effect of various fuel pellet design changes on the behaviour of the fuel element under irradiation. Two clusters, one each of UO 2 and MOX, have been successfully irradiated to the required burn-up level and are now awaiting post irradiation examinations. The third MOX cluster is still undergoing irradiation. Fabrication of these fuel elements involved considerable amount of developing work related to the fabrication of the MOX fuel pellets and the element welding technique and is reported in detail in this report. (author)
Extraction of Uranium Using Nitrogen Dioxide and Carbon Dioxide for Spent Fuel Reprocessing
Energy Technology Data Exchange (ETDEWEB)
Kayo Sawada; Daisuke Hirabayashi; Youichi Enokida [EcoTopia Science Institute, Nagoya University, Nagoya, 464-8603 (Japan)
2008-07-01
For the reprocessing of spent nuclear fuels, a new method to extract actinides from spent fuel using highly compressed gases, nitrogen dioxide and carbon dioxide was proposed. Uranium extraction from broken pieces, whose average grain size was 5 mm, of uranium dioxide pellet with nitrogen dioxide and carbon dioxide was demonstrated in the present study. (authors)
The SLOWPOKE-2 reactor with low enrichment uranium oxide fuel
International Nuclear Information System (INIS)
Townes, B.M.; Hilborn, J.W.
1985-06-01
A SLOWPOKE-2 reactor core contains less than 1 kg of highly enriched uranium (HEU) and the proliferation risk is very low. However, to overcome proliferation concerns a new low enrichment uranium (LEU) fuelled reactor core has been designed. This core contains approximately 180 fuel elements based on the Zircaloy-4 clad UOsub(2) CANDU fuel element, but with a smaller outside diameter. The physics characteristics of this new reactor core ensure the inherent safety of the reactor under all conceivable conditions and thus the basic SLOWPOKE safety philosophy which permits unattended operation is not affected
International Nuclear Information System (INIS)
Wachs, G.W.
1997-11-01
The Department of Energy (DOE) Fissile Materials Disposition Materials Disposition Program (FMDP) has announced that reactor irradiation of MOX fuel is one of the preferred alternatives for disposal of surplus weapons-usable plutonium (Pu). MOX fuel has been utilized domestically in test reactors and on an experimental basis in a number of Commercial Light Water Reactors (CLWRs). Most of this experience has been with Pu derived from spent low enriched uranium (LEU) fuel, known as reactor grade (RG) Pu. The MOX fuel test will be irradiated in the ATR to provide preliminary data to demonstrate that the unique properties of surplus weapons-derived or weapons-grade (WG) plutonium (Pu) do not compromise the applicability of this MOX experience base. In addition, the test will contribute experience with irradiation of gallium-containing fuel to the data base required for resolution of generic CLWR fuel design issues (ORNL/MD/LTR-76). This Fabrication, Inspection, and Test Plan (FITP) is a level 2 document as defined in the FMDP LWR MOX Fuel Irradiation Test Project Plan (ORNL/MD/LTR-78)
Irradiation of TZM: Uranium dioxide fuel pin at 1700 K
Mcdonald, G. E.
1973-01-01
A fuel pin clad with TZM and containing solid pellets of uranium dioxide was fission heated in a static helium-cooled capsule at a maximum surface temperature of 1700 K for approximately 1000 hr and to a total burnup of 2.0 percent of the uranium-235. The results of the postirradiation examination indicated: (1) A transverse, intergranular failure of the fuel pin occurred when the fuel pin reached 2.0-percent burnup. This corresponds to 1330 kW-hr/cu cm, where the volume is the sum of the fuel, clad, and void volumes in the fuel region. (2) The maximum swelling of the fuel pin was less than 1.5 percent on the fuel-pin diameter. (3) There was no visible interaction between the TZM clad and the UO2. (4) Irradiation at 1700 K produced a course-grained structure, with an average grain diameter of 0.02 centimeter and with some of the grains extending one-half of the thickness of the clad. (5) Below approximately 1500 K, the irradiation of the clad produced a moderately fine-grained structure, with an average grain diameter of 0.004 centimeter.
Mixed U/Pu oxide fabrication facility for gel-sphere-pac fuel
International Nuclear Information System (INIS)
1978-09-01
This paper describes a conceptual plant which uses the gel-sphere-pac process to fabricate mixed oxide (MOX) fuel and covers (1) fabrication of co-processed MOX fuel and (2) fabrication of co-processed spiked MOX fuel, using 60 Co. The report describes: the fuel fabrication process and plant layout, including scrap and waste processing; and maintenance safety and ventilation measures. A description of the conversion of U and Pu nitrate using a gel sphere process is given in Appendix A
Quantities of actinides in nuclear reactor fuel cycles
International Nuclear Information System (INIS)
Ang, K.P.
1975-01-01
The quantities of plutonium and other fuel actinides have been calculated for equilibrium fuel cycles for 1000 MW reactors of the following types: water reactors fueled with slightly enriched uranium, water reactors fueled with plutonium and natural uranium, fast-breeder reactors, gas-cooled reactors fueled with thorium and highly enriched uranium, and gas-cooled reactors fueled with thorium, plutonium, and recycled uranium. The radioactivity levels of plutonium, americium, and curium processed yearly in these fuel cycles are greatest for the water reactors fueled with natural uranium and recycled plutonium. The total amount of actinides processed is calculated for the predicted future growth of the United States nuclear power industry. For the same total installed nuclear power capacity, the introduction of the plutonium breeder has little effect upon the total amount of plutonium processed in this century. The estimated amount of plutonium in the low-level process wastes in the plutonium fuel cycles is comparable to the amount of plutonium in the high-level fission product wastes. The amount of plutonium processed in the nuclear fuel cycles can be considerably reduced by using gas-cooled reactors to consume plutonium produced in uranium-fueled water reactors. These, and other reactors dedicated for plutonium utilization, could be co-located with facilities for fuel reprocessing and fuel fabrication to eliminate the off-site transport of separated plutonium. (U.S.)
Design study and evaluation of fuel fabrication systems for FR fuel cycle
International Nuclear Information System (INIS)
Namekawa, Takashi; Tanaka, Kenya; Kawaguchi, Koichi; Koike, Kazuhiro; Shimuta, Hiroshi; Suzuki, Yoshihiro
2004-01-01
The plant concept for each FBR fuel fabrication system has been constructed and evaluated, which achieves economical improvement, decrease in the environmental burden, better resource utilization, and proliferation resistance by the various innovative techniques employed. The results are as follows: (1) For oxide fuels, the simplified pelletizing method has a high technical feasibility, and it is possible to apply this method to practical process at early stage, because this method is based on wealth results of a conventional method. (2) For oxide fuels, the sphere packing fuel fabrication system by gelation and vibro-compaction processes has the advantage of lesser dispersion of the fine powder due to the use of solution and granule in the process. However this system shoulders additional cost for the liquid waste treatment process to dispose a large bulk of process liquid waste. (3) For the metal fuel, the casting system is generally expected to have high economical efficiency even for small-scale facilities, although verification for fabrication of the TRU alloy slug is required. (author)
Radioactive decay properties of CANDU fuel. Volume 1: the natural uranium fuel cycle
International Nuclear Information System (INIS)
Clegg, L.J.; Coady, J.R.
1977-01-01
The two books of Volume 1 comprise the first in a three-volume series of compilations on the radioactive decay propertis of CANDU fuel and deal with the natural uranium fuel cycle. Succeeding volumes will deal with fuel cycles based on plutonium recycle and thorium. In Volume 1 which is divided into three parts, the computer code CANIGEN was used to obtain the mass, activity, decay heat and toxicity of CANDU fuel and its component isotopes. Data are also presented on gamma spectra and neutron emissions. Part 3 contains the data relating to the plutonium product and the high level wastes produced during fuel reprocessing. (author)
Licensed fuel facility status report. Inventory difference data, January-June 1985. Volume 6, No. 1
International Nuclear Information System (INIS)
1986-02-01
NRC is committed to the periodic publication of licensed fuel facilities' inventory difference data, following agency review of the information and completion of any related investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, plutonium, or uranium-233
Licensed fuel facility status report. Inventory difference data, July-December 1985. Volume 6, No. 2
International Nuclear Information System (INIS)
1986-08-01
NRC is committed to the periodic publication of licensed fuel facilities' inventory difference data, following agency review of the information and completion of any related investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, plutonium, or uranium-233
Licensed fuel facility status report. Inventory difference data, January-June 1984. Volume 5, No. 1
International Nuclear Information System (INIS)
1985-04-01
NRC is committed to the periodic publication of licensed fuel facilities' inventory difference data, following agency review of the information and completion of any related investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, plutonium, or Uranium-233
Evaluation of thorium based nuclear fuel. Extended summary
International Nuclear Information System (INIS)
Franken, W.M.P.; Bultman, J.H.; Konings, R.J.M.; Wichers, V.A.
1995-04-01
Application of thorium based nuclear fuels has been evaluated with emphasis on possible reduction of the actinide waste. As a result three ECN-reports are published, discussing in detail: - The reactor physics aspects, by comparing the operation characteristics of the cores of Pressurized Water Reactors and Heavy Water Reactors with different fuel types, including equilibrium thorium/uranium free, once-through uranium fuel and equilibrium uranium/plutonium fuel, - the chemical aspects of thorium based fuel cycles with emphasis on fuel (re)fabrication and fuel reprocessing, - the possible reduction in actinide waste as analysed for Heavy Water Reactors with various types of thorium based fuels in once-through operation and with reprocessing. These results are summarized in this report together with a short discussion on non-proliferation and uranium resource utilization. It has been concluded that a substantial reduction of actinide radiotoxicity of the disposed waste may be achieved by using thorium based fuels, if very efficient partitioning and multiple recycling of uranium and thorium can be realized. This will, however, require large efforts to develop the technology to the necessary industrial scale of operation. (orig.)
Management of reprocessed uranium. Current status and future prospects
International Nuclear Information System (INIS)
2007-02-01
There is worldwide interest in developing advanced and innovative technologies for nuclear fuel cycles, minimizing waste and environmental impacts. As of the beginning of 2003, about 171000 tonnes heavy metal spent nuclear fuel is in storage, while smaller amounts have been reprocessed. In several countries, including France, India, Japan and the Russian Federation, spent fuel has been viewed as a national energy resource. Some countries hold reprocessed uranium as the result of their commercial reprocessing service contracts for reprocessing the spent fuel of others. Reprocessed uranium has a potential value for recycling either directly or after appropriate treatment. This report analyses the existing options, approaches and developments in the management of reprocessed uranium. It includes the technical issues involved in managing reprocessed uranium which are RepU arisings, storage, chemical conversion, re-enrichment, fuel fabrication, transport, reactor irradiation, subsequent reprocessing and disposal options, as well as assessment of holistic environmental impacts. The objective of this document is to overview the information on the current status and future trends in the management of RepU and to identify major issues to be considered for future projects
Method of chemical reprocessing of irradiated nuclear fuels (especially fuels containing uranium)
International Nuclear Information System (INIS)
Koch, G.
1975-01-01
The invention deals with a method for the extraction especially of fast breeder fuels of high burn-up. A quaternary ammonium nitrate of high molecular weight is put into an organic diluting medium as extraction agent, corresponding to the general formula NRR'R''R'''NO 3 where R,R' and R'' are aliphatic radicals, R''' a methyl radical and the sum of the C atoms is greater than 16. After the extraction of the aqueous nitric acid containing nuclear fuel solution with this extracting agent, uranium, plutonium (or also thorium) can be found to a very high percentage in the organic phase and can be practically quantitatively back-extracted by means of diluted nitric acid, sulphuric acid or acetic acid. By using 30 volume percent tricapryl methyl ammonium nitrate in diethyl benzene for example, a distribution coefficient of 10.3 is obtained for uranium. (RB/LH) [de
Uranium kernel formation via internal gelation
International Nuclear Information System (INIS)
Hunt, R.D.; Collins, J.L.
2004-01-01
In the 1970s and 1980s, U.S. Department of Energy (DOE) conducted numerous studies on the fabrication of nuclear fuel particles using the internal gelation process. These amorphous kernels were prone to flaking or breaking when gases tried to escape from the kernels during calcination and sintering. These earlier kernels would not meet today's proposed specifications for reactor fuel. In the interim, the internal gelation process has been used to create hydrous metal oxide microspheres for the treatment of nuclear waste. With the renewed interest in advanced nuclear fuel by the DOE, the lessons learned from the nuclear waste studies were recently applied to the fabrication of uranium kernels, which will become tri-isotropic (TRISO) fuel particles. These process improvements included equipment modifications, small changes to the feed formulations, and a new temperature profile for the calcination and sintering. The modifications to the laboratory-scale equipment and its operation as well as small changes to the feed composition increased the product yield from 60% to 80%-99%. The new kernels were substantially less glassy, and no evidence of flaking was found. Finally, key process parameters were identified, and their effects on the uranium microspheres and kernels are discussed. (orig.)
Licensed-fuel-facility status report, inventory difference data January 1981-June 1981
International Nuclear Information System (INIS)
1982-07-01
NRC is committed to the periodic release of inventory difference data from the licensed fuel facilities after the agency has had an opportunity to review the data and has performed any related investigations associated with the data. Information included in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, plutonium, or uranium-233
Development of metal uranium fuel and testing of construction materials (I-VI); Part I
International Nuclear Information System (INIS)
Mihajlovic, A.
1965-11-01
This project includes the following tasks: Study of crystallisation of metal melt and beta-alpha transforms in uranium and uranium alloys; Study of the thermal treatment influence on phase transformations and texture in uranium alloys; Radiation damage of metal uranium; Project related to irradiation of metal uranium in the reactor; Development of fuel element for nuclear reactors
Uranium loss from BISO-coated weak-acid-resin HTGR fuel
International Nuclear Information System (INIS)
Pearson, R.L.; Lindemer, T.B.
1977-02-01
Recycle fuel for the High-Temperature Gas-Cooled Reactor (HTGR) contains a weak-acid-resin (WAR) kernel, which consists of a mixture of UC 2 , UO 2 , and free carbon. At 1900 0 C, BISO-coated WAR UC 2 or UC 2 -UO 2 kernels lose a significant portion of their uranium in several hundred hours. The UC 2 decomposes and uranium diffuses through the pyrolytic coating. The rate of escape of the uranium is dependent on the temperature and the surface area of the UC 2 , but not on a temperature gradient. The apparent activation energy for uranium loss, ΔH, is approximately 90 kcal/mole. Calculations indicate that uranium loss from the kernel would be insignificant under conditions to be expected in an HTGR
International Nuclear Information System (INIS)
Sridhar, T.S.
1988-01-01
A one-step, continuous, thermochemical calcination process, called the FLASHCAL (Flash Calcination) process has been developed for the production of single- and mixed-oxide powders of fuel metals (uranium, thorium and plutonium) from the respective nitrate solutions using the Whiteshell Roto-Spray Calciner (RSC). The metal-nitrate feed solution, either by itself or mixed with a suitable chemical reactant or additive, is converted to its oxide powder in the RSC at temperatures between 300 and 600 0 C. Rapid denitration takes place in the calciner, yielding the metal-oxide powders while simultaneously destroying any excess chemical additive and reaction by-products. In the production of precursor oxide powders suitable for fuel fabrication, the FLASHCAL process has advantages over batch calcination and other processes that involve precipitation and filtration steps because fewer processing and handling operations are needed. Results obtained with thorium nitrate and uranium nitrate-thorium nitrate mixtures indicate that some measure of control over the size distribution and morphology of the oxide product powders is possible in this process with the proper selection of chemical additive, as well as the operating parameters of the calciner
The manufacture of LEU fuel elements at Dounreay
Energy Technology Data Exchange (ETDEWEB)
Gibson, J.
1997-08-01
Two LEU test elements are being manufactured at Dounreay for test irradiation in the HFR at Petten, The Netherlands. This paper describes the installation of equipment and the development of the fabrication and inspection techniques necessary for the manufacture of LEU fuel plates. The author`s experience in overcoming the technical problems of stray fuel particles, dog-boning, uranium homogeneity and the measurement of uranium distribution is also described.
Fabrication of FFTF fuel pin wire wrap
International Nuclear Information System (INIS)
Epperson, E.M.
1980-06-01
Lateral spacing between FFTF fuel pins is required to provide a passageway for the sodium coolant to flow over each pin to remove heat generated by the fission process. This spacing is provided by wrapping each fuel pin with type 316 stainless steel wire. This wire has a 1.435mm (0.0565 in.) to 1.448mm (0.0570 in.) diameter, contains 17 +- 2% cold work and was fabricated and tested to exacting RDT Standards. About 500 kg (1100 lbs) or 39 Km (24 miles) of fuel pin wrap wire is used in each core loading. Fabrication procedures and quality assurance tests are described
Fabrication of internally instrumented reactor fuel rods
International Nuclear Information System (INIS)
Schmutz, J.D.; Meservey, R.H.
1975-01-01
Procedures are outlined for fabricating internally instrumented reactor fuel rods while maintaining the original quality assurance level of the rods. Instrumented fuel rods described contain fuel centerline thermocouples, ultrasonic thermometers, and pressure tubes for internal rod gas pressure measurements. Descriptions of the thermocouples and ultrasonic thermometers are also contained
U.S. technology for mechanized/automated fabrication of fast reactor fuel
International Nuclear Information System (INIS)
Nyman, D.H.; Bennett, D.W.; Claudson, T.T.; Dahl, R.E.; Graham, R.A.; Keating, J.J.; Yatabe, J.M.
1978-01-01
The status of the U.S. fast reactor Fuel Fabrication Development Program is discussed. The objectives of the program are to develop and evaluate a high throughput pilot fuel fabrication line including close-coupled chemistry and wet scrap recycle operations. The goals of the program are to demonstrate by mechanized/automated and remote processes: reduced personnel exposure, enhanced safegurads/accountability, improved fuel performance, representative fabrication rates and reduced fuel costs
Penetrate-leach dissolution of zirconium-clad uranium and uranium dioxide fuels
International Nuclear Information System (INIS)
Harmon, H.D.
1975-01-01
A new decladding-dissolution process was developed for zirconium-clad uranium metal and UO 2 fuels. The proposed penetrate-leach process consists of penetrating the zirconium cladding with Alniflex solution (2M HF--1M HNO 3 --1M Al(NO 3 ) 3 --0.1M K 2 Cr 2 O 7 ) and of leaching the exposed core with 10M HNO 3 . Undissolved cladding pieces are discarded as solid waste. Periodic HF and HNO 3 additions, efficient agitation, and in-line zirconium analyses are required for successful control of ZrF 4 and/or AlF 3 precipitation during the cladding-penetration step. Preliminary solvent extraction studies indicated complete recovery of uranium with 30 vol. percent tributyl phosphate (TBP) from both Alniflex solution and blended Alniflex-HNO 3 leach solutions. With 7.5 vol. percent TBP, high extractant/feed flow ratios and low scrub flows are required for satisfactory uranium recovery from Alniflex solution. Modified waste-handling procedures may be required for Alniflex waste, because it cannot be evaporated before neutralization and large quantities of solids are generated on neutralization. The effect of unstable UZr 3 (epsilon phase of uranium-zirconium system) on the safety of penetrate-leach dissolution was investigated
International Nuclear Information System (INIS)
1984-08-01
NRC is committed to the periodic publication of licensed fuel facilities inventory difference data, following agency review of the information and completion of any related investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, plutonium, or uranium-233
International Nuclear Information System (INIS)
1983-07-01
NRC is committed to the periodic publication of licensed fuel facilities inventory difference data, following agency review of the information and completion of any related investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, plutonium, or uranium-233
International Nuclear Information System (INIS)
1985-10-01
NRC is committed to the periodic publication of licensed fuel facilities' inventory difference data, following agency review of the information and completion of any related investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, plutonium, or uranium-233
Destruction of plutonium using non-uranium fuels in pressurized water reactor peripheral assemblies
International Nuclear Information System (INIS)
Chodak, P. III
1996-05-01
This thesis examines and confirms the feasibility of using non-uranium fuel in a pressurized water reactor (PWR) radial blanket to eliminate plutonium of both weapons and civilian origin. In the equilibrium cycle, the periphery of the PWR is loaded with alternating fresh and once burned non-uranium fuel assemblies, with the interior of the core comprised of conventional three batch UO 2 assemblies. Plutonium throughput is such that there is no net plutonium production: production in the interior is offset by destruction in the periphery. Using this approach a 50 MT WGPu inventory could be eliminated in approximately 400 reactor years of operation. Assuming all other existing constraints were removed, the 72 operating US PWRs could disposition 50 MT of WGPu in 5.6 years. Use of a low fissile loading plutonium-erbium inert-oxide-matrix composition in the peripheral assemblies essentially destroys 100% of the 239 Pu and ≥90% total Pu over two 18 month fuel cycles. Core radial power peaking, reactivity vs EFPD profiles and core average reactivity coefficients were found to be comparable to standard PWR values. Hence, minimal impact on reload licensing is anticipated. Examination of potential candidate fuel matrices based on the existing experience base and thermo-physical properties resulted in the recommendation of three inert fuel matrix compositions for further study: zirconia, alumina and TRISO particle fuels. Objective metrics for quantifying the inherent proliferation resistance of plutonium host waste and fuel forms are proposed and were applied to compare the proposed spent WGPu non-uranium fuel to spent WGPu MOX fuels and WGPu borosilicate glass logs. The elimination disposition option spent non-uranium fuel product was found to present significantly greater barriers to proliferation than other plutonium disposal products
Destruction of plutonium using non-uranium fuels in pressurized water reactor peripheral assemblies
Energy Technology Data Exchange (ETDEWEB)
Chodak, III, Paul [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)
1996-05-01
This thesis examines and confirms the feasibility of using non-uranium fuel in a pressurized water reactor (PWR) radial blanket to eliminate plutonium of both weapons and civilian origin. In the equilibrium cycle, the periphery of the PWR is loaded with alternating fresh and once burned non-uranium fuel assemblies, with the interior of the core comprised of conventional three batch UO2 assemblies. Plutonium throughput is such that there is no net plutonium production: production in the interior is offset by destruction in the periphery. Using this approach a 50 MT WGPu inventory could be eliminated in approximately 400 reactor years of operation. Assuming all other existing constraints were removed, the 72 operating US PWRs could disposition 50 MT of WGPu in 5.6 years. Use of a low fissile loading plutonium-erbium inert-oxide-matrix composition in the peripheral assemblies essentially destroys 100% of the 239Pu and ≥90% {sub total}Pu over two 18 month fuel cycles. Core radial power peaking, reactivity vs EFPD profiles and core average reactivity coefficients were found to be comparable to standard PWR values. Hence, minimal impact on reload licensing is anticipated. Examination of potential candidate fuel matrices based on the existing experience base and thermo-physical properties resulted in the recommendation of three inert fuel matrix compositions for further study: zirconia, alumina and TRISO particle fuels. Objective metrics for quantifying the inherent proliferation resistance of plutonium host waste and fuel forms are proposed and were applied to compare the proposed spent WGPu non-uranium fuel to spent WGPu MOX fuels and WGPu borosilicate glass logs. The elimination disposition option spent non-uranium fuel product was found to present significantly greater barriers to proliferation than other plutonium disposal products.
Characterization of uranium silicide powder using XRD
International Nuclear Information System (INIS)
Garcia, Rafael H.L.; Saliba-Silva, Adonis M.; Carvalho, Elita F.U.; Lima, Nelson B.; Ichikawa, Rodrigo U.; Martinez, Luiz G.
2013-01-01
Uranium silicide (U 3 Si 2 ) is an intermetallic used as nuclear fuel in most modern MTR - Materials Test Reactor. Dispersed in aluminum, this fuel allows high uranium densities, up to 4.8 gU/cm 3 . At IPEN, the fabrication of fuel elements based on U 3 Si 2 for the IEA-R1 reactor is carried out in the Nuclear Fuel Center (CCN), by vacuum induction melting of uranium and silicon, followed by grinding. Before employed in a nuclear reactor, U 3 Si 2 must be submitted to a strict quality control, which includes granulometry, density, X-ray radiography for dispersion homogeneity, chemical and crystallographic characterization. Concerning phase composition for a qualified fuel, the fraction of U 3 Si 2 should be higher than 80wt.%. Aiming at the development of a routine methodology for quantification of phases via analysis of XRD data using the Rietved method, six samples from two production baths of CCN were submitted to X-ray diffraction. The data were analyzed using software GSAS and line profile analysis methods. The results suggest that fusion product have preferred orientation and grinding step is important for a better refinement. (author)
International Nuclear Information System (INIS)
Wehner, E.; Sohnius, B.
1992-01-01
This research work is aimed at the assessment of new procedures in the framework of the decommissioning of the NUKEM-A facility, a plant used for over 30 years for the fabrication of Material Test Reactor (MTR) and Thorium High Temperature Reactor (THTR) fuel elements. Important issues in this work are the preparation of detailed uranium and thorium contamination distribution maps in walls and floors, the execution of various dismantling and decontamination operations under health physics control, the large-scale treatment of arising primary waste and the minimization of secondary waste
International Nuclear Information System (INIS)
Borgonovi, G.M.; McCartin, T.J.; McDaniel, T.; Miller, C.L.; Nguyen, T.
1978-01-01
This report describes the measurement techniques, the instrumentation, and the procedures used in accountability and control of nuclear materials, as they apply to fuel fabrication facilities. A general discussion is given of instrumentation and measurement techniques which are presently used being considered for fuel fabrication facilities. Those aspects which are most significant from the point of view of satisfying regulatory constraints have been emphasized. Sensors and measurement devices have been discussed, together with their interfacing into a computerized system designed to permit real-time data collection and analysis. Estimates of accuracy and precision of measurement techniques have been given, and, where applicable, estimates of associated costs have been presented. A general description of material control and accounting is also included. In this section, the general principles of nuclear material accounting have been reviewed first (closure of material balance). After a discussion of the most current techniques used to calculate the limit of error on inventory difference, a number of advanced statistical techniques are reviewed. The rest of the section deals with some regulatory aspects of data collection and analysis, for accountability purposes, and with the overall effectiveness of accountability in detecting diversion attempts in fuel fabrication facilities. A specific example of application of the accountability methods to a model fuel fabrication facility is given. The effect of random and systematic errors on the total material uncertainty has been discussed, together with the effect on uncertainty of the length of the accounting period
Energy Technology Data Exchange (ETDEWEB)
Borgonovi, G.M.; McCartin, T.J.; McDaniel, T.; Miller, C.L.; Nguyen, T.
1978-01-01
This report describes the measurement techniques, the instrumentation, and the procedures used in accountability and control of nuclear materials, as they apply to fuel fabrication facilities. A general discussion is given of instrumentation and measurement techniques which are presently used being considered for fuel fabrication facilities. Those aspects which are most significant from the point of view of satisfying regulatory constraints have been emphasized. Sensors and measurement devices have been discussed, together with their interfacing into a computerized system designed to permit real-time data collection and analysis. Estimates of accuracy and precision of measurement techniques have been given, and, where applicable, estimates of associated costs have been presented. A general description of material control and accounting is also included. In this section, the general principles of nuclear material accounting have been reviewed first (closure of material balance). After a discussion of the most current techniques used to calculate the limit of error on inventory difference, a number of advanced statistical techniques are reviewed. The rest of the section deals with some regulatory aspects of data collection and analysis, for accountability purposes, and with the overall effectiveness of accountability in detecting diversion attempts in fuel fabrication facilities. A specific example of application of the accountability methods to a model fuel fabrication facility is given. The effect of random and systematic errors on the total material uncertainty has been discussed, together with the effect on uncertainty of the length of the accounting period.
Optimization of fuel cycle strategies with constraints on uranium availability
International Nuclear Information System (INIS)
Silvennoinen, P.; Vira, J.; Westerberg, R.
1982-01-01
Optimization of nuclear reactor and fuel cycle strategies is studied under the influence of reduced availability of uranium. The analysis is separated in two distinct steps. First, the global situation is considered within given high and low projections of the installed capacity up to the year 2025. Uranium is regarded as an exhaustible resource whose production cost would increase proportionally to increasing cumulative exploitation. Based on the estimates obtained for the uranium cost, a global strategy is derived by splitting the installed capacity between light water reactor (LWR) once-through, LWR recycle, and fast breeder reactor (FBR) alternatives. In the second phase, the nuclear program of an individual utility is optimized within the constraints imposed from the global scenario. Results from the global scenarios indicate that in a reference case the uranium price would triple by the year 2000, and the price escalation would continue throughout the planning period. In a pessimistic growth scenario where the global nuclear capacity would not exceed 600 GW(electric) in 2025, the uranium price would almost double by 2000. In both global scenarios, FBRs would be introduced, in the reference case after 2000 and in the pessimistic case after 2010. In spite of the increases in the uranium prices, the levelized power production cost would increase only by 45% up to 2025 in the utility case provided that the plutonium is incinerated as a substitute fuel
Development of the fabrication technology of the simulated DUPIC fuel
International Nuclear Information System (INIS)
Kang, Kweon Ho; Yang, M. S.; Bae, K. K. and others
2000-06-01
It is important to get basic data to analysis physical properties, behavior in reactor and performance of the DUPIC fuel because physical properties of the DUPIC fuel is different from the commercial UO 2 fuel. But what directly measures physical properties et al. of DUPIC fuel being resinterred simulated spent fuel through OREOX process is very difficult in laboratory owing to its high level radiation. Then fabrication of simulated DUPIC fuel is needed to measure its properties. In this study, processes on powder treatment, OREOX, compaction and sintering to fabricate simulated DUPIC fuel using simulated spent fuel are discribed. To fabricate simulated DUPIC fuel, the powder from 3 times OREOX and 5 times attrition milling simulated spent fuel is compacted with 1.3 ton/cm 2 . Pellets are sintered in 100% H 2 atmosphere over 10 h at 1800 deg C. Sintered densities of pellets are 10.2-10.5 g/cm 3
Implications of using alternate fuel cycles to meet Ontario's nuclear power demand
International Nuclear Information System (INIS)
Lau, J.H.K.
1978-08-01
The use of alternate fuel cycles to meet an assumed nuclear capacity growth rate in Ontario is examined. Two criteria are used: the ability of the alternate fuel cycles to lessen the uranium demand; and the ease of commercialization. The nuclear strategies considered assume the use of the natural uranium cycle and, starting in the year 2000, the gradual introduction of an alternate fuel cycle. The alternate fuel cycles reviewed are enriched uranium, mixed oxides, and a variety of thorium cycles. The cumulative uranium requirement to the year 2070, and the growth and size of the reprocessing and fuel fabrication industries are discussed in detail. Sensitivity analyses on nuclear capacity growth rate, recycling loss and delay time are also described. (auth)
Fuel fabrication and post-irradiation examination
Energy Technology Data Exchange (ETDEWEB)
Venter, P J; Aspeling, J C [Atomic Energy Corporation of South Africa Ltd., Pretoria (South Africa)
1990-06-01
This paper provides an overview of the A/c's Bevan and Eldopar facilities for the fabrication of nuclear fuel. It also describes the sophisticated Hot Cell Complex, which is capable of accommodating pressurised water reactor fuel and various other irradiated samples. Some interesting problems and their solutions are discussed. (author)
Fuel fabrication and post-irradiation examination
International Nuclear Information System (INIS)
Venter, P.J.; Aspeling, J.C.
1990-01-01
This paper provides an overview of the A/c's Bevan and Eldopar facilities for the fabrication of nuclear fuel. It also describes the sophisticated Hot Cell Complex, which is capable of accommodating pressurised water reactor fuel and various other irradiated samples. Some interesting problems and their solutions are discussed. (author)
Uranium-plutonium fuel for fast reactors
International Nuclear Information System (INIS)
Antipov, S.A.; Astafiev, V.A.; Clouchenkov, A.E.; Gustchin, K.I.; Menshikova, T.S.
1996-01-01
Technology was established for fabrication of MOX fuel pellets from co-precipitated and mechanically blended mixed oxides. Both processes ensure the homogeneous structure of pellets readily dissolvable in nitric acid upon reprocessing. In order to increase the plutonium charge in a reactor-burner a process was tested for producing MOX fuel with higher content of plutonium and an inert diluent. It was shown that it is feasible to produce fuel having homogeneous structure and the content of plutonium up to 45% mass
International Nuclear Information System (INIS)
Narasimha Murty, B.; Bharati Misra, U.; Yadav, R.B.; Srivastava, R.K.
2005-01-01
This paper describes quantitatively the effect of in-situ generated uranium metal (that could be formed due to the conducive manufacturing conditions) in a sintered uranium dioxide fuel pellet on the experimentally determined O/U ratio using analytical methods involving dissolution of the pellet material. To quantify the effect of in-situ generated uranium metal in the fuel pellet, a mathematical expression is derived for the actual O/U ratio in terms of the O/U ratio as determined by an experiment involving dissolution of the material and the quantity of uranium metal present in the uranium dioxide pellet. The utility of this derived mathematical expression is demonstrated by tabulating the calculated actual O/U ratios for varying amounts of uranium metal (from 5 to 95% in 5% intervals) and different O/U ratio values (from 2.001 to 2.015 in 0.001 intervals). This paper brings out the necessity of care to be exercised while interpreting the experimentally determined O/U ratio and emphasizes the fact that it is always safer to produce the nuclear fuel with oxygen to uranium ratios well below the specified maximum limit of 2.015. (author)
Safety aspects of LWR fuel reprocessing and mixed oxide fuel fabrication plants
International Nuclear Information System (INIS)
Fischer, M.; Leichsenring, C.H.; Herrmann, G.W.; Schueller, W.; Hagenberg, W.; Stoll, W.
1977-01-01
The paper is focused on the safety and the control of the consequences of credible accidents in LWR fuel reprocessing plants and in mixed oxide fuel fabrication plants. Each of these plants serve for many power reactor (about 50.000 Mwel) thus the contribution to the overall risk of nuclear energy is correspondingly low. Because of basic functional differences between reprocessing plants, fuel fabrication plants and nuclear power reactors, the structure and safety systems of these plants are different in many respects. The most important differences that influence safety systems are: (1) Both fuel reprocessing and fabrication plants do not have the high system pressure that is associated with power reactors. (2) A considerable amount of the radioactivity of the fuel, which is in the form of short-lived radionuclides has decayed. Therefore, fuel reprocessing plants and mixed oxide fuel fabrication plants are designed with multiple confinement barriers for control of radioactive materials, but do not require the high-pressure containment systems that are used in LWR plants. The consequences of accidents which may lead to the dispersion of radioactive materials such as chemical explosions, nuclear excursions, fires and failure of cooling systems are considered. A reasonable high reliability of the multiple confinement approach can be assured by design. In fuel reprocessing plants, forced cooling is necessary only in systems where fission products are accumulated. However, the control of radioactive materials can be maintained during normal operation and during the above mentioned accidents, if the dissolver off-gas and vessel off-gas treatment systems provide for effective removal of radioactive iodine, radioactive particulates, nitrogen oxides, tritium and krypton 85. In addition, the following incidents in the dissolver off-gas system itself must be controlled: failures of iodine filters, hydrogen explosion in O 2 - and NOsub(x)-reduction component, decomposition of
Advanced fuel cycles: a rationale and strategy for adopting the low-enriched-uranium fuel cycle
International Nuclear Information System (INIS)
James, R.A.
1980-01-01
A two-year study of alternatives to the natural uranium fuel cycle in CANDU reactors is summarized. The possible advanced cycles are briefly described. Selection criteria for choosing a cycle for development include resource utilization, economics, ease of implementaton, and social acceptability. It is recommended that a detailed study should be made with a view to the early implementation of the low-enriched uranium cycle. (LL)
International Nuclear Information System (INIS)
Lucas, N.
1978-01-01
The French energy plan is described under the following heads: structure of supply/use; nuclear energy; oil consumption; coal consumption; gas consumption; energy conservation; renewable resources. The section on nuclear energy covers: nuclear programmes, uranium supplies (mining and enrichment), fuel fabrication and reprocessing, finance and economics, political and administrative aspects, nuclear trade, development of fast breeder reactor. (U.K.)
Processing of irradiated, enriched uranium fuels at the Savannah River Plant
Energy Technology Data Exchange (ETDEWEB)
Hyder, M L; Perkins, W C; Thompson, M C; Burney, G A; Russell, E R; Holcomb, H P; Landon, L F
1979-04-01
Uranium fuels containing /sup 235/U at enrichments from 1.1% to 94% are processed and recovered, along with neptunium and plutonium byproducts. The fuels to be processed are dissolved in nitric acid. Aluminum-clad fuels are disssolved using a mercury catalyst to give a solution rich in aluminum. Fuels clad in more resistant materials are dissolved in an electrolytic dissolver. The resulting solutions are subjected to head-end treatment, including clarification and adjustment of acid and uranium concentration before being fed to solvent extraction. Uranium, neptunium, and plutonium are separated from fission products and from one another by multistage countercurrent solvent extraction with dilute tri-n-butyl phosphate in kerosene. Nitric acid is used as the salting agent in addition to aluminum or other metal nitrates present in the feed solution. Nuclear safety is maintained through conservative process design and the use of monitoring devices as secondary controls. The enriched uranium is recovered as a dilute solution and shipped off-site for further processing. Neptunium is concentrated and sent to HB-Line for recovery from solution. The relatively small quantities of plutonium present are normally discarded in aqueous waste, unless the content of /sup 238/Pu is high enough to make its recovery desirable. Most of the /sup 238/Pu can be recovered by batch extraction of the waste solution, purified by counter-current solvent extraction, and converted to oxide in HB-Line. By modifying the flowsheet, /sup 239/Pu can be recovered from low-enriched uranium in the extraction cycle; neptunium is then not recovered. The solvent is subjected to an alkaline wash before reuse to remove degraded solvent and fission products. The aqueous waste is concentrated and partially deacidified by evaporation before being neutralized and sent to the waste tanks; nitric acid from the overheads is recovered for reuse.
Processing of irradiated, enriched uranium fuels at the Savannah River Plant
International Nuclear Information System (INIS)
Hyder, M.L.; Perkins, W.C.; Thompson, M.C.; Burney, G.A.; Russell, E.R.; Holcomb, H.P.; Landon, L.F.
1979-04-01
Uranium fuels containing 235 U at enrichments from 1.1% to 94% are processed and recovered, along with neptunium and plutonium byproducts. The fuels to be processed are dissolved in nitric acid. Aluminum-clad fuels are disssolved using a mercury catalyst to give a solution rich in aluminum. Fuels clad in more resistant materials are dissolved in an electrolytic dissolver. The resulting solutions are subjected to head-end treatment, including clarification and adjustment of acid and uranium concentration before being fed to solvent extraction. Uranium, neptunium, and plutonium are separated from fission products and from one another by multistage countercurrent solvent extraction with dilute tri-n-butyl phosphate in kerosene. Nitric acid is used as the salting agent in addition to aluminum or other metal nitrates present in the feed solution. Nuclear safety is maintained through conservative process design and the use of monitoring devices as secondary controls. The enriched uranium is recovered as a dilute solution and shipped off-site for further processing. Neptunium is concentrated and sent to HB-Line for recovery from solution. The relatively small quantities of plutonium present are normally discarded in aqueous waste, unless the content of 238 Pu is high enough to make its recovery desirable. Most of the 238 Pu can be recovered by batch extraction of the waste solution, purified by counter-current solvent extraction, and converted to oxide in HB-Line. By modifying the flowsheet, 239 Pu can be recovered from low-enriched uranium in the extraction cycle; neptunium is then not recovered. The solvent is subjected to an alkaline wash before reuse to remove degraded solvent and fission products. The aqueous waste is concentrated and partially deacidified by evaporation before being neutralized and sent to the waste tanks; nitric acid from the overheads is recovered for reuse
Processing used nuclear fuel with nanoscale control of uranium and ultrafiltration
Energy Technology Data Exchange (ETDEWEB)
Wylie, Ernest M.; Peruski, Kathryn M.; Prizio, Sarah E. [Department of Civil and Environmental Engineering and Earth Sciences, University of Notre Dame, Notre Dame, IN 46556 (United States); Bridges, Andrea N.A.; Rudisill, Tracy S.; Hobbs, David T. [Savannah River National Laboratory, Aiken, SC 29808 (United States); Phillip, William A. [Department of Chemical and Biomolecular Engineering, University of Notre Dame, Notre Dame, IN 46556 (United States); Burns, Peter C., E-mail: pburns@nd.edu [Department of Civil and Environmental Engineering and Earth Sciences, University of Notre Dame, Notre Dame, IN 46556 (United States); Department of Chemistry and Biochemistry, University of Notre Dame, Notre Dame, IN 46556 (United States)
2016-05-15
Current separation and purification technologies utilized in the nuclear fuel cycle rely primarily on liquid–liquid extraction and ion-exchange processes. Here, we report a laboratory-scale aqueous process that demonstrates nanoscale control for the recovery of uranium from simulated used nuclear fuel (SIMFUEL). The selective, hydrogen peroxide induced oxidative dissolution of SIMFUEL material results in the rapid assembly of persistent uranyl peroxide nanocluster species that can be separated and recovered at moderate to high yield from other process-soluble constituents using sequestration-assisted ultrafiltration. Implementation of size-selective physical processes like filtration could results in an overall simplification of nuclear fuel cycle technology, improving the environmental consequences of nuclear energy and reducing costs of processing. - Highlights: • Nanoscale control in irradiated fuel reprocessing. • Ultrafiltration to recover uranyl cage clusters. • Alternative to solvent extraction for uranium purification.
Development of vacuum continuous casting technology for uranium
International Nuclear Information System (INIS)
Lee, Y.S.; Kim, C. K.; Kim, K. H.; Lee, D. B.; Kim, J. D.; Jang, S. J.; Ahn, H. S.; Shin, Y. J.
2001-02-01
The spent fuel disposal process of new dry storage concept has been developed in KAERI, in which the uranium metal abstracted by Li-reduction of spent fuel will be formed to long rods and then the rods will be arranged uniformly in canister. The objective of this study is to review the feasibility of applying the continuous casting method to cast a long rod with modifying the vacuum high-frequency induction furnace to vacuum continuous casting system, which was normally used to cast the uranium. The results are as follows. With the nozzle size of 3mm and the withdrawal speed of 3.5 mm/sec, the length of 160mm, diameter of 30 mm continuous casting uranium bar was successfully cast. This result shows there might be a possibility of continuous casting of uranium and helps the design and fabrication of new continuous casting equipment
LEU fuel fabrication program for the RECH-1 reactor. Status report
International Nuclear Information System (INIS)
Chavez, J.C.; Barrera, M.; Jimenez, O.; Lisboa, J.; Marin, J.
2000-01-01
In 1995 a 50 LEU U 3 Si 2 fuel elements fabrication program for the RECH-1 research reactor was established at the Comision Chilena de Energia Nuclear, CCHEN. After a fabrication process qualification stage, in 1998, four elements were early delivered to the reactor in order to start an irradiation qualification stage. The irradiation has reached an estimated 10% burn-up and no fabrication problems have been detected up to this burn-up level. During 1999 and up to the first quarter of 2000, 19 fuel elements were produced and 7 fuel elements are expected for the end of 2000. This report presents an updated summary of the main results obtained in this fuel fabrication program. A summary of other activities generated by this program, such as in core follow-up of the four leader fuel elements, ISO 9001 implementation for the fabrication process and a fabrication and qualification optimization planning, is also presented here. (author)
IAEA Activities on Uranium Resources and Production, and Databases for the Nuclear Fuel Cycle
Energy Technology Data Exchange (ETDEWEB)
Ganguly, C.; Slezak, J. [Divison of Nuclear Fuel Cycle and Waste Technology, International Atomic Energy Agency, Vienna (Austria)
2014-05-15
In recent years rising expectation for nuclear power has led to a significant increase in the demand for uranium and in turn dramatic increases in uranium exploration, mining and ore processing activities worldwide. Several new countries, often with limited experience, have also embarked on these activities. The ultimate goal of the uranium raw material industry is to provide an adequate supply of uranium that can be delivered to the market place at a competitive price by environmentally sound, mining and milling practices. The IAEA’s programme on uranium raw material encompass all aspects of uranium geology and deposits, exploration, resources, supply and demand, uranium mining and ore processing, environmental issues in the uranium production cycle and databases for the uranium fuel cycle. Radiological safety and environmental protection are major challenges in uranium mines and mills and their remediation. The IAEA has revived its programme for the Uranium Production Site Appraisal Team (UPSAT) to assist Member States to improve operational and safety performances at uranium mines and mill sites. The present paper summarizes the ongoing activities of IAEA on uranium raw material, highlighting the status of global uranium resources, their supply and demand, the IAEA database on world uranium deposit (UDEPO) and nuclear fuel cycle information system (NFCIS), recent IAEA Technical Meetings (TM) and related ongoing Technical Cooperation (TC) projects. (author)
Uranium chloride extraction of transuranium elements from LWR fuel
International Nuclear Information System (INIS)
Miller, W.E.; Ackerman, J.P.; Battles, J.E.; Johnson, T.R.; Pierce, R.D.
1992-01-01
A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels containing rare earth and noble metal fission products as well as other fission products is disclosed. The oxide fuel is reduced with Ca metal in the presence of Ca chloride and a U-Fe alloy which is liquid at about 800 C to dissolve uranium metal and the noble metal fission product metals and transuranium actinide metals and rare earth fission product metals leaving Ca chloride having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein. The Ca chloride and CaO and the fission products contained therein are separated from the U-Fe alloy and the metal values dissolved therein. The U-Fe alloy having dissolved therein reduced metals from the spent nuclear fuel is contacted with a mixture of one or more alkali metal or alkaline earth metal halides selected from the class consisting of alkali metal or alkaline earth metal and Fe or U halide or a combination thereof to transfer transuranium actinide metals and rare earth metals to the halide salt leaving the uranium and some noble metal fission products in the U-Fe alloy and thereafter separating the halide salt and the transuranium metals dissolved therein from the U-Fe alloy and the metals dissolved therein. 1 figure
The conversion of NRU from HEU to LEU fuel
International Nuclear Information System (INIS)
Sears, D.F.; Atfield, M.D.; Kennedy, I.C.
1990-01-01
The program at Chalk River Nuclear Laboratories (CRNL) to develop and test low-enriched uranium fuel (LEU, 3 Si, USiAl, USi Al and U 3 Si 2 (U-3.96 wt% Si; U-3.5 wt% Si-1.5 wt% AL; U-3.2 wt%; Si-3 wt% Al; U-7.3 wt% Si, respectively). Fuel elements were fabricated with uranium loadings suitable for NRU, 3.15 gU/cm 3 , and for NRX, 4.5 gU/cm 3 , and were irradiated under normal fuel-operating conditions. Eight experimental irradiations involving 100 mini-elements and 84 full-length elements (7X12-element rods) were completed to qualify the LEU fuel and the fabrication technology. Post irradiation examinations confirmed that the performance of the LEU fuel, and that of a medium enrichment uranium (MEU, 45% U-235) alloy fuel tested as a back-up, was comparable to the HEU fuel. The uranium silicide dispersion fuel swelling was approximately linear up to burnups exceeding NRU's design terminal burnup (80 at%). NRU was partially converted to LEU fuel when the first 31 prototype fuel rods manufactured with industrial scale production equipment were installed in the reactor. The rods were loaded in NRU at a fuelling rate of about two rods per week over the period 1988 September to December. This partial LEU core (one third of a full NRU core) has allowed the reactor engineers and physicists to evaluate the bulk effects of the LEU conversion on NRU operations. As expected, the irradiation is proceeding without incident
Control in fabrication of PWR and BWR type reactor fuel elements
International Nuclear Information System (INIS)
Gorskij, V.V.
1981-01-01
Both destructive and non-destructive testing methods now in use in fabrication of BWR and PWR type reactor fuel elements at foreign plants are reviewed. Technological procedures applied in fabrication of fuel elements and fuel assemblies are described. Major attention is paid to radiographic, ultrasonic, metallographic, visual and autoclavic testings. A correspondence of the methods applied to the ASTM standards is discussed. The most part of the countries are concluded the apply similar testing methods enabling one to reliably evaluate the quality of primary materials and fabricated fuel elements and thus meeting the demands to contemporary PWR and BWR type reactor fuel elements. Practically all fuel element and pipe fabrication plants in Western Europe, Asia and America use the ASTM standards as the basis for the quality contr [ru
Recycling of MOX fuel for LWRs
International Nuclear Information System (INIS)
Joo, Hyung Kook; Oh, Soo Youl
1992-01-01
The status and issues related to the thermal recycling of reprocessed nuclear fuels have been reviewed. It is focused on the use of reprecessed plutonium in the form of mixed oxide (MOX) for a light water reactor and the review on reprocessing and fabrication processes is beyond the scope. In spite of the difference in the nuclear characteristics between plutonium and uranium isotopes, the neutronics behavior in a core with MOX fuels is similar to that with normal uranium fuels. However, since the neutron spectrum is hardened in a core with MOX, the Doppler, viod, and moderator temperature coefficients become more negative and the control rod and boron worths are slightly reduced. Therefore, the safety will be evaluated carefully in addition to the core neutronics analysis. The MOX fuel rod behavior related to the rod performance such as the pellet to clad interaction and fission gas release is also similar to that of uranium rods, and no specific problem arises. Substituting MOX fuels for a portion of uranium fuels, it is estimated that the savings be about 25% in uranium ore and 10% in uranium enrichment service requirements. The use of MOX fuel in LWRs has been commercialized in European countries including Germany, France, Belgium, etc., and a demonstration program has been pursued in Japan for the commercial utilization in the late 1990s. Such a worldwide trend indicates that the utilization of MOX fuel in LWRs is a proven technology and meets economics criteria. (Author)
Uranium resource utilization improvements in the once-through PWR fuel cycle
International Nuclear Information System (INIS)
Matzie, R.A.
1980-04-01
In support of the Nonproliferation Alternative Systems Assessment Program (NASAP), Combustion Engineering, Inc. performed a comprehensive analytical study of potential uranium utilization improvement options that can be backfit into existing PWRs operating on the once-through uranium fuel cycle. A large number of potential improvement options were examined as part of a preliminary survey of candidate options. The most attractive of these, from the standpoint of uranium utilization improvement, economic viability, and ease of implementation, were then selected for detailed analysis and were included in a single composite improvement case. This composite case represents an estimate of the total savings in U 3 O 8 consumption that can be achieved in current-design PWRs by implementing improvements which can be developed and demonstrated in the near term. The improvement options which were evaluated in detail and included in the composite case were a new five-batch, extended-burnup fuel management scheme, low-leakage fuel management, modified lattice designs, axial blankets, reinsertion of initial core batches, and end-of-cycle stretchout
Water reactor fuel element fabrication, with special emphasis on its effects on fuel performance
International Nuclear Information System (INIS)
1979-01-01
Full text: The performance of nuclear fuel has improved over the years and is now a minor cause of outages and of power limitations in nuclear power plants. On the other hand, an increasing number of countries are in the process of developing or implementing their own capability for manufacturing fuel elements. In this context, the IAEA International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) advised that a symposium be organized devoted to the relationship between fuel fabrication and performance The Czechoslovak Atomic Energy Commission agreed to co-operate in the organization of this symposium and to host it in Prague. Those factors which influence fuel fabrication requirements are now well ascertained: as little reactor primary circuit contamination as possible, the tendency to increased burnups, reactor manoeuverability to match power grid demands, the desirability of an autonomous fabrication capability. It is the general experience of fuel element suppliers that fuel quality and performance has increased over the years, the importance of quality assurance and process monitoring has been decisive in this respect The ever increasing mass-production aspect of nuclear fuel leads to some processing steps being revised and alternatives being developed. The relation between fabrication processes and fuel performance characteristics, although generally well perceived, are still the subject of a large amount of experiment and assessment in most countries, both industrial and developing This evidence is most encouraging; it means indeed that nuclear power, which is already amongst the cheapest and safest sources of energy, will continue to be improved. The performance of Zircaloy fuel cladding - presently the material used in most water reactors - is under particular consideration. Better understanding of this quite recent alloy will pave the way for broader fuel utilization limits in the future. The panel discussion, which noted some
Energy Technology Data Exchange (ETDEWEB)
Cook, David Howard [ORNL; Freels, James D [ORNL; Ilas, Germina [ORNL; Jolly, Brian C [ORNL; Miller, James Henry [ORNL; Primm, Trent [ORNL; Renfro, David G [ORNL; Sease, John D [ORNL; Pinkston, Daniel [ORNL
2011-02-01
This report documents progress made during FY 2010 in studies of converting the High Flux Isotope Reactor (HFIR) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current level. Studies are reported of support to a thermal hydraulic test loop design, the implementation of finite element, thermal hydraulic analysis capability, and infrastructure tasks at HFIR to upgrade the facility for operation at 100 MW. A discussion of difficulties with preparing a fuel specification for the uranium-molybdenum alloy is provided. Continuing development in the definition of the fuel fabrication process is described.
Energy Technology Data Exchange (ETDEWEB)
Milosevic, S; Momcilovic, I [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)
1961-12-15
Based on the uranium microstructure in the fuel element, i.e. grain size, presence of other phases, porosity and the position of metal connections, some data could be obtained about the mechanical and thermal treatment of the uranium during fuel element fabrication. The mentioned characteristics could indicate the behaviour of the fuel under reactor operating conditions taking into account its stable dimensions and roughness of its surfaces. Na osnovu mikrostrukture urana u gorivnom elementu, odnosno velicine i oblika njegovih kristalnih zrna, prisustva drugih faza, pozoziteta kao i rasporeda metalnih ukljucaka, mogu se dobiti izvesni podaci o tome koje je postupke mehanickog i termickog tretiranja pretrpeo uran u pripremi za gorivni elemenat. Prethodne karakteristike takodje mogu da daju predstavu o tome kako ce se gorivo ponasati u uslovima rada reaktora, s obzirom na njegovu dimenzionu stabilnost i ogrubljivanje njegove povrsine (author)
Remote fabrication of breeder reactor fuel
International Nuclear Information System (INIS)
Gerber, E.W.; Hoitink, N.C.; Graham, R.A.
1984-06-01
The Secure Automated Fabrication (SAF) Line, a remotely operable plutonium fuel fabrication facility, incorporates advanced automation techniques. Included in the plant are 24 robots used to perform complex operations, and to enhance equipment standardization and ease of maintenance. Automated equipment is controlled remotely from centrally located supervisory computer control consoles or alternatively from control consoles dedicated to individual systems
International Nuclear Information System (INIS)
Choi, J; Ebbinghaus, B; Meiers, T; Ahn, J
2006-01-01
The US National Energy Policy of 2001 advocated the development of advanced fuel and fuel cycle technologies that are cleaner, more efficient, less waste-intensive, and more proliferation resistant. The need for advanced fuel development is emphasized in on-going DOE-supported programs, e.g., Global Nuclear Energy Initiative (GNEI), Advanced Fuel Cycle Initiative (AFCI), and GEN-IV Technology Development. The Directorates of Energy and Environment (E and E) and Chemistry and Material Sciences (C and MS) at Lawrence Livermore National Laboratory (LLNL) are interested in advanced fuel research and manufacturing using its multi-disciplinary capability and facilities to support a design concept of a small, secure, transportable, and autonomous reactor (SSTAR). The E and E and C and MS Directorates co-sponsored this Laboratory Directed Research and Development (LDRD) Project on Mono-Uranium Nitride Fuel Development for SSTAR and Space Applications. In fact, three out of the six GEN-IV reactor concepts consider using the nitride-based fuel, as shown in Table 1. SSTAR is a liquid-metal cooled, fast reactor. It uses nitride fuel in a sealed reactor vessel that could be shipped to the user and returned to the supplier having never been opened in its long operating lifetime. This sealed reactor concept envisions no fuel refueling nor on-site storage of spent fuel, and as a result, can greatly enhance proliferation resistance. However, the requirement for a sealed, long-life core imposes great challenges to research and development of the nitride fuel and its cladding. Cladding is an important interface between the fuel and coolant and a barrier to prevent fission gas release during normal and accidental conditions. In fabricating the nitride fuel rods and assemblies, the cladding material should be selected based on its the coolant-side corrosion properties, the chemical/physical interaction with the nitride fuel, as well as their thermal and neutronic properties. The US
International Nuclear Information System (INIS)
Sease, J.D.; Primm, R.T. III; Miller, J.H.
2007-01-01
The U.S. nonproliferation policy 'to minimize, and to the extent possible, eliminate the use of HEU in civil nuclear programs throughout the world' has resulted in the conversion (or scheduled conversion) of many of the U.S. research reactors from high-enriched uranium (HEU) to low-enriched uranium (LEU). A foil fuel appears to offer the best option for using a LEU fuel in the High Flux Isotope Reactor (HFIR) without degrading the performance of the reactor. The purpose of this document is to outline a proposed conceptual fabrication process flow sheet for a new, foil-type, 19.75%-enriched fuel for HFIR. The preparation of the flow sheet allows a better understanding of the costs of infrastructure modifications, operating costs, and implementation schedule issues associated with the fabrication of LEU fuel for HFIR. Preparation of a reference flow sheet is one of the first planning steps needed in the development of a new manufacturing capacity for low enriched fuels for U.S. research and test reactors. The flow sheet can be used to develop a work breakdown structure (WBS), a critical path schedule, and identify development needs. The reference flow sheet presented in this report is specifically for production of LEU foil fuel for the HFIR. The need for an overall reference flow sheet for production of fuel for all High Performance Research Reactors (HPRR) has been identified by the national program office. This report could provide a starting point for the development of such a reference flow sheet for a foil-based fuel for all HPRRs. The reference flow sheet presented is based on processes currently being developed by the national program for the LEU foil fuel when available, processes used historically in the manufacture of other nuclear fuels and materials, and processes used in other manufacturing industries producing a product configuration similar to the form required in manufacturing a foil fuel. The processes in the reference flow sheet are within the
Energy Technology Data Exchange (ETDEWEB)
Sease, J.D.; Primm, R.T. III; Miller, J.H.
2007-09-30
The U.S. nonproliferation policy 'to minimize, and to the extent possible, eliminate the use of HEU in civil nuclear programs throughout the world' has resulted in the conversion (or scheduled conversion) of many of the U.S. research reactors from high-enriched uranium (HEU) to low-enriched uranium (LEU). A foil fuel appears to offer the best option for using a LEU fuel in the High Flux Isotope Reactor (HFIR) without degrading the performance of the reactor. The purpose of this document is to outline a proposed conceptual fabrication process flow sheet for a new, foil-type, 19.75%-enriched fuel for HFIR. The preparation of the flow sheet allows a better understanding of the costs of infrastructure modifications, operating costs, and implementation schedule issues associated with the fabrication of LEU fuel for HFIR. Preparation of a reference flow sheet is one of the first planning steps needed in the development of a new manufacturing capacity for low enriched fuels for U.S. research and test reactors. The flow sheet can be used to develop a work breakdown structure (WBS), a critical path schedule, and identify development needs. The reference flow sheet presented in this report is specifically for production of LEU foil fuel for the HFIR. The need for an overall reference flow sheet for production of fuel for all High Performance Research Reactors (HPRR) has been identified by the national program office. This report could provide a starting point for the development of such a reference flow sheet for a foil-based fuel for all HPRRs. The reference flow sheet presented is based on processes currently being developed by the national program for the LEU foil fuel when available, processes used historically in the manufacture of other nuclear fuels and materials, and processes used in other manufacturing industries producing a product configuration similar to the form required in manufacturing a foil fuel. The processes in the reference flow sheet are