WorldWideScience

Sample records for advanced fissile fuel

  1. Development of a Fissile Materials Irradiation Capability for Advanced Fuel Testing at the MIT Research Reactor

    International Nuclear Information System (INIS)

    A fissile materials irradiation capability has been developed at the Massachusetts Institute of Technology (MIT) Research Reactor (MITR) to support nuclear engineering studies in the area of advanced fuels. The focus of the expected research is to investigate the basic properties of advanced nuclear fuels using small aggregates of fissile material. As such, this program is intended to complement the ongoing fuel evaluation programs at test reactors. Candidates for study at the MITR include vibration-packed annular fuel for light water reactors and microparticle fuels for high-temperature gas reactors. Technical considerations that pertain to the design of the MITR facility are enumerated including those specified by 10 CFR 50 concerning the definition of a research reactor and those contained in a separate license amendment that was issued by the U.S. Nuclear Regulatory Commission to MIT for these types of experiments. The former includes limits on the cross-sectional area of the experiment, the physical form of the irradiated material, and the removal of heat. The latter addresses experiment reactivity worth, thermal-hydraulic considerations, avoidance of fission product release, and experiment specific temperature scrams

  2. Electronuclear fissile fuel production

    International Nuclear Information System (INIS)

    A linear accelerator fuel generator is proposed to enrich naturally occurring fertile U-238 or thorium 232 with fissile Pu-239 or U-233 for use in LWR power reactors. High energy proton beams in the range of 1 to 3 GeV energy are made to impinge on a centrally located dispersed liquid lead target producing spallation neutrons which are then absorbed by a surrounding assembly of fabricated LWR fuel elements. The accelerator-target design is reviewed and a typical fuel cycle system and economic analysis is presented. One 300 MW beam (300 ma-1 GeV) linear accelerator fuel regenerator can provide fuel for three 1000 MW(e) LWR power reactors overs its 30-year lifetime. There is a significant saving in natural uranium requirement which is a factor of 4.5 over the present LWR fuel requirement assuming the restraint of no fissile fuel recovery by reprocessing. A modest increase (approximately 10%) in fuel cycle and power production cost is incurred over the present LWR fuel cycle cost. The linear accelerator fuel regenerator and producer assures a long-term supply of fuel for the LWR power economy even with the restraint of the non-proliferation policy of no reprocessing. It can also supply hot-denatured thorium U-233 fuel operating in a secured reprocessing fuel center. (orig.)

  3. Fissile-fuel factories: fact or fantasy

    International Nuclear Information System (INIS)

    Some of the major issues associated with development and possible implementation of fissile-fuel breeders are discussed. Hybrids are first defined and categorized, their potential roles in the economy and relation to other fissile-fuel and power producers are described, results of some hybrid studies are presented, and physics and technology problems associated with hybrids are summarized. It is concluded that fissile fuel breeders with external neutron sources (hybrids, accelerator breeders) could supply moderate-cost fissile fuel for fission power reactors, could supply moderate-cost fissile fuel for initial fast breeder reactor inventories, or could operate solely in a power production mode with the possibility of no reprocessing requirements. In addition, hybrids could serve as a test bed for pure fusion reactor operation, as well as provide relatively near-term benefits from the fusion process. Hybrids offer the potential of cheaper system power costs than the fast breeder or accelerator breeder, although the fast breeder is substantially further along in its development than the hybrid, and the accelerator breeder is further along in its development than the magnetically-confined or small pellet inertially-confined hybrid but at about the same level of development as the large pellet inertially-confined hybrid

  4. Recovery of fissile materials from plutonium residues, miscellaneous spent nuclear fuel, and uranium fissile wastes

    International Nuclear Information System (INIS)

    A new process is proposed that converts complex feeds containing fissile materials into a chemical form that allows the use of existing technologies (such as PUREX and ion exchange) to recover the fissile materials and convert the resultant wastes to glass. Potential feed materials include (1) plutonium scrap and residue, (2) miscellaneous spent nuclear fuel, and (3) uranium fissile wastes. The initial feed materials may contain mixtures of metals, ceramics, amorphous solids, halides, and organics. 14 refs., 4 figs

  5. Accelerating fissile fuel breeding in FBR with natural safety features

    International Nuclear Information System (INIS)

    To guarantee the rapid growth of the Chinese economy in 21st century, nuclear energy should be fully exploited, together with other renewable energies to replace coal and other depletive fossil fuels. Unfortunately, the major Chinese nuclear power plants under construction are mostly PWRs that would consume a lot of natural uranium during their operations. The availability of cheap natural uranium could seriously constraint the Chinese nuclear power development, unless artificial fissile fuel-plutonium is supplied from fast breeder reactors. The fissile nuclei production rate in the traditional fast breeders, however, seems too slow to match the rapid growth of nuclear power. A concept of the advanced fast breeder (AFBR) is introduced, therefore, to greatly accelerating the fissile fuel breeding process. In said breeder, the spherical hollow fuel elements and on-line refueling are adopted. The coolant flows horizontally through the annual fuel bed and bring out the full-fission power with natural circulation generated with the temperature difference between the inlet and outlet coolant flow. In such case, the fuel breeding rate could be greatly increased as the specific thermal power of the fuel inside the reactor core. The said AFBR is an improvement to the Russia-designed BREST FBR. It could transport, without the lead pump and high-elevation lead pool, all the fission heat solely via natural circulation under full power, as well as the decay heat after reactor shut-down, working just as a nuclear hot spring. No radioactivity would be released to the environments under any outside disasters. (authors)

  6. Fissile fuel breeding with peaceful nuclear explosives

    International Nuclear Information System (INIS)

    Neutron physics analysis of a dual purpose modified PACER concept has been conducted. A protective liquid droplet jet zone of 2 m thickness is considered as coolant, energy carrier, and fusile and fissile breeder. Flibe as the main constituent is mixed with increased mole-fractions of heavy metal salt (ThF4 and UF4) starting by 2 up to 12 mol.%. The neutronic model assumed a 30 m radius underground spherical geometry cavity with a 1 cm thick SS-304 stainless steel liner attached to the excavated rock wall. By a self-sufficient tritium breeding of 1.05 with 5 mol.% ThF4, or 9 mol.% UF4 an excess nuclear fuel breeding rate of 1900 kg/year of 233U or 3000 kg/year 239Pu of extremely high isotopic purity can be realized. This precious fuel can be considered for special applications, such as spacecraft reactors or other compact reactors. The heavy metal constituents in jet zone acts as an energy amplifier, leading to an energy multiplication of M=1.27 or 1.65 for 5 mol.% ThF4, or 9 mol.% UF4, respectively. As an immediate result of the strong neutron attenuation in the jet zone, radiation damage with dpa<1.4 and He<7 ppm after a plant operation period of 30 years will be well below the damage limit values. The site could essentially be abandoned, or the cavity could be used as a shallow burial site for other qualified materials upon decommissioning. Finally, the totality of the site with all nuclear peripheral sections must be internationally safeguarded carefully

  7. Novel reactor designs to burn non-fissile fuels

    International Nuclear Information System (INIS)

    Intellectual Ventures is leading a private initiative to explore new approaches to nuclear power. We have used advanced computational tools and modeling technology to reexamine some of the fundamental assumptions-such as the requirement for a continuous supply of enriched fuel-that have long driven the design of fission reactor cores. Our highest goal is to simplify nuclear infrastructure in multiple ways that will help fission power meet the world's growing need for affordable and sustainable energy by directly addressing the issues of emissions, proliferation resistance, safely, and long- term fuel disposal. Our concepts build on the ideas for a 'traveling wave' reactor (TWR) first presented by Edward Teller et al. in the early 1990's'. Initial work suggested that traveling-wave reactor designs could offer several important advantages over other fission reactor designs. After startup, for example, a TWR could run for its entire design life without refueling with fissile material or alternatively, might allow for refueling without a pause in operation. A TWR plant would also require little or no enriched uranium after an initial 'ignition' phase; instead it would burn natural uranium, depleted uranium (a low-cost byproduct of uranium enrichment), or possibly even thorium for most of its life. These advantages derive from the unique way in which a TWR burns its fuel. A nuclear deflagration wave is first formed using enriched uranium or some other fissile material. The slow-moving wave front is allowed to expand into adjacent sections of the core that contain only fertile material, where it breeds its own fuel and advances at a slow and self-limiting pace. The cost of energy delivered by LWRs has been increasing due to the rapidly rising cost of enriched uranium fuel. TWRs fueled primarily by natural uranium or depleted uranium (huge stockpiles of which exist) could, if widely deployed, extend the remaining economic resources of uranium from a century to several

  8. Simple economics parametric analysis of fissile fuel production by fusion--fision reactors

    International Nuclear Information System (INIS)

    A simplified but general analytic model is formulated and evaluated to relate all major elements of fissile-fuel production by fusion-fission (hybrid) reactors on basis of simple economic constraints. The hybrid reactor performance is examined in terms of its fissile-fuel conversion ratio, blanket multiplication, and intrinsic efficiency. A stationary, equilibrium fissile-fuel/energy market is assumed, and the economically constrained performance is evaluated parametrically as a function of burner-converter conversion ratio, plant capital costs, and ratio of fissile-fuel to energy costs. The model and results presented herein can be applied to other means of fissile-fuel production

  9. An analysis of tritium and fissile fuel exchange in fusion-fission systems

    International Nuclear Information System (INIS)

    A model is developed to describe the tritium and fissile fuel flow in a fusion-fission system which consists of a fusion (hybrid) reactor, tritium production (fission) reactors, and (fission) power reactors. The fusion reactor provides all of the fissile fuel for the tritium production and power reactors. Tritium production reactors assure that the system is self sufficient in tritium even if the fusion reactor is not self sufficient. Studies were performed to determine the changes in the cost of electricity from the system as the tritium breeding responsibility varies between the fusion reactor and the tritium production reactors. Allowance for system growth was accomplished with the use of a compound doubling time parameter. Results indicate that the cost of electricity from certain fusion-fission systems may be comparable to that from other advanced systems expected in the same era

  10. Fissile fuel breeding in DT fusion reactor blankets

    International Nuclear Information System (INIS)

    Results of neutronic evaluations of fissile fuel breeding in a variety of DT fusion hybrid-reactor blankets are presented. The blankets are of the fast-fission or fission-suppressed rather than fission-enhanced designs, i.e. in the blankets considered emphasis is on fissile fuel rather than power production. For 233U breeding, when Li metal is the coolant for the first wall and the graphite moderator and the tritium breeding constituent of the blanket, the number of atoms of 233U produced per fusion in blankets that could be of practical interest is in the range 0.5 - 0.68, with the lower value applying to water-cooled ThO2 fertile fuel, the upper to gas-cooled Th-metal fuel located next to the reactor first wall. Neutron multipliers like Pb or Be can increase the production to about 0.74. For 239Pu breeding, the production ratio in practical blankets is 0.6 - 1.64, with the best results being for gas, Na- or Li-metal-cooled U-metal fuels located adjacent to the first wall (the U is depleted uranium). Gas-cooled U-Th-metal blankets, optimized for 233U breeding, yield 0.76 atoms of 233U and 0.38 atoms of 239Pu. The blanket energy multiplication factors are in the range 1.6 - 2.5 for Th blankets, 2.5 - 9.0 for U blankets and approximately 5.5 for the U-Th-metal blanket. The tritium breeding ratio in all blankets is 1.075. Blankets with other first wall, coolant and tritium breeding constituents are also considered. The fusion power requirements of hybrids that could supply the fuel needs of thorium-burning CANDU power reactors, and the allowed costs for building the hybrids are indicated

  11. Economic evaluation of fissile fuel production using resistive magnet tokamaks

    International Nuclear Information System (INIS)

    The application of resistive magnet tokamaks to fissile fuel production has been studied. Resistive magnets offer potential advantages over superconducting magnets in terms of robustness, less technology development required and possibility of demountable joints. Optimization studies within conservatively specified constraints for a compact machine result in a major radius of 3.81 m and 618 MW fusion power and a blanket space envelope of 0.35 m inboard and 0.75 m outboard. This machine is called the Resistive magnet Tokamak Fusion Breeder (RTFB). A computer code was developed to estimate the cost of the resistive magnet tokamak breeder. This code scales from STARFIRE values where appropriate and calculates costs of other systems directly. The estimated cost of the RTFB is $3.01 B in 1984 dollars. The cost of electricity on the same basis as STARFIRE is 42.4 mills/kWhre vs 44.9 mills/kWhre for STARFIRE (this does not include the fuel value or fuel cycle costs for the RTFB). The breakeven cost of U3O8 is $150/lb when compared to a PWR on the once through uranium fuel cycle with no inflation and escalation. On the same basis, the breakeven cost for superconducting tokamak and tandem mirror fusion breeders is $160/lb and $175/lb. Thus, the RTFB appears to be competitive in breakeven U3O8 cost with superconducting magnet fusion breeders and offers the potential advantages of resistive magnet technology

  12. Irradiation behaviour of coated fuel particles for the fissile/fertile particle system

    International Nuclear Information System (INIS)

    The cross evaluation of 30 irradiation experiments which were carried out in the last ten years in order to test fuel particles for the separate use of high enriched uranium in fissile particles and thorium in fertile particles, led to the following results: An oxide-based fissile/fertile particle system (UO2 fissile kernel/ThO2 fertile kernel) can be used as well as the American carbide/oxide particle system (UC2 fissile kernel/ThO2 fertile kernel) under the operation conditions of a high temperature reactor with spherical fuel elements. The swelling of fissile kernels as a consequence of fission gas pores is much more pronounced in UO2 than in UC2 fissile kernels but the buffer layer copes with the swelling without any problems. Ceramic kernel additives (e.g. Al2O3)as well as carbon additives proved not to be suitable because they deteriorate the mechanical properties of the fissile kernels. Kernel migration in a temperature gradient ('amoeba effect') is observed during irradiation of UO2 fissile particles but this does not cause any coating failure. The amoeba effect is suppressed completely by 10% UC2 additives to the UO2 kernel. The silicon carbide interlayer is absolutely necessary for an efficient retention of the solid fission products and has also proved successfully for fertile particles. A measurable fraction of defective particle coatings was not observed before exceeding the target values of burnup and fast neutron fluence. The irradiation-induced dimensional changes of the graphite matrix are independent of the fuel volume loading and have no influence on the irradiation behaviour of the embedded fissile and fertile particles. (orig./IHOE)

  13. Development of NDA reactivity measurement method to determine the fissile contents for spent fuel

    International Nuclear Information System (INIS)

    NDA reactivity measurement method was developed and simulated for the possible use in determination of fissile content and its reactivity characteristics of spent fuel material. This is necessary and practical for the purpose of safeguards implementation and timely control of process line under the condition of very high radiation environment. The neutron detector response due to induced fission from fissile material in spent fuel is calculated by using the MCNP model. The effects of detector response on fissile content, burnup and initial enrichment of fuel are investigated. It shows that the detector response of spent fuel contained 239Pu and fission products is slightly higher than that of flesh fuel with coresponding 235U enrichment

  14. Advanced Fuel Bundles for PHWRS

    International Nuclear Information System (INIS)

    The fuel used by NPCIL presently is natural uranium dioxide in the form of 19- element fuel bundles for 220 MWe PHWRs and 37-element fuel bundles for the TAPP-3&4 540 MWe units. The new 700 MWe PHWRs also use 37-element fuel bundles. These bundles are of short 0.5 m length of circular geometry. The cladding is of collapsible type made of Zircaloy-4 material. PHWRs containing a string of short length fuel bundles and the on-power refueling permit flexibility in using different advanced fuel designs and in core fuel management schemes. Using this flexibility, alternative fuel concepts are tried in Indian PHWRs. The advances in PHWR fuel designs are governed by the desire to use resources other than uranium, improve fuel economics by increasing fuel burnup and reduce overall spent nuclear fuel waste and improve reactor safety. The rising uranium prices are leading to a relook into the Thorium based fuel designs and reprocessed Uranium based and Plutonium based MOX designs and are expected to play a major role in future. The requirement of synergism between different type of reactors also plays a role. Increase in fuel burnup beyond 15 000 MW∙d/TeU in PHWRs, using higher fissile content materials like slightly enriched uranium, Mixed Oxide and Thorium Oxide in place of natural uranium in fuel elements, was studied many PHWR operating countries. The work includes reactor physics studies and test irradiation in research reactors and power reactors. Due to higher fissile content these bundles will be capable of delivering higher burnup than the natural uranium bundles. In India the fuel cycle flexibility of PHWRs is demonstrated by converting this type of technical flexibility to the real economy by irradiating these different types of advanced fuel materials namely Thorium, MOX, SEU, etc. The paper gives a review of the different advanced fuel design concepts studied for Indian PHWRs. (author)

  15. Determination of fissile fraction in MOX (mixed U + Pu oxides) fuels for different burnup values

    Energy Technology Data Exchange (ETDEWEB)

    Ozdemir, Levent, E-mail: levent.ozdemir@taek.gov.tr [Department of Nuclear Engineering, Hacettepe University, 06800 Beytepe, Ankara (Turkey); Acar, Banu Bulut; Zabunoglu, Okan H. [Department of Nuclear Engineering, Hacettepe University, 06800 Beytepe, Ankara (Turkey)

    2011-02-15

    When spent Light Water Reactor fuels are processed by the standard Purex method of reprocessing, plutonium (Pu) and uranium (U) in spent fuel are obtained as pure and separate streams. The recovered Pu has a fissile content (consisting of {sup 239}Pu and {sup 241}Pu) greater than 60% typically (although it mainly depends on discharge burnup of spent fuel). The recovered Pu can be recycled as mixed-oxide (MOX) fuel after being blended with a fertile U makeup in a MOX fabrication plant. The burnup that can be obtained from MOX fuel depends on: (1) isotopic composition of Pu, which is closely related to the discharge burnup of spent fuel from which Pu is recovered; (2) the type of fertile U makeup material used (depleted U, natural U, or recovered U); and (3) fraction of makeup material in the mix (blending ratio), which in turn determines the total fissile fraction of MOX. Using the Non-linear Reactivity Model and the code MONTEBURNS, a step-by-step procedure for computing the total fissile content of MOX is introduced. As was intended, the resulting expression is simple enough for quick/hand calculations of total fissile content of MOX required to reach a desired burnup for a given discharge burnup of spent fuel and for a specified fertile U makeup. In any case, due to non-fissile (parasitic) content of recovered Pu, a greater fissile fraction in MOX than that in fresh U is required to obtain the same burnup as can be obtained by the fresh U fuel.

  16. Optimisation Studies of Accelerator Driven Fertile to Fissile Conversion Rates in Thorium Fuel Cycle

    OpenAIRE

    Bungau, Cristian; Barlow, Roger; Cywinski, R.

    2012-01-01

    The need for proliferation-resistance, longer fuel cycles, higher burn up and improvedwaste form characteristics has led to a renewed worldwide interest in thorium-based fuels and fuel cycles. In this paper the GEANT4 Monte Carlo code has been used to simulate the Thorium-Uranium fuel cycle. The accelerator driven fertile to fissile conversion rates have been calculated for various geometries. Several new classes have been added by the authors to the GEANT4 simulation ...

  17. If Fossil and Fissile Fuels Falter, We've Got. . .

    Science.gov (United States)

    Klaus, Robert L.

    1977-01-01

    Alternative energy sources and the new systems and techniques required for their development are described: fuel cells, magnetohydrodynamics, thermionics, geothermal, wind, tides, waste consersion, biomass, and ocean thermal energy conversion. (MF)

  18. Physical concept on the nuclear reactor with constellation type fissile fuels

    International Nuclear Information System (INIS)

    Under a prolong time neutron irradiation of 232Th by a strong constant neutron flux, a part of 232Th atoms will be converted into a series other nuclides as a result of successive neutron interaction. These neutron born daughter nuclides of 232Th consist of fissile nuclides such as 233U, 235U, 239Pu, 241Pu, ... and fertile nuclides, such as 233Pa, 2'34U, 236U, 237Np, 238Pu etc. As a rule the concentration of each daughter nuclide starts from zero and then increases gradually as the irradiation proceeds until reaching a saturated value and then decreases at a similar rate as the decreasing rate of 232Th. A reactor fueled with thorium and its whole family of neutron born daughter nuclides, with each daughter nuclide at its own saturated concentration, may possess some interesting properties. A preliminary study of the feasibility of such constellation type fissile fuels reactor is presented

  19. The neutron emission method for determination of fissile materials within the spent fuel equipment optimization

    International Nuclear Information System (INIS)

    A nondestructive assay method using neutron technique for determination of the fissile isotopes content along the irradiated fuel rods of MARIA reactor is presented. This method is based on detection of the fission neutrons emitted from external neutron source and multiplied by the fissile isotopes U-235, Pu-239, and Pu-241 within the fuel rod. Neutrons emitted from the spent fuel originate mainly from induced fission in the fissile material and source neutrons penetrating the fuel rod without interaction. Additionally, the neutrons from (α, n) reaction and spontaneous fission of actinide isotopes contribute in the total population of emitted ones. The method gives a chance to perform an experimental calibration of the equipment using two points: fresh fuel rod (maximum signal plus background) and its mock-up (background). The Monte Carlo code has been used for the geometrical simulation and optimization of the measuring equipment: neutron source, moderating container, collimator, and the neutron detector. The results of the calculation show that the moderating container of 30 cm length and 32 cm diameter and a collimator of 26 cm length, 6.8 cm width, and 2 cm height are the optimal configuration. With respect to the fission chamber position, the number of neutrons has been calculated as a function of distance from the fuel rod surface in the case of fresh fuel and its mock-up. The distance, at which the ratio of the signal to background has its maximum, has been found at 4.5 cm far from the outer surface of the fuel. (author)

  20. On international criticality codes for fuel pellets in fissile solution

    International Nuclear Information System (INIS)

    The reference calculations, based on the APOLLO-Pic method implemented in the framework of this study, demonstrated that the actual reactivity variation (benchmark n0 20) is a monotonic decrease with pellet dissolution. At the opposite of the contributor's results, based on the international criticality code SCALE, the reactivity loss with dissolution is weak. The discrepancy is mainly due to 238U resonant absorption which can induce, in this fuel double heterogeneity problem n0 20, as much as -30 000 pcm K∞ underestimation. It was pointed out that design-oriented transport codes must be improved by accurate deterministic formalisms: PIC equivalence method, subgroup theory (WIMSE), ultrafine slowing-down calculation (ROLAIDS). Ultimate confirmation of the reference results presented in this paper should be provided by a set of critical experiments which mock-up hypothetical dissolver geometries. Finally it should be noted that thanks to the interest and the efforts of the OECD/NEA Criticality Working Group in performing the international benchmark exercise and in pursuing the explanation of the discrepancies, a potentially dangerous inadequacy in criticality calculation methods was exposed and resolved

  1. US Light-Water Reactor spent fuel inventory-fissile distribution

    International Nuclear Information System (INIS)

    Those conducting waste management studies to reduce the potential for a nuclear criticality accident in a future geological repository must examine the quantities and distribution of fissile isotopes that are present in discharged boiling-water reactor (BWR) and pressurized-water reactor (PWR) spent nuclear fuel (SNF) scheduled for disposition. The major fissile isotopes present in LWR fuels that impact criticality safety are the nuclides, 235U, 239Pu, and 24lPu. The sum of the quantities of these three nuclides, expressed as a percentage of the total amount of all U and Pu isotopes present in a batch of discharged fuel, determines the final enrichment of the fuel batch under consideration. The final enrichment provides an approximate measure of the nuclear criticality potential. As the final enrichment increases, the mass, geometry, or administrative controls that must be in place to prevent nuclear criticality become more stringent. Below an enrichment of about 0.7%, however, criticality is no longer a concern because the infinite multiplication factor for any heterogeneous or homogeneous mixture of fuel and water, even under conditions of optimum moderation, is less than unity. The current study examines the distribution of the final enrichment of the LWR SNF which was discharged through December 31, 1993, and which currently resides in the fuel storage pools of the various utilities or in one of several AFR facilities

  2. Neutronic evaluation of fissile fuel breeding blankets for the fission-suppressed Tandem-Mirror Hybrid Reactor

    International Nuclear Information System (INIS)

    A computational study was performed on the blanket design of the Lawrence Livermore National Laboratory (LLNL) fission-suppressed Tandem Mirror Hybrid Reactor (TMHR) to qualify the methods and data bases available at Oak Ridge National Laboratory (ORNL) for use in analyzing the neutronic performance of fissile fuel breeding blankets. The eventual goal of the study was to establish the capability for analysis and optimization of advanced fissile fuel production blanket designs. Discrete ordinates radiation transport calculations were performed in one-dimensional cylindrical geometry to obtain the blanket spatial distribution and energy spectra of the neutron and gamma-ray fluxes resulting from the monoenergetic (14.1 MeV) fusion first wall source. Key macroscopic cross sections of the blanket materials were then folded with the flux spectra to obtain reaction rates critical to evaluating blanket feasibility. Finally, a time-dependent depletion analysis was performed to evaluate the blanket performance during equilibrium cycle conditions. The results of the study are presented both as graphs and tables

  3. Physics concept on the constellation type fissile fuels and its application to the prospective Th-232U Reactor

    International Nuclear Information System (INIS)

    In contrast with the conventional nuclear reactor which usually fuelled with on single fissile nuclide, a constellation type fissile fuels reactor consists of a parent nuclide such as 232Th or 238U and its whole family of neutron generated daughter nuclides. All of them are regarded as fissile fuels but of quite different fission ability. The concentration of each daughter nuclide is determined by its saturate concentration ratio with the parent nuclide. In such fuel system, the whole fuel consumed by neutron reaction almost completely results in fission products. In this article, some properties of such fuel system, determination of the saturate concentration of each daughter nuclide and applicability to Th-233U fueled reactor will be discussed. 3 refs., 1 tab., 2 figs

  4. FISSILE FUEL BREEDING IN THE ARIES-ST FUSION REACTOR BY USING MOLTEN SALT WITH UF4

    OpenAIRE

    ÜBEYLİ, Mustafa

    2010-01-01

    ABSTRACTFissile fuel breeding in the ARIES-ST of the 1000 MWel power plant is investigated by using molten salt containing UF4 Calculations are done with the aid of one-dimensional code of SCALE4.3. In this hybrid model, a substantial amount of fissile fuel can be produced with a fissile fuel breeding ratio of 239 Pu = 0.115 per incident neutron at start-up conditions, that corresponds to 3558 kg 239 Pu/year by a full fusion power of 2740 MW. Tritium breeding ratio is found as 1.14 so that tr...

  5. Determination of dependence of fissile fraction in MOX fuels on spent fuel storage period for different burnup values

    International Nuclear Information System (INIS)

    Highlights: ► In a previous study, an expression to calculate fissile fraction of MOX for various burnups was obtained for 5-year cooled SF. ► In this follow-up study, a correction factor for spent fuel storage periods other than 5 years is derived. ► Thus, one major restriction on use of the expression derived in the initial study is eliminated. - Abstract: The purpose of this technical note is to remove one of the limitations of a derived expression in a previously published article (Özdemir et al., 2011). The original article focused on deriving (computationally) an expression for calculating total fissile fraction of mixed oxid (MOX) fuels depending on discharge burnup of spent fuel and desired burnup of MOX fuel; consequently, such an expression was obtained and put forward, together with its limitations. One of the limitations has been that all the computations and therefore the resulting expression are based on the assumption of a spent fuel storage period of 5 years. This follow-up study simply aims to obtain a correction factor for spent fuel storage periods other than 5 years; thus to remove one major restriction on use of the expression derived in the original article

  6. Fissile compound - Inert matrix compatibility studies for the development of gas cooled fast reactor fuels

    International Nuclear Information System (INIS)

    Helium-Cooled High-Temperature Fast Reactors have a high potential for transmutation of minor actinides (Pu, Am, Cm... ). In this kind of reactor, the fuel temperature would be 1200 deg C in use and the inert matrix should retain the fission products in the fuel structure up to 1600 deg C. The fissile compound would be (U,Pu)C or (U,Pu)N owing to their high density, good thermal conductivity and refractory behavior. SiC, TiC, ZrC and TiN, ZrN would be the inert matrix surrounding (U,Pu)C or (U,Pu)N fissile compounds. This study is devoted to the chemical compatibility between UC or UN and inert matrix in the 1200 deg C - 2000 deg C temperature range. In order to achieve a limited number of specific experiments, thermodynamic calculations are realized using the thermodynamic data provided either by the Thermodata database or from the literature. (authors)

  7. Monte Carlo simulations of differential die-away instrument for determination of fissile content in spent fuel assemblies

    Science.gov (United States)

    Lee, Tae-Hoon; Menlove, Howard O.; Swinhoe, Martyn T.; Tobin, Stephen J.

    2011-10-01

    The differential die-away (DDA) technique has been simulated by using the MCNPX code to quantify its capability of measuring the fissile content in spent fuel assemblies. For 64 different spent fuel cases of various initial enrichment, burnup and cooling time, the count rate and signal to background ratios of the DDA system were obtained, where neutron backgrounds are mainly coming from the 244Cm of the spent fuel. To quantify the total fissile mass of spent fuel, a concept of the effective 239Pu mass was introduced by weighing the relative contribution to the signal of 235U and 241Pu compared to 239Pu and the calibration curves of DDA count rate vs. 239Pu eff were obtained by using the MCNPX code. With a deuterium-tritium (DT) neutron generator of 10 9 n/s strength, signal to background ratios of sufficient magnitude are acquired for a DDA system with the spent fuel assembly in water.

  8. MCNP-to-TORT radiation transport calculations in support of mixed oxide fuels testing for the Fissile Materials Disposition Program

    International Nuclear Information System (INIS)

    The US (US) Department of Energy Fissile Materials Disposition Program has begun studies for disposal of surplus weapons-grade plutonium (WG-Pu) as mixed uranium plutonium oxide (MOX) fuel for commercial light water reactors (LWRs). Currently MOX fuel is used commercially in a number of foreign countries, but is not in the US. Most of the experience is with reactor grade plutonium (RG-Pu) in MOX fuel. Therefore, to use WG-Pu in MOX fuel, one must demonstrate that the experience with RG-Pu is relevant. As a first step in this program, the utilization of WG-Pu in a LWR environment must be demonstrated. To accomplish this, a test is to be conducted to investigate some of the unresolved issues. The initial tests will be made in an I-hole of the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL). Initial radiation transport calculations of the test specimens were made at INEEL using the MCNP Monte Carlo radiation transport code. These calculations were made to determine the linear heating rates in the fuel specimens. Unfortunately, the results of the calculations could not show the detailed high and low power density spots in the specimens. However, a discrete ordinates radiation transport code could pinpoint these spatial details. Therefore, INEEL was tasked with producing a MCNP source at the boundary of a rectangular parallelepiped enclosing the ATR I-hole, and Oak Ridge National Laboratory was tasked with transforming this boundary source into a discrete ordinates boundary source for the Three-dimensional Oak Ridge radiation Transport (TORT) code. Thus, the TORT results not only complemented, but also were in agreement with the MCNP results

  9. MCNP-to-TORT Radiation Transport Calculations in Support of Mixed Oxide Fuels Testing for the Fissile Materials Disposition Program

    International Nuclear Information System (INIS)

    The United States (US) Department of Energy Fissile Materials Disposition Program (FMDP) began studies for disposal of surplus weapons-grade plutonium (WG-Pu) as mixed uranium-plutonium oxide (M40X) fuel for commercial light-water reactors(LWRS). As a first step in this program, a test of the utilization of WG-Pu in a LWR environment is being conducted in an I-hole of the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL). Initial radiation transport calculations of the test specimens were made at INEEL using the MCNP Monte Carlo radiation transport code to determine the linear heating rates in the fuel specimens. Unfortunately, the results of the calculations could not show the detailed high and low power-density spots in the specimens. Therefore, INEEL produced an MCNP source at the boundary of a rectangular parallelepiped enclosing the ATR I-hole, and Oak Ridge National Laboratory (ORNL) transformed this boundary source into a discrete -ordinates boundary source for the Three-dimensional Oak Ridge radiation Transport (TORT) code to pinpoint spatial detail. Agreement with average MCNP results were within 5%

  10. Neutron physical and thermohydraulic investigations on the fissile fuel production with spallation neutrons

    International Nuclear Information System (INIS)

    In case of an enhanced utilisation of nuclear energy from fission reactors it is necessary to preserve the natural deposits of fissile material. One possibility for this preservation could be the breeding of synthetic fuel by the means of spallation neutrons. The configuration discussed here consists of an integral spallation target/breeding blanket pebble bed of hollow Thorium spheres, which is cooled alternatively with helium gas or liquid sodium or lithium. With helium cooling the neutron yield is 35 neutrons par initial proton and the breeding gain is 775 kg U-233/a. With sodium as coolant the n-yields is 25 n/p, and the fuel production is 520 kg U-233/a. If liquid lithium is used as coolant the neutron yield is 30 n/p with a production of 440 kg Uranium-233 per year. Supposing a fuel requirement of a high converting HTR of 90 kg U-233/GW-a, 5 to 9 1-GWsub(e)-reactors could be supplied depending on the coolant. (orig.)

  11. Feasibility of fissile mass assay of spent nuclear fuel using 252Cf-source-driven frequency-analysis

    International Nuclear Information System (INIS)

    The feasibility was evaluated using MCNP-DSP, an analog Monte Carlo transport cod to simulate source-driven measurements. Models of an isolated Westinghouse 17x17 PWR fuel assembly in a 1500-ppM borated water storage pool were used. In the models, the fuel burnup profile was represented using seven axial burnup zones, each with isotopics estimated by the PDQ code. Four different fuel assemblies with average burnups from fresh to 32 GWd/MTU were modeled and analyzed. Analysis of the fuel assemblies was simulated by inducing fission in the fuel using a 252Cf source adjacent to the assembly and correlating source fissions with the response of a bank of 3He detectors adjacent to the assembly opposite the source. This analysis was performed at 7 different axial positions on each of the 4 assemblies, and the source-detector cross-spectrum signature was calculated for each of these 28 simulated measurements. The magnitude of the cross-spectrum signature follows a smooth upward trend with increasing fissile material (235U and 239Pu) content, and the signature is independent of the concentration of spontaneously fissioning isotopes (e.g., 244Cm) and (α,n) sources. Furthermore, the cross-spectrum signature is highly sensitive to changes in fissile material content. This feasibility study indicated that the signature would increase ∼100% in response to an increase of only 0.1 g/cm3 of fissile material

  12. Fissile fuel production and usage of thermal reactor waste fueled with UO2 by means of hybrid reactor system

    International Nuclear Information System (INIS)

    The use of Fast Breeder Reactors to produce fissile fuel from nuclear waste and the operation of these reactors with a new neutron source are becoming today' topic. In the thermonuclear reactors, it is possible to use 2.45-14.1 MeV - neutrons which can be obtained by D-T, D-D Semicatalyzed (D-D) and other fusion reactions. To be able to do these, Hybrid Reactor System, which still has experimental and theoretical studies, have to be taken into consideration.In this study, neutronic analysis of hybrid blanket with grafit reflector, is performed. D-T driven fusion reaction is surrounded by UO2 fuel layer and the production of ''2''3''9Pu fissile fuel from waste ''2''3''8U is analyzed. It is also compared to the other possible fusion reactions. The results show that 815.8 kg/year ''2''3''8Pu with D-T reaction and 1431.6 kg/year ''2''3''8Pu with semicatalyzed (D-D) reaction can be produced for 1000 MW fusion power. This means production of 2.8/ year and 4.94/ year LWR respectively. In addition, 1000 MW fusion flower is is multiplicated to 3415 MW and 4274 MW for D-T and semicatalyzed (D-D) reactions respectively. The system works subcritical and these values are 0.4115 and 0.312 in order. The calculations, ANISN-ORNL code, S16-P3 approach and DLC36 data library are used

  13. Advanced fuel cycles in CANDU reactors

    International Nuclear Information System (INIS)

    This paper re-examines the rationale for advanced nuclear fuel cycles in general, and for CANDU advanced fuel cycles in particular. The traditional resource-related arguments for more uranium nuclear fuel cycles are currently clouded by record-low prices for uranium. However, the total known conventional uranium resources can support projected uranium requirements for only another 50 years or so, less if a major revival of the nuclear option occurs as part of the solution to the world's environmental problems. While the extent of the uranium resource in the earth's crust and oceans is very large, uncertainty in the availability and price of uranium is the prime resource-related motivation for advanced fuel cycles. There are other important reasons for pursuing advanced fuel cycles. The three R's of the environmental movement, reduce, recycle, reuse, can be achieved in nuclear energy production through the employment of advanced fuel cycles. The adoption of more uranium-conserving fuel cycles would reduce the amount of uranium which needs to be mined, and the environmental impact of that mining. Environmental concerns over the back end of the fuel cycle can be mitigated as well. Higher fuel burnup reduces the volume of spent fuels which needs to be disposed of. The transmutation of actinides and long-lived fission products into short-lived fission products would reduce the radiological hazard of the waste from thousands to hundreds of years. Recycling of uranium and/or plutonium in spent fuel reuses valuable fissile material, leaving only true waste to be disposed of. Advanced fuel cycles have an economical benefit as well, enabling a ceiling to be put on fuel cycle costs, which are

  14. Status and prospects of advanced fissile fuel breeders

    International Nuclear Information System (INIS)

    Fusion--fission hybrid systems, fast breeder systems, and accelerator breeder systems were compared on a common basis using a simple economic model. Electricity prices based on system capital costs only were computed, and were plotted as functions of five key breeder system parameters. Nominally, hybrid system electricity costs were about twenty-five percent lower than fast breeder system electricity costs, and fast breeder system electricity costs were about forty percent lower than accelerator breeder system electricity costs. In addition, hybrid system electricity costs were very insensitive to key parameter variations on the average, fast breeder system electricity costs were moderately sensitive to key parameter variations on the average, and accelerator breeder system electricity costs were the most sensitive to key parameter variations on the average

  15. Development of the fuel of gas fast future reactors. Study of the compatibility between fissile compound and inert matrix

    International Nuclear Information System (INIS)

    The gas coolant fast future reactors require to develop new types of fuels. The fissile compounds (U, Pu, AM)C and (U, Pu, AM)N are retained, on account of their high density, their refractory character and their good thermal conductivity. The choice for the inert matrix, which will contain the fissile compound, bears on ceramics: SiC, TiC and ZrC in the case of actinides monocarbide and SiC, TiN and ZrN in the case of actinides mono-nitrides. The chemical compatibility between the main fissile compound (UN or UC) and the inert matrices SiC and TiN has been studied in the range of 1200 C (nominal running conditions) and 2000 C (accidental running conditions). The thermodynamic compatibility calculations allow to determine particularly at which temperature a liquid phase risks to appear by reaction between the fissile compound and the inert matrix and to limit the number of tests to carry out. This approach is completed by specific interaction tests carried out under different atmospheres. (O.M.)

  16. LSDS Development for Isotopic Fissile Content Assay

    International Nuclear Information System (INIS)

    Atomic Energy Research Institute(KAERI). LSDS has the feature to analyze the isotopic fissile material contents in a near real time without an interference of neutron and gamma rays background from spent fuels. The optimized geometry was setup and several system parameters were calculated for the spectrometer working. Sensitivity calculations were done on the selected geometry. In the paper, the multiplication effect is introduced and theneutron resolution broadening is analyzed in the slowing down medium. The energy resolution is important factor to extract the individual fission signal from the mixture of U235, Pu239 and Pu241. The advanced isotopic fissile assay technology will give an increased international proliferation resistance on the spent fuel re-utilization and management

  17. Monte Carlo simulations of differential die-away instrument for determination of fissile content in spent fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Tae-Hoon, E-mail: typhoon@kaeri.re.kr [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Korea Atomic Energy Research Institute, 150-1 Dukjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Menlove, Howard O.; Swinhoe, Martyn T.; Tobin, Stephen J. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

    2011-10-01

    The differential die-away (DDA) technique has been simulated by using the MCNPX code to quantify its capability of measuring the fissile content in spent fuel assemblies. For 64 different spent fuel cases of various initial enrichment, burnup and cooling time, the count rate and signal to background ratios of the DDA system were obtained, where neutron backgrounds are mainly coming from the {sup 244}Cm of the spent fuel. To quantify the total fissile mass of spent fuel, a concept of the effective {sup 239}Pu mass was introduced by weighing the relative contribution to the signal of {sup 235}U and {sup 241}Pu compared to {sup 239}Pu and the calibration curves of DDA count rate vs. {sup 239}Pu{sub eff} were obtained by using the MCNPX code. With a deuterium-tritium (DT) neutron generator of 10{sup 9} n/s strength, signal to background ratios of sufficient magnitude are acquired for a DDA system with the spent fuel assembly in water.

  18. Predicting fissile content of spent nuclear fuel assemblies with the Passive Neutron Albedo Reactivity technique and Monte Carlo code emulation

    International Nuclear Information System (INIS)

    There is a great need in the safeguards community to be able to nondestructively quantify the mass of plutonium of a spent nuclear fuel assembly. As part of the Next Generation of Safeguards Initiative, we are investigating several techniques, or detector systems, which, when integrated, will be capable of quantifying the plutonium mass of a spent fuel assembly without dismantling the assembly. This paper reports on the simulation of one of these techniques, the Passive Neutron Albedo Reactivity with Fission Chambers (PNAR-FC) system. The response of this system over a wide range of spent fuel assemblies with different burnup, initial enrichment, and cooling time characteristics is shown. A Monte Carlo method of using these modeled results to estimate the fissile content of a spent fuel assembly has been developed. A few numerical simulations of using this method are shown. Finally, additional developments still needed and being worked on are discussed. (author)

  19. Predicting fissile content of spent nuclear fuel assemblies with the passive neutron Albedo reactivity technique and Monte Carlo code emulation

    International Nuclear Information System (INIS)

    There is a great need in the safeguards community to be able to nondestructively quantify the mass of plutonium of a spent nuclear fuel assembly. As part of the Next Generation of Safeguards Initiative, we are investigating several techniques, or detector systems, which, when integrated, will be capable of quantifying the plutonium mass of a spent fuel assembly without dismantling the assembly. This paper reports on the simulation of one of these techniques, the Passive Neutron Albedo Reactivity with Fission Chambers (PNAR-FC) system. The response of this system over a wide range of spent fuel assemblies with different burnup, initial enrichment, and cooling time characteristics is shown. A Monte Carlo method of using these modeled results to estimate the fissile content of a spent fuel assembly has been developed. A few numerical simulations of using this method are shown. Finally, additional developments still needed and being worked on are discussed.

  20. Experimental evaluation of the differential die-away pulsed-neutron technique for the fissile assay of hot irradiated fuel waste

    International Nuclear Information System (INIS)

    Hot irradiated fuel waste packages containing self neutron sources up to 107 n/s may be assayed successfully with a 25-mg 239Pu detection sensitivity using a differential die-away fissile assay system based on a Sandia-designed 14-MeV neutron generator. With a minor design change, our system can accommodate contact gamma dose rates up to 104 R/h with no loss of fissile assay sensitivity

  1. LIGHTBRIDGE corporation advanced metallic fuel

    International Nuclear Information System (INIS)

    Lightbridge Corporation is developing a metallic nuclear fuel which utilizes an innovative fuel rod geometry and composition to improve power plant economics and enhance the performance and safety of commercial light water reactors. The versatile fuel can utilize uranium or plutonium as the fissile component. The fuel is fully compatible with existing light water reactor designs and requires no major changes to reactor operations. The metallic fuel provides a durable solution that is also capable of operating at higher power density than existing fuels allowing for increased power output and cycle length compared to conventional oxide fuels. Lightbridge patented nuclear fuel technologies are designed to significantly enhance nuclear power industry economics and increase power output by: 1) extending fuel cycle length to 24 months or longer while simultaneously increasing power output by 10% or increasing power output by up to 17% with 18-month fuel cycles in existing pressurized water reactors (PWRs); 2) enabling increased reactor power output of up to 30% without changing core size in new build PWRs; and 3) reducing the volume of used fuel per kilowatt-hour as well as enhancing proliferation resistance of spent fuel. (author)

  2. Advanced coated particle fuels

    International Nuclear Information System (INIS)

    The coated particle fuel (cpf) has been developed for use in high-temperature gas-cooled reactors, but it may find applications in other types of reactors. In JAERI, besides the development of cpf for High Temperature Engineering Test Reactor, conceptual studies of the cpf applications in actinide burner reactors and space reactors have been made. The conceptual design studies as well as the research and development of advanced coatings, ZrC and TiN, are reviewed. (author)

  3. Irradiation performance of (Th,U)02 fuel designed for advanced fuel cycle applications

    International Nuclear Information System (INIS)

    The reference fabrication route for Advanced Cycle thoria-based fuel is conventional in that it produces cold-pressed and sintered pellets. However we are also evaluating alternative fuels which offer the potential for simpler fabrication in a remote facility, and in some cases improved high burnup performance. These alternatives are impregnated, spherepac, and extruded thoria-based fuels. Spherepac fuel has been irradiated at a linear power of 50-60 kW/m to about 180 MW.h/kg H.E. There have been unexplained defects in fuel with both free-standing and collapsible cladding. Impregnated fuel has operated to 650 MW.h/kg H.E. at 50-60 KW/m. An experiment examining fuel from the sol-gel extrusion process has reached 450 MW.h/kg H.E. at a maximum linear power of 60 KW/m. The latter two experiments have operated without defects and with fission gas release less than that for U02 under identical conditions. The extruded fuel has a pellet geometry similar to that for conventional fuel and is AECL's first practical demonstration of thoria-based fuel with the fissile component distributed homogeneously on an atomic scale. We will continue monitoring the extruded fuel to a burnup approaching 1000 MW.h/kg H.E., as an indicator for the performance expected from co-precipitated (Th,U)02 or mechanically-mixed (Th,U)02 with good fissile homogeneity

  4. CANDU advanced fuel cycles

    International Nuclear Information System (INIS)

    This report is based on informal lectures and presentations made on CANDU Advanced Fuel Cycles over the past year or so, and discusses the future role of CANDU in the changing environment for the Canadian and international nuclear power industry. The changing perspectives of the past decade lead to the conclusion that a significant future market for a CANDU advanced thermal reactor will exist for many decades. Such a reactor could operate in a stand-alone strategy or integrate with a mixed CANDU-LWR or CANDU-FBR strategy. The consistent design focus of CANDU on enhanced efficiency of resource utilization combined with a simple technology to achieve economic targets, will provide sufficient flexibility to maintain CANDU as a viable power producer for both the medium- and long-term future

  5. A catalogue of advanced fuel cycles in CANDU-PHW reactors

    International Nuclear Information System (INIS)

    A catalogue raisonne is presented of various advanced fuel cycle options which have the potential of substantially improving the uranium utilization for CANDU-PHW reactors. Three categories of cycles are: once-through cycles without recovery of fissile materials, cycles that depend on the recovery and recycle of fissile materials in thorium or uranium, cycles that depend primarily on the production of fissile material in a fertile blanket by means of an intense neutron source other than fission, such as an accelerator breeder. Detailed tables are given of the isotopic compositions of the feed and discharge fuels, the logistics of materials and processes required to sustain each of the cycles, and tables of fuel cycle costs based on a method of continuous discounting of cash flow

  6. Advanced Fuels Campaign 2012 Accomplishments

    Energy Technology Data Exchange (ETDEWEB)

    Not Listed

    2012-11-01

    The Advanced Fuels Campaign (AFC) under the Fuel Cycle Research and Development (FCRD) program is responsible for developing fuels technologies to support the various fuel cycle options defined in the DOE Nuclear Energy Research and Development Roadmap, Report to Congress, April 2010. The fiscal year 2012 (FY 2012) accomplishments are highlighted below. Kemal Pasamehmetoglu is the National Technical Director for AFC.

  7. Thorium - denatured uranium fuel cycles in PHWR-pressure tube type using low enriched uranium as annual externally supplied fissile material

    International Nuclear Information System (INIS)

    The use of denatured uranium as the initial fissile inventory of the thorium-uranium cycles could be straight-forward. The use of denatured uranium as annual externally supplied fissile material could be not applicable in the case of homogenous HWR fuel bundle concept if it is intended to avoid the shift from the Th/U cycle to U/Pu cycle or the reenrichment of the recycled uranium containing U-232. The paper presents a heterogenous fuel concept for HWR which permits the use of denatured uranium without the above-mentioned shift. According to this concept the annual externally supplied fissile material is introduced in distinct fuel rods separable at the front end of the reprocessing or as distinct fuel bundles. In these cases the normal reenrichment could be applied, this part of the fuel being free of U-233, and therefore free of U-232. The resource utilization penalties in addition to those introduced by the denaturing of the initial core were evaluated. At 3% enrichment these penalties rise with about 40% the annual natural uranium requirements. It is concluded that for these Th/U cycles in HWR, it is possible to avoid the presence of highly enriched uranium at the fuel fabrication step

  8. General overview of CANDU advanced fuel cycles program

    International Nuclear Information System (INIS)

    The R and D program for CANDU advanced fuel cycles may be roughly divided into two components which have a near-and long-term focus, respectively. The near-term focus is on the technology to implement improved once-through cycles and mixed oxide (plutonium-uranium oxides) recycle in CANDU and on technologies to separate zirconium isotopes. Included is work on those technologies which would allow a CANDU-LWR strategy to be developed in a growing nuclear power system. For the longer-term, activities are focused on those technologies and fuel cycles which would be appropriate in a period when nuclear fuel demand significantly exceeds mined uranium supplies. Fuel cycles and systems under study are thorium recycle, CANDU fast breeder systems and electro-nuclear fissile breeders. The paper will discuss the rationale underlying these activities, together with a brief description of activities currently under way in each of the fuel cycle technology areas

  9. Development of LSDT Spectrometer for Nuclear Fissile Assay

    International Nuclear Information System (INIS)

    Recently, the price of resources rapidly increased and the stable supply of resources is a big issue internationally. Specially, countries which have highly dependency on the oil face more severe situation in the energy supply. Moreover, from the current environmental change, air pollution by carbon dioxide, drain of oil resource, increasing demand of electricity, and energy independence, nuclear energy is reconsidered and is the most feasible option in meeting the expanding world energy demand, as a sustainable manner. However, nuclear power plant has an inevitable situation in the production of radioactive wastes. Therefore, the management for spent fuel is an urgent problem and issue to be resolved for a safe next generation. The nuclear reactors related to nuclear fuel cycle are developed as a generation IV type. Nuclear fuel cycle facility will increase through GNEP and advanced nuclear fuel cycle program to expand the use of nuclear energy. pyro-processing is developed to reduce the volume and the radiotoxicity of spent fuels for a permanent storage, and to reuse fissionable elements in a spent fuel. The pyro-processing produces uranium and uranium-TRU (neptunium, plutonium, and americium) mixture. The uranium-TRU is fabricated for the fuel of sodium fast reactor (SFR), enhanced proliferation resistance. The development for a practical application of the assay of nuclear fissile materials was initiated in KAERI for pyro-process related SFR fuel cycle, compact and integrated spent fuel storage design and safety enhanced fuel recycle. The isotopic fissile assay is very important for reuse of fissile material. Specially, the accuracy of fissile assay in a recycled fuel including TRU enhances safety and economy in a nuclear power plant operation. Lead slowing down time spectrometer (LSDTS) is the most proper tool and being developed in KAERI. It has good feature to analyze the isotopic fissile contents in a near real time without an interference of neutron and

  10. Comparison of fission gas products accumulated in the irradiated UO2 fuel with the radiogenic gas trapped in minerals containing fissile atoms

    International Nuclear Information System (INIS)

    Both the old mineral materials containing fissile isotopes and the neutron irradiated uranium dioxide fuel hold fairly a lot of noble gases. Appreciable fraction of the inert gases is immobilized in the uranium dioxide fuel after irradiation exposure in excess of 1015 fissions/cm3. Spontaneous fission fluency for the minerals containing fissile atoms of about 2 billions years old is about 1015 fissions/cm3. Alike the old minerals containing fissile isotopes and the irradiated UO2 fuel reveal during annealing three peaks of gas release in the low temperature range. In the range only several percent of the total amount of gas as well in the mineral materials and the uranium dioxide fuel is released. It still holds significant part of noble gases after annealing at a temperature about 1400 deg C. Noble gas atoms can be trapped by the point defects, clusters of the point defects, clathrates and small bubbles in the irradiated material. It is suggested that above a fission fluency threshold about 1015 fissions/cm3 occurs an additional trapping process based on the irradiation induced chemical bonding process. It is inferred further that as well in the minerals and in the irradiated UO2 fuel the gas release kinetics is determined by the kinetics of thermal recovery of the radiation induced defects in the low temperature range. It is surmised that the same amount of damage caused by alpha decay of uranium is much less effective than caused by fission of uranium for the immobilization of noble gases since only during fissioning there are formed conditions for strong chemical binding between the noble gas atoms and the mineral materials. Finally it is concluded identical nature of noble gas accumulation mechanisms in the irradiated UO2 fuel and in the old mineral materials containing fissile atoms. (author)

  11. Advanced fuel developments to improve fuel cycle cost in PWR

    International Nuclear Information System (INIS)

    Increasingly lower fuel cycle costs and higher plant availability factors have been two crucial components in keeping the overall cost of electricity produced by nuclear low and competitive with respect to other energy sources. The continuous quest to reduce fuel cycle cost has resulted in some consolidated trends in LWR fuel management schemes: smaller number of feed fuel assemblies with longer residence time; longer cycles, with 18-month cycle as the predominant option, and some plants already operating on, or considering, 24-month refueling intervals; higher power ratings with many plants undergoing power uprates. In order to maintain or improve fuel utilization for the longer cycles and/or higher power ratings, the licensed limits in fuel fissile content (5.0 w/o U235 enrichment) and discharge burnup (62 GWd/tHM for the peak pin) have been approached. In addition, Zr-based fuel cladding materials are also being challenged by the resulting increased duty. For the above reasons further improvements in fuel cycle cost have to overcome one or more of the current limits. This paper discusses an option to break through this 'stalemate', i.e. uranium nitride (UN) fuel with SiC clad. In UN the higher density of the nitride with respect to the oxide fuel leads to higher fissile content and reduction in the number of feed assemblies, improved fuel utilization and potentially higher specific powers. The SiC clad, among other benefits, enables higher clad irradiation, thereby exploiting the full potential of UN fuel. An alternative to employing UN fuel is to maintain UO2 fuel but boost the fissile content increasing the U235 enrichment beyond the 5 w/o limit. The paper describes and compares the potential benefits on fuel cycle cost of either option using realistic full-core calculations and ensuing economic analysis performed using Westinghouse in-house reactor physics tools and methodologies. (author)

  12. Advanced fuel technology and performance

    International Nuclear Information System (INIS)

    The purpose of the Advisory Group Meeting on Advanced Fuel Technology and Performance was to review the experience of advanced fuel fabrication technology, its performance, peculiarities of the back-end of the nuclear fuel cycle with regard to all types of reactors and to outline the future trends. As a result of the meeting recommendations were made for the future conduct of work on advanced fuel technology and performance. A separate abstract was prepared for each of the 20 papers in this issue

  13. On-Going Comparison of Advanced Fuel Cycle Options

    International Nuclear Information System (INIS)

    The Advanced Fuel Cycle Initiative (AFCI) program is addressing key issues associated with critical national needs. This paper compares the major options with these major ''outcome'' objectives - waste geological repository capacity and cost, energy security and sustainability, proliferation resistance, fuel cycle economics, and safety as well as ''process'' objectives associated with readiness to proceed and adaptability and robustness in the face of uncertainties. Working together, separation, transmutation, and fuel technologies provide complete energy systems that can improve waste management compared to the current ''once-through/no separation'' approach. Future work will further increase confidence in potential solutions, optimize solutions for the mixtures of objectives, and develop attractive development and deployment paths for selected options. This will allow the nation to address nearer-term issues such as avoiding the need for additional geological repositories while making nuclear energy a more sustainable energy option for the long-term. While the Generation IV Initiative is exploring multiple reactor options for future nuclear energy for both electricity generation and additional applications, the AFCI is assessing fuel cycles options for either a continuation or expansion of nuclear energy in the United States. This report compares strategies and technology options for managing the associated spent fuel. There are four major potential strategies, as follows: (smbullet) The current U.S. strategy is once through: standard nuclear power plants, standard fuel burnup, direct geological disposal of spent fuel. Variants include higher burnup fuels in water-cooled power plants, once-through gas-cooled power plants, and separation (without recycling) of spent fuel to reduce the number and cost of geological waste packages. (smbullet) The second strategy is thermal recycle, recycling some fuel components in thermal reactors. This strategy extends the useful

  14. On-Going Comparison of Advanced Fuel Cycle Options

    Energy Technology Data Exchange (ETDEWEB)

    Steven J. Piet; Ralph G. Bennett; Brent W. Dixon; J. Stephen Herring; David E. Shropshire; Mark Roth; J. D. Smith; Robert Hill; James Laidler; Kemal Pasamehmetoglu

    2004-10-01

    The Advanced Fuel Cycle Initiative (AFCI) program is addressing key issues associated with critical national needs. This paper compares the major options with these major “outcome” objectives - waste geological repository capacity and cost, energy security and sustainability, proliferation resistance, fuel cycle economics, and safety as well as “process” objectives associated with readiness to proceed and adaptability and robustness in the face of uncertainties. Working together, separation, transmutation, and fuel technologies provide complete energy systems that can improve waste management compared to the current “once-through/no separation” approach. Future work will further increase confidence in potential solutions, optimize solutions for the mixtures of objectives, and develop attractive development and deployment paths for selected options. This will allow the nation to address nearer-term issues such as avoiding the need for additional geological repositories while making nuclear energy a more sustainable energy option for the long-term. While the Generation IV Initiative is exploring multiple reactor options for future nuclear energy for both electricity generation and additional applications, the AFCI is assessing fuel cycles options for either a continuation or expansion of nuclear energy in the United States. This report compares strategies and technology options for managing the associated spent fuel. There are four major potential strategies, as follows: · The current U.S. strategy is once through: standard nuclear power plants, standard fuel burnup, direct geological disposal of spent fuel. Variants include higher burnup fuels in water-cooled power plants, once-through gas-cooled power plants, and separation (without recycling) of spent fuel to reduce the number and cost of geological waste packages. · The second strategy is thermal recycle, recycling some fuel components in thermal reactors. This strategy extends the useful life of

  15. Overview of advanced fuel fusion

    International Nuclear Information System (INIS)

    The status and issues related to the development of advanced fuel fusion are discussed. D-3He is a key advanced fuel since it has the potential of igniting in a variety of confinement concepts. However, to obtain a plentiful source of 3He, either lunar mining or breeding becomes necessary. Highly non-Maxwellian plasmas, such as might occur in beam-beam fusion concepts, are necessary to address fuels like p-11B which have the added advantages of a more aneutronic character and plentiful fuel supply. Such plasmas appear very difficult to achieve but several possible approaches such as electrostatic confinement are noted. 52 refs., 13 figs, 5 tabs

  16. Advanced fuels for fast reactors

    International Nuclear Information System (INIS)

    Full text: In addition to traditional fast reactor fuels that contain Uranium and Plutonium, the advanced fast reactor fuels are likely to include the minor actinides [Neptunium (Np), Americium (Am) and Curium (Cm)]. Such fuels are also referred to as transmutation fuels. The goal of transmutation fuel development programs is to develop and qualify a nuclear fuel system that performs all of the functions of a traditional fast spectrum nuclear fuel while destroying recycled actinides. Oxide, metal, nitride, and carbide fuels are candidates under consideration for this application, based on historical knowledge of fast reactor fuel development and specific fuel tests currently being conducted in international transmutation fuel development programs. Early fast reactor developers originally favored metal alloy fuel due to its high density and potential for breeder operation. The focus of pressurized water reactor development on oxide fuel and the subsequent adoption by the commercial nuclear power industry, however, along with early issues with low burnup potential of metal fuel (now resolved), led later fast reactor development programs to favor oxide fuels. Carbide and nitride fuels have also been investigated but are at a much lower state of development than metal and oxide fuels, with limited large scale reactor irradiation experience. Experience with both metal and oxide fuels has established that either fuel type will meet performance and reliability goals for a plutonium fueled fast spectrum test reactor, both demonstrating burnup capability of up to 20 at.% under normal operating conditions, when clad with modified austenitic or ferritic martensitic stainless steel alloys. Both metal and oxide fuels have been shown to exhibit sufficient margin to failure under transient conditions for successful reactor operation. Summary of selected fuel material properties taken are provided in the paper. The main challenge for the development of transmutation fast reactor

  17. ARPA advanced fuel cell development

    Energy Technology Data Exchange (ETDEWEB)

    Dubois, L.H.

    1995-08-01

    Fuel cell technology is currently being developed at the Advanced Research Projects Agency (ARPA) for several Department of Defense applications where its inherent advantages such as environmental compatibility, high efficiency, and low noise and vibration are overwhelmingly important. These applications range from man-portable power systems of only a few watts output (e.g., for microclimate cooling and as direct battery replacements) to multimegawatt fixed base systems. The ultimate goal of the ARPA program is to develop an efficient, low-temperature fuel cell power system that operates directly on a military logistics fuel (e.g., DF-2 or JP-8). The absence of a fuel reformer will reduce the size, weight, cost, and complexity of such a unit as well as increase its reliability. In order to reach this goal, ARPA is taking a two-fold, intermediate time-frame approach to: (1) develop a viable, low-temperature proton exchange membrane (PEM) fuel cell that operates directly on a simple hydrocarbon fuel (e.g., methanol or trimethoxymethane) and (2) demonstrate a thermally integrated fuel processor/fuel cell power system operating on a military logistics fuel. This latter program involves solid oxide (SOFC), molten carbonate (MCFC), and phosphoric acid (PAFC) fuel cell technologies and concentrates on the development of efficient fuel processors, impurity scrubbers, and systems integration. A complementary program to develop high performance, light weight H{sub 2}/air PEM and SOFC fuel cell stacks is also underway. Several recent successes of these programs will be highlighted.

  18. Advanced Recycling Reactor with Minor Actinide Fuel

    International Nuclear Information System (INIS)

    The Advanced Recycling Reactor (ARR) with minor actinide fuel has been studied. This paper presents the pre-conceptual design of the ARR proposed by the International Nuclear Recycling Alliance (INRA) for FOA study sponsored by DOE of the United States of America (U.S.). Although the basic reactor concept is technically mature, it is not suitable for commercial use due to the need to reduce capital costs. As a result of INRA's extensive experience, it is anticipated that a non-commercial ARR1 will be viable and meet U.S. requirements by 2025. Commercial Advanced Recycling Reactor (ARR) operations are expected to be feasible in competition with LWRs by 2050, based on construction of ARR2 in 2035. The ARR based on the Japan Sodium-cooled Fast Reactor (JSFR) is a loop-typed sodium cooled reactor with MOX fuel that is selected because of much experience of SFRs in the world. Major features of key technology enhancements incorporated into the ARR are the following: Decay heat can be removed by natural circulation to improve safety. The primary cooling system consists of two-loop system and the integrated IHX/Pump to improve economics. The steam generator with the straight double-walled tube is used to improve reliability. The reactor core of the ARR1 is 70 cm high and the volume fraction of fuel is 31.6%. The conversion ratio of fissile is set up less than 0.65 and the amount of burned TRU is 45-51 kg/TWeh. According to survey of more effective TRU burning core, the oxide fuel core containing high TRU (MA 15%, Pu 35% average) with moderate pins of 12% arranged driver fuel assemblies can decrease TRU conversion ratio to 0.33 and improve TRU burning capability to 67 kg/TWeh. The moderator can enhance TRU burning, while increasing the Doppler effect and reducing the positive sodium void effect. High TRU fraction promotes TRU burning by curbing plutonium production. High Am fraction and Am blanket promote Am transmutation. The ARR1 consists of a reactor building (including

  19. CANDU: Shortest path to advanced fuel cycles

    International Nuclear Information System (INIS)

    Full text: The global nuclear renaissance exhibiting itself in the form of new reactor build programs is rapidly gaining momentum. Many countries are seeking to expand the use of economical and carbon-free nuclear energy to meet growing electricity demand and manage global climate change challenges. Nuclear power construction programs that are being proposed in many countries will dramatically increase the demand on uranium resources. The projected life-long uranium consumption rates for these reactors will surpass confirmed uranium reserves. Therefore, securing sufficient uranium resources and taking corresponding measures to ensure the availability of long-term and stable fuel resources for these nuclear power plants is a fundamental requirement for business success. Increasing the utilization of existing uranium fuel resources and implementing the use of alternate fuels in CANDU reactors is an important element to meet this challenge. The CANDU heavy water reactor has unequalled flexibility for using a variety of fuels, such as Natural Uranium (NU), Low Enriched Uranium (LEU), Recycled Uranium (RU), Mixed Oxide (MOX), and thorium. This CANDU feature has not been used to date simply due to lack of commercial drivers. The capability is anchored around a versatile pressure tube design, simple fuel bundle, on-power refuelling, and high neutron economy of the CANDU concept. Atomic Energy of Canada Limited (AECL) has carried out theoretical and experimental investigations on various advanced fuel cycles, including thorium, over many years. Two fuels are selected as the subject of this paper: Natural Uranium Equivalent (NUE) and thorium. NUE fuel is developed by combining RU and depleted uranium (DU) in such a manner that the resulting NUE fuel is neutronically equivalent to NU fuel. RU is recovered from reprocessed light water reactor (LWR) fuel and has a nominal 235U concentration of approximately 0.9 wt%. This concentration is higher than NU used in CANDU reactors

  20. Advanced PWR fuel design concepts

    International Nuclear Information System (INIS)

    For nearly 15 years, Combustion Engineering has provided pressurized water reactor fuel with the features most suppliers are now introducing in their advanced fuel designs. Zircaloy grids, removable upper end fittings, large fission gas plenum, high burnup, integral burnable poisons and sophisticated analytical methods are all features of C-E standard fuel which have been well proven by reactor performance. C-E's next generation fuel for pressurized water reactors features 24-month operating cycles, optimal lattice burnable poisons, increased resistance to common industry fuel rod failure mechanisms, and hardware and methodology for operating margin improvements. Application of these various improvements offer continued improvement in fuel cycle economics, plant operation and maintenance. (author)

  1. Advanced fuel chemistry for advanced engines.

    Energy Technology Data Exchange (ETDEWEB)

    Taatjes, Craig A.; Jusinski, Leonard E.; Zador, Judit; Fernandes, Ravi X.; Miller, James A.

    2009-09-01

    Autoignition chemistry is central to predictive modeling of many advanced engine designs that combine high efficiency and low inherent pollutant emissions. This chemistry, and especially its pressure dependence, is poorly known for fuels derived from heavy petroleum and for biofuels, both of which are becoming increasingly prominent in the nation's fuel stream. We have investigated the pressure dependence of key ignition reactions for a series of molecules representative of non-traditional and alternative fuels. These investigations combined experimental characterization of hydroxyl radical production in well-controlled photolytically initiated oxidation and a hybrid modeling strategy that linked detailed quantum chemistry and computational kinetics of critical reactions with rate-equation models of the global chemical system. Comprehensive mechanisms for autoignition generally ignore the pressure dependence of branching fractions in the important alkyl + O{sub 2} reaction systems; however we have demonstrated that pressure-dependent 'formally direct' pathways persist at in-cylinder pressures.

  2. Advanced thermally stable jet fuels

    Energy Technology Data Exchange (ETDEWEB)

    Schobert, H.H.

    1999-01-31

    The Pennsylvania State University program in advanced thermally stable coal-based jet fuels has five broad objectives: (1) Development of mechanisms of degradation and solids formation; (2) Quantitative measurement of growth of sub-micrometer and micrometer-sized particles suspended in fuels during thermal stressing; (3) Characterization of carbonaceous deposits by various instrumental and microscopic methods; (4) Elucidation of the role of additives in retarding the formation of carbonaceous solids; (5) Assessment of the potential of production of high yields of cycloalkanes by direct liquefaction of coal. Future high-Mach aircraft will place severe thermal demands on jet fuels, requiring the development of novel, hybrid fuel mixtures capable of withstanding temperatures in the range of 400--500 C. In the new aircraft, jet fuel will serve as both an energy source and a heat sink for cooling the airframe, engine, and system components. The ultimate development of such advanced fuels requires a thorough understanding of the thermal decomposition behavior of jet fuels under supercritical conditions. Considering that jet fuels consist of hundreds of compounds, this task must begin with a study of the thermal degradation behavior of select model compounds under supercritical conditions. The research performed by The Pennsylvania State University was focused on five major tasks that reflect the objectives stated above: Task 1: Investigation of the Quantitative Degradation of Fuels; Task 2: Investigation of Incipient Deposition; Task 3: Characterization of Solid Gums, Sediments, and Carbonaceous Deposits; Task 4: Coal-Based Fuel Stabilization Studies; and Task 5: Exploratory Studies on the Direct Conversion of Coal to High Quality Jet Fuels. The major findings of each of these tasks are presented in this executive summary. A description of the sub-tasks performed under each of these tasks and the findings of those studies are provided in the remainder of this volume

  3. ADVANCED FUELS CAMPAIGN 2013 ACCOMPLISHMENTS

    Energy Technology Data Exchange (ETDEWEB)

    Not Listed

    2013-10-01

    The mission of the Advanced Fuels Campaign (AFC) is to perform Research, Development, and Demonstration (RD&D) activities for advanced fuel forms (including cladding) to enhance the performance and safety of the nation’s current and future reactors; enhance proliferation resistance of nuclear fuel; effectively utilize nuclear energy resources; and address the longer-term waste management challenges. This includes development of a state-of-the art Research and Development (R&D) infrastructure to support the use of “goal-oriented science-based approach.” In support of the Fuel Cycle Research and Development (FCRD) program, AFC is responsible for developing advanced fuels technologies to support the various fuel cycle options defined in the Department of Energy (DOE) Nuclear Energy Research and Development Roadmap, Report to Congress, April 2010. Accomplishments made during fiscal year (FY) 2013 are highlighted in this report, which focuses on completed work and results. The process details leading up to the results are not included; however, the technical contact is provided for each section.

  4. Advanced Fuels Campaign Execution Plan

    Energy Technology Data Exchange (ETDEWEB)

    Kemal Pasamehmetoglu

    2011-09-01

    The purpose of the Advanced Fuels Campaign (AFC) Execution Plan is to communicate the structure and management of research, development, and demonstration (RD&D) activities within the Fuel Cycle Research and Development (FCRD) program. Included in this document is an overview of the FCRD program, a description of the difference between revolutionary and evolutionary approaches to nuclear fuel development, the meaning of science-based development of nuclear fuels, and the 'Grand Challenge' for the AFC that would, if achieved, provide a transformational technology to the nuclear industry in the form of a high performance, high reliability nuclear fuel system. The activities that will be conducted by the AFC to achieve success towards this grand challenge are described and the goals and milestones over the next 20 to 40 year period of research and development are established.

  5. Advanced Fuels Campaign Execution Plan

    Energy Technology Data Exchange (ETDEWEB)

    Kemal Pasamehmetoglu

    2010-10-01

    The purpose of the Advanced Fuels Campaign (AFC) Execution Plan is to communicate the structure and management of research, development, and demonstration (RD&D) activities within the Fuel Cycle Research and Development (FCRD) program. Included in this document is an overview of the FCRD program, a description of the difference between revolutionary and evolutionary approaches to nuclear fuel development, the meaning of science-based development of nuclear fuels, and the “Grand Challenge” for the AFC that would, if achieved, provide a transformational technology to the nuclear industry in the form of a high performance, high reliability nuclear fuel system. The activities that will be conducted by the AFC to achieve success towards this grand challenge are described and the goals and milestones over the next 20 to 40 year period of research and development are established.

  6. Advanced Fuel Cycle Cost Basis

    Energy Technology Data Exchange (ETDEWEB)

    D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert

    2007-04-01

    This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 26 cost modules—24 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, and high-level waste.

  7. Advanced Fuel Cycle Cost Basis

    Energy Technology Data Exchange (ETDEWEB)

    D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert; E. Schneider

    2009-12-01

    This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 25 cost modules—23 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste.

  8. Advanced Fuel Cycle Cost Basis

    Energy Technology Data Exchange (ETDEWEB)

    D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert; E. Schneider

    2008-03-01

    This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 25 cost modules—23 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste.

  9. Nuclear propulsion technology advanced fuels technology

    Science.gov (United States)

    Stark, Walter A., Jr.

    1993-01-01

    Viewgraphs on advanced fuels technology are presented. Topics covered include: nuclear thermal propulsion reactor and fuel requirements; propulsion efficiency and temperature; uranium fuel compounds; melting point experiments; fabrication techniques; and sintered microspheres.

  10. Advanced Thermally Stable Jet Fuels

    Energy Technology Data Exchange (ETDEWEB)

    A. Boehman; C. Song; H. H. Schobert; M. M. Coleman; P. G. Hatcher; S. Eser

    1998-01-01

    The Penn State program in advanced thermally stable jet fuels has five components: 1) development of mechanisms of degradation and solids formation; 2) quantitative measurement of growth of sub-micrometer and micrometer-sized particles during thermal stressing; 3) characterization of carbonaceous deposits by various instrumental and microscopic methods; 4) elucidation of the role of additives in retarding the formation of carbonaceous solids; and 5) assessment of the potential of producing high yields of cycloalkanes and hydroaromatics from coal.

  11. Advanced Fuel Pellet Materials and Fuel Rod Design for Water Cooled Reactors. Proceedings of a Technical Committee Meeting

    International Nuclear Information System (INIS)

    The economics of current nuclear power plants have improved through increased fuel burnup and longer fuel cycles, i.e. increasing the effective time that fuel remains in the reactor core and the amount of energy it generates. Efficient consumption of fissile material in the fuel element before it is discharged from the reactor means that less fuel is required over the reactor's life cycle, which results in lower amounts of fresh fuel, lower spent fuel storage costs, and less waste for ultimate disposal. Better utilization of fissile nuclear materials, as well as more flexible power manoeuvring, place challenging operational demands on materials used in reactor components, and first of all, on fuel and cladding materials. It entails increased attention to measures ensuring desired in-pile fuel performance parameters that require adequate improvements in fuel material properties and fuel rod designs. These are the main reasons that motivated the IAEA Technical Working Group on Fuel Performance and Technology (TWG-FPT) to recommend the organization of a Technical Committee Meeting on Advanced Fuel Pellet Materials and Fuel Rod Designs for Power Reactors. The proposal was supported by the IAEA TWGs on Advanced Technologies for Light and Heavy Water-Cooled Reactors (TWG-LWR and TWG-HWR), and the meeting was held at the invitation of the Government of Switzerland at the Paul Scherrer Institute in Villigen, from 23 to 26 November 2009. This was the third IAEA meeting on these subjects (the first was held in 1996 in Tokyo, Japan, and the second in 2003 in Brussels, Belgium), which reflects the continuous interest in the above issues among Member States. The purpose of the meeting was to review the current status in the development of fuel pellet materials and to explore recent improvements in fuel rod designs for light and heavy water cooled power reactors. The meeting was attended by 45 specialists representing fuel vendors, nuclear utilities, research and development

  12. Zirconia-magnesia inert matrix fuel and waste form: Synthesis, characterization and chemical performance in an advanced fuel cycle

    Science.gov (United States)

    Holliday, Kiel Steven

    There is a significant buildup in plutonium stockpiles throughout the world, because of spent nuclear fuel and the dismantling of weapons. The radiotoxicity of this material and proliferation risk has led to a desire for destroying excess plutonium. To do this effectively, it must be fissioned in a reactor as part of a uranium free fuel to eliminate the generation of more plutonium. This requires an inert matrix to volumetrically dilute the fissile plutonium. Zirconia-magnesia dual phase ceramic has been demonstrated to be a favorable material for this task. It is neutron transparent, zirconia is chemically robust, magnesia has good thermal conductivity and the ceramic has been calculated to conform to current economic and safety standards. This dissertation contributes to the knowledge of zirconia-magnesia as an inert matrix fuel to establish behavior of the material containing a fissile component. First, the zirconia-magnesia inert matrix is synthesized in a dual phase ceramic containing a fissile component and a burnable poison. The chemical constitution of the ceramic is then determined. Next, the material performance is assessed under conditions relevant to an advanced fuel cycle. Reactor conditions were assessed with high temperature, high pressure water. Various acid solutions were used in an effort to dissolve the material for reprocessing. The ceramic was also tested as a waste form under environmental conditions, should it go directly to a repository as a spent fuel. The applicability of zirconia-magnesia as an inert matrix fuel and waste form was tested and found to be a promising material for such applications.

  13. Fuel for advanced CANDU reactors

    International Nuclear Information System (INIS)

    The CANDU reactor system has proven itself to be a world leader in terms of station availability and low total unit energy cost. In 1985 for example, four of the top ten reactor units in the world were CANDU reactors operating in South Korea and Canada. This excellent operating record requires an equivalent performance record of the low-cost, natural uranium fuel. Future CANDU reactors will be an evolution of the present design. Engineering work is under way to refine the existing CANDU 600 and to incorporate state-of-the-art technology, reducing the capital cost and construction schedule. In addition, a smaller CANDU 300 plant has been designed using proven CANDU 600 technology and components but with an innovative new plant layout that makes it cost competitive with coal fired plants. For the long term, work on advanced fuel cycles and major system improvements is underway ensuring that CANDU plants will stay competitive well into the next century

  14. Accelerating fissile fuel breeding in FBR with natural safety features%加速增产核燃料的天然安全“核热泉”快中子增殖堆

    Institute of Scientific and Technical Information of China (English)

    吕应中

    2012-01-01

    为保证21世纪中国经济的持续稳定地高速增长,必须充分发挥核能的巨大潜力,使之配合其他可再生能源同步增长,及早大规模替代煤炭等化石能源.由于目前国内大量兴建的核电站以压水堆为主,需要消费大量天然铀资源,倚靠廉价铀供应难于维持长期增长,必须依靠快中子增殖生产人造裂变燃料——钚,才能摆脱天然铀原料短缺的束缚.然而,传统的快中子增殖堆的核燃料增产速度较慢,难于配合中国核电的高速增长.本文介绍一种先进快中于增殖堆(AFBR)方案,其中利用在线连续换料的空心球形燃料元件,依靠载热剂的出人口之间的温度差实现满功率自然循环,可以成倍地提高燃料比功率与核燃料增殖速度.本快中子增殖堆改进了俄罗斯称为“天然安全”的BREST铅冷快堆设计方案,成为无须人为控制的“核热泉”,它能在不设置加压泵及高位铅池的情况下,自动按外部负荷需要供应必要的热量,完全依靠自然循环将全部裂变热能及停堆后堆芯余热散出,不至对环境产生放射性污染.%To guarantee the rapid growth of the Chinese economy in 21" century, nuclear energy should be fully exploited, together with other renewable energies to replace coal and other depletive fossil fuels. Unfortunately, the major Chinese nuclear power plants under construction are mostly PWRs that would consume a lot of natural uranium during their operations. The availability of cheap natural uranium could seriously constraint the Chinese nuclear power development, unless artificial fissile fuel-plutonium is supplied from fast breeder reactors. The fissile nuclei production rate in the traditional fast breeders, however, seems too slow to match the rapid growth of nuclear power. A concept of the advanced fast breeder (AFBR) is introduced, therefore, to greatly accelerating the fissile fuel breeding process. In said breeder, the spherical hollow

  15. Advanced research reactor fuel development

    International Nuclear Information System (INIS)

    The fabrication technology of the U3Si fuel dispersed in aluminum for the localization of HANARO driver fuel has been launches. The increase of production yield of LEU metal, the establishment of measurement method of homogeneity, and electron beam welding process were performed. Irradiation test under normal operation condition, had been carried out and any clues of the fuel assembly breakdown was not detected. The 2nd test fuel assembly has been irradiated at HANARO reactor since 17th June 1999. The quality assurance system has been re-established and the eddy current test technique has been developed. The irradiation test for U3Si2 dispersed fuels at HANARO reactor has been carried out in order to compare the in-pile performance of between the two types of U3Si2 fuels, prepared by both the atomization and comminution processes. KAERI has also conducted all safety-related works such as the design and the fabrication of irradiation rig, the analysis of irradiation behavior, thermal hydraulic characteristics, stress analysis for irradiation rig, and thermal analysis fuel plate, for the mini-plate prepared by international research cooperation being irradiated safely at HANARO. Pressure drop test, vibration test and endurance test were performed. The characterization on powders of U-(5.4 ∼ 10 wt%) Mo alloy depending on Mo content prepared by rotating disk centrifugal atomization process was carried out in order to investigate the phase stability of the atomized U-Mo alloy system. The γ-U phase stability and the thermal compatibility of atomized U-16at.%Mo and U-14at.%Mo-2at.%X(: Ru, Os) dispersion fuel meats at an elevated temperature have been investigated. The volume increases of U-Mo compatibility specimens were almost the same as or smaller than those of U3Si2. However the atomized alloy fuel exhibited a better irradiation performance than the comminuted alloy. The RERTR-3 irradiation test of nano-plates will be conducted in the Advanced Test Reactor(ATR). 49

  16. Advanced research reactor fuel development

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Kyu; Pak, H. D.; Kim, K. H. [and others

    2000-05-01

    -plates will be conducted in the Advanced Test Reactor(ATR). 49 compacts with a uranium density of 8 gU/cc consist of 7 different atomized uranium-molybdenum alloy powders. The tensile strength increased and the elongation decreased with increasing the volume fraction of U-10Mo powders in dispersion fuel. The tensile strength was lower and elongation was larger in dispersion fuel using atomized U-10Mo powders than that using comminuted fuel powders. The green strength of the comminuted powder compacts was about twice as large as that of the atomized powder compacts. It is suggested that the compacting condition required to fabricate the atomized powder compacts is over the 350MPa. The comminuted irregular shaped particles and smaller particle size of fuel powders showed improved homogeneity of powder mixture. The homogeneity of powder mixtures increased to a minimum at approximately 0.10 wt% moisture and then decreased with moisture content.

  17. Advanced Fuels Campaign FY 2015 Accomplishments Report

    Energy Technology Data Exchange (ETDEWEB)

    Braase, Lori Ann [Idaho National Lab. (INL), Idaho Falls, ID (United States); Carmack, William Jonathan [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-10-29

    The mission of the Advanced Fuels Campaign (AFC) is to perform research, development, and demonstration (RD&D) activities for advanced fuel forms (including cladding) to enhance the performance and safety of the nation’s current and future reactors; enhance proliferation resistance of nuclear fuel; effectively utilize nuclear energy resources; and address the longer-term waste management challenges. This report is a compilation of technical accomplishment summaries for FY-15. Emphasis is on advanced accident-tolerant LWR fuel systems, advanced transmutation fuels technologies, and capability development.

  18. A decade of advances in metallic fuel

    International Nuclear Information System (INIS)

    Significant advances in the understanding of behavior and performance of metallic fuels to high burnup have been achieved over the past four decades. Metallic fuels were the first fuels for liquid-metal-cooled fast reactors (LMR) but in the late 1960's worldwide interest turned toward ceramic fuels before the full potential of metallic fuel could be achieved. Now metallic fuels are recognized as a preferred viable option with regard to safety, integral fuel cycle, waste minimization and deployment economics. This paper reviews the key advances in the last decade and highlights the behavior and performance features which have demonstrated a much greater potential than previously expected

  19. Experiences and Trends of Manufacturing Technology of Advanced Nuclear Fuels

    International Nuclear Information System (INIS)

    The 'Atoms for Peace' mission initiated in the mid-1950s paved the way for the development and deployment of nuclear fission reactors as a source of heat energy for electricity generation in nuclear power reactors and as a source of neutrons in non-power reactors for research, materials irradiation, and testing and production of radioisotopes. The fuels for nuclear reactors are manufactured from natural uranium (∼99.3% 238U + ∼0.7% 235U) and natural thorium (∼100% 232Th) resources. Currently, most power and research reactors use 235U, the only fissile isotope found in nature, as fuel. The fertile isotopes 238U and 232Th are transmuted in the reactor to human-made 239Pu and 233U fissile isotopes, respectively. Likewise, minor actinides (MA) (Np, Am and Cm) and other plutonium isotopes are also formed by a series of neutron capture reactions with 238U and 235U. Long term sustainability of nuclear power will depend to a great extent on the efficient, safe and secure utilization of fissile and fertile materials. Light water reactors (LWRs) account for more than 82% of the operating reactors, followed by pressurized heavy water reactors (PHWRs), which constitute ∼10% of reactors. LWRs will continue to dominate the nuclear power market for several decades, as long as economically viable natural uranium resources are available. Currently, the plutonium obtained from spent nuclear fuel is subjected to mono recycling in LWRs as uranium-plutonium mixed oxide (MOX), containing up to 12% PuO2, in a very limited way. The reprocessed uranium (RepU) is also re-enriched and recycled in LWRs in a few countries. Unfortunately, the utilization of natural uranium resources in thermal neutron reactors is 2 and MOX fuel technology has matured during the past five decades. These fuels are now being manufactured, used and reprocessed on an industrial scale. Mixed uranium- plutonium monocarbide (MC), mononitride (MN) and U-Pu-Zr alloys are recognized as advanced fuels for sodium

  20. The viability of plutonium-consuming reactors and their associated fuel cycles for the disposition of surplus fissile material

    International Nuclear Information System (INIS)

    The National Academy of Sciences (NAS) has suggested that surplus plutonium should be made to satisfy the spent fuel standard, i.e., that plutonium be made as inaccessible as the existing and growing inventory of plutonium in spent commercial nuclear fuel currently being generated domestically. The utilization of plutonium-based fuels in reactors is one way in which the spent fuel standard can be achieved. For this reason and others, the NAS recommended that plutonium-based fuels be considered for the disposition of plutonium declared surplus to national security requirements. The NAS gave special emphasis to water-cooled reactors using mixed oxide (MOX) fuels. This paper examines the technical viability of water-cooled reactors to use MOX fuels, considering proposed new light water reactors in the US, operating light water reactors in the US, and operating heavy-water moderated CANDU reactors in Canada. This paper focuses primarily on the effects of MOX utilization on reactor performance and operations and on the cradle-to-grave infrastructure that was predicated on low enriched uranium (LEU) fuels. The paper will describe the operations and facilities that are required to convert existing plutonium inventories into feed material for fuel fabrication and the fuel cycle that relates to utilizing MOX fuels

  1. Neutron emission tomography for nuclear fissile materials safeguards

    International Nuclear Information System (INIS)

    Any nondestructive method for fissile assay in spent nuclear fuel must sample a substantial portion of the total number of reactor fuel elements to account for the radial and axial asymmetry of fissile material distribution that may arise during fuel burnup. Cross-sectional scans of the spatial distribution of the fissile isotopes can be tomographically reconstructed while a spent fuel element is being assayed in a lead slowing-down-time (SDT) spectrometer by employing image reconstruction techniques. The SDT approach employs threshold detectors (fission chambers), which measure the induced prompt fast fission neutrons emitted from the fissile isotopes in spent fuel while being interrogated with a slowing pulse of source neutrons. The objectives of this work were to examine to what extent a cross section representing an array of partially void (missing pins) and intact fuel pins could be reconstructed and how sensitive this method is to diversion, spatial burnup distribution, and gross fissile contents consistency

  2. Chemical Kinetic Modeling of Advanced Transportation Fuels

    Energy Technology Data Exchange (ETDEWEB)

    PItz, W J; Westbrook, C K; Herbinet, O

    2009-01-20

    Development of detailed chemical kinetic models for advanced petroleum-based and nonpetroleum based fuels is a difficult challenge because of the hundreds to thousands of different components in these fuels and because some of these fuels contain components that have not been considered in the past. It is important to develop detailed chemical kinetic models for these fuels since the models can be put into engine simulation codes used for optimizing engine design for maximum efficiency and minimal pollutant emissions. For example, these chemistry-enabled engine codes can be used to optimize combustion chamber shape and fuel injection timing. They also allow insight into how the composition of advanced petroleum-based and non-petroleum based fuels affect engine performance characteristics. Additionally, chemical kinetic models can be used separately to interpret important in-cylinder experimental data and gain insight into advanced engine combustion processes such as HCCI and lean burn engines. The objectives are: (1) Develop detailed chemical kinetic reaction models for components of advanced petroleum-based and non-petroleum based fuels. These fuels models include components from vegetable-oil-derived biodiesel, oil-sand derived fuel, alcohol fuels and other advanced bio-based and alternative fuels. (2) Develop detailed chemical kinetic reaction models for mixtures of non-petroleum and petroleum-based components to represent real fuels and lead to efficient reduced combustion models needed for engine modeling codes. (3) Characterize the role of fuel composition on efficiency and pollutant emissions from practical automotive engines.

  3. Advanced fuel development at AECL: What does the future hold for CANDU fuels/fuel cycles?

    International Nuclear Information System (INIS)

    This paper outlines advanced fuel development at AECL. It discusses expanding the limits of fuel utilization, deploy alternate fuel cycles, increase fuel flexibility, employ recycled fuels; increase safety and reliability, decrease environmental impact and develop proliferation resistant fuel and fuel cycle.

  4. Disposal costs for advanced CANDU fuel cycles

    International Nuclear Information System (INIS)

    The CANDU reactor can 'burn' a wide range of fuels without modification to the reactor system, including natural uranium, slightly enriched uranium, mixed oxide and spent LWR fuels. The economic feasibility of the advanced fuel cycles requires consideration of their disposal costs. Preliminary cost analyses for the disposal of spent CANDU-SEU (Slightly Enriched Uranium) and CANDU-DUPIC (Direct Use of spent PWR fuel In CANDU) fuels have been performed and compared to the internationally published costs for the direct disposal of spent CANDU and LWR fuels. The analyses show significant economic advantages in the disposal costs of CANDU-SEU and CANDU-DUPIC fuels. (author)

  5. Uncertainty Analyses of Advanced Fuel Cycles

    International Nuclear Information System (INIS)

    The Department of Energy is developing technology, experimental protocols, computational methods, systems analysis software, and many other capabilities in order to advance the nuclear power infrastructure through the Advanced Fuel Cycle Initiative (AFDI). Our project, is intended to facilitate will-informed decision making for the selection of fuel cycle options and facilities for development

  6. Uncertainty Analyses of Advanced Fuel Cycles

    Energy Technology Data Exchange (ETDEWEB)

    Laurence F. Miller; J. Preston; G. Sweder; T. Anderson; S. Janson; M. Humberstone; J. MConn; J. Clark

    2008-12-12

    The Department of Energy is developing technology, experimental protocols, computational methods, systems analysis software, and many other capabilities in order to advance the nuclear power infrastructure through the Advanced Fuel Cycle Initiative (AFDI). Our project, is intended to facilitate will-informed decision making for the selection of fuel cycle options and facilities for development.

  7. Advanced fuel cycles for CANDU reactors

    International Nuclear Information System (INIS)

    The current natural uranium-fuelled CANDU system is a world leader, both in terms of overall performance and uranium utilization. Moreover, the CANDU reactor is capable of using many different advanced fuel cycles, with improved uranium utilization relative to the natural uranium one-through cycle. This versatility would enable CANDU to maintain its competitive edge in uranium utilization as improvements are made by the competition. Several CANDU fuel cycles are symbiotic with LWRs, providing an economical vehicle for the recycle of uranium and/or plutonium from discharges LWR fuel. The slightly enriched uranium (SEU) fuel cycle is economically attractive now, and this economic benefit will increase with anticipated increases in the cost of natural uranium, and decreases in the cost of fuel enrichment. The CANFLEX fuel bundle, an advanced 43-element design, will ensure that the full benefits of SEU, and other advanced fuel cycles, can be achieved in the CANDU reactor. 25 refs

  8. Upgrading of raw oil into advanced fuel

    Energy Technology Data Exchange (ETDEWEB)

    1991-10-01

    The overall objective of the research effort is the determination of the minimum processing requirements to produce high energy density fuels (HEDF) having acceptable fuel specifications. The program encompasses assessing current technology capability; selecting acceptable processing and refining schemes; and generating samples of advanced test fuels. The Phase I Baseline Program is intended to explore the processing alternatives for producing advanced HEDF from two raw synfuel feedstocks, one from Mild Coal Gasification as exemplified by the COALITE process and one from Colorado shale oil. Eight key tasks have been identified as follows: (1) Planning and Environmental Permitting; (2) Transporting and Storage of Raw Fuel Sources and Products; (3) Screening of Processing and Upgrading Schemes; (4) Proposed Upgrading Schemes for Advanced Fuel; (5) Upgrading of Raw Oil into Advanced Fuel (6) Packaging and Shipment of Advanced Fuels; (7) Updated Technical and Economic Assessment; and, (8) Final Report of Phase I Efforts. This topical report summarizes the operations and results of the Phase I Task 5 sample preparation program. The specific objectives of Task 5 were to: Perform laboratory characterization tests on the raw COALITE feed, the intermediate liquids to the required hydroprocessing units and final advanced fuels and byproducts; and produce a minimum of 25-gal of Category I test fuel for evaluation by DOE and its contractors.

  9. Characterization of a suspect nuclear fuel rod in a case of illegal international traffic of fissile material.

    Science.gov (United States)

    Capannesi, G; Vicini, C; Rosada, A; Avino, P

    2010-06-15

    This case study describes the characterization of a suspect rod of nuclear fuel seized in Italy: on request of the coroner, the characterization concerned the kind and the conditions of the rod, the amount and the specific characteristics of the species present in it, with particular attention to their possible use chemical and/or nuclear plants. The methodology used was based on radiochemical analyses (gammagraphic and gamma-spectrometry) whereas the comparison was performed by means of a fuel reference element working in the TRIGA nuclear reactor at Research Center of ENEA-Casaccia. The results show clearly how the exhibit was an element of nuclear fuel, how long it was irradiated, and the amount of (239)Pu produced and the (235)U consumed. Finally, even if the seized rod was briefly radiated at the "zero power" and traces of fission products and plutonium were found, it would be still usable as "fresh" fuel in a reactor type TRIGA if it had not been intercepted by Italian police authorities. PMID:20223608

  10. Fuels for Advanced Nuclear Energy Systems

    International Nuclear Information System (INIS)

    Fuels for advanced nuclear reactors differ greatly from conventional light water reactor fuels and vary widely between the different concepts, due differences in reactor architecture and deployment. Functional requirements of all fuel designs include (1) retention of fission products and fuel nuclides, (2) dimensional stability, and (3) maintaining a coolable geometry. In all cases, the anticipated fuel performance under normal or off-normal conditions is the limiting factor in reactor system design, and cumulative effects of increased exposure to higher burnup degrades fuel performance. In high-temperature (thermal) gas reactor systems, fuel particles of uranium dioxide or uranium oxycarbide particles are coated with layers of carbon and SiC (or ZrC). Such fuels have been used successfully to very high burnup (10-20% of heavy-metal atoms) and can withstand transient temperatures up to 1600 C. Oxide (pellet-type) and metal (pin-type) fuels clad in stainless steel tubes have been successfully used in liquid metal cooled fast reactors, attaining burnup of 20% or more of heavy-metal atoms. Those fuel designs are being adapted for actinide management missions, requiring greater contents of minor actinides (e.g. Am, Np, Cm). The current status of each fuel system is reviewed and technical challenges confronting the implementation of each fuel in the context of the entire advanced reactor fuel cycle (fabrication, reactor performance, recycle) are discussed

  11. Development of nuclear fuel. Development of CANDU advanced fuel bundle

    International Nuclear Information System (INIS)

    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)

  12. Advances in ultrasonic fuel cleaning

    International Nuclear Information System (INIS)

    Ultrasonic fuel cleaning has been demonstrated to be effective for avoiding or postponing the axial offset anomaly (AOA) in high-duty PWRs. In addition, the reduction in corrosion product inventory achieved by this method of fuel cleaning has been shown to reduce ex-core dose rates, and hence reduce personnel exposure. Through laboratory mock-up tests as well as tests on discharged fuel in the Callaway spent fuel pool, the EPRI ultrasonic fuel cleaning system was demonstrated to clean crud from fuel rods throughout the assembly without deleterious effects on fuel integrity. This paper presents details of the evaluation tests at AmerenUE's Callaway PWR, as well as the full-reload ultrasonic cleaning for Cycle 12. Ninety-six once-burned fuel assemblies were cleaned via the EPRI ultrasonic process during refueling outage 11 in April 2001. Fuel cleaning activities presented no special problems in or around the fuel pool, neither in terms of activity nor in terms of turbidity of the pool water, nor were special radiological situations encountered by personnel working in the area. The corrosion products were captured on cartridge filters designed to avoid loss of material into the fuel pool water during interim storage. Activity levels on the cartridges were maintained sufficiently low for ease of handling, processing, and shipment in Radwaste. The fuel cleaning operation was completed within a time window of approximately 48 hours. The cleaned fuel was returned to service in May 2001 for Cycle 12. To date, this fuel cycle remains free of AOA (Axial Offset Anomaly). This result is in contrast to earlier cycles of similar design and power duty, but without reload fuel cleaning. These cycles were operated at a reduced TAVE in order to mitigate AOA. Nevertheless, Cycle 11 experienced AOA starting as early as 6 GWD/MTU. Based on the favorable Cycle 12 data, the utility intends to institute fuel cleaning as a routine outage activity, thus expecting to recover the TAVE

  13. Thermochemistry of nuclear fuels in advanced reactors

    International Nuclear Information System (INIS)

    The presence of a large number of elements, accompanied with steep temperature gradient results in dynamic chemistry during nuclear fuel burn-up. Understanding this chemistry is very important for efficient and safe usage of nuclear fuels. The radioactive nature of these fuels puts lot of constraint on regulatory bodies to ensure their accident free operation in the reactors. One of the common aims of advanced fuels is to achieve high burn-up. As burn-up of the fuel increases, chemistry of fission-products becomes increasingly more important. To understand different phenomenon taking place in-pile, many out of-pile experiments are carried out. Extensive studies of thermodynamic properties, phase analysis, thermophysical property evaluation, fuel-fission product clad compatibility are carried out with relevant compounds and simulated fuels (SIMFUEL). All these data are compiled and jointly evaluated using different computational methods to predict fuel behaviour during burn-up. Only when this combined experimental and theoretical information confirms safe operation of the pin, a test pin is prepared and burnt in a test reactor. Every fuel has a different chemistry and different constraints associated with it. In this talk, various thermo-chemical aspects of some of the advanced fuels, mixed carbide, mixed nitride, 'Pu' rich MOX, 'Th' based AHWR fuels and metallic fuels will be discussed. (author)

  14. Advanced Fuel Cycle Economic Sensitivity Analysis

    Energy Technology Data Exchange (ETDEWEB)

    David Shropshire; Kent Williams; J.D. Smith; Brent Boore

    2006-12-01

    A fuel cycle economic analysis was performed on four fuel cycles to provide a baseline for initial cost comparison using the Gen IV Economic Modeling Work Group G4 ECON spreadsheet model, Decision Programming Language software, the 2006 Advanced Fuel Cycle Cost Basis report, industry cost data, international papers, the nuclear power related cost study from MIT, Harvard, and the University of Chicago. The analysis developed and compared the fuel cycle cost component of the total cost of energy for a wide range of fuel cycles including: once through, thermal with fast recycle, continuous fast recycle, and thermal recycle.

  15. Advanced compressed hydrogen fuel storage systems

    International Nuclear Information System (INIS)

    Dynetek was established in 1991 by a group of private investors, and since that time efforts have been focused on designing, improving, manufacturing and marketing advanced compressed fuel storage systems. The primary market for Dynetek fuel systems has been Natural Gas, however as the automotive industry investigates the possibility of using hydrogen as the fuel source solution in Alternative Energy Vehicles, there is a growing demand for hydrogen storage on -board. Dynetek is striving to meet the needs of the industry, by working towards developing a fuel storage system that will be efficient, economical, lightweight and eventually capable of storing enough hydrogen to match the driving range of the current gasoline fueled vehicles

  16. Advanced Fuel Cycle Economic Sensitivity Analysis

    International Nuclear Information System (INIS)

    A fuel cycle economic analysis was performed on four fuel cycles to provide a baseline for initial cost comparison using the Gen IV Economic Modeling Work Group G4 ECON spreadsheet model, Decision Programming Language software, the 2006 Advanced Fuel Cycle Cost Basis report, industry cost data, international papers, the nuclear power related cost study from MIT, Harvard, and the University of Chicago. The analysis developed and compared the fuel cycle cost component of the total cost of energy for a wide range of fuel cycles including: once through, thermal with fast recycle, continuous fast recycle, and thermal recycle

  17. Advanced Fuels Campaign FY 2011 Accomplishments Report

    Energy Technology Data Exchange (ETDEWEB)

    Not Listed

    2011-11-01

    One of the major research and development (R&D) areas under the Fuel Cycle Research and Development (FCRD) program is advanced fuels development. The Advanced Fuels Campaign (AFC) has the responsibility to develop advanced fuel technologies for the Department of Energy (DOE) using a science-based approach focusing on developing a microstructural understanding of nuclear fuels and materials. Accomplishments made during fiscal year (FY 20) 2011 are highlighted in this report, which focuses on completed work and results. The process details leading up to the results are not included; however, the technical contact is provided for each section. The order of the accomplishments in this report is consistent with the AFC work breakdown structure (WBS).

  18. Advanced fuel cycles. Proceedings of the workshop

    Energy Technology Data Exchange (ETDEWEB)

    Ospina, C.; Stanculescu, A. [eds.

    1995-12-31

    The proceedings enclose the papers presented at the workshop sessions on strategies concerning reactors and fuel cycles, on increased plutonium utilisation in LWRs, on advanced systems, complemented by the workshop summaries and recommendations. figs., tabs., refs.

  19. Advanced fuel cycles. Proceedings of the workshop

    International Nuclear Information System (INIS)

    The proceedings enclose the papers presented at the workshop sessions on strategies concerning reactors and fuel cycles, on increased plutonium utilisation in LWRs, on advanced systems, complemented by the workshop summaries and recommendations. figs., tabs., refs

  20. the effect of advanced fuel designs on fuel utilization

    International Nuclear Information System (INIS)

    Fuel management is one of the key topic in nuclear engineering. It is possible to increase fuel burnup and reactor lifetime by using advanced fuel management strategies. In order to increase the cycle lifetime, required amount of excess reactivity must be added to system. Burnable poisons can be used to compensate this excess reactivity. Usually gadolinium (Gd) is used as burnable poison. But the use of Gd presents some difficulties that have not been encountered with the use of boron

  1. US advanced LMFBR fuels development program

    International Nuclear Information System (INIS)

    Following the oil crisis in 1974, a national Advanced LMFBR Fuels Development Program was initiated in the U. S. This program was developed on the basis of the experience obtained during the exploratory years. As a result, most aspects of advanced fuels development have been expanded in the U. S. in a unified national program. The experience obtained during the exploratory phase has been summarized previously. The purpose of this paper is to describe the new program and to summarize recent major findings

  2. Westinghouse advanced fuel management system

    International Nuclear Information System (INIS)

    The Westinghouse Incore Fuel Management analysis methodology has been qualified and implemented for a broad range of fuel management strategies and operating conditions. Monitoring of the performance and additions to the qualification database provide a continuous process for methodology upgrades. Additions to the experience database include performance for other NSSS and fuel assembly designs, including large guide thimble and off-center assembly core configuration, Mixed-Oxide reload design, and Gadolinia burnable absorber. Several other functional interfaces are required for the design, safety evaluation, licensing, operation support and core monitoring of PWR cores. In-core fuel management methods need to provide the data required for these interfaces in a consistent manner (format and accuracy) so as to avoid unnecessary conservatisms that penalize operational margins. This paper reviews recent results in the Westinghouse Incore Fuel Management methodology. (author). 3 refs, 5 figs, 2 tabs

  3. Advanced Fuels Campaign Cladding & Coatings Meeting Summary

    Energy Technology Data Exchange (ETDEWEB)

    Not Listed

    2013-03-01

    The Fuel Cycle Research and Development (FCRD) Advanced Fuels Campaign (AFC) organized a Cladding and Coatings operational meeting February 12-13, 2013, at Oak Ridge National Laboratory (ORNL). Representatives from the U.S. Department of Energy (DOE), national laboratories, industry, and universities attended the two-day meeting. The purpose of the meeting was to discuss advanced cladding and cladding coating research and development (R&D); review experimental testing capabilities for assessing accident tolerant fuels; and review industry/university plans and experience in light water reactor (LWR) cladding and coating R&D.

  4. Advanced Fuels Campaign FY 2010 Accomplishments Report

    Energy Technology Data Exchange (ETDEWEB)

    Lori Braase

    2010-12-01

    The Fuel Cycle Research and Development (FCRD) Advanced Fuels Campaign (AFC) Accomplishment Report documents the high-level research and development results achieved in fiscal year 2010. The AFC program has been given responsibility to develop advanced fuel technologies for the Department of Energy (DOE) using a science-based approach focusing on developing a microstructural understanding of nuclear fuels and materials. The science-based approach combines theory, experiments, and multi-scale modeling and simulation aimed at a fundamental understanding of the fuel fabrication processes and fuel and clad performance under irradiation. The scope of the AFC includes evaluation and development of multiple fuel forms to support the three fuel cycle options described in the Sustainable Fuel Cycle Implementation Plan4: Once-Through Cycle, Modified-Open Cycle, and Continuous Recycle. The word “fuel” is used generically to include fuels, targets, and their associated cladding materials. This document includes a brief overview of the management and integration activities; but is primarily focused on the technical accomplishments for FY-10. Each technical section provides a high level overview of the activity, results, technical points of contact, and applicable references.

  5. Equipment system for advanced nuclear fuel development

    International Nuclear Information System (INIS)

    The purpose of the settlement of equipment system for nuclear Fuel Technology Development Facility(FTDF) is to build a seismic designed facility that can accommodate handling of nuclear materials including <20% enriched Uranium and produce HANARO fuel commercially, and also to establish the advanced common research equipment essential for the research on advanced fuel development. For this purpose, this research works were performed for the settlement of radiation protection system and facility special equipment for the FTDF, and the advanced common research equipment for the fuel fabrication and research. As a result, 11 kinds of radiation protection systems such as criticality detection and alarm system, 5 kinds of facility special equipment such as environmental pollution protection system and 5 kinds of common research equipment such as electron-beam welding machine were established. By the settlement of exclusive domestic facility for the research of advanced fuel, the fabrication and supply of HANARO fuel is possible and also can export KAERI-invented centrifugal dispersion fuel materials and its technology to the nations having research reactors in operation. For the future, the utilization of the facility will be expanded to universities, industries and other research institutes

  6. LANL's Role in the U.S. Fissile Material Disposition Program

    Energy Technology Data Exchange (ETDEWEB)

    Whitworth, Julia [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kay, Virginia [NA-233

    2015-02-18

    The process of Fissile Material Disposition is in part a result of the Advanced Recovery and Integrated Extraction System (ARIES), which is an agreement between the U.S. and Russia to dispose of excess plutonium used to make weapons. LANL is one sight that aides in the process of dismantling, storage and repurposing of the plutonium gathered from dismantled weapons. Some uses for the repurposed plutonium is fuel for commercial nuclear reactors which will provide energy for citizens.

  7. Fuel behavior in advanced water reactors

    International Nuclear Information System (INIS)

    Fuel rod behavior of advanced pressurized water reactors under steady state conditions has been investigated in this study. System-80+ and Westinghouse Vantage-5 fuels have been considered as advanced pressurized water reactor fuels to be analyzed. The purpose of this study is to analyze the sensitivity of ditferent models and the effect of selected design parameters on the overall fuel behavior. FRAPCON-II computer code has been used for the analyses. Different modelling options of FRAPCON-II have also been considered in these analyses. Analyses have been performed in two main parts. In the first part, effects of operating conditions on fuel behavior have been investigated. First, fuel rod response under normal operating conditions has been analyzed. Then, fuel rod response to different fuel ratings has been calculated. In the second part, in order to estimate the effect of design parameters on fuel behavior, parametric analyses have been performed. In this part, the effects of initial gap thickness, as fabricated fuel density, and initial fill gas pressure on fuel behavior have been analyzed. The computations showed that both of the fuel rods used in this study operate within the safety limits. However, FRAPCON-II modelling options have been resulted in different behavior due to their modelling characteristics. Hence, with the absence of experimental data, it is difficult to make assesment for the best fuel parameters. It is also difficult to estimate error associated with the results. To improve the performance of the code, it is necessary to develop better experimental correlations for material properties in order to analyze the eftect ot considerably different design parameters rather than nominal rod parameters

  8. Advances and highlights of the CNEA qualification program as high density fuel manufacturer for research reactors

    International Nuclear Information System (INIS)

    One of the main objectives of CNEA regarding the fuel for research reactors is the development and qualification of the manufacturing of LEU high-density fuels. The qualification programs for both types of fuels, Silicide fuel and U- x Mo fuel, are similar. They include the following activities: development and set up of the fissile compound manufacturing technology, set up of fuel plate manufacturing, fabrication and irradiation of mini plates and plates, design and fabrication of fuel assembly prototypes for irradiation, post-irradiation examination and feedback for manufacturing improvements. This paper describes the different activities performed within each program during the last year and the main advances and achievements of the programs within this period. The main achievements may be summarized in the following activities: Continuation of the irradiation of the first silicide fuel element in the R A3. Completion of the manufacturing of the second silicide fuel element, licensing and beginning of its irradiation in the R A3. Development of the HMD Process to manufacture U-Mo powder (pUMA project). Set up of fuel plates manufacturing at industrial level using U-Mo powder. Preliminary studies and the design for the irradiation of mini plates, plates and full scale fuel elements with U-Mo and 7 g U/cm3. PIE destructive studies for the P-04 silicide fuel prototype (accurate burnup determination through chemical analysis, metallography and SEM of samples from the irradiated fuel plates). Improvement and development of new characterization techniques for high density fuel plates quality control including US testing and densitometric analysis of X-ray examinations. The results obtained in this period are encouraging and also allow to foresee a wider participation of CNEA in the international effort to qualify U-Mo as a new material for the manufacturing of research reactor fuels. (author)

  9. The Conflux Fuel bundle: An Economic and Pragmatic Route to the use of Advanced Fuel Cycles in CANDU Reactors

    International Nuclear Information System (INIS)

    The CANFLEX1 bundle is being developed jointly by AECL and KAERI as a vehicle for introducing the use of enrichment and advanced fuel cycles in CANDU2 reactors. The bundle design uses smaller diameter fuel elements in the outer ring of a 43-element bundle to reduce the maximum element ratings in a CANDU fuel bundle by 20% compared to the 37-element bundle currently in use. This facilitates burnups of greater than 21,000 MW d/TAU to optimize the economic benefit available from the use of enrichment and advanced fuel cycles. A combination of this lower fuel rating, plus development work underway at Aecl to enhance the thermalhydraulic characteristics of the bundle (including both CHF3 and bundle. This provides extra flexibility in the fuel management procedures required for fuel bundles with higher fissile contents. The different bundle geometry requires flow tests to demonstrate acceptable vibration and fretting behavior of the Conflux bundle. A program to undertake the necessary range of flow tests has started at KAERI, involving the fabrication of the required bundles, and setting up for the actual tests. A program to study the fuel management requirements for slightly enriched (0.9 wt % 235 in total U) Conflux fuel has been undertaken by both Aecl and KAERI staff, and further work has started for higher enrichments. Irradiation testing of the Conflux bundle started in the NUR reactor in 1989, and a second irradiation test is due to start shortly. This paper describes the program, and reviews the status of key parts of the program

  10. Thermochemical modelling of advanced CANDU reactor fuel

    Science.gov (United States)

    Corcoran, Emily Catherine

    2009-04-01

    With an aging fleet of nuclear generating facilities, the imperative to limit the use of non-renewal fossil fuels and the inevitable need for additional electricity to power Canada's economy, a renaissance in the use of nuclear technology in Canada is at hand. The experience and knowledge of over 40 years of CANDU research, development and operation in Ontario and elsewhere has been applied to a new generation of CANDU, the Advanced CANDU Reactor (ACR). Improved fuel design allows for an extended burnup, which is a significant improvement, enhancing the safety and the economies of the ACR. The use of a Burnable Neutron Absorber (BNA) material and Low Enriched Uranium (LEU) fuel has created a need to understand better these novel materials and fuel types. This thesis documents a work to advance the scientific and technological knowledge of the ACR fuel design with respect to thermodynamic phase stability and fuel oxidation modelling. For the BNA material, a new (BNA) model is created based on the fundamental first principles of Gibbs energy minimization applied to material phase stability. For LEU fuel, the methodology used for the BNA model is applied to the oxidation of irradiated fuel. The pertinent knowledge base for uranium, oxygen and the major fission products is reviewed, updated and integrated to create a model that is applicable to current and future CANDU fuel designs. As part of this thesis, X-Ray Diffraction (XRD) and Coulombic Titration (CT) experiments are compared to the BNA and LEU models, respectively. From the analysis of the CT results, a number of improvements are proposed to enhance the LEU model and provide confidence in its application to ACR fuel. A number of applications for the potential use of these models are proposed and discussed. Keywords: CANDU Fuel, Gibbs Energy Mimimization, Low Enriched Uranium (LEU) Fuel, Burnable Neutron Absorber (BNA) Material, Coulometric Titration, X-Ray Diffraction

  11. Development of advanced HTGR fuel, 1

    International Nuclear Information System (INIS)

    Aiming at advancement of HTGR fuel ability a fabrication test of monolithic fuel rods, of which a fuel-embedded zone was surrounded by a fuel-free zone, was conducted by a low temperature isostatic press, in order to investigate fabrication conditions and properties of the monolithic fuel rods. Dimensions of the monolithic fuel rods tested in the present experiment were 14-16 mm in diameter and about 130 mm in length. Three different rubber molds were applied for fabrication of the fuel rods; (1) a mold without any device, (2) a mold, inside of which was evacuated and (3) a mold with a valve to release an inner air. In the fuels fabricated by the mold(1) many damages were involved, while producibility by the mold(2) was good, and high density graphite-matrix was obtained. The mold(3) gave a good result in producing the fuel rod. Furthermore, the properties of the graphite-matrix could be improved by fabricating with granulated graphite powder. Failure of the coated particles during pressing and sintering processes was not detected. The monolithic fuel rods has a uniform properties along an axial direction in density and Young's modulus. (author)

  12. Technical meeting on advanced fuel pellet materials and fuel rod designs for water cooled reactors. Book of abstracts

    International Nuclear Information System (INIS)

    The economics of current nuclear power plants have improved through increasing fuel burnups and fuel cycles, i.e. the effective time that fuel remains in the reactor core and the amount of energy it generates. Increasing the consumption of fissile material in the fuel element before it is discharged from the reactor means less fuel is required over the reactor's life cycle, which results in lower amounts of fresh fuel, lower spent fuel storage costs, and less waste for ultimate disposal. There has been a continuous historical increase in fuel burnup from 20-25 GWd/tU in Generation I reactors to 50-60 GWd/tU in today's light water reactors and this tendency continues in as much as technological and operational improvements make it possible. In parallel, higher enrichments are discussed, leading to a higher energy yield. For heavy water reactors slight enrichment of fuels and correspondingly growing burnups in CANDU/PHWR are driven by the same economical incentives. Higher burnups and better utilization of fissile nuclear materials (including use of MOX fuel and burnable neutron absorbers), as well as more flexible power manoeuvring, place challenging operational demands on materials used in reactor components, and first of all on fuel and cladding materials. It defines a need for increased attention to measures ensuring compliance to safety criteria related to fission gas release (to limit the internal rod pressure), pellet-cladding interaction (to avoid clad cracking combined with stress and aggressive chemical environment) and pellet-cladding mechanical interaction (to avoid clad mechanical fracture). These measures that secure desired in-pile fuel performance parameters include adequate improvements in fuel material properties and fuel rod designs. That is why the subject Technical Meeting was recommended to the IAEA in 2007 by the Technical Working Group on Fuel Performance and Technology (TWGFPT), and the recommendation was supported by the TWG on Light and

  13. Advanced and sustainable fuel cycles for innovative reactor systems

    International Nuclear Information System (INIS)

    The key objective of nuclear energy systems of the future as defined by the Generation IV road map is to provide a sustainable energy generation for the future. It includes the requirement to minimize the nuclear waste produced and thereby notably reduce the long term stewardship burden in the future. It is therefore evident that the corresponding fuel cycles will play a central role in trying to achieve these goals by creating clean waste streams which contain almost exclusively the fission products. A new concept based on a grouped separation of actinides is widely discussed in this context, but it is of course a real challenge to achieve this type of separation since technologies available today have been developed to separate actinides from each other. In France, the CEA has launched extensive research programs in the ATALANTE facility in Marcoule to develop the advanced fuel cycles for new generation reactor systems. In this so called global actinide management (GAM) concept, the actinides are extracted in a sequence of chemical reactions (grouped actinide extraction (GANEX)) and immediately reintroduced in the fuel fabrication process is to use all actinides in the energy production process. The new group separation processes can be derived as in this case from aqueous techniques but also from so-called pyrochemical partitioning processes. Significant progress was made in recent years for both routes in the frame of the European research projects PARTNEW, PYROREP and EUROPART, mainly devoted to the separation of minor actinides in the frame of partitioning and transmutation (P and T) studies. The fuels used in the new generation reactors will be significantly different from the commercial fuels of today. Because of the fuel type and the very high burn-ups reached, pyrometallurgical reprocessing could be the preferred method. The limited solubility of some of the fuel materials in acidic aqueous solutions, the possibility to have an integrated irradiation and

  14. Dynamic Simulations of Advanced Fuel Cycles

    Energy Technology Data Exchange (ETDEWEB)

    Steven J. Piet; Brent W. Dixon; Jacob J. Jacobson; Gretchen E. Matthern; David E. Shropshire

    2011-03-01

    Years of performing dynamic simulations of advanced nuclear fuel cycle options provide insights into how they could work and how one might transition from the current once-through fuel cycle. This paper summarizes those insights from the context of the 2005 objectives and goals of the U.S. Advanced Fuel Cycle Initiative (AFCI). Our intent is not to compare options, assess options versus those objectives and goals, nor recommend changes to those objectives and goals. Rather, we organize what we have learned from dynamic simulations in the context of the AFCI objectives for waste management, proliferation resistance, uranium utilization, and economics. Thus, we do not merely describe “lessons learned” from dynamic simulations but attempt to answer the “so what” question by using this context. The analyses have been performed using the Verifiable Fuel Cycle Simulation of Nuclear Fuel Cycle Dynamics (VISION). We observe that the 2005 objectives and goals do not address many of the inherently dynamic discriminators among advanced fuel cycle options and transitions thereof.

  15. Future Transient Testing of Advanced Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Jon Carmack

    2009-09-01

    The transient in-reactor fuels testing workshop was held on May 4–5, 2009 at Idaho National Laboratory. The purpose of this meeting was to provide a forum where technical experts in transient testing of nuclear fuels could meet directly with technical instrumentation experts and nuclear fuel modeling and simulation experts to discuss needed advancements in transient testing to support a basic understanding of nuclear fuel behavior under off-normal conditions. The workshop was attended by representatives from Commissariat à l'Énergie Atomique CEA, Japanese Atomic Energy Agency (JAEA), Department of Energy (DOE), AREVA, General Electric – Global Nuclear Fuels (GE-GNF), Westinghouse, Electric Power Research Institute (EPRI), universities, and several DOE national laboratories. Transient testing of fuels and materials generates information required for advanced fuels in future nuclear power plants. Future nuclear power plants will rely heavily on advanced computer modeling and simulation that describes fuel behavior under off-normal conditions. TREAT is an ideal facility for this testing because of its flexibility, proven operation and material condition. The opportunity exists to develop advanced instrumentation and data collection that can support modeling and simulation needs much better than was possible in the past. In order to take advantage of these opportunities, test programs must be carefully designed to yield basic information to support modeling before conducting integral performance tests. An early start of TREAT and operation at low power would provide significant dividends in training, development of instrumentation, and checkout of reactor systems. Early start of TREAT (2015) is needed to support the requirements of potential users of TREAT and include the testing of full length fuel irradiated in the FFTF reactor. The capabilities provided by TREAT are needed for the development of nuclear power and the following benefits will be realized by

  16. Advances in HTGR spent fuel treatment technology

    International Nuclear Information System (INIS)

    GA Technologies, Inc. has been investigating the burning of spent reactor graphite under Department of Energy sponsorship since 1969. Several deep fluidized bed burners have been used at the GA pilot plant to develop graphite burning techniques for both spent fuel recovery and volume reduction for waste disposal. Since 1982 this technology has been extended to include more efficient circulating bed burners. This paper includes updates on high-temperature gas-cooled reactor fuel cycle options and current results of spent fuel treatment testing for fluidized and advanced circulating bed burners

  17. 49 CFR 173.420 - Uranium hexafluoride (fissile, fissile excepted and non-fissile).

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 2 2010-10-01 2010-10-01 false Uranium hexafluoride (fissile, fissile excepted and non-fissile). 173.420 Section 173.420 Transportation Other Regulations Relating to Transportation PIPELINE AND HAZARDOUS MATERIALS SAFETY ADMINISTRATION, DEPARTMENT OF TRANSPORTATION HAZARDOUS MATERIALS REGULATIONS SHIPPERS-GENERAL REQUIREMENTS...

  18. CANFLEX - an advanced fuel bundle for CANDU

    International Nuclear Information System (INIS)

    The performance of CANDU pressurized heavy-water reactors, in terms of lifetime load factors, is excellent. More than 600 000 bundles containing natural-uranium fuel have been irradiated, with a low defect rate; reactor unavailability due to fuel incidents is typically zero. To maintain and improve CANDU's competitive position, Atomic Energy of Canada Limited (AECL) has an ongoing program comprising design, safety and availability improvements, advanced fuel concepts and schemes to reduce construction time. One key finding is that the introduction of slightly-enriched uranium (SEU, less than 1.5 wt% U-235 in U) offers immediate benefits for CANDU, in terms of fuelling and back-end disposal costs. The use of SEU places more demands on the fuel because of extended burnup, and an anticipated capability to load-follow also adds to the performance requirements. To ensure that the duty-cycle targets for SEU and load-following are achieved, AECL is developing a new fuel bundle, termed CANFLEX (CANdu FLEXible), where flexible refers to the versatility of the bundle with respect to operational and fuel-cycle options. Though the initial purpose of the new 43-element bundle is to introduce SEU into CANDU, CANFLEX is extremely versatile in its application, and is compatible with other fuel cycles of interest: natural uranium in existing CANDU reactors, recycled uranium and mixed-oxides from light-water reactors, and thoria-based fuels. Capability with a variety of fuel cycles is the key to future CANDU success in the international market. The improved performance of CANFLEX, particularly at high burnups, will ensure that the full economic benefits of advanced fuels cycles are achieved. A proof-tested CANFLEX bundle design will be available in 1993 for large-scale commercial-reactor demonstration

  19. Advances in fuel cell vehicle design

    Science.gov (United States)

    Bauman, Jennifer

    Factors such as global warming, dwindling fossil fuel reserves, and energy security concerns combine to indicate that a replacement for the internal combustion engine (ICE) vehicle is needed. Fuel cell vehicles have the potential to address the problems surrounding the ICE vehicle without imposing any significant restrictions on vehicle performance, driving range, or refuelling time. Though there are currently some obstacles to overcome before attaining the widespread commercialization of fuel cell vehicles, such as improvements in fuel cell and battery durability, development of a hydrogen infrastructure, and reduction of high costs, the fundamental concept of the fuel cell vehicle is strong: it is efficient, emits zero harmful emissions, and the hydrogen fuel can be produced from various renewable sources. Therefore, research on fuel cell vehicle design is imperative in order to improve vehicle performance and durability, increase efficiency, and reduce costs. This thesis makes a number of key contributions to the advancement of fuel cell vehicle design within two main research areas: powertrain design and DC/DC converters. With regards to powertrain design, this research first analyzes various powertrain topologies and energy storage system types. Then, a novel fuel cell-battery-ultracapacitor topology is presented which shows reduced mass and cost, and increased efficiency, over other promising topologies found in the literature. A detailed vehicle simulator is created in MATLAB/Simulink in order to simulate and compare the novel topology with other fuel cell vehicle powertrain options. A parametric study is performed to optimize each powertrain and general conclusions for optimal topologies, as well as component types and sizes, for fuel cell vehicles are presented. Next, an analytical method to optimize the novel battery-ultracapacitor energy storage system based on maximizing efficiency, and minimizing cost and mass, is developed. This method can be applied

  20. Modeling of advanced fossil fuel power plants

    Science.gov (United States)

    Zabihian, Farshid

    The first part of this thesis deals with greenhouse gas (GHG) emissions from fossil fuel-fired power stations. The GHG emission estimation from fossil fuel power generation industry signifies that emissions from this industry can be significantly reduced by fuel switching and adaption of advanced power generation technologies. In the second part of the thesis, steady-state models of some of the advanced fossil fuel power generation technologies are presented. The impacts of various parameters on the solid oxide fuel cell (SOFC) overpotentials and outputs are investigated. The detail analyses of operation of the hybrid SOFC-gas turbine (GT) cycle when fuelled with methane and syngas demonstrate that the efficiencies of the cycles with and without anode exhaust recirculation are close, but the specific power of the former is much higher. The parametric analysis of the performance of the hybrid SOFC-GT cycle indicates that increasing the system operating pressure and SOFC operating temperature and fuel utilization factor improves cycle efficiency, but the effects of the increasing SOFC current density and turbine inlet temperature are not favourable. The analysis of the operation of the system when fuelled with a wide range of fuel types demonstrates that the hybrid SOFC-GT cycle efficiency can be between 59% and 75%, depending on the inlet fuel type. Then, the system performance is investigated when methane as a reference fuel is replaced with various species that can be found in the fuel, i.e., H2, CO2, CO, and N 2. The results point out that influence of various species can be significant and different for each case. The experimental and numerical analyses of a biodiesel fuelled micro gas turbine indicate that fuel switching from petrodiesel to biodiesel can influence operational parameters of the system. The modeling results of gas turbine-based power plants signify that relatively simple models can predict plant performance with acceptable accuracy. The unique

  1. Corrosion of spent Advanced Test Reactor fuel

    International Nuclear Information System (INIS)

    The results of a study of the condition of spent nuclear fuel elements from the Advanced Test Reactor (ATR) currently being stored underwater at the Idaho National Engineering Laboratory (INEL) are presented. This study was motivated by a need to estimate the corrosion behavior of dried, spent ATR fuel elements during dry storage for periods up to 50 years. The study indicated that the condition of spent ATR fuel elements currently stored underwater at the INEL is not very well known. Based on the limited data and observed corrosion behavior in the reactor and in underwater storage, it was concluded that many of the fuel elements currently stored under water in the facility called ICPP-603 FSF are in a degraded condition, and it is probable that many have breached cladding. The anticipated dehydration behavior of corroded spent ATR fuel elements was also studied, and a list of issues to be addressed by fuel element characterization before and after forced drying of the fuel elements and during dry storage is presented

  2. Economic projection for advanced fuel fabrication

    International Nuclear Information System (INIS)

    This paper presents a fabrication cost evaluation based on the status of MTR fuel development at NUKEM as of November 1980, and includes advanced chemical and mechanical processes starting with UF6 (including chemical scrap recovery). It also assumes that there will be just one fabrication line for each fuel type. UAlx-Al fuel with a uranium density of 1.0 g/cm3 was taken as the 100% cost reference. Excluding enrichment for the moment, the 100% reference values for all fuels are individually different for each reactor because of the different fuel element designs. Plate type, thickness of meat, length of meat, number of plates per element, bent or flat plates/tubes, upper and lower components, assembly design, materials, element quantity per order, etc., are all important in determining the final costs. New activity in research and development (R and D) has been started at NUKEM. In the very beginning, all the steps involved in fabrication were examined, starting with UF6 conversion and chemical treatment of the fuel (mainly, the oxide fuels), and continuing through powder production and plate fabrication, including intermediate chemical scrap recovery and final scrap recovery. The latter is very important because these procedures can also be used in reprocessing later on. If the mechanical or technical capability and/or the costs no longer makes sense, that R and D activity is stopped if there is an alternative fuel. At the moment, the prospects for alternative fuels are promising. The R and D on a particular fuel is stopped if its cost to the reactor operator would be increased by more than 30% over the current fuel cost. For this reason, the R and D on UAl alloy fuel was stopped about two years ago at a uranium density of roughly 1.2 g/cm3. Similarly, the R and D on UAIx fuel is now being stopped at around 2.2 g U/cm3 because the 30% limit has been reached. The R and D on U3O8 fuel is continuing up to about 3.2 g U/cm3, but there is some possibility of achieving

  3. Computational Design of Advanced Nuclear Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Savrasov, Sergey [Univ. of California, Davis, CA (United States); Kotliar, Gabriel [Rutgers Univ., Piscataway, NJ (United States); Haule, Kristjan [Rutgers Univ., Piscataway, NJ (United States)

    2014-06-03

    The objective of the project was to develop a method for theoretical understanding of nuclear fuel materials whose physical and thermophysical properties can be predicted from first principles using a novel dynamical mean field method for electronic structure calculations. We concentrated our study on uranium, plutonium, their oxides, nitrides, carbides, as well as some rare earth materials whose 4f eletrons provide a simplified framework for understanding complex behavior of the f electrons. We addressed the issues connected to the electronic structure, lattice instabilities, phonon and magnon dynamics as well as thermal conductivity. This allowed us to evaluate characteristics of advanced nuclear fuel systems using computer based simulations and avoid costly experiments.

  4. The benefits of an advanced fast reactor fuel cycle for plutonium management

    International Nuclear Information System (INIS)

    The United States has no program to investigate advanced nuclear fuel cycles for the large-scale consumption of plutonium from military and civilian sources. The official U.S. position has been to focus on means to bury spent nuclear fuel from civilian reactors and to achieve the spent fuel standard for excess separated plutonium, which is considered by policy makers to be an urgent international priority. Recently, the National Research Council published a long awaited report on its study of potential separation and transmutation technologies (STATS), which concluded that in the nuclear energy phase-out scenario that they evaluated, transmutation of plutonium and long-lived radioisotopes would not be worth the cost. However, at the American Nuclear Society Annual Meeting in June, 1996, the STATS panelists endorsed further study of partitioning to achieve superior waste forms for burial, and suggested that any further consideration of transmutation should be in the context of energy production, not of waste management. 2048 The U.S. Department of Energy (DOE) has an active program for the short-term disposition of excess fissile material and a 'focus area' for safe, secure stabilization, storage and disposition of plutonium, but has no current programs for fast reactor development. Nevertheless, sufficient data exist to identify the potential advantages of an advanced fast reactor metallic fuel cycle for the long-term management of plutonium. Advantages are discussed

  5. Advances in HTGR fuel performance models

    International Nuclear Information System (INIS)

    Advances in HTGR fuel performance models have improved the agreement between observed and predicted performance and contributed to an enhanced position of the HTGR with regard to investment risk and passive safety. Heavy metal contamination is the source of about 55% of the circulating activity in the HTGR during normal operation, and the remainder comes primarily from particles which failed because of defective or missing buffer coatings. These failed particles make up about 5 x 10-4 fraction of the total core inventory. In addition to prediction of fuel performance during normal operation, the models are used to determine fuel failure and fission product release during core heat-up accident conditions. The mechanistic nature of the models, which incorporate all important failure modes, permits the prediction of performance from the relatively modest accident temperatures of a passively safe HTGR to the much more severe accident conditions of the larger 2240-MW/t HTGR. (author)

  6. Development of advanced spent fuel management process

    International Nuclear Information System (INIS)

    This study is to develop an advanced spent fuel management process for countries which have not yet decided a back-end nuclear fuel cycle policy. The aims of this process development based on the pyroreduction technology of PWR spent fuels with molten lithium, are to reduce the storage volume by a quarter and to reduce the storage cooling load in half by the preferential removal of highly radioactive decay-heat elements such as Cs-137 and Sr-90 only. From the experimental results which confirm the feasibility of metallization technology, it is concluded that there are no problems in aspects of reaction kinetics and equilibrium. However, the operating performance test of each equipment on an engineering scale still remain and will be conducted in 1999. (author). 21 refs., 45 tabs., 119 figs

  7. Advances in nuclear fuel technology. 3. Development of advanced nuclear fuel recycle systems

    International Nuclear Information System (INIS)

    Fast breeder reactor (FBR) cycle technology has a technical characteristics flexibly easy to apply to diverse fuel compositions such as plutonium, minor actinides, and so on and fuel configurations. By using this characteristics, various feasibilities on effective application of uranium resources based on breeding of uranium of plutonium for original mission of FBR, contribution to radioactive wastes problems based on amounts reduction of transuranium elements (TRU) in high level radioactive wastes, upgrading of nuclear diffusion resistance, extremely upgrading of economical efficiency, and so on. In this paper, were introduced from these viewpoints, on practice strategy survey study on FBR cycle performed by cooperation of the Japan Nuclear Cycle Development Institute (JNC) with electric business companies and so on, and on technical development on advanced nuclear fuel recycle systems carried out at the Central Research Institute of Electric Power Industry, Japan Atomic Energy Research Institute, and so on. Here were explained under a vision on new type of fuels such as nitride fuels, metal fuels, and so on as well as oxide fuels, a new recycle system making possible to use actinides except uranium and plutonium, an 'advanced nuclear fuel cycle technology', containing improvement of conventional wet Purex method reprocessing technology, fuel manufacturing technology, and so on. (G.K.)

  8. Advanced mixed oxide fuel assemblies with higher plutonium content for light water reactors

    International Nuclear Information System (INIS)

    The MOX introduction in LWRs (PWR and BWR) was started in Germany with initial steps of design in the early 70's. The process of commercial utilisation of Pu recycling was based on these designs and initial MOX insertion at the Obrigheim-plant KWO (PWR) and the Gundremmingen-plant Unit A KRB-A (BWR). The needs of testing and validation of used methods could mainly be fulfilled by the insertion of initial MOX-FA reloads. Experiments with these early MOX-FAs have been conducted under fully realistic power reactor conditions for neutronic/nuclear and fuel/technological aspects. Optimising fuel cycle costs by increasing the final burnup leads to reduced generation of plutonium. Under properly defined boundary conditions thermal recycling in MOX-FAs further reduces the amount of Pu which has to be disposed of to final storage. Increasing the final burnup requires higher initial enrichments of U-fuel to be matched by an advanced design of MOX-FAs with higher Pu contents. The neutronic design of these MOX-FAs has to consider the licensing status of NPPs concerning the use of MOX fuel and the evolution of U fuel enrichment and burnup level. The Siemens Nuclear Fuel Cycle Division, with more than 20 year's experience in the production of MOX fuel, has designed several advanced MOX FAs of different types for PWRs (14x14 to 18x18) as well as for BWRs (9x9 and 10x10) with averaged contents of fissile plutonium up to 5.85 w/o. Some reloads of this kind are at present under irradiation in different NPPs. (author)

  9. Advanced technologies for power and fuel production

    Energy Technology Data Exchange (ETDEWEB)

    Watts, J.U.; Mann, A.N. [US Department of Energy/National Energy Technology Lab., Pittsburgh, PA (United States)

    2001-07-01

    The Clean Coal Technology Program (CCT) being conducted by the United States Department of Energy (DOE) is a government and industry co-funded effort. The program's purpose is to demonstrate new generation of innovative, environmentally friendly processes that enhance the utilization of coal to meet increasing demand for electric power and fuels. Program demonstration areas include environmental control, advanced power generation, fuels production, and industrial applications. The CCT Program has now grown to maturity, with over 50% of the projects selected having successfully completed their demonstration goals and objectives. Under the CCT Program, nine advanced electric power generation projects and five coal processing for clean fuels projects were selected for full scale commercial demonstration. This paper provides the status, accomplishments and results of the most widely accepted technologies currently being commercialized under these two categories. The projects are (1) Atmospheric Fluidized-Bed Combustion (AFBC) at Jacksonville Electric Authority; (2) Integrated Gasification Combined-cycle (IGCC) at Wabash River, Tampa Electric and Kentucky Pioneer; and (3) Eastman Chemical's production of methanol via coal gasification using the LPMEOH{trademark} process. 7 figs., 7 tabs.

  10. Advanced fuel technology - A UK perspective

    International Nuclear Information System (INIS)

    The nuclear power industry in the United Kingdom is perhaps more diverse than in any other country. The diversity in design of stations is matched by a diversity in operating responsibility. The SGHWR and PFR are operated by the United Kingdom Atomic Energy Authority (UKAEA), 2 of the Magnox stations are owned and run by BNFL, 2 of the AGR stations and 1 Magnox station are controlled by the South of Scotland Electricity Board (SSEB), and the remaining reactors (including the Sizewell 'B' PWR) currently come under the responsibility of the Central Electricity Generating Board (CEGB) but will pass into the control of a new state-run company when the rest of the CEGB is privatized in 1990. Against this background of a variety of designs and operational responsibilities, there is clearly a great deal of scope for advances in fuel and fuel component technology. It should be noted, however, that the nuclear energy policy within the United Kingdom, particularly with regard to PWR plants, has been to adopt well proven designs wherever possible. Emphasis is therefore directed towards achieving the successful operation of conservative systems, with research and development work on advanced options for future implementation. The following sections give an overview of the areas where advanced designs are either in production or under development for each of the UK reactor systems in turn, together with an indication of possible future developments

  11. Electronuclear conversion of fertile to fissile material

    International Nuclear Information System (INIS)

    The electronuclear conversion of fertile to fissile material by accelerator-produced neutrons is discussed. Experimental and theoretical results obtained in the MTA program (1949--1954) on the production of low-energy (less than 20-MeV) neutrons by high-energy proton, deuteron, and neutron bombardment of target materials are briefly reviewed. More recent calculations of the cascade process, by which the low-energy neutrons are produced, are discussed. A system is described by which 500- to 600-MeV deuterons incident on a lithium primary target can be converted to high-energy neutrons, which can be multiplied by spallation cascades and nuclear excitation to produce low-energy neutrons in a depleted-uranium or thorium secondary target. Fission events producing heat and additional neutrons are produced. The evaporation and fission neutrons would be captured, and fissile material would be produced. The production rates for 239Pu and 233U are estimated for 0.25-A and 0.375-A deuteron beams from an Alvarez linac. The capital and operating costs are estimated, and the resulting costs of fissile materials are calculated. The cost of generating power in reactors using the fissile material so produced as make-up fuel is also estimated. The energy multiplication (power generated in reactors so fueled/power consumed by the accelerator) ranges from about 10 to about 50 depending upon the make-up of the secondary target; depleted uranium, thorium, or a combination of the two. An experimental and theoretical program to facilitate optimization of the parameters of a production installation is described. 13 figures, 14 tables

  12. Practice and prospect of advanced fuel management and fuel technology application in PWR in China

    International Nuclear Information System (INIS)

    Since Daya Bay nuclear power plant implemented 18-month refueling strategy in 2001, China has completed a series of innovative fuel management and fuel technology projects, including the Ling Ao Advanced Fuel Management (AFM) project (high-burnup quarter core refueling) and the Ningde 18-month refueling project with gadolinium-bearing fuel in initial core. First, this paper gives brief introduction to China's advanced fuel management and fuel technology experience. Second, it introduces practices of the advanced fuel management in China in detail, which mainly focuses on the implementation and progress of the Ningde 18-month refueling project with gadolinium-bearing fuel in initial core. Finally, the paper introduces the practices of advanced fuel technology in China and gives the outlook of the future advanced fuel management and fuel technology in this field. (author)

  13. Overview of the CANDU fuel handling system for advanced fuel cycles

    International Nuclear Information System (INIS)

    Because of its neutron economies and on-power re-fuelling capabilities the CANDU system is ideally suited for implementing advanced fuel cycles because it can be adapted to burn these alternative fuels without major changes to the reactor. The fuel handling system is adaptable to implement advanced fuel cycles with some minor changes. Each individual advanced fuel cycle imposes some new set of special requirements on the fuel handling system that is different from the requirements usually encountered in handling the traditional natural uranium fuel. These changes are minor from an overall plant point of view but will require some interesting design and operating changes to the fuel handling system. Some preliminary conceptual design has been done on the fuel handling system in support of these fuel cycles. Some fuel handling details were studies in depth for some of the advanced fuel cycles. This paper provides an overview of the concepts and design challenges. (author)

  14. Options for treatment of legacy and advanced nuclear fuels

    OpenAIRE

    Maher, Christopher John

    2014-01-01

    The treatment of advanced nuclear fuels is relevant to the stabilisation of legacy spent fuels or nuclear materials and fuels from future nuclear reactors. Historically, spent fuel reprocessing has been driven to recover uranium and plutonium for reuse. Future fuel cycles may also recover the minor actinides neptunium, americium and perhaps curium. These actinides would be fabricated into new reactor fuel to produce energy and for transmutation of the minor actinides. This has the potential t...

  15. Evolutionary developments of advanced PWR nuclear fuels and cladding materials

    International Nuclear Information System (INIS)

    Highlights: • PWR fuel and cladding materials development processes are provided. • Evolution of PWR advanced fuel in U.S.A. and in Korea is described. • Cutting-edge design features against grid-to-rod fretting and debris are explained. • High performance data of advanced grids, debris filters and claddings are given. -- Abstract: The evolutionary developments of advanced PWR fuels and cladding materials are explained with outstanding design features of nuclear fuel assembly components and zirconium-base cladding materials. The advanced PWR fuel and cladding materials development processes are also provided along with verification tests, which can be used as guidelines for newcomers planning to develop an advanced fuel for the first time. The up-to-date advanced fuels with the advanced cladding materials may provide a high level of economic utilization and reliable performance even under current and upcoming aggressive operating conditions. To be specific, nuclear fuel vendors may achieve high fuel burnup capability of between 45,000 and 65,000 MWD/MTU batch average, overpower thermal margin of as much as 15% and longer cycle length up to 24 months on the one hand and fuel failure rates of around 10−6 on the other hand. However, there is still a need for better understanding of grid-to-rod fretting wear mechanisms leading to major PWR fuel defects in the world and subsequently a driving force for developing innovative spacer grid designs with zero fretting wear-induced fuel failure

  16. Evolutionary developments of advanced PWR nuclear fuels and cladding materials

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyu-Tae, E-mail: ktkim@dongguk.ac.kr

    2013-10-15

    Highlights: • PWR fuel and cladding materials development processes are provided. • Evolution of PWR advanced fuel in U.S.A. and in Korea is described. • Cutting-edge design features against grid-to-rod fretting and debris are explained. • High performance data of advanced grids, debris filters and claddings are given. -- Abstract: The evolutionary developments of advanced PWR fuels and cladding materials are explained with outstanding design features of nuclear fuel assembly components and zirconium-base cladding materials. The advanced PWR fuel and cladding materials development processes are also provided along with verification tests, which can be used as guidelines for newcomers planning to develop an advanced fuel for the first time. The up-to-date advanced fuels with the advanced cladding materials may provide a high level of economic utilization and reliable performance even under current and upcoming aggressive operating conditions. To be specific, nuclear fuel vendors may achieve high fuel burnup capability of between 45,000 and 65,000 MWD/MTU batch average, overpower thermal margin of as much as 15% and longer cycle length up to 24 months on the one hand and fuel failure rates of around 10{sup −6} on the other hand. However, there is still a need for better understanding of grid-to-rod fretting wear mechanisms leading to major PWR fuel defects in the world and subsequently a driving force for developing innovative spacer grid designs with zero fretting wear-induced fuel failure.

  17. Advanced ceramic cladding for water reactor fuel

    International Nuclear Information System (INIS)

    Under the US Department of Energy's Nuclear Energy Research Initiatives (NERI) program, continuous fiber ceramic composites (CFCCs) are being developed as cladding for water reactor fuel elements. The purpose is to substantially increase the passive safety of water reactors. A development effort was initiated in 1991 to fabricate CFCC-clad tubes using commercially available fibers and a sol-gel process developed by McDermott Technologies. Two small-diameter CFCC tubes were fabricated using pure alumina and alumina-zirconia fibers in an alumina matrix. Densities of approximately 60% of theoretical were achieved. Higher densities are required to guarantee fission gas containment. This NERI work has just begun, and only preliminary results are presented herein. Should the work prove successful, further development is required to evaluate CFCC cladding and performance, including in-pile tests containing fuel and exploring a marriage of CFCC cladding materials with suitable advanced fuel and core designs. The possibility of much higher temperature core designs, possibly cooled with supercritical water, and achievement of plant efficiencies ge50% would be examined

  18. Advanced Coal-Fueled Gas Turbine Program

    Energy Technology Data Exchange (ETDEWEB)

    Horner, M.W.; Ekstedt, E.E.; Gal, E.; Jackson, M.R.; Kimura, S.G.; Lavigne, R.G.; Lucas, C.; Rairden, J.R.; Sabla, P.E.; Savelli, J.F.; Slaughter, D.M.; Spiro, C.L.; Staub, F.W.

    1989-02-01

    The objective of the original Request for Proposal was to establish the technological bases necessary for the subsequent commercial development and deployment of advanced coal-fueled gas turbine power systems by the private sector. The offeror was to identify the specific application or applications, toward which his development efforts would be directed; define and substantiate the technical, economic, and environmental criteria for the selected application; and conduct such component design, development, integration, and tests as deemed necessary to fulfill this objective. Specifically, the offeror was to choose a system through which ingenious methods of grouping subcomponents into integrated systems accomplishes the following: (1) Preserve the inherent power density and performance advantages of gas turbine systems. (2) System must be capable of meeting or exceeding existing and expected environmental regulations for the proposed application. (3) System must offer a considerable improvement over coal-fueled systems which are commercial, have been demonstrated, or are being demonstrated. (4) System proposed must be an integrated gas turbine concept, i.e., all fuel conditioning, all expansion gas conditioning, or post-expansion gas cleaning, must be integrated into the gas turbine system.

  19. Advanced fuel cycles options for LWRs and IMF benchmark definition

    International Nuclear Information System (INIS)

    In the paper, different advanced nuclear fuel cycles including thorium-based fuel and inert-matrix fuel are examined under light water reactor conditions, especially VVER-440, and compared. Two investigated thorium based fuels include one solely plutonium-thorium based fuel and the second one plutonium-thorium based fuel with initial uranium content. Both of them are used to carry and burn or transmute plutonium created in the classical UOX cycle. The inert-matrix fuel consist of plutonium and minor actinides separated from spent UOX fuel fixed in Yttria-stabilised zirconia matrix. The article shows analysed fuel cycles and their short description. The conclusion is concentrated on the rate of Pu transmutation and Pu with minor actinides cumulating in the spent advanced thorium fuel and its comparison to UOX open fuel cycle. Definition of IMF benchmark based on presented scenario is given. (authors)

  20. Development of the advanced PHWR technology -Verification tests for CANDU advanced fuel-

    International Nuclear Information System (INIS)

    This is final report of the CANDU advanced fuel (CANFLEX fuel) verification test project. This report describes performance verification tests performed for the development of the CANFLEX-NU bundle. The test items described in the report are as follows. - Fuel channel pressure drop test, -Fuel strength tests, - Fuel impact test, - Fuel endurance test (vibration test), - Compatibility test with fueling machine, - Critical heat flux test. 58 tabs., 60 figs., 32 refs. (Author)

  1. US DOE Advanced Nuclear Fuel Development Programme Overview

    International Nuclear Information System (INIS)

    The Advanced Fuels Campaign (AFC) has been given the responsibility to develop advanced nuclear fuel technologies for the Department of Energy (DOE) Fuel Cycle Research and Development Program using a science based approach, focusing on developing a microstructural understanding of nuclear fuels and materials. The science based approach combines theory, experiment and multiscale modelling and simulation to develop a fundamental understanding of the fuel fabrication processes and fuel and cladding performance under irradiation. The objective is to use a predictive approach to design fuels and cladding to achieve the desired performance (in contrast to more empirical observation based approaches traditionally used in fuel development). The AFC programme conducts research and development of innovative, enhanced, accident tolerant, next generation LWRs and transmutation fuel systems for sustainable fuel cycles. The major areas of research include enhancing the accident tolerance of fuels and materials, improving the fuel system’s ability to achieve significantly higher fuel and plant performance, and developing innovations that provide for major increases in burnup and performance. The AFC programme is interested in advanced nuclear fuels and materials technologies that are robust, have high performance capability, and are more tolerant to accident conditions than traditional fuel systems. The scope of the AFC includes evaluation and development of multiple fuel forms to support the objectives described in the DOE Strategic Plan and the DOE’s Office of Nuclear Energy Research and Development Roadmap. The word ‘fuel’ is used generically to include fuels, targets and their associated cladding materials. (author)

  2. Conceptual design study of advanced fuel fabrication systems

    International Nuclear Information System (INIS)

    The fuel fabrication plant images based on the advanced equipment with availability to operate in hot-cell facility are constructed. The characteristics of each fuel fabrication system for economical and environmental are evaluated roughly. The advanced fuel fabrication routes such as simplified pelletizing, vibration compaction and casting process would have the potential for reducing plant construction cost and minimizing the radioactive waste generated from fuel fabrication process. (author)

  3. Advanced Multiphysics Modeling of Fast Reactor Fuel Behavior

    International Nuclear Information System (INIS)

    Evaluation of fast reactor fuel thermo-mechanical performance using fuel performance codes is a key aspect of advanced fast reactors designs. Those fuel performance codes capture the multiphysics nature of fuel behavior during irradiation where different, mostly interdependent, phenomena are taking place. Existing fuel performance codes do not fully capture those interdependencies and present the different phenomena through de-coupled models. Recent developments in multiphysics simulation capabilities and availability of advanced computing platforms led to advancements in simulation of nuclear fuel behavior. This paper presents current experiences in applying different multiphysics simulation platforms to evaluation of fast reactors metallic fuel behavior. Full 3D finite element simulation platforms that include capabilities to fully couple key fuel behavior models are discussed. Issues associated with coupling metallic fuels phenomena, such as fission gas models and constituent distribution models, with thermo-mechanical finite element platforms, as well as different coupling schemes are also discussed. (author)

  4. IEA-Advanced Motor Fuels Annual Report 2010

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-12-02

    The annual report from the IEA implementing agreement on Advanced Motor Fuels (AMF) describes the agreement, activities, and projects for the year. A section on the global situation for Advanced Motor Fuels includes country reports from each participating AMF member. A status report on each active annex for the agreement is also included, as is a message from the AMF Chairman. Final sections include an Outlook for Advanced Motor Fuels, further information, and a glossary of terms.

  5. Advanced membrane electrode assemblies for fuel cells

    Science.gov (United States)

    Kim, Yu Seung; Pivovar, Bryan S

    2014-02-25

    A method of preparing advanced membrane electrode assemblies (MEA) for use in fuel cells. A base polymer is selected for a base membrane. An electrode composition is selected to optimize properties exhibited by the membrane electrode assembly based on the selection of the base polymer. A property-tuning coating layer composition is selected based on compatibility with the base polymer and the electrode composition. A solvent is selected based on the interaction of the solvent with the base polymer and the property-tuning coating layer composition. The MEA is assembled by preparing the base membrane and then applying the property-tuning coating layer to form a composite membrane. Finally, a catalyst is applied to the composite membrane.

  6. Advanced fuel system technology for utilizing broadened property aircraft fuels

    Science.gov (United States)

    Reck, G. M.

    1980-01-01

    Possible changes in fuel properties are identified based on current trends and projections. The effect of those changes with respect to the aircraft fuel system are examined and some technological approaches to utilizing those fuels are described.

  7. Development of fabrication technology for CANDU advanced fuel -Development of the advanced CANDU technology-

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Chang Beom; Kim, Hyeong Soo; Kim, Sang Won; Seok, Ho Cheon; Shim, Ki Seop; Byeon, Taek Sang; Jang, Ho Il; Kim, Sang Sik; Choi, Il Kwon; Cho, Dae Sik; Sheo, Seung Won; Lee, Soo Cheol; Kim, Yoon Hoi; Park, Choon Ho; Jeong, Seong Hoon; Kang, Myeong Soo; Park, Kwang Seok; Oh, Hee Kwan; Jang, Hong Seop; Kim, Yang Kon; Shin, Won Cheol; Lee, Do Yeon; Beon, Yeong Cheol; Lee, Sang Uh; Sho, Dal Yeong; Han, Eun Deok; Kim, Bong Soon; Park, Cheol Joo; Lee, Kyu Am; Yeon, Jin Yeong; Choi, Seok Mo; Shon, Jae Moon [Korea Atomic Energy Res. Inst., Taejon (Korea, Republic of)

    1994-07-01

    The present study is to develop the advanced CANDU fuel fabrication technologies by means of applying the R and D results and experiences gained from localization of mass production technologies of CANDU fuels. The annual portion of this year study includes following: 1. manufacturing of demo-fuel bundles for out-of-pile testing 2. development of technologies for the fabrication and inspection of advanced fuels 3. design and munufacturing of fuel fabrication facilities 4. performance of fundamental studies related to the development of advanced fuel fabrication technology.

  8. Strategies in development of advanced fuels for LMFBR

    International Nuclear Information System (INIS)

    Overseas strategies in development of advanced fuels for LMFBR are reviewed. Recent irradiation experiment and out-of-pile test data of the fuels are given in detail. The present status of development of oxide fueled LMFBR is also treated. (auth.)

  9. Recent advances in fuel product and manufacturing process development

    International Nuclear Information System (INIS)

    This paper discusses advancements in commercial nuclear fuel products and manufacturing made by the Westinghouse Electric Corporation in response to the commercial nuclear fuel industry's demand for high reliability, increased plant availability and improved operating flexibility. The features and benefits of Westinghouse's most advanced fuel products--VANTAGE 5 for PWR plants and QUAD+ for BWR plants--are described, as well as high performance fuel concepts now under development for delivery in the late 1980s. The paper also discusses the importance of in-process quality control throughout manufacturing towards reducing product variability and improving fuel reliability

  10. Recent advances in fuel product and manufacturing process development

    International Nuclear Information System (INIS)

    This paper discusses advancements in commercial nuclear fuel products and manufacturing made by the Westinghouse Electric Corporation in response to the commercial nuclear fuel industry's demand for high reliability, increased plant availability and improved operating flexibility. The features and benefits of Westinghouse's most advanced fuel products--VANTAGE 5 for PWR plants and QUAD+ for BWR plants--are described, as well as 'high performance' fuel concepts now under development for delivery in the late 1980s. The paper also disusses the importance of in-process quality control throughout manufacturing towards reducing product variability and improving fuel reliability. (author)

  11. Advanced methods of solid oxide fuel cell modeling

    CERN Document Server

    Milewski, Jaroslaw; Santarelli, Massimo; Leone, Pierluigi

    2011-01-01

    Fuel cells are widely regarded as the future of the power and transportation industries. Intensive research in this area now requires new methods of fuel cell operation modeling and cell design. Typical mathematical models are based on the physical process description of fuel cells and require a detailed knowledge of the microscopic properties that govern both chemical and electrochemical reactions. ""Advanced Methods of Solid Oxide Fuel Cell Modeling"" proposes the alternative methodology of generalized artificial neural networks (ANN) solid oxide fuel cell (SOFC) modeling. ""Advanced Methods

  12. Advanced Combustion and Fuels; NREL (National Renewable Energy Laboratory)

    Energy Technology Data Exchange (ETDEWEB)

    Zigler, Brad

    2015-06-08

    Presented at the U.S. Department of Energy Vehicle Technologies Office 2015 Annual Merit Review and Peer Evaluation Meeting, held June 8-12, 2015, in Arlington, Virginia. It addresses technical barriers of inadequate data and predictive tools for fuel and lubricant effects on advanced combustion engines, with the strategy being through collaboration, develop techniques, tools, and data to quantify critical fuel physico-chemical effects to enable development of advanced combustion engines that use alternative fuels.

  13. New glass material oxidation and dissolution system facility: Direct conversion of surplus fissile materials, spent nuclear fuel, and other material to high-level-waste glass. Storage and disposition of weapons-usable fissile materials programmatic environmental impact statement data report: Predecisional draft

    International Nuclear Information System (INIS)

    With the end of the Cold War, countries have excess plutonium and other materials from the reductions in inventories of nuclear weapons. It has been recommended that these surplus fissile materials (SFMs) be processed so that they are no more accessible than plutonium in spent nuclear fuel (SNF). This SNF standard, if adopted worldwide, would prevent rapid recovery of SFMs for the manufacture of nuclear weapons. This report provides for the PEIS the necessary input data on a new method for the disposition of SFMs: the simultaneous conversion of SFMs, SNF, and other highly radioactive materials into high-level-waste (HLW) glass. The SFMs include plutonium, neptunium, americium, and 233U. The primary SFM is plutonium. The preferred SNF is degraded SNF, which may require processing before it can be accepted by a geological repository for disposal. The primary form of this SNF is Hanford-N SNF with preirradiation uranium enrichments between 0.95 and 1.08%. The final product is a plutonium, low-enriched-uranium, HLW, borosilicate glass for disposition in a geological repository. The proposed conversion process is the Glass Material Oxidation and Dissolution System (GMODS), which is a new process. The initial analysis of the GMODS process indicates that a MODS facility for this application would be similar in size and environmental impact to the Defense Waste Processing Facility (DWPF) at the Savannah River Site. Because of this, the detailed information available on DWPF was used as the basis for much of the GMODS input into the SFMs PEIS

  14. 1980 Annual status report: fissile materials control and management

    International Nuclear Information System (INIS)

    The R and D activities of the JRC in the field of Fissile Material Control and Management are oriented to the development of safeguards systems in the European Community nuclear fuel cycle and to provide means for a more efficient nuclear material management within the nuclear industry

  15. Upgrading of raw oil into advanced fuel. Task 5

    Energy Technology Data Exchange (ETDEWEB)

    1991-10-01

    The overall objective of the research effort is the determination of the minimum processing requirements to produce high energy density fuels (HEDF) having acceptable fuel specifications. The program encompasses assessing current technology capability; selecting acceptable processing and refining schemes; and generating samples of advanced test fuels. The Phase I Baseline Program is intended to explore the processing alternatives for producing advanced HEDF from two raw synfuel feedstocks, one from Mild Coal Gasification as exemplified by the COALITE process and one from Colorado shale oil. Eight key tasks have been identified as follows: (1) Planning and Environmental Permitting; (2) Transporting and Storage of Raw Fuel Sources and Products; (3) Screening of Processing and Upgrading Schemes; (4) Proposed Upgrading Schemes for Advanced Fuel; (5) Upgrading of Raw Oil into Advanced Fuel (6) Packaging and Shipment of Advanced Fuels; (7) Updated Technical and Economic Assessment; and, (8) Final Report of Phase I Efforts. This topical report summarizes the operations and results of the Phase I Task 5 sample preparation program. The specific objectives of Task 5 were to: Perform laboratory characterization tests on the raw COALITE feed, the intermediate liquids to the required hydroprocessing units and final advanced fuels and byproducts; and produce a minimum of 25-gal of Category I test fuel for evaluation by DOE and its contractors.

  16. Development of advanced LWR fuel cladding

    International Nuclear Information System (INIS)

    This report describes the results from evaluating the preliminary Zr-based alloys to develop the advanced Zr-based alloys for the nuclear fuel claddings, which should have good corrosion resistance and mechanical properties at high burn-up over 70,000MWD/MTU. It also includes the results from the basic studies for optimizing the processes which are involved in the development of the advanced Zr-based alloys. Ten(10) kinds of candidates for the alloys of which performance is over that of the existing Zircaloy-4 or ZIRLO alloy were selected out of the preliminary alloys of 150 kinds which were newly designed and repeatedly manufactured and evaluated to find out the promising alloys. First of all, the corrosion tests on the preliminary alloys were carried out to evaluate their performance in both pure water and LiOH solution at 360 deg C and in steam at 400 deg C. The tensile tests were performed on the alloys which proved to be good in the corrosion resistance. The creep behaviors were tested at 400 deg C for 10 days with the application of constant load on the samples which showed good performance in the corrosion resistance and tensile properties. The effect of the final heat treatment and A-parameters as well as Sn or Nb on the corrosion resistance, tensile properties, hardness, microstructures of the alloys was evaluated for some alloys interested. The other basic researches on the oxides, electrochemical properties, corrosion mechanism, and the establishment of the phase diagrams of some alloys were also carried out

  17. Development of advanced LWR fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Yong Hwan; Park, S. Y.; Lee, M. H. [and others

    2000-04-01

    This report describes the results from evaluating the preliminary Zr-based alloys to develop the advanced Zr-based alloys for the nuclear fuel claddings, which should have good corrosion resistance and mechanical properties at high burn-up over 70,000MWD/MTU. It also includes the results from the basic studies for optimizing the processes which are involved in the development of the advanced Zr-based alloys. Ten(10) kinds of candidates for the alloys of which performance is over that of the existing Zircaloy-4 or ZIRLO alloy were selected out of the preliminary alloys of 150 kinds which were newly designed and repeatedly manufactured and evaluated to find out the promising alloys. First of all, the corrosion tests on the preliminary alloys were carried out to evaluate their performance in both pure water and LiOH solution at 360 deg C and in steam at 400 deg C. The tensile tests were performed on the alloys which proved to be good in the corrosion resistance. The creep behaviors were tested at 400 deg C for 10 days with the application of constant load on the samples which showed good performance in the corrosion resistance and tensile properties. The effect of the final heat treatment and A-parameters as well as Sn or Nb on the corrosion resistance, tensile properties, hardness, microstructures of the alloys was evaluated for some alloys interested. The other basic researches on the oxides, electrochemical properties, corrosion mechanism, and the establishment of the phase diagrams of some alloys were also carried out.

  18. Advanced fuel cycles of WWER-1000 reactors

    International Nuclear Information System (INIS)

    The present paper considers characteristics of fuel cycles for the WWER-1000 reactor satisfying the following conditions: duration of the campaign at the nominal power is extended from 250 EFPD up to 470 and more ones; fuel enrichment does not exceed 5 wt.%; fuel assemblies maximum burnup does not exceed 55 MWd/kgHM. Along with uranium fuel, the use of mixed Uranium-Plutonium fuel is considered. Calculations were conducted by codes TVS-M, BIPR-7A and PERMAK-A developed in the RRC Kurchatov Institute, verified for the calculations of uranium fuel and certified by GAN RF

  19. Development of Advanced Spent Fuel Management Process

    International Nuclear Information System (INIS)

    As a part of research efforts to develop an advanced spent fuel management process, this project focused on the electrochemical reduction technology which can replace the original Li reduction technology of ANL, and we have successfully built a 20 kgHM/batch scale demonstration system. The performance tests of the system in the ACPF hot cell showed more than a 99% reduction yield of SIMFUEL, a current density of 100 mA/cm2 and a current efficiency of 80%. For an optimization of the process, the prevention of a voltage drop in an integrated cathode, a minimization of the anodic effect and an improvement of the hot cell operability by a modulation and simplization of the unit apparatuses were achieved. Basic research using a bench-scale system was also carried out by focusing on a measurement of the electrochemical reduction rate of the surrogates, an elucidation of the reaction mechanism, collecting data on the partition coefficients of the major nuclides, quantitative measurement of mass transfer rates and diffusion coefficients of oxygen and metal ions in molten salts. When compared to the PYROX process of INL, the electrochemical reduction system developed in this project has comparative advantages in its application of a flexible reaction mechanism, relatively short reaction times and increased process yields

  20. TALSPEAK Chemistry in Advanced Nuclear Fuel Cycles

    International Nuclear Information System (INIS)

    The separation of trivalent transplutonium actinides from fission product lanthanide ions represents a challenging aspect of advanced nuclear fuel partitioning schemes. The challenge of this separation could be amplified in the context of the AFCI-UREX+1a process, as Np and Pu will accompany the minor actinides to this stage of separation. At present, the baseline lanthanide-actinide separation method is the TALSPEAK (Trivalent Actinide - Lanthanide Separation by Phosphorus reagent Extraction from Aqueous complexes) process. TALSPEAK was developed in the late 1960's at Oak Ridge National Laboratory and has been demonstrated at pilot scale. This process relies on the complex interaction between an organic and an aqueous phase both containing complexants for selectively separating the trivalent actinide. The 3 complexing components are: the di(2-ethylhexyl) phosphoric acid (HDEHP), the lactic acid (HL) and the diethylenetriamine-N,N,N',N'',N''-pentaacetic acid (DTPA). In this report we discuss observations on kinetic and thermodynamic features described in the prior literature and describe some results of our ongoing research on basic chemical features of this system. The information presented indicates that the lactic acid buffer participates in the net operation of the TALSPEAK process in a manner that is not explained by existing information on the thermodynamic features if the known Eu(III)-lactate species. (authors)

  1. Development of Advanced Spent Fuel Management Process

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Chung Seok; Choi, I. K.; Kwon, S. G. (and others)

    2007-06-15

    As a part of research efforts to develop an advanced spent fuel management process, this project focused on the electrochemical reduction technology which can replace the original Li reduction technology of ANL, and we have successfully built a 20 kgHM/batch scale demonstration system. The performance tests of the system in the ACPF hot cell showed more than a 99% reduction yield of SIMFUEL, a current density of 100 mA/cm{sup 2} and a current efficiency of 80%. For an optimization of the process, the prevention of a voltage drop in an integrated cathode, a minimization of the anodic effect and an improvement of the hot cell operability by a modulation and simplization of the unit apparatuses were achieved. Basic research using a bench-scale system was also carried out by focusing on a measurement of the electrochemical reduction rate of the surrogates, an elucidation of the reaction mechanism, collecting data on the partition coefficients of the major nuclides, quantitative measurement of mass transfer rates and diffusion coefficients of oxygen and metal ions in molten salts. When compared to the PYROX process of INL, the electrochemical reduction system developed in this project has comparative advantages in its application of a flexible reaction mechanism, relatively short reaction times and increased process yields.

  2. Advanced fuel cycles and burnup increase of WWER-440 fuel

    International Nuclear Information System (INIS)

    Analyses of operational experience of 4.4% enriched fuel in the 5-year fuel cycle at Kola NPP Unit 3 and fuel assemblies with Uranium-Gadolinium fuel at Kola NPP Unit 4 are made. The operability of WWER-440 fuel under high burnup is studied. The obtained results indicate that the fuel rods of WWER-440 assemblies intended for operation within six years of the reviewed fuel cycle totally preserve their operability. Performed analyses have demonstrated the possibility of the fuel rod operability during the fuel cycle. 12 assemblies were loaded into the reactor unit of Kola 3 in 2001. The predicted burnup in six assemblies was 59.2 MWd/kgU. Calculated values of the burnup after operation for working fuel assemblies were ∼57 MWd/kgU, for fuel rods - up to ∼61 MWd/kgU. Data on the coolant activity, specific activity of the benchmark iodine radionuclides of the reactor primary circuit, control of the integrity of fuel rods of the assemblies that were operated for six years indicate that not a single assembly has reached the criterion for the early discharge

  3. Waste management planned for the advanced fuel cycle facility

    International Nuclear Information System (INIS)

    The U.S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP) program has been proposed to develop and employ advanced technologies to increase the proliferation resistance of spent nuclear fuels, recover and reuse nuclear fuel resources, and reduce the amount of wastes requiring permanent geological disposal. In the initial GNEP fuel cycle concept, spent nuclear fuel is to be reprocessed to separate re-usable transuranic elements and uranium from waste fission products, for fabricating new fuel for fast reactors. The separated wastes would be converted to robust waste forms for disposal. The Advanced Fuel Cycle Facility (AFCF) is proposed by DOE for developing and demonstrating spent nuclear fuel recycling technologies and systems. The AFCF will include capabilities for receiving and reprocessing spent fuel and fabricating new nuclear fuel from the reprocessed spent fuel. Reprocessing and fuel fabrication activities will generate a variety of radioactive and mixed waste streams. Some of these waste streams are unique and unprecedented. The GNEP vision challenges traditional U.S. radioactive waste policies and regulations. Product and waste streams have been identified during conceptual design. Waste treatment technologies have been proposed based on the characteristics of the waste streams and the expected requirements for the final waste forms. Results of AFCF operations will advance new technologies that will contribute to safe and economical commercial spent fuel reprocessing facilities needed to meet the GNEP vision. As conceptual design work and research and design continues, the waste management strategies for the AFCF are expected to also evolve. (authors)

  4. Delayed gamma technique for fissile material assay

    Energy Technology Data Exchange (ETDEWEB)

    Mozin, Vladimir [Los Alamos National Laboratory; Tobin, Stephen [Los Alamos National Laboratory; Vujie, Jasmina [UC BERKELEY; Hunt, Alan [IDAHO ACCELERATOR CENTER

    2010-01-01

    Research sponsored by the Next Generation Safeguards Initiative are investigating several non-destructive assay techniques for the quantification of fissile plutonium mass in spent nuclear fuel assemblies. AppHcation of the delayed gamma signatures is investigated in this context. The objective of the research is to assess whether the delayed gamma assay instrument can provide sufficient sensitivity, isotope specificity and accuracy as required in nuclear material safeguards. This effort includes theoretical and experimental components for the optimal combination of interrogation parameters. A new modeling algorithm offering a high level of detail was developed specifically for this purpose and was validated in series of benchmark experiments. Preliminary modeling of the delayed gamma response from spent fuel assemblies was accomplished offering a future direction for the design process.

  5. Technology Readiness Levels for Advanced Nuclear Fuels and Materials Development

    Energy Technology Data Exchange (ETDEWEB)

    Jon Carmack

    2014-01-01

    The Technology Readiness Level (TRL) process is used to quantitatively assess the maturity of a given technology. The TRL process has been developed and successfully used by the Department of Defense (DOD) for development and deployment of new technology and systems for defense applications. In addition, NASA has also successfully used the TRL process to develop and deploy new systems for space applications. Advanced nuclear fuels and materials development is a critical technology needed for closing the nuclear fuel cycle. Because the deployment of a new nuclear fuel forms requires a lengthy and expensive research, development, and demonstration program, applying the TRL concept to the advanced fuel development program is very useful as a management and tracking tool. This report provides definition of the technology readiness level assessment process as defined for use in assessing nuclear fuel technology development for the Advanced Fuel Campaign (AFC).

  6. Fuel rod bundles proposed for advanced pressure tube nuclear reactors

    International Nuclear Information System (INIS)

    The paper aims to be a general presentation for fuel bundles to be used in Advanced Pressure Tube Nuclear Reactors (APTNR). The characteristics of such a nuclear reactor resemble those of known advanced pressure tube nuclear reactors like: Advanced CANDU Reactor (ACRTM-1000, pertaining to AECL) and Indian Advanced Heavy Water Reactor (AHWR). We have also developed a fuel bundle proposal which will be referred as ASEU-43 (Advanced Slightly Enriched Uranium with 43 rods). The ASEU-43 main design along with a few neutronic and thermalhydraulic characteristics are presented in the paper versus similar ones from INR Pitesti SEU-43 and CANDU-37 standard fuel bundles. General remarks regarding the advantages of each fuel bundle and their suitability to be burned in an APTNR reactor are also revealed. (authors)

  7. Metallic fuels: The EBR-II legacy and recent advances

    Energy Technology Data Exchange (ETDEWEB)

    Douglas L. Porter; Steven L. Hayes; J. Rory Kennedy

    2012-09-01

    Experimental Breeder Reactor – II (EBR-II) metallic fuel was qualified for high burnup to approximately 10 atomic per cent. Subsequently, the electrometallurgical treatment of this fuel was demonstrated. Advanced metallic fuels are now investigated for increased performance, including ultra-high burnup and actinide burning. Advances include additives to mitigate the fuel/cladding chemical interaction and uranium alloys that combine Mo, Ti and Zr to improve alloy performance. The impacts of the advances—on fabrication, waste streams, electrorefining, etc.—are found to be minimal and beneficial. Owing to extensive research literature and computational methods, only a modest effort is required to complete their development.

  8. Advanced Fuel Cell System Thermal Management for NASA Exploration Missions

    Science.gov (United States)

    Burke, Kenneth A.

    2009-01-01

    The NASA Glenn Research Center is developing advanced passive thermal management technology to reduce the mass and improve the reliability of space fuel cell systems for the NASA exploration program. An analysis of a state-of-the-art fuel cell cooling systems was done to benchmark the portion of a fuel cell system s mass that is dedicated to thermal management. Additional analysis was done to determine the key performance targets of the advanced passive thermal management technology that would substantially reduce fuel cell system mass.

  9. Romanian concern for advanced fuels development

    International Nuclear Information System (INIS)

    The Institute for Nuclear Research (ICN), a subsidiary of Romanian Authority for Nuclear Activities, at Pitesti - Romania, has developed a preliminary design of a fuel bundle with 43 elements named SEU 43 for high burnup in CANDU Reactor. A very high experience in nuclear fuels manufacturing and control has also been accumulated. Additionally, on the nuclear site Pitesti there is the Nuclear Fuel Plant (NFP) qualified to manufacturing CANDU 6 type fuel, the main fuel supplier for NPP Cernavoda. A very good collaboration of ICN with NFP can lead to a low cost upgrading the facilities which ensure at present the CANDU standard fuel fabrication to be able of manufacturing also SEU 43 fuel for extended burnup. The financial founds are allocated by Romanian Authority for Nuclear Activities of the Ministry of Industry and Resources to sustain the departmental R and D program 'Nuclear Fuel'. This Program has the main objective to establish a technology for manufacturing a new CANDU fuel type destined for extended burnup. It is studied the possibility to use the Recovered Uranium (RU) resulted from LWR spent fuel reprocessing facility existing in stockpiles. The International Agency for Atomic Energy (IAEA) sustains also this program. By ROM/4/025/ Model Project, IAEA helps ICN to solve the problems regarding materials (RU, Zircaloy 4 tubes) purchasing, devices' upgrading and personnel training. The paper presents the main actions needing to be create the technical base for SEU 43 fuel bundle manufacturing. First step, the technological experiments and experimental fuel element manufacturing, will be accomplished in ICN installations. Second step, the industrial scale, need thorough studies for each installation from NFP to determine tools and technology modification imposed by the new CANDU fuel bundle manufacturing. All modifications must be done such as to the NFP, standard CANDU and SEU fuel bundles to be manufactured alternatively. (author)

  10. Advances in AGR fuel fabrication - now and the future

    International Nuclear Information System (INIS)

    To date, over 3 million AGR fuel pins have been manufactured at Springfields for the UK AGR programme. During this time, AGR fuel design and manufacture has developed and evolved in response to the needs of the reactor operators to enhance fuel reliability and performance. More recently, major advances have been made in the systems and organisational culture which support fuel manufacture at Fuel Division. The introduction of MRP II in 1989 into Fuel Division enabled significant reductions in stock and work-in-progress, together with reductions in manufacturing lead times. Other successful initiatives introduced into Fuel Division have been Just-in-Time (JIT) and AST (Additional Skills Training) which have built on the success of MRP II. All of these initiatives are evidence of Fuel Division's ''Total Quality'' approach to fabricating fuel. Fuel Division is currently in the final stages of commissioning the New Oxide Fuels Complex (NOFC) where both AGR and PWR fuel will be manufactured to the highest standards of quality, safety and environmental protection. NOFC is a totally integrated plant which represents a Pound 200M investment, demonstrating Fuel Division's commitment to building on its 40+ years of fuel fabrication experience and ensuring secure supply of fuel to its customers for years to come. (author)

  11. CANDU RU fuel manufacturing basic technology development and advanced fuel verification tests

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Chang Hwan; Chang, S.K.; Hong, S.D. [and others

    1999-04-01

    A PHWR advanced fuel named the CANFLEX fuel has been developed through a KAERI/AECL joint Program. The KAERI made fuel bundle was tested at the KAERI Hot Test Loop for the performance verification of the bundle design. The major test activities were the fuel bundle cross-flow test, the endurance fretting/vibration test, the freon CHF test, and the fuel bundle heat-up test. KAERI also has developing a more advanced PHWR fuel, the CANFLEX-RU fuel, using recovered uranium to extend fuel burn-up in the CANDU reactors. For the purpose of proving safety of the RU handling techniques and appraising feasibility of the CANFLEX-RU fuel fabrication in near future, a physical, chemical and radiological characterization of the RU powder and pellets was performed. (author). 54 refs., 46 tabs., 62 figs.

  12. CANDU RU fuel manufacturing basic technology development and advanced fuel verification tests

    International Nuclear Information System (INIS)

    A PHWR advanced fuel named the CANFLEX fuel has been developed through a KAERI/AECL joint Program. The KAERI made fuel bundle was tested at the KAERI Hot Test Loop for the performance verification of the bundle design. The major test activities were the fuel bundle cross-flow test, the endurance fretting/vibration test, the freon CHF test, and the fuel bundle heat-up test. KAERI also has developing a more advanced PHWR fuel, the CANFLEX-RU fuel, using recovered uranium to extend fuel burn-up in the CANDU reactors. For the purpose of proving safety of the RU handling techniques and appraising feasibility of the CANFLEX-RU fuel fabrication in near future, a physical, chemical and radiological characterization of the RU powder and pellets was performed. (author). 54 refs., 46 tabs., 62 figs

  13. Advanced Fuels Campaign FY 2014 Accomplishments Report

    Energy Technology Data Exchange (ETDEWEB)

    Lori Braase; W. Edgar May

    2014-10-01

    The overall goal of ATF development is to identify alternative fuel system technologies to further enhance the safety, competitiveness, and economics of commercial nuclear power. The complex multiphysics behavior of LWR nuclear fuel in the integrated reactor system makes defining specific material or design improvements difficult; as such, establishing desirable performance attributes is critical in guiding the design and development of fuels and cladding with enhanced accident tolerance.

  14. Nonproliferation characteristics of advanced fuel cycle concepts

    International Nuclear Information System (INIS)

    The purpose of this study is to comment on the proliferation characteristic profiles of some of the proposed fuel cycle alternatives to help ensure that nonproliferation concerns are introduced into the early stages of a fuel cycle concept development program, and to perhaps aid in the more effective implementation of the international nonproliferation regime initiatives and safeguards methods and systems. Alternative cycle concepts proposed by several countries involve the recycle of spent fuel without the separation of plutonium from uranium and fission products

  15. Advancing the Limits of Dual Fuel Combustion

    OpenAIRE

    Königsson, Fredrik

    2012-01-01

    There is a growing interest in alternative transport fuels. There are two underlying reasons for this interest; the desire to decrease the environmental impact of transports and the need to compensate for the declining availability of petroleum. In the light of both these factors the Diesel Dual Fuel, DDF, engine is an attractive concept. The primary fuel of the DDF engine is methane, which can be derived both from renewables and from fossil sources. Methane from organic waste; commonly refer...

  16. Management of recyclable fissile and fertile materials

    International Nuclear Information System (INIS)

    The possibility to recycle fuel is a very attractive - nearly unique - feature of nuclear energy systems. Fissile and fertile materials that are contained in spent nuclear fuels and enrichment plant tails for example may be retrieved and re-used to provide additional energy and reduce the amount and toxicity of ultimate waste to be sent to repositories. While recycling becomes increasingly attractive in the context of renewed interest for nuclear energy and of sustainable development goals, extended interim storage and direct disposal of recyclable materials remain options favoured by many countries. Recyclable materials which are not intended to be re-used may be disposed of in a safe way guaranteeing their isolation from the biosphere over very long periods of time until they become harmless for humans and the environment. The study on recyclable fissile and fertile materials published in 2007 by the OECD was carried out by its Nuclear Energy Agency (NEA) in order to review technical, strategic and policy issues raised by the management of those materials and to provide some insights on the opportunities and challenges offered by alternative options in this regard. The materials considered include: spent fuel; depleted uranium from enrichment plant tails; separated uranium and plutonium from commercial reprocessing plants; ex-military materials (highly enriched uranium and plutonium) declared excess to national security by the Russian Federation and the United States; and thorium inventories. The present paper is based on the analyses, findings and conclusions from the NEA study. It provides overview on the quantities and potential energetic value of recyclable materials available worldwide. The main advantages and drawbacks of the two management options that may be adopted are described. Some findings and conclusions are drawn from the identification and analysis of economic, social and environmental indicators associated with each option that need to be taken

  17. Advanced Safeguards Approaches for New TRU Fuel Fabrication Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Durst, Philip C.; Ehinger, Michael H.; Boyer, Brian; Therios, Ike; Bean, Robert; Dougan, A.; Tolk, K.

    2007-12-15

    This second report in a series of three reviews possible safeguards approaches for the new transuranic (TRU) fuel fabrication processes to be deployed at AFCF – specifically, the ceramic TRU (MOX) fuel fabrication line and the metallic (pyroprocessing) line. The most common TRU fuel has been fuel composed of mixed plutonium and uranium dioxide, referred to as “MOX”. However, under the Advanced Fuel Cycle projects custom-made fuels with higher contents of neptunium, americium, and curium may also be produced to evaluate if these “minor actinides” can be effectively burned and transmuted through irradiation in the ABR. A third and final report in this series will evaluate and review the advanced safeguards approach options for the ABR. In reviewing and developing the advanced safeguards approach for the new TRU fuel fabrication processes envisioned for AFCF, the existing international (IAEA) safeguards approach at the Plutonium Fuel Production Facility (PFPF) and the conceptual approach planned for the new J-MOX facility in Japan have been considered as a starting point of reference. The pyro-metallurgical reprocessing and fuel fabrication process at EBR-II near Idaho Falls also provided insight for safeguarding the additional metallic pyroprocessing fuel fabrication line planned for AFCF.

  18. Fast Reactor Fuel Cycle Cost Estimates for Advanced Fuel Cycle Studies

    International Nuclear Information System (INIS)

    Presentation Outline: • Why Do I Need a Cost Basis?; • History of the Advanced Fuel Cycle Cost Basis; • Description of the Cost Basis; • Current Work; • Fast Reactor Fuel Cycle Applications; • Sample Fuel Cycle Cost Estimate Analysis; • Future Work

  19. Radioactive waste management and advanced nuclear fuel cycle technologies

    International Nuclear Information System (INIS)

    In 2007 ENEA's Department of Nuclear Fusion and Fission, and Related Technologies acted according to national policy and the role assigned to ENEA FPN by Law 257/2003 regarding radioactive waste management and advanced nuclear fuel cycle technologies

  20. CANDU-6 fuel bundle fabrication and advanced fuels development in China

    International Nuclear Information System (INIS)

    In recent years, China North Nuclear Fuel Corporation (CNNFC) has introduced several modifications to the manufacturing processes and the production line equipment. This has been beneficial in achieving a very high level of quality in the production of fuel bundles. Since 2008 CNNFC has participated in a multi party project with the goal of developing advanced fuels for use in CANDU reactors. Other project team members include the Nuclear Power Institute of China (NPIC), Third Qinshan Nuclear Power Company (TQNPC) and Atomic Energy of Canada Ltd (AECL). This paper will present the improvements developed during the manufacture of natural fuel bundles and advanced fuels. (author)

  1. Advanced fuel cycle development at Chalk River Laboratories

    International Nuclear Information System (INIS)

    Chalk River Laboratories (CRL) has a mandate from the Canadian government to develop nuclear technologies that support generation of clean, safe energy. This includes the development of advanced nuclear fuel technologies to ensure sustainable energy sources for Canadians. The Fuel Development Branch leads CRL's development of advanced nuclear-reactor fuels. CRL capabilities include fuel fabrication development, irradiation testing, post-irradiation examination (PIE), materials characterization and code development (modeling). This paper provides an overview of these capabilities and describes recent development activities that support fuel-cycle flexibility in heavy-water reactors. This includes a review of irradiation testing and PIE for mixed-oxide, thoria, high-burnup UO2 and low-void reactivity fuels and burnable neutron absorbers. Fabrication development, material characterizations and modeling associated with these tests are also described. (author)

  2. Advanced fuel in the Budapest research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hargitai, T.; Vidovsky, I. [KFKI Atomic Energy Research Inst., Budapest (Hungary)

    1997-07-01

    The Budapest Research Reactor, the first nuclear facility of Hungary, started to operate in 1959. The main goal of the reactor is to serve neutron research, but applications as neutron radiography, radioisotope production, pressure vessel surveillance test, etc. are important as well. The Budapest Research Reactor is a tank type reactor, moderated and cooled by light water. After a reconstruction and upgrading in 1967 the VVR-SM type fuel elements were used in it. These fuel elements provided a thermal power of 5 MW in the period 1967-1986 and 10 MW after the reconstruction from 1992. In the late eighties the Russian vendor changed the fuel elements slightly, i.e. the main parameters of the fuel remained unchanged, however a higher uranium content was reached. This new fuel is called VVR-M2. The geometry of VVR-SM and VVR-M2 are identical, allowing the use to load old and new fuel assemblies together to the active core. The first new type fuel assemblies were loaded to the Budapest Research Reactor in 1996. The present paper describes the operational experience with the new type of fuel elements in Hungary. (author)

  3. Basic research and industrialization of CANDU advanced fuel

    International Nuclear Information System (INIS)

    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

  4. 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

  5. Epithermal interrogation of fissile waste

    Energy Technology Data Exchange (ETDEWEB)

    Coop, K.L.; Hollas, C.L.

    1996-09-01

    Self-shielding of interrogating thermal neutrons in lumps of fissile material can be a major source of error in transuranic waste assay using the widely employed differential dieaway technique. We are developing a new instrument, the combined thermal/epithermal neutron (CTEN) interrogation instrument to detect the occurrence of self- shielding and mitigate its effects. Neutrons are moderated in the graphite walls of the CTEN instrument to provide an interrogating flux of epithermal and thermal neutrons. The induced prompt fission neutrons are detected in proportional counters. We report the results of measurements made with the CTEN instrument, using minimal and highly self-shielding plutonium and uranium sources in 55 gallon drums containing a variety of mock waste matrices. Fissile isotopes and waste forms for which the method is most applicable, and limitations associated with the hydrogen content of the waste package/matrix are described.

  6. Epithermal interrogation of fissile waste

    International Nuclear Information System (INIS)

    Self-shielding of interrogating thermal neutrons in lumps of fissile material can be a major source of error in transuranic waste assay using the widely employed differential dieaway technique. We are developing a new instrument, the combined thermal/epithermal neutron (CTEN) interrogation instrument to detect the occurrence of self- shielding and mitigate its effects. Neutrons are moderated in the graphite walls of the CTEN instrument to provide an interrogating flux of epithermal and thermal neutrons. The induced prompt fission neutrons are detected in proportional counters. We report the results of measurements made with the CTEN instrument, using minimal and highly self-shielding plutonium and uranium sources in 55 gallon drums containing a variety of mock waste matrices. Fissile isotopes and waste forms for which the method is most applicable, and limitations associated with the hydrogen content of the waste package/matrix are described

  7. Nuclear fuel cycles of WWER-1000 at Kozloduy NPP: a program for transition to advanced fuel

    International Nuclear Information System (INIS)

    A systematical approach for WWER nuclear fuel utilization improvement is applied at the Kozloduy NPP while observing safety requirements. The fuel assemblies utilization for 4 years, higher burnup and reducing of high-activity radioactive waste will be achieved by realization of Activities program for transition of Units 5 and 6 of Kozloduy NPP to operating with Alternative Fuel Assemblies (AFA). The specified results from advanced fuel introduction also will improve the economic indices of Units 5 and 6 operation at Kozloduy NPP

  8. Development of advanced mixed oxide fuels for plutonium management

    International Nuclear Information System (INIS)

    A number of advanced Mixed Oxide (MOX) fuel forms are currently being investigated at Los Alamos National Laboratory that have the potential to be effective plutonium management tools. Evolutionary Mixed Oxide (EMOX) fuel is a slight perturbation on standard MOX fuel, but achieves greater plutonium destruction rates by employing a fractional nonfertile component. A pure nonfertile fuel is also being studied. Initial calculations show that the fuel can be utilized in existing light water reactors and tailored to address different plutonium management goals (i.e., stabilization or reduction of plutonium inventories residing in spent nuclear fuel). In parallel, experiments are being performed to determine the feasibility of fabrication of such fuels. Initial EMOX pellets have successfully been fabricated using weapons-grade plutonium

  9. The DOE Advanced Gas Reactor Fuel Development and Qualification Program

    International Nuclear Information System (INIS)

    The high outlet temperatures and high thermal-energy conversion efficiency of modular High Temperature Gas-cooled Reactors (HTGRs) enable an efficient and cost effective integration of the reactor system with non-electricity generation applications, such as process heat and/or hydrogen production, for the many petrochemical and other industrial processes that require temperatures between 300 C and 900 C. The Department of Energy (DOE) has selected the HTGR concept for the Next Generation Nuclear Plant (NGNP) Project as a transformative application of nuclear energy that will demonstrate emissions-free nuclear-assisted electricity, process heat, and hydrogen production, thereby reducing greenhouse-gas emissions and enhancing energy security. The objective of the DOE Advanced Gas Reactor (AGR) Fuel Development and Qualification program is to qualify tristructural isotropic (TRISO)-coated particle fuel for use in HTGRs. The Advanced Gas Reactor Fuel Development and Qualification Program consists of five elements: fuel manufacture, fuel and materials irradiations, post-irradiation examination (PIE) and safety testing, fuel performance modeling, and fission-product transport and source term evaluation. An underlying theme for the fuel development work is the need to develop a more complete, fundamental understanding of the relationship between the fuel fabrication process and key fuel properties, the irradiation and accident safety performance of the fuel, and the release and transport of fission products in the NGNP primary coolant system. An overview of the program and recent progress is presented.

  10. Advanced nuclear fuel for VVER reactors. Status and operation experience

    International Nuclear Information System (INIS)

    The paper discusses the major VVER fuel trends, aimed at the enhancement of FAs' effectiveness and reliability, flexibility of their operating performances and fuel cycle efficiency, specifically: (i) Fuel burnup increasing is one of the major objectives during the development of improved nuclear fuel and fuel cycles. At present, the achieved fuel rod burn up is 65 MWdays/kgU. The tasks are set and the activities are carried out to achieve fuel rod burnup up to 70 MWdays/kgU and burnup of discharged batch of FAs - up to 60 MWdays/kgU. (ii) Improvement of FA rigidity enables to increase operating reliability of fuel due to gaps reducing between FAs and, as a result, the fall of peak load coefficients. FA geometric stability enables to optimize the speed of handling procedures with fuel. (iii) Increasing of uranium content of FA is aimed at extension of fuel cycles' duration. Fuel weight increase in FA is achieved both due to fuel column height extension and to changes of pellet geometrical size. (iv) Extension of FA service live satisfies the up-to-date NPP requirements for fuel cycles of various duration from 4x320 eff. days to 5x320 eff. days and 3x480 eff. days. (v) The development of new-generation FAs with increased strength characteristics has required the zirconium alloys' improvement. Advanced zirconium alloys shall provide safety and effectiveness of FA and fuel rods during long-life operation up to 40 000 eff. hours. (vi) Utilization of reprocessed uranium enables to use spent nuclear fuel in cycle and to create the partly complete fuel cycle for VVER reactors. This paper summarizes the major operating results of LTAs, which meet the modern and prospective requirements for VVER fuel, at Russian NPPs with VVER-440 and VVER-1000 reactors. (author)

  11. Using Advanced Fuel Bundles in CANDU Reactors

    International Nuclear Information System (INIS)

    Improving the exit fuel burnup in CANDU reactors was a long-time challenge for both bundle designers and performance analysts. Therefore, the 43-element design together with several fuel compositions was studied, in the aim of assessing new reliable, economic and proliferation-resistant solutions. Recovered Uranium (RU) fuel is intended to be used in CANDU reactors, given the important amount of slightly enriched Uranium (~0.96% w/o U235) that might be provided by the spent LWR fuel recovery plants. Though this fuel has a far too small U235 enrichment to be used in LWR's, it can be still used to fuel CANDU reactors. Plutonium based mixtures are also considered, with both natural and depleted Uranium, either for peacefully using the military grade dispositioned Plutonium or for better using Plutonium from LWR reprocessing plants. The proposed Thorium-LEU mixtures are intended to reduce the Uranium consumption per produced MW. The positive void reactivity is a major concern of any CANDU safety assessment, therefore reducing it was also a task for the present analysis. Using the 43-element bundle with a certain amount of burnable poison (e.g. Dysprosium) dissolved in the 8 innermost elements may lead to significantly reducing the void reactivity. The expected outcomes of these design improvements are: higher exit burnup, smooth/uniform radial bundle power distribution and reduced void reactivity. Since the improved fuel bundles are intended to be loaded in existing CANDU reactors, we found interesting to estimate the local reactivity effects of a mechanical control absorber (MCA) on the surrounding fuel cells. Cell parameters and neutron flux distributions, as well as macroscopic cross-sections were estimated using the transport code DRAGON and a 172-group updated nuclear data library. (author)

  12. Advanced-fuel reversed-field pinch reactor (RFPR)

    International Nuclear Information System (INIS)

    The utilization of deuterium-based fuels offers the potential advantages of greater flexibility in blanket design, significantly reduced tritium inventory, potential reduction in radioactivity level, and utilization of an inexhaustible fuel supply. The conventional DT-fueled Reversed-Field Pinch Reactor (RFPR) designs are reviewed, and the recent extension of these devices to advanced-fuel (catalyzed-DD) operation is presented. Attractive and economically competitive DD/RFPR systems are identified having power densities and plasma parameters comparable to the DT systems. Converting an RFP reactor from DT to DD primarily requires increasing the magnetic field levels a factor of two, still requiring only modest magnet coil fields (less than or equal to 4 T). When compared to the mainline tokamak, the unique advantages of the RFP (e.g., high beta, low fields at the coils, high ohmic-heating power densities, unrestricted aspect ratio) are particularly apparent for the utilization of advanced fuels

  13. Advances in High Temperature Gas Cooled Reactor Fuel Technology

    International Nuclear Information System (INIS)

    This publication reports on the results of a coordinated research project on advances in high temperature gas cooled reactor (HTGR) fuel technology and describes the findings of research activities on coated particle developments. These comprise two specific benchmark exercises with the application of HTGR fuel performance and fission product release codes, which helped compare the quality and validity of the computer models against experimental data. The project participants also examined techniques for fuel characterization and advanced quality assessment/quality control. The key exercise included a round-robin experimental study on the measurements of fuel kernel and particle coating properties of recent Korean, South African and US coated particle productions applying the respective qualification measures of each participating Member State. The summary report documents the results and conclusions achieved by the project and underlines the added value to contemporary knowledge on HTGR fuel.

  14. Analysis of Advanced Fuel Cycle Strategies: New Insights

    International Nuclear Information System (INIS)

    Nuclear power is a crucial component of future energy portfolios for expanding worldwide energy demand in the context of anticipated resource and emission constraints. Fuel resource management, spent fuel management, and material non-proliferation, have been identified as items that have to be addressed for nuclear power to fulfill this role. This paper reviews the current fuel cycles operating internationally and the advanced fuel cycle strategies that are proposed to ensure the nuclear future. Perspectives on these strategies are discussed to identify the capabilities and limitations of the nuclear systems and fuel cycle configurations. Results of transition scenario studies from the currently operating systems to advanced systems are also summarized. International proposals designed to curtail the spread of weapons-usable materials in an expanding nuclear future are also briefly discussed. (authors)

  15. PNC`s proposal on the Advanced Fuel Recycle concept

    Energy Technology Data Exchange (ETDEWEB)

    Kamiya, Masayoshi; Shinoda, Yoshihiko; Ojima, Hisao [Power Reactor and Nuclear Fuel Development Corp., Tokai, Ibaraki (Japan). Tokai Works

    1998-03-01

    MOX fuel for FBR is allowed to contain impurities within several thousand ppm, which means less than 1000 of decontamination factor (DF) in reprocessing is enough for Pu and U recycle use. The Advanced Fuel Recycle proposed by PNC is on this basis. The concept consists of innovations on both MOX fuel fabrication and aqueous reprocessing technologies based on the Purex process and it is believed that successful optimization of fuel cycle interface condition is the key issue to realize the concept. The lower DF such as 1000 can be easily obtained by the simplified Purex flowsheet which has no purification steps. However, new subject arises in MOX fuel fabrication, that is, fabrication is conducted in the shielding cell using equipment which is maintained remotely. A simplified fabrication technology becomes essential to establish the remote maintenance system and is one of the critical path for achieving the Advanced Fuel Recycle. The PNC`s proposal on the advanced fuel recycle concept consists of modified PUREX process having single extraction cycle and crystallization, Remote fuel fabrication such as gelation and vibro-packing. In the Advanced Fuel Recycle concept, as it is low DF cycle system, all processes should be installed in remote maintenance cells. Then both reprocessing and fabrication facility would be able to be integrated into a same building. Integrated fuel cycle plant has several merits. No transportation of nuclear material between reprocessing and fabrication enhances non-proriferation aspect in addition to the low-DF concept. Cost performance is also improved because of optimization and rationalization of auxiliary equipment, and so on. (author)

  16. PNC's proposal on the Advanced Fuel Recycle concept

    International Nuclear Information System (INIS)

    MOX fuel for FBR is allowed to contain impurities within several thousand ppm, which means less than 1000 of decontamination factor (DF) in reprocessing is enough for Pu and U recycle use. The Advanced Fuel Recycle proposed by PNC is on this basis. The concept consists of innovations on both MOX fuel fabrication and aqueous reprocessing technologies based on the Purex process and it is believed that successful optimization of fuel cycle interface condition is the key issue to realize the concept. The lower DF such as 1000 can be easily obtained by the simplified Purex flowsheet which has no purification steps. However, new subject arises in MOX fuel fabrication, that is, fabrication is conducted in the shielding cell using equipment which is maintained remotely. A simplified fabrication technology becomes essential to establish the remote maintenance system and is one of the critical path for achieving the Advanced Fuel Recycle. The PNC's proposal on the advanced fuel recycle concept consists of modified PUREX process having single extraction cycle and crystallization, Remote fuel fabrication such as gelation and vibro-packing. In the Advanced Fuel Recycle concept, as it is low DF cycle system, all processes should be installed in remote maintenance cells. Then both reprocessing and fabrication facility would be able to be integrated into a same building. Integrated fuel cycle plant has several merits. No transportation of nuclear material between reprocessing and fabrication enhances non-proliferation aspect in addition to the low-DF concept. Cost performance is also improved because of optimization and rationalization of auxiliary equipment, and so on. (author)

  17. Cermet-fueled reactors for advanced space applications

    International Nuclear Information System (INIS)

    Cermet-fueled nuclear reactors are attractive candidates for high-performance advanced space power systems. The cermet consists of a hexagonal matrix of a refractory metal and a ceramic fuel, with multiple tubular flow channels. The high performance characteristics of the fuel matrix come from its high strength at elevated temperatures and its high thermal conductivity. The cermet fuel concept evolved in the 1960s with the objective of developing a reactor design that could be used for a wide range of mobile power generating sytems, including both Brayton and Rankine power conversion cycles. High temperature thermal cycling tests for the cermet fuel were carried out by General Electric as part of the 710 Project (General Electric 1966), and by Argonne National Laboratory in the Direct Nuclear Rocket Program (1965). Development programs for cermet fuel are currently under way at Argonne National Laboratory and Pacific Northwest Laboratory. The high temperature qualification tests from the 1960s have provided a base for the incorporation of cermet fuel in advanced space applications. The status of the cermet fuel development activities and descriptions of the key features of the cermet-fueled reactor design are summarized in this paper

  18. Romanian progress in the advanced CANDU fuel manufacturing

    Energy Technology Data Exchange (ETDEWEB)

    Ohai, D.; Benga, D. [RAAN, Inst. for Nuclear Research, Pitesti- Mioveni (Romania)]. E-mail: dohai@nuclear.ro

    2005-07-01

    The initial concept in developing an advanced fuel compatible with CANDU 6 Reactor, using part of Nuclear Fuel Plant (FCN) Pitesti facilities [1] should be revised. New aspects were considered: working within FCN area, a technological transfer suspicion appears (inobservance of AECL-FCN confidentiality agreement), and the enriched Uranium use on FCN area is prohibited (IAEA requirement). Under these conditions, the Institute for Nuclear Research (ICN) decided to develop or modernize its own facilities for nuclear fuel (CANDU type) manufacturing. The intention was to cover the main technological steps in fuel manufacturing, beginning with powder manufacturing and ending up with fuel bundle assembling. The development or modernization of own facilities for the nuclear fuel manufacturing open the possibilities for the collaboration with other entities interested in advanced fuel development. Having a Research Reactor for material testing and a Post Irradiation+ Facility, ICN can complete the irradiation and post-irradiation services with experimental fuel elements manufacturing, the services being completed. This can be a possibility to eliminate the interstates transport of nuclear materials. The new international requirements for the transport of the nuclear materials are drastic and need a lot of time and money for obtaining authorizations and for transport. It is financially advantageous to manufacture experimental fuel elements on the same site with the irradiation and post-irradiation facilities. (author)

  19. Surrogate Model Development for Fuels for Advanced Combustion Engines

    Energy Technology Data Exchange (ETDEWEB)

    Anand, Krishnasamy [University of Wisconsin, Madison; Ra, youngchul [University of Wisconsin, Madison; Reitz, Rolf [University of Wisconsin; Bunting, Bruce G [ORNL

    2011-01-01

    The fuels used in internal-combustion engines are complex mixtures of a multitude of different types of hydrocarbon species. Attempting numerical simulations of combustion of real fuels with all of the hydrocarbon species included is highly unrealistic. Thus, a surrogate model approach is generally adopted, which involves choosing a few representative hydrocarbon species whose overall behavior mimics the characteristics of the target fuel. The present study proposes surrogate models for the nine fuels for advanced combustion engines (FACE) that have been developed for studying low-emission, high-efficiency advanced diesel engine concepts. The surrogate compositions for the fuels are arrived at by simulating their distillation profiles to within a maximum absolute error of 4% using a discrete multi-component (DMC) fuel model that has been incorporated in the multi-dimensional computational fluid dynamics (CFD) code, KIVA-ERC-CHEMKIN. The simulated surrogate compositions cover the range and measured concentrations of the various hydrocarbon classes present in the fuels. The fidelity of the surrogate fuel models is judged on the basis of matching their specific gravity, lower heating value, hydrogen/carbon (H/C) ratio, cetane number, and cetane index with the measured data for all nine FACE fuels.

  20. Advanced sipping facilities for fuel elements

    International Nuclear Information System (INIS)

    The sipping facilities for BWR type plants and PWR type plants of the Russian type WWER-440 are equipped with a bell instead of caps, which is used above the opened reactor, moved by the fuel handling machine, and covers up to eight fuel elements in the core during inspection. In all sipping facilities, the complete inspection sequence is controlled by a desk switchboard near the fuel element storage pool or the reactor well. Siemens' sipping facilities are used in all Siemens-built nuclear power plants and in many others by different manufacturers. Part of them has been in operation already for more than 20 years with a high degree of reliability. Inspection safety is more than 99.5%. (orig./DG)

  1. Recycled uranium: An advanced fuel for CANDU reactors

    International Nuclear Information System (INIS)

    The use of recycled uranium (RU) fuel offers significant benefits to CANDU reactor operators particularly if used in conjunction with advanced fuel bundle designs that have enhanced performance characteristics. Furthermore, these benefits can be realised using existing fuel production technologies and practices and with almost negligible change to fuel receipt and handling procedures at the reactor. The paper will demonstrate that the supply of RU as a ceramic-grade UO2 powder will increasingly become available as a secure option to virgin natural uranium and slightly enriched uranium(SEU). In the context of RU use in Canadian CANDU reactors, existing national and international transport regulations and arrangements adequately allow all material movements between the reprocessor, RU powder supplier, Canadian CANDU fuel manufacturer and Canadian CANDU reactor operator. Studies have been undertaken of the impact on personnel dose during fuel manufacturing operations from the increased specific activity of the RU compared to natural uranium. These studies have shown that this impact can be readily minimised without significant cost penalty to the acceptable levels recognised in modem standards for fuel manufacturing operations. The successful and extensive use of RU, arising from spent Magnox fuel, in British Energy's Advanced Gas-Cooled reactors is cited as relevant practical commercial scale experience. The CANFLEX fuel bundle design has been developed by AECL (Canada) and KAERI (Korea) to facilitate the achievement of higher bum-ups and greater fuel performance margins necessary if the full economic potential of advanced CANDU fuel cycles are to be achieved. The manufacture of a CANFLEX fuel bundle containing RU pellets derived from irradiated PWR fuel reprocessed in the THORP plant of BNFL is described. This provided a very practical verification of dose modelling calculations and also demonstrated that the increase of external activity is unlikely to require any

  2. Fuel cell and advanced turbine power cycle

    Energy Technology Data Exchange (ETDEWEB)

    White, D.J. [Solar Turbines, Inc., San Diego, CA (United States)

    1995-10-19

    Solar Turbines, Incorporated (Solar) has a vested interest in the integration of gas turbines and high temperature fuel cells and in particular, solid oxide fuel cells (SOFCs). Solar has identified a parallel path approach to the technology developments needed for future products. The primary approach is to move away from the simple cycle industrial machines of the past and develop as a first step more efficient recuperated engines. This move was prompted by the recognition that the simple cycle machines were rapidly approaching their efficiency limits. Improving the efficiency of simple cycle machines is and will become increasingly more costly. Each efficiency increment will be progressively more costly than the previous step.

  3. Characterisation of fuels for advanced pressurised combustion

    Energy Technology Data Exchange (ETDEWEB)

    Zevenhoven, R.; Hupa, M.; Backman, P.; Forssen, M.; Karlsson, M.; Kullberg, M.; Sorvari, V.; Uusikartano, T. [Aabo Akademi, Turku (Finland). Combustion Chemistry Research Group; Nurk, M. [Tallinskij Politekhnicheskij Inst., Tallinn (Estonia)

    1997-10-01

    The objective of the research was to determine a set of fuel characteristics which quantify the behaviour of a fuel in a typical pressurised combustor or gasifier environment, especially in hybrid processes such as second generation PFBC. One specific aspect was to cover a wide range of fuels, including several coal types and several grades of peat and biomasses: 7 types of coal, 2 types of peat, 2 types of wood, 2 types of black liquor, Estonian oil shale and Venezuelan Orimulsion were studied. The laboratory facilities used are a pressurised thermogravimetric reactor (PTGR), a pressurised grid heater (PGH) and an atmospheric entrained flow quartz tube reactor, with gas analysis, which can be operated as a fixed bed reactor. A major part of the work was related to fuel devolatilisation in the PGH and sequential devolatilisation and char gasification (with carbon dioxide or steam) in the PTGR. The final part of that work is reported here, with the combustion of Estonian oil shale at AFBC or PFBC conditions as additional subject. Devolatilisation of the fuels at atmospheric pressure in nitrogen while monitoring gaseous exhausts, followed by ultimate analysis of the chars has been reported earlier. Here, results on the analysis of the reduction of NO (with and without CO) on chars at atmospheric pressure in a fixed bed reactor are reported. Finally, a comparison is given between experimental results and direct numerical simulation with several computer codes, i.e. PyroSim, developed at TU Graz, Austria, and the codes Partikkeli, Pisara and Cogas, which were provided by VTT Energy, Jyvaeskylae

  4. Research and Development Strategy of France towards Sustainable Nuclear Fuel Cycles - Plenary session on 'Advanced Nuclear Fuel Cycles: Which Options? Which Strategies?'

    International Nuclear Information System (INIS)

    France is reprocessing the spent fuel of its PWR nuclear generating fleet since 1976. This strategy affords retrieving re-usable nuclear materials and segregating fission products with minor actinides (currently) for a safe packaging as vitrified waste and interim storage until disposal in a geological repository. Plutonium and reprocessed uranium (partly) are recycled once in PWRs (respectively as MOX and Rep-UO2 fuel) thus reducing by almost 15 % needs for uranium and enrichment separative work. Spent MOX fuel subassemblies are stored as fissile resource for future fast reactors. Scenario studies of nuclear power deployment and utilization of uranium worldwide lead to anticipate an industrial deployment of fast neutron reactors around 2040 in the French generating fleet. This sets an overall time frame for research and pilot-scale demonstrations on next generation fast reactors and advanced recycling modes to prepare their industrial deployment in time. Besides, the French bill of June 28, 2006 on 'A sustainable management of nuclear materials and radioactive waste' institutes a strategy for the nuclear fuel back-end in France with plans to open a high level long lived radioactive waste repository by 2025. Moreover, it establishes a link between continuing research on partitioning and transmutation to further decrease the long term burden of current waste packages, and research on Generation IV fast neutron systems with closed fuel cycles, thus acknowledging future fast power reactors as most likely nuclear systems to perform transmutation at industrial scale. This bill calls for identifying by 2012 recycling modes that have most promising industrial prospects and to proceed with their demonstration in a prototype fast reactor in the 2020's. Subsequently, French nuclear stakeholders (CEA, AREVA and EDF) currently conduct active research on advanced fast reactors, fuel and fuel cycle technologies for screening promising design features for a new generation of fast

  5. Development of challengeable reprocessing and fuel fabrication technologies for advanced fast reactor fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Nomura, S.; Aoshima, T.; Myochin, M. [Japan Nuclear Cycle Development Institute, Tokai Works (Japan)

    2001-07-01

    R and D in the next five years in Feasibility Study Phase-2 are focused on selected key technologies for the advanced fuel cycle. These are the reference technology of simplified aqueous extraction and fuel pellet short process based on the oxide fuel and the innovative technology of oxide-electrowinning and metal- electrorefining process and their direct particle/metal fuel fabrication methods in a hot cell. Automatic and remote handling system operation in both reprocessing and fuel manufacturing can handle MA and LLFP concurrently with Pu and U attaining the highest recovery and an accurate accountability of these materials. (author)

  6. Ultraclean Fuels Production and Utilization for the Twenty-First Century: Advances toward Sustainable Transportation Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Fox, Elise B.; Liu, Zhong-Wen; Liu, Zhao-Tie

    2013-11-21

    Ultraclean fuels production has become increasingly important as a method to help decrease emissions and allow the introduction of alternative feed stocks for transportation fuels. Established methods, such as Fischer-Tropsch, have seen a resurgence of interest as natural gas prices drop and existing petroleum resources require more intensive clean-up and purification to meet stringent environmental standards. This review covers some of the advances in deep desulfurization, synthesis gas conversion into fuels and feed stocks that were presented at the 245th American Chemical Society Spring Annual Meeting in New Orleans, LA in the Division of Energy and Fuels symposium on "Ultraclean Fuels Production and Utilization".

  7. Development of challengeable reprocessing and fuel fabrication technologies for advanced fast reactor fuel cycle

    International Nuclear Information System (INIS)

    R and D in the next five years in Feasibility Study Phase-2 are focused on selected key technologies for the advanced fuel cycle. These are the reference technology of simplified aqueous extraction and fuel pellet short process based on the oxide fuel and the innovative technology of oxide-electrowinning and metal- electrorefining process and their direct particle/metal fuel fabrication methods in a hot cell. Automatic and remote handling system operation in both reprocessing and fuel manufacturing can handle MA and LLFP concurrently with Pu and U attaining the highest recovery and an accurate accountability of these materials. (author)

  8. LWR spent fuel storage technology: Advances and experience

    International Nuclear Information System (INIS)

    By 2003, the year the US Department of Energy (DOE) currently predicts a repository will be available, 58 domestic commercial nuclear-power plant units are expected to run out of wet storage space for LWR spent fuel. To alleviate this problem, utilities implemented advances in storage methods that increased storage capacity as well as reduced the rate of generating spent fuel. Those advances include (1) transhipping spent-fuel assemblies between pools within the same utility system, (2) reracking pools to accommodate additional spent-fuel assemblies, (3) taking credit for fuel burnup in pool storage rack designs, (4) extending fuel burnup, (5) rod consolidation, and (6) dry storage. The focus of this paper is on advances in rod consolidation and dry storage. Wet storage continues to be the predominant US spent-fuel management technology, but as a measure to enhance at-reactor storage capacity, the Nuclear Waste Policy Act of 1982 authorized DOE to assist utilities with licensing at-reactor dry storage. Information exchanges with other nations, laboratory testing and modeling, and cask tests cooperatively funded by US utilities and DOE produced a strong technical basis to develop confidence that LWR spent fuel can be stored safely for several decades in both wet and dry modes. Licensed dry storage of spent fuel in an inert atmosphere was first achieved in the US in 1986. Studies are underway in several countries to determine acceptable conditions for storing LWR spent fuel in air. Rod-consolidation technology is being developed and demonstrated to enhance the capacity for both wet and dry storage. Large-scale commercial implementation is awaiting optimization of practical and economical mechanical systems. 22 refs., 1 fig

  9. Strategic research of advanced fuel cycle technologies in JNC

    International Nuclear Information System (INIS)

    Key technologies for the future nuclear fuel cycle have been proposed and are being reviewed in JNC as a part of the Feasibility Study for an Advanced Fuel Cycle, which is to achieve a more flexible energy choice to satisfy a sustainable energy security and global environmental protection. The candidate reprocessing technologies are: 1) aqueous simplified PUREX process, 2) oxide or metallic electrowinning, and 3) fluoride volatilization for oxide, metal, or nitride fuels. The fuel fabrication methods being investigated are: 1) simplified pellet process, 2) sphere/vibro-packed process for MOX/MN fuel, and 3) casting for metal fuel. These candidate technologies are currently being compared based on past experiences, technical issues to be solved, industrial applicability for future plants, feasible options for MA/LLFP separation, and nonproliferation aspects. Alter two years of the present reviewing process, selected key technologies will be developed over the next five years to evaluate industrial applicability of reprocessing and fuel manufacturing processes for the advanced fuel cycle. (authors)

  10. Methodology for the Weapons-Grade MOX Fuel Burnup Analysis in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    A UNIX BASH (Bourne Again SHell) script CMO has been written and validated at the Idaho National Laboratory (INL) to couple the Monte Carlo transport code MCNP with the depletion and buildup code ORIGEN-2 (CMO). The new Monte Carlo burnup analysis methodology in this paper consists of MCNP coupling through CMO with ORIGEN-2, and is therefore called the MCWO. MCWO is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN-2. MCWO is capable of handling a large number of fuel burnup and material loading specifications, Advanced Test Reactor (ATR) lobe powers, and irradiation time intervals. MCWO processes user input that specifies the system geometry, initial material compositions, feed/removal specifications, and other code-specific parameters. Calculated results from MCNP, ORIGEN-2, and data process module calculations are output in succession as MCWO executes. The principal function of MCWO is to transfer one-group cross-section and flux values from MCNP to ORIGEN-2, and then transfer the resulting material compositions (after irradiation and/or decay) from ORIGEN-2 back to MCNP in a repeated, cyclic fashion. The basic requirements of MCWO are a working MCNP input file and some additional input parameters; all interaction with ORIGEN-2 as well as other calculations are performed by CMO. This paper presents the MCWO-calculated results for the Reduced Enrichment Research and Test Reactor (RERTR) experiments RERTR-1 and RERTR-2 as well as the Weapons-Grade Mixed Oxide (WG-MOX) fuel testing in ATR. Calculations performed for the WG-MOX test irradiation, which is managed by the Oak Ridge National Laboratory (ORNL), supports the DOE Fissile Materials Disposition Program (FMDP). The MCWO-calculated results are compared with measured data

  11. Natural Gas for Advanced Dual-Fuel Combustion Strategies

    Science.gov (United States)

    Walker, Nicholas Ryan

    Natural gas fuels represent the next evolution of low-carbon energy feedstocks powering human activity worldwide. The internal combustion engine, the energy conversion device widely used by society for more than one century, is capable of utilizing advanced combustion strategies in pursuit of ultra-high efficiency and ultra-low emissions. Yet many emerging advanced combustion strategies depend upon traditional petroleum-based fuels for their operation. In this research the use of natural gas, namely methane, is applied to both conventional and advanced dual-fuel combustion strategies. In the first part of this work both computational and experimental studies are undertaken to examine the viability of utilizing methane as the premixed low reactivity fuel in reactivity controlled compression ignition, a leading advanced dual-fuel combustion strategy. As a result, methane is shown to be capable of significantly extending the load limits for dual-fuel reactivity controlled compression ignition in both light- and heavy-duty engines. In the second part of this work heavy-duty single-cylinder engine experiments are performed to research the performance of both conventional dual-fuel (diesel pilot ignition) and advanced dual-fuel (reactivity controlled compression ignition) combustion strategies using methane as the premixed low reactivity fuel. Both strategies are strongly influenced by equivalence ratio; diesel pilot ignition offers best performance at higher equivalence ratios and higher premixed methane ratios, whereas reactivity controlled compression ignition offers superior performance at lower equivalence ratios and lower premixed methane ratios. In the third part of this work experiments are performed in order to determine the dominant mode of heat release for both dual-fuel combustion strategies. By studying the dual-fuel homogeneous charge compression ignition and single-fuel spark ignition, strategies representative of autoignition and flame propagation

  12. Non-proliferation, safeguards, and security for the fissile materials disposition program immobilization alternatives

    Energy Technology Data Exchange (ETDEWEB)

    Duggan, R.A.; Jaeger, C.D.; Tolk, K.M. [Sandia National Labs., Albuquerque, NM (United States); Moore, L.R. [Lawrence Livermore National Lab., CA (United States)

    1996-05-01

    The Department of Energy is analyzing long-term storage and disposition alternatives for surplus weapons-usable fissile materials. A number of different disposition alternatives are being considered. These include facilities for storage, conversion and stabilization of fissile materials, immobilization in glass or ceramic material, fabrication of fissile material into mixed oxide (MOX) fuel for reactors, use of reactor based technologies to convert material into spent fuel, and disposal of fissile material using geologic alternatives. This paper will focus on how the objectives of reducing security and proliferation risks are being considered, and the possible facility impacts. Some of the areas discussed in this paper include: (1) domestic and international safeguards requirements, (2) non-proliferation criteria and measures, (3) the threats, and (4) potential proliferation, safeguards, and security issues and impacts on the facilities. Issues applicable to all of the possible disposition alternatives will be discussed in this paper. However, particular attention is given to the plutonium immobilization alternatives.

  13. Characterisation of fuels for advanced pressurized combustion

    Energy Technology Data Exchange (ETDEWEB)

    Zevenhoven, R.; Hupa, M.; Backman, P.; Karlsson, M.; Kullberg, M.; Sorvari, V. [Aabo Akademi, Turku (Finland); Nurk, M. [Tallinn Univ. (Estonia)

    1996-12-01

    After 2 of the 3 years for this EU Joule 2 extension project, a rough comparison on the devolatilisation behaviour and char reactivity of 11 fossil fuels and 4 biofuels has been obtained. The experimental plan for 1995 has been completed, the laboratory facilities appeared to be well suited for the broad range of analyses presented here. A vast amount of devolatilisation tests in nitrogen at atmospheric pressure with gas analysis and char analysis gave a lot of information on the release of carbon, sulphur, nitrogen and also sodium, chloride and some other elements. Also first-order rate parameters could be determined. Solid pyrolysis yield measurements with the pressurised grid heater show a very good reproducibility except for the fuels with high carbonate content and those with very small char yield. Problems have to be solved considering lower heating rates and the use of folded grids. Fuel pyrolysis followed by gasification (with carbon dioxide or water as oxidising agent) at various temperatures and pressures shows that in general char solid yields and gasification reactivities are higher at elevated pressure. The design and construction of a pressurized single particle reactor, to be operational early 1996 is currently being negotiated. Numerical modelling of coal devolatilisation shows that even for atmospheric pressures the results differ significantly from experimental findings. (author)

  14. The AMP (Advanced MultiPhysics) Nuclear Fuel Performance code

    International Nuclear Information System (INIS)

    Highlights: ► New, three-dimensional, parallel, multi-physics code to simulate fuel behavior in nominal operation. ► Fully-coupled thermomechanics for nominal operation and operation during transients. ► Isotopic depletion using Scale/ORIGEN-S within a fuel performance code. ► Leveraging of existing, validated material models from existing fuel performance codes. ► Initial validation evaluation of an advanced modeling and simulation code for fuel performance. - Abstract: The AMP (Advanced MultiPhysics) Nuclear Fuel Performance code is a new, three-dimensional, multi-physics tool that uses state-of-the-art solution methods and validated nuclear fuel models to simulate the nominal operation and anticipated operational transients of nuclear fuel. The AMP Nuclear Fuel Performance code leverages existing validated material models from traditional fuel performance codes and the Scale/ORIGEN-S spent-fuel characterization code to provide an initial capability that is shown to be sufficiently accurate for a single benchmark problem and anticipated to be accurate for a broad range of problems. The thermomechanics foundation can be solved in a time-dependent or quasi-static approach with any variation of operator-split or fully-coupled solutions at each time step through interoperable interfaces to leading computational mathematics tools, including PETSc, Trilinos, and SUNDIALS. A baseline validation of the AMP Nuclear Fuel Performance code has been performed through the modeling of an experiment in the Halden Reactor Project (IFA-432) that demonstrates the integrated capability and provides a baseline of the initial accuracy of the software.

  15. Alternative Fuel and Advanced Technology Commercial Lawn Equipment (Brochure)

    Energy Technology Data Exchange (ETDEWEB)

    2014-10-01

    The U.S. Department of Energy's Clean Cities program produced this guide to help inform the commercial mowing industry about product options and potential benefits. This guide provides information about equipment powered by propane, ethanol, compressed natural gas, biodiesel, and electricity, as well as advanced engine technology. In addition to providing an overview for organizations considering alternative fuel lawn equipment, this guide may also be helpful for organizations that want to consider using additional alternative fueled equipment.

  16. Alternative Fuel and Advanced Technology Commercial Lawn Equipment

    Energy Technology Data Exchange (ETDEWEB)

    None

    2014-10-10

    The U.S. Department of Energy's Clean Cities program produced this guide to help inform the commercial mowing industry about product options and potential benefits. This guide provides information about equipment powered by propane, ethanol, compressed natural gas, biodiesel, and electricity, as well as advanced engine technology. In addition to providing an overview for organizations considering alternative fuel lawn equipment, this guide may also be helpful for organizations that want to consider using additional alternative fueled equipment.

  17. Advancing liquid metal reactor technology with nitride fuels

    International Nuclear Information System (INIS)

    A review of the use of nitride fuels in liquid metal fast reactors is presented. Past studies indicate that both uranium nitride and uranium/plutonium nitride possess characteristics that may offer enhanced performance, particularly in the area of passive safety. To further quantify these effects, the analysis of a mixed-nitride fuel system utilizing the geometry and power level of the US Advanced Liquid Metal Reactor as a reference is described. 18 refs., 2 figs., 2 tabs

  18. Lessons Learned From Dynamic Simulations of Advanced Fuel Cycles

    Energy Technology Data Exchange (ETDEWEB)

    Steven J. Piet; Brent W. Dixon; Jacob J. Jacobson; Gretchen E. Matthern; David E. Shropshire

    2009-04-01

    Years of performing dynamic simulations of advanced nuclear fuel cycle options provide insights into how they could work and how one might transition from the current once-through fuel cycle. This paper summarizes those insights from the context of the 2005 objectives and goals of the Advanced Fuel Cycle Initiative (AFCI). Our intent is not to compare options, assess options versus those objectives and goals, nor recommend changes to those objectives and goals. Rather, we organize what we have learned from dynamic simulations in the context of the AFCI objectives for waste management, proliferation resistance, uranium utilization, and economics. Thus, we do not merely describe “lessons learned” from dynamic simulations but attempt to answer the “so what” question by using this context. The analyses have been performed using the Verifiable Fuel Cycle Simulation of Nuclear Fuel Cycle Dynamics (VISION). We observe that the 2005 objectives and goals do not address many of the inherently dynamic discriminators among advanced fuel cycle options and transitions thereof.

  19. Advanced fuel technology and performance: Current status and trends

    International Nuclear Information System (INIS)

    During the last years the Nuclear Fuel Cycle and Waste Management Division of the IAEA has been giving great attention to the collection, analysis and exchange of information in the field of reactor fuel technology. Most of these activities are being conducted in the framework of the International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT). The purpose of this Advisory Group Meeting on Advanced Fuel Technology and Performance was to update and to continue the previous work, and to review the experience of advanced fuel technology, its performance with regard to all types of reactors and to outline the future trends on the basis of national experience and discussions during the meeting. As a result of the meeting a Summary Report was prepared which reflected the status of the advanced nuclear fuel technology up to 1990. The 10 papers presented by participants of this meeting are also published here. A separate abstract was prepared for each of these papers. Refs, figs and tabs

  20. Design study on advanced reprocessing systems for FR fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, H.; Kawamura, F.; Nishimura, T.; Kamiya, M. [Japan Nuclear Cycle Development Institute, Ibaraki (Japan)

    2001-07-01

    A design study has been carried out for four advanced reprocessing technologies for the future fast rector (FR) recycle systems (advanced aqueous, and three non-aqueous systems based on oxide electrowinning, metal electrorefining, and fluoride volatility methods). The systems were evaluated mainly from the viewpoint of economics. It has been shown that, for MOX fuel reprocessing, all the systems with a capacity of 200 t/y attains the economical target, whereas for such a small capacity as 50 t/y, only the non-aqueous systems have potential to attain the target. For metallic and nitride fuel, a metal electrorefining system has been shown to be advantageous. (author)

  1. IEA-Advanced Motor Fuels Annual Report 2012

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-06-15

    The annual report from the IEA implementing agreement on Advanced Motor Fuels (AMF) describes what the agreement is about, how to join, various activities of the agreement, a message from the Chairman, and projects/annexes active for the year. An annual section covers the global situation for the topic of advanced motor fuels. Another section includes highlights coming from each country participating in AMF, and major sections relaying activities on each of the ongoing annexes. Information regarding participating delegations, contact information, publications resulting from AMF, and upcoming meetings rounds out the report.

  2. IEA-Advanced Motor Fuels Annual Report 2011

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-07-01

    The annual report from the IEA implementing agreement on Advanced Motor Fuels (AMF) describes what the agreement is about, how to join, various activities of the agreement, a message from the Chairman, and projects/annexes active for the year. An annual section covers the global situation for the topic of advanced motor fuels. Another section includes highlights coming from each country participating in AMF, and major sections relaying activities on each of the ongoing annexes. Information regarding participating delegations, contact information, publications resulting from AMF, and upcoming meetings rounds out the report.

  3. Design study on advanced reprocessing systems for FR fuel cycle

    International Nuclear Information System (INIS)

    A design study has been carried out for four advanced reprocessing technologies for the future fast rector (FR) recycle systems (advanced aqueous, and three non-aqueous systems based on oxide electrowinning, metal electrorefining, and fluoride volatility methods). The systems were evaluated mainly from the viewpoint of economics. It has been shown that, for MOX fuel reprocessing, all the systems with a capacity of 200 t/y attains the economical target, whereas for such a small capacity as 50 t/y, only the non-aqueous systems have potential to attain the target. For metallic and nitride fuel, a metal electrorefining system has been shown to be advantageous. (author)

  4. Non-proliferation issues for the disposition of fissile materials using reactor alternatives

    International Nuclear Information System (INIS)

    The Department of Energy (DOE) is analyzing long-term storage on options for excess weapons-usable fissile materials. A number of the disposition alternatives are being considered which involve the use of reactors. The various reactor alternatives are all very similar and include front-end processes that could convert plutonium to a usable form for fuel fabrication, a MOX fuel fab facility, reactors to bum the MOX fuel and ultimate disposal of spent fuel in some geologic repository. They include existing, partially completed, advanced or evolutionary light water reactors and Canadian deuterium uranium (CANDU) reactors. In addition to the differences in the type of reactors, other variants on these alternatives are being evaluated to include the location and number of the reactors, the location of the mixed oxide (MOX) fabrication facility, the ownership of the facilities (private or government) and the colocation and/or separation of these facilities. All of these alternatives and their variants must be evaluated with respect to non-proliferation resistance. Both domestic and international safeguards support are being provided to DOE's Fissile Materials Disposition Program (FMDP) and includes such areas as physical protection, nuclear materials accountability and material containment and surveillance. This paper will focus on how the non-proliferation objective of reducing security risks and strengthening arms reduction will be accomplished and what some of the nonproliferation issues are for the reactor alternatives. Proliferation risk has been defined in terms of material form, physical environment, and the level of security and safeguards that is applied to the material. Metrics have been developed for each of these factors. The reactor alternatives will be evaluated with respect to these proliferation risk factors at each of the unit process locations in the alternative

  5. Advanced PWR fuel assembly development programs in Korea

    International Nuclear Information System (INIS)

    Both KNFC and Westinghouse have continued to focus on developing products that will meet the challenge of increasing fuel duty requirements in Korea. These higher duty conditions include higher energy core designs through improved plant capacity factors, power uprate, extended fuel burnup, peaking factor increases, and more severe coolant chemistry (including high lithium concentration). Recent advanced fuel development activities in Korea include implementation of the 17x17 Robust Fuel Assembly (RFA), which is currently in operation with excellent performance in the United States and Europe, as well as the 16x16 PLUS7TM fuel assembly for use in KSNP plants. KNFC and Westinghouse are jointly developing advanced fuel that will meet future fuel duty challenges of 17x17 and 16x16 Westinghouse type plants. This paper focuses on advanced fuel assembly development programs that are underway and how these designs demonstrate improved margins under high duty plant operating conditions. In designing for these high duty conditions key design considerations for the various operational modes (i.e. power uprating, high burnup, long cycles, etc.) must be identified. These design considerations will include the traditional factors such as safety margin (DNB and LOCA), fuel rod design margin (e.g. corrosion, internal pressure, etc.) and mechanical design margins, among others. In addressing these design considerations, the fundamental approach is to provide additional design margin through materials, mechanical, and thermal performance enhancements, to assure flawless fuel performance. The foundation of all fuel designs is the product development process used to meet the demands of modern high duty operation including power uprating, high burnup, longer cycles, and high-lithium coolant chemistries. These advanced fuel assembly designs incorporate features that provide improved mechanical design margin, as well as thermal performance margin (DNB). Enhanced grid designs result in a

  6. Selection and development of advanced nuclear fuel products

    International Nuclear Information System (INIS)

    The highly competitive international marketplace requires a continuing product development commitment, short development cycle times and timely, on-target product development to assure customer satisfaction and continuing business. Westinghouse has maintained its leadership position within the nuclear fuel industry with continuous developments and improvements to fuel assembly materials and design. This paper presents a discussion of the processes used by Westinghouse in the selection and refinement of advanced concepts for deployment in the highly competitive US and international nuclear fuel fabrication marketplace. (author)

  7. Criticality Control Fissile of Materials. Proceedings of the Symposium on Criticality Control of Fissile Materials

    International Nuclear Information System (INIS)

    Criticality control comprises all the administrative and technical procedures which enable the storage and processing of fissile material to be carried out under conditions of nuclear safety. It is of particular importance in the safe design and operation of chemical and metallurgical plants processing fissile material, in the handling and storage of enriched fuel for reactors, and in transportation of fissile material. The growth of nuclear power, with its increasing use of fissile material and production of plutonium, is leading to an ever widening need for this discipline. This Symposium was held 4½ years after the only other international meeting on this topic, at which the first broad exchange of ideas and theories enabled a comparison to be drawn between the various ways in which the subject is handled in the different countries. That meeting showed that criticality safety was often achieved by procedures known to be ultra-safe, as there was a great lack of useful experimental data with which to check theoretical models. Since that time the quantities of material being processed have increased, and with the now urgent necessity of achieving economic, and hence commercially competitive, operation, the procedure of using arbitrary factors of safety is no longer adequate. Plant Managers now require good data on the basis of which they can choose a suitable factor of safety, and design a process to be safe under any foreseeable circumstances. The present Symposium showed the great increase in the amount of available experimental data and its importance in checking the now highly sophisticated computer calculations. There are many diagrams in these Proceedings with curves from which critical parameters for various configurations can be taken. The dearth of data for plutonium systems is causing some difficulty in plutonium processing plants, which are becoming commercially important. The excellent safety record of the atomic energy industry was highlighted, and a

  8. Uranium-plutonium carbide as an LMFBR advanced fuel

    International Nuclear Information System (INIS)

    Uranium-plutonium carbide offers an improved fuel system for advanced breeder reactors. The high thermal conductivity and density of carbide fuels permit superior breeding performance and high specific power operation. These advantages combine to increase plutonium production, reduce fuel cycle and power costs, and lower plant capital costs. The carbide advantages are obtained at conservative fuel sytem design and operating conditions. Carbide fabrication technology has been demonstrated by the production of quality-assured fuel elements for irradiation testing. The carbide irradiation test program has demonstrated that high burnup can be achieved with several designs and that the consequences of postulated off-normal operating events are benign. Design bases to support helium- and sodium-bonded carbide fuel pin test irradiations in the Fast Flux Test Facility have been developed in the Experimental Breeder Reactor-II and the Transient Reactor irradiation experiments. Important issues regarding safety, reprocessing, and commercial-scale fabrication remain to be addressed in the continuing development of carbide fuels. Fiscal and historical circumstances have combined to preclude this development. This report reviews these circumstances and the state of the technology in general and advances a rationale for why development should be continued

  9. Structural evaluation of Siemens advanced fuel channel under accident loadings

    International Nuclear Information System (INIS)

    As a part of an effort to develop an advanced BWR fuel channel design, Siemens Power Corporation (SPC) and the Siemens AG Power Generation Group (KWU) performed structural analyses to verify the acceptability of the fuel channel design under combined seismic/LOCA (Loss Of. Coolant Accident) loadings. The results of the analyses give some interesting insights into the problem: 1) fluid-structure interaction (FSI) effects are significant and should be considered, 2) the problem may simplified by using a linear analysis despite non-linear features (gaps) between interfacing components, and 3) sufficient accuracy may be obtained by using only the first mode of vibration. The channeled fuel assembly can be considered to be a beam where the flexural stiffness is primarily determined by the fuel channel and the mass is given by the fuel assembly. The results from the analyses show the advanced fuel channel design meets applicable design criteria with adequate margins while at the same time exhibiting superior nuclear performance compared to a conventional BWR fuel channel. (author)

  10. Development of advanced zirconium fuel cladding

    International Nuclear Information System (INIS)

    This report includes a series of the characterization results of candidate alloys, the manufacturing description of the advanced sample cladding tubes, and both the summary of out-of pile tests and the overview plan of in-pile test for them. Ten(10) kinds of the second candidate alloys, which had been selected at the first stage of the project, were comprehensively tested for their out-of pile performance. Six(6) kinds of the alloys were selected of the second ones as the final candidates through the screening tests. The out-of pile performance of the final candidates were superior to that of zircaloy-4. The advanced sample cladding tubes were made of the final candidates and tested for their out-of pile performances. The corrosion behaviors of the tubes were evaluated though the corrosion tests in water at 360 .deg. C, steam at 400 .deg. C and LiOH solution at 360 .deg. C. The mechanical properties such as creep, tensile and burst were also evaluated for each tube. The textures, microstructures, precipitates and hydrides of each tube were analyzed as well as the phase transformation was studied for each tube. In general, the test results showed that the performance of the advanced sample cladding tubes was improved over 30% in corrosion and 20% in mechanical property than that of zircaloy-4. The in-pile test of the tubes for the first phase was arranged from January 2003 to March 2007

  11. A Consistent Comparative Study of Advanced Sodium-cooled Fast Burner Cores loaded with Thorium and Uranium-based Metallic Fuels

    International Nuclear Information System (INIS)

    We considered uranium-based metallic fuel of TRU-U-10Zr for driver fuel and thorium was considered as blanket because thorium blanket produces less amount of TRU than uranium blanket and use of thorium blanket leads to smaller sodium void worth than the use of uranium blanket due to the fact that the η-value increases much less with energy for 233U than for 239Pu and 232Th is less fissile than 238U. However, these cores using thorium blanket still have a large amount of TRU production from the driver fuels because the driver fuels contain a large amount of depleted uranium which leads to the production of TRU through neutron capture. The objective of this work is to consistently compare the neutronic performances of advanced sodium cooled fast reactor cores loaded with thorium and uraniumbased metallic fuels as driver fuel for TRU burning. Our main emphasis is given on the analyses of the differences in the core performance parameters. For consistent comparison, we used the same core configuration and all the same design parameters except for the fact that depleted uranium in uraniumbased fuel is replaced with thorium. We considered the cores having no thorium blanket and the cores having thorium blanket that were designed in our previous works

  12. Advanced methods for fabrication of PHWR and LMFBR fuels

    International Nuclear Information System (INIS)

    For self-reliance in nuclear power, the Department of Atomic Energy (DAE), India is pursuing two specific reactor systems, namely the pressurised heavy water reactors (PHWR) and the liquid metal cooled fast breeder reactors (LMFBR). The reference fuel for PHWR is zircaloy-4 clad high density (≤ 96 per cent T.D.) natural UO2 pellet-pins. The advanced PHWR fuels are UO2-PuO2 (≤ 2 per cent), ThO2-PuO2 (≤ 4 per cent) and ThO2-U233O2 (≤ 2 per cent). Similarly, low density (≤ 85 per cent T.D.) (UPu)O2 pellets clad in SS 316 or D9 is the reference fuel for the first generation of prototype and commercial LMFBRs all over the world. However, (UPu)C and (UPu)N are considered as advanced fuels for LMFBRs mainly because of their shorter doubling time. The conventional method of fabrication of both high and low density oxide, carbide and nitride fuel pellets starting from UO2, PuO2 and ThO2 powders is 'powder metallurgy (P/M)'. The P/M route has, however, the disadvantage of generation and handling of fine powder particles of the fuel and the associated problem of 'radiotoxic dust hazard'. The present paper summarises the state-of-the-art of advanced methods of fabrication of oxide, carbide and nitride fuels and highlights the author's experience on sol-gel-microsphere-pelletisation (SGMP) route for preparation of these materials. The SGMP process uses sol gel derived, dust-free and free-flowing microspheres of oxides, carbide or nitride for direct pelletisation and sintering. Fuel pellets of both low and high density, excellent microhomogeneity and controlled 'open' or 'closed' porosity could be fabricated via the SGMP route. (author). 5 tables, 14 figs., 15 refs

  13. Candu advanced fuel cycles: key to energy sustainability

    International Nuclear Information System (INIS)

    A primary rationale for Indonesia to proceed with a nuclear power program is to diversity its energy sources and achieve freedom from future resource constraints. While other considerations, such as economy of power supply, hedging against potential future increases in the price of fossil fuels, fostering the technological development of the Indonesia economy and minimizing greenhouse and other gaseous are important, the strategic resource issue is key. In considering candidate nuclear power technologies upon which to base such a program, a major consideration will be the potential for those technologies to be economically sustained in the face of large future increases in demand for nuclear fuels. the technology or technologies selected should be amenable to evaluation in a rapidly changing technical, economic, resource and environmental policy environment. the world's proven uranium resources which can be economically recovered represent a fairly modest energy resource if utilization is based on the currently commercialized fuel cycles, even with the use of recovered plutonium in mixed oxide fuels. In the long term, fuel cycles relying solely on the use of light water reactors will encounter increasing fuel supply constraints. Because of its outstanding neutron economy and the flexibility of on-power refueling, Candu reactors are the most fuel resource efficient commercial reactors and offer the potential for accommodating an almost unlimited variety of advanced and even more fuel efficient cycles. Most of these cycles utilize nuclear fuel which are too low grade to be used in light water reactors, including many products now considered to be waste, such as spent light water reactor fuel and reprocessing products such as recovered uranium. The fuel-cycle flexibility of the Candu reactor provides a ready path to sustainable energy development in both the short and the long terms. Most of the potential Candu fuel cycle developments can be accommodated in existing

  14. Advanced methods of quality control in nuclear fuel fabrication

    International Nuclear Information System (INIS)

    Under pressure of current economic and electricity market situation utilities implement more demanding fuel utilization schemes including higher burn ups and thermal rates, longer fuel cycles and usage of Mo fuel. Therefore, fuel vendors have recently initiated new R and D programmes aimed at improving fuel quality, design and materials to produce robust and reliable fuel. In the beginning of commercial fuel fabrication, emphasis was given to advancements in Quality Control/Quality Assurance related mainly to product itself. During recent years, emphasis was transferred to improvements in process control and to implementation of overall Total Quality Management (TQM) programmes. In the area of fuel quality control, statistical control methods are now widely implemented replacing 100% inspection. This evolution, some practical examples and IAEA activities are described in the paper. The paper presents major findings of the latest IAEA Technical Meetings (TMs) and training courses in the area with emphasis on information received at the TM and training course held in 1999 and other latest publications to provide an overview of new developments in process/quality control, their implementation and results obtained including new approaches to QC

  15. System dynamics studies of advanced fuel cycle scenarios

    International Nuclear Information System (INIS)

    This work describes dynamic analysis studies of possible U.S. deployment scenarios of advanced nuclear energy systems. Different scenarios of future nuclear energy demand and different spent nuclear fuel management strategies to respond to those demands are considered. The management strategies include once-through, limited recycling, and transitional and sustained recycling strategies. The scenarios descriptions, data, timeline, and analysis are provided. Comparisons between the once-through and the recycling strategies show that the continuation of the current once-through fuel cycle practice can lead to unfavorable consequences as the demand for nuclear energy increase in the US. Those consequences include substantial increase in the number of geologic repository sites, continued accumulation of weapons-usable materials, and inefficient use of limited uranium resources. The analysis presented here shows that those concerns can only be addressed by employing an advanced fuel cycle. (author)

  16. Effect of advanced fuel cycles on waste management policies

    International Nuclear Information System (INIS)

    The study aims at analysing a range of future fuel cycle options from the perspective of their impact on waste repository demand and specification. The study would focus on: Assessment of the characteristics of radioactive wastes arising from advanced nuclear fuel cycle options, repository performance analysis studies using source terms for waste arising from such advanced nuclear fuel cycles, identification of new options for waste management and disposal. Three families of fuel cycles having increasing recycling capabilities are assessed. Each cycle is composed of waste generating and management processes. Examples of waste generating processes are fuel factories (7 types) and reprocessing plants (7 types). Packaging and conditioning plants (7) and disposal facilities are examples of waste management processes. The characteristic of all these processes have been described and then total waste flows are summarised. In order to simplify the situation, three waste categories have been defined based on the IAEA definitions in order to emphasize the major effects of different types of waste. These categories are: short-life waste for surface or sub-surface disposal, long-life low heat producing waste for geological disposal, high-level waste for geological disposal. The feasibilities of the fuel cycles are compared in terms of economics, primary resource consumption and amount of waste generated. The effect of high-level waste composition for the repository performance is one of the tools in these comparisons. The results of this will be published as an NEA publication before the end of 2005. (authors)

  17. Fuel-cycle greenhouse gas emissions impacts of alternative transportation fuels and advanced vehicle technologies

    International Nuclear Information System (INIS)

    At an international conference on global warming, held in Kyoto, Japan, in December 1997, the United States committed to reduce its greenhouse gas (GHG) emissions by 7% over its 1990 level by the year 2012. To help achieve that goal, transportation GHG emissions need to be reduced. Using Argonne's fuel-cycle model, I estimated GHG emissions reduction potentials of various near- and long-term transportation technologies. The estimated per-mile GHG emissions results show that alternative transportation fuels and advanced vehicle technologies can help significantly reduce transportation GHG emissions. Of the near-term technologies evaluated in this study, electric vehicles; hybrid electric vehicles; compression-ignition, direct-injection vehicles; and E85 flexible fuel vehicles can reduce fuel-cycle GHG emissions by more than 25%, on the fuel-cycle basis. Electric vehicles powered by electricity generated primarily from nuclear and renewable sources can reduce GHG emissions by 80%. Other alternative fuels, such as compressed natural gas and liquefied petroleum gas, offer limited, but positive, GHG emission reduction benefits. Among the long-term technologies evaluated in this study, conventional spark ignition and compression ignition engines powered by alternative fuels and gasoline- and diesel-powered advanced vehicles can reduce GHG emissions by 10% to 30%. Ethanol dedicated vehicles, electric vehicles, hybrid electric vehicles, and fuel-cell vehicles can reduce GHG emissions by over 40%. Spark ignition engines and fuel-cell vehicles powered by cellulosic ethanol and solar hydrogen (for fuel-cell vehicles only) can reduce GHG emissions by over 80%. In conclusion, both near- and long-term alternative fuels and advanced transportation technologies can play a role in reducing the United States GHG emissions

  18. Fissile solution dynamics: Student research

    Energy Technology Data Exchange (ETDEWEB)

    Hetrick, D.L.

    1994-09-01

    There are two research projects in criticality safety at the University of Arizona: one in dynamic simulation of hypothetical criticality accidents in fissile solutions, and one in criticality benchmarks using transport theory. We have used the data from nuclear excursions in KEWB, CRAC, and SILENE to help in building models for solution excursions. An equation of state for liquids containing gas bubbles has been developed and coupled to point-reactor dynamics in an attempt to predict fission rate, yield, pressure, and kinetic energy. It appears that radiolytic gas is unimportant until after the first peak, but that it does strongly affect the shape of the subsequent power decrease and also the dynamic pressure.

  19. Advanced materials for solid oxide fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Armstrong, T.R.; Stevenson, J.

    1995-08-01

    The purpose of this research is to improve the properties of the current state-of-the-art materials used for solid oxide fuel cells (SOFCs). The objectives are to: (1) develop materials based on modifications of the state-of-the-art materials; (2) minimize or eliminate stability problems in the cathode, anode, and interconnect; (3) Electrochemically evaluate (in reproducible and controlled laboratory tests) the current state-of-the-art air electrode materials and cathode/electrolyte interfacial properties; (4) Develop accelerated electrochemical test methods to evaluate the performance of SOFCs under controlled and reproducible conditions; and (5) Develop and test materials for use in low-temperature SOFCs. The goal is to modify and improve the current state-of-the-art materials and minimize the total number of cations in each material to avoid negative effects on the materials properties. Materials to reduce potential deleterious interactions, (3) improve thermal, electrical, and electrochemical properties, (4) develop methods to synthesize both state-of-the-art and alternative materials for the simultaneous fabricatoin and consolidation in air of the interconnections and electrodes with the solid electrolyte, and (5) understand electrochemical reactions at materials interfaces and the effects of component composition and processing on those reactions.

  20. Development of advanced zirconium fuel cladding

    International Nuclear Information System (INIS)

    This report includes the manufacturing technology developed for HANATM claddings, a series of their characterization results as well as the results of their in-pile and out-of pile performances tests which were carried out to develop some fuel claddings for a high burn-up (70,000MWd/mtU) which are competitive in the world market. Some of the HANATM claddings, which had been manufactured based on the results from the 1st and 2nd phases of the project, have been tested in a research reactor in Halden of Norway for an in-pile performance qualification. The results of the in-pile test showed that the performance of the HANATM claddings for corrosion and creep was better than 50% compared to that of Zircaloy-4 or A cladding. It was also found that the out-of pile performance of the HANATM claddings for such as LOCA and RIA in some accident conditions corrosion creep, tensile, burst and fatigue was superior or equivalent to that of the Zircaloy-4 or A cladding. The project also produced the other many data which were required to get a license for an in-pile test of HANATM claddings in a commercial reactor. The data for the qualification or characterization were provided for KNFC to assist their activities to get the license for the in-pile test of HANATM Lead Test Rods(LTR) in a commercial reactor

  1. Advanced teleoperation in nuclear applications: consolidated fuel reprocessing program

    International Nuclear Information System (INIS)

    A new generation of integrated remote maintenance systems is being developed to meet the needs of future nuclear fuel reprocessing at the Oak Ridge National Laboratory. Development activities cover all aspects of an advanced teleoperated maintenance system with particular emphasis on a new force-reflecting servomanipulator concept. The new manipulator, called the advanced servomanipulator, is microprocessor controlled and is designed to achieve force-reflection performance near that of mechanical master/slave manipulators. The advanced servomanipulator uses a gear-drive transmission which permits modularization for remote maintainability (by other advanced servomanipulators) and increases reliability. Human factors analysis has been used to develop an improved man/machine interface concept based upon colographic displays and menu-driven touch screens. Initial test and evaluation of two advanced servomanipulator slave arms and several other development components have begun. 9 references, 5 figures

  2. Advances in nuclear fuel technology. 2. Advances in nuclear fuel technology for LWRs

    International Nuclear Information System (INIS)

    From a viewpoint of upgrading on economical efficiency, some developments aiming at reduction of fuel cycle cost and used fuel forming amounts and response to long term operation cycle are carried out. These developments are required for back-fitness for already established reactors, so are progressed under limited changing tolerances. On fuels for BWRs, under confirming their used results, stepwise planning on upgrading of burnup such as steps 1, 2 and 3 is examined. For example, on a new 8 x 8 zirconium liner fuel (step 1), by adapting a zirconium liner cladding tube, its PCI (fuel pellet-cladding interaction) resistance feature is largely improved, to reach about 33 GWd/t in average discharge burnup. And, a high burnup 8 x 8 fuel (step 2) is intended to upgrade high burnup by increasing concentration degree as well as to improve design on fuel assembly structural element, to further upgrade its economical efficiency. At present, on a 9 x 9 type fuel (type 3) begun on its practical use, array of fuel rods is made by nine rows and nine columns, to increase to 45 GWd/t in average discharge burnup and 55 GWd/t in highest assembly burnup. Furthermore, on future fuel, a wide high burnup over limitation on improved 9 x 9 type and fuel cycle is investigated, to promote developments on improved fuel pellet and new alloys for structural materials. Here were introduced design and production based on upgradings of reliability and economical efficiency on recent commercial LWRs, and trends on their R and D at every fields. (G.K.)

  3. Carbide and nitride fuels for advanced burner reactor

    International Nuclear Information System (INIS)

    Full text: Under the U.S. fast reactor program, reference and alternative 1000 MWth Advanced Burner Reactor (ABR) core concepts were developed using ternary metallic (U-TRU-Zr) and mixed oxide (UO2+TRUO2) fuels. Recently, mixed carbide and nitride fuels have been considered as fast reactor fuels on the basis of their high density, compatibility with coolant, high melting temperature, and excellent thermal conductivity although they are ceramic fuel like a mixed oxide fuel. Thus, the performance of the ABR core loaded with carbide and nitride fuels was evaluated in this study with an expectation that the carbide and nitride fuels can mitigate disadvantages of both metallic and oxide fuels in the ABR: favorable passive safety features in a severe accident compared to the oxide core, a higher discharge burnup compared to the metallic core, and a potential to increase thermal efficiency. All calculations performed in this study were focused on the neutronics characteristics, although the fabrication and irradiation experiences for carbide and nitride fuels are limited and some problems were observed in the reprocessing and irradiation of these fuels. The mixed monocarbide and mixed mononitride fuels were selected as the alternative fuel forms and the ABR core concepts with these fuels were developed based on the reference 1000 MWth ABR core concepts. For consistency, the potential design goals used in the reference ABR core concepts were also employed in this study: a 1000 MWth power rating, medium TRU conversion ratio of ∼0.75, a compact core, one-year operational cycle length at least with a capacity factor of 90%, sufficient shutdown margin with a limited maximum single control assembly fault, and possible use of either metallic or any ceramic fuels in the same core layout. The core layout and outer assembly dimensions of the reference 1000 MWth ABR core were kept, but the intra assembly design parameters were varied to maximize the discharge burnup within the

  4. RU fuel development program for an advanced fuel cycle in Korea

    International Nuclear Information System (INIS)

    Korea is a unique country, having both PWR and CANDU reactors. Korea can therefore exploit the natural synergism between the two reactor types to minimize overall waste production, and maximize energy derived from the fuel, by ultimately burning the spent fuel from its PWR reactors in CANDU reactors. As one of the possible fuel cycles, Recovered Uranium (RU) fuel offers a very attractive alternative to the use of Natural Uranium (NU) and slightly enriched uranium (SEU) in CANDU reactors. Potential benefits can be derived from a number of stages in the fuel cycle: no enrichment required, therefore no enrichment tails, direct conversion to UO2, lower sensitivity to 234U and 236U absorption in the CANDU reactor, and expected lower cost relative to NU and SEU. These benefits all fit well with the PWR-CANDU fuel cycle synergy. RU arising from the conventional reprocessing of European and Japanese oxide spent fuel by 2000 is projected to be approaching 25,000 te. The use of RU fuel in a CANDU 6 reactor should result in no serious radiological difficulties and no requirements for special precautions and should not require any new technologies for the fuel fabrication and handling. The use of the CANDU Flexible Fueling (CANFLEX) bundle as the carrier for RU will be fully compatible with the reactor design, current safety and operational requirements, and there will be improved fuel performance compared with the CANDU 37-element NU fuel bundle. Compared with the 37-element NU bundle, the RU fuel has significantly improved fuel cycle economics derived from increased burnups, a large reduction in both fuel requirements and spent fuel, arisings, and the potential lower cost for RU material. There is the potential for annual fuel cost savings in the range of one-third to two-thirds, with enhanced operating margins using RU in the CANFLEX bundle design. These benefits provide the rationale for justifying R and D efforts on the use of RU fuel for advanced fuel cycles in CANDU

  5. Assessment of Research Needs for Advanced Fuel Cells

    Energy Technology Data Exchange (ETDEWEB)

    Penner, S.S.

    1985-11-01

    The DOE Advanced Fuel Cell Working Group (AFCWG) was formed and asked to perform a scientific evaluation of the current status of fuel cells, with emphasis on identification of long-range research that may have a significant impact on the practical utilization of fuel cells in a variety of applications. The AFCWG held six meetings at locations throughout the country where fuel cell research and development are in progress, for presentations by experts on the status of fuel cell research and development efforts, as well as for inputs on research needs. Subsequent discussions by the AFCWG have resulted in the identification of priority research areas that should be explored over the long term in order to advance the design and performance of fuel cells of all types. Surveys describing the salient features of individual fuel cell types are presented in Chapters 2 to 6 and include elaborations of long-term research needs relating to the expeditious introduction of improved fuel cells. The Introduction and the Summary (Chapter 1) were prepared by AFCWG. They were repeatedly revised in response to comments and criticism. The present version represents the closest approach to a consensus that we were able to reach, which should not be interpreted to mean that each member of AFCWG endorses every statement and every unexpressed deletion. The Introduction and Summary always represent a majority view and, occasionally, a unanimous judgment. Chapters 2 to 6 provide background information and carry the names of identified authors. The identified authors of Chapters 2 to 6, rather than AFCWG as a whole, bear full responsibility for the scientific and technical contents of these chapters.

  6. Toxicity of irradiated advanced heavy water reactor fuels.

    Science.gov (United States)

    Priest, N D; Richardson, R B; Edwards, G W R

    2013-02-01

    The good neutron economy and online refueling capability of the CANDU® heavy water moderated reactor (HWR) enable it to use many different fuels such as low enriched uranium (LEU), plutonium, or thorium, in addition to its traditional natural uranium (NU) fuel. The toxicity and radiological protection methods for these proposed fuels, unlike those for NU, are not well established. This study uses software to compare the fuel composition and toxicity of irradiated NU fuel against those of two irradiated advanced HWR fuel bundles as a function of post-irradiation time. The first bundle investigated is a CANFLEX® low void reactor fuel (LVRF), of which only the dysprosium-poisoned central element, and not the outer 42 LEU elements, is specifically analyzed. The second bundle investigated is a heterogeneous high-burnup (LEU,Th)O(2) fuelled bundle, whose two components (LEU in the outer 35 elements and thorium in the central eight elements) are analyzed separately. The LVRF central element was estimated to have a much lower toxicity than that of NU at all times after shutdown. Both the high burnup LEU and the thorium fuel had similar toxicity to NU at shutdown, but due to the creation of such inhalation hazards as (238)Pu, (240)Pu, (242)Am, (242)Cm, and (244)Cm (in high burnup LEU), and (232)U and (228)Th (in irradiated thorium), the toxicity of these fuels was almost double that of irradiated NU after 2,700 d of cooling. New urine bioassay methods for higher actinoids and the analysis of thorium in fecal samples are recommended to assess the internal dose from these two fuels. PMID:23274823

  7. Development of the advanced CANDU technology -Development of CANDU advanced fuel fabrication technology-

    International Nuclear Information System (INIS)

    This project is carrying out jointly with AECL to develop CANFLEX fuel which can enhance reactor safety, fuel economy and can be used with various fuel cycles (natural U, slightly enriched U, other advanced fuel). The final goal of this research is to load the CANFLEX fuel in commercial CANDU reactor for demonstration irradiation. The annual portion of research activities performed during this year are followings ; The detail design of CANFLEX-NU fuel was determined. Based on this design, various fabrication drawings and process specifications were revised. The seventeen CANFLEX-NU fuel bundles for reactivity test in ZED-2 and out-pile test, two CANFLEX-SEU fuel bundles for demo-irradiation in NRU were fabricated. Advanced tack welding machine was designed and sequence control software of automatic assembly welder was developed. The basic researches related to fabrication processes, such as weld evaluation by ECT, effect of additives in UO2, thermal stabilities of Zr based metallic glasses, were curried out. 51 figs, 22 tabs, 42 refs. (Author)

  8. Development of the advanced CANDU technology -Development of CANDU advanced fuel fabrication technology-

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Chang Bum; Park, Choon Hoh; Park, Chul Joo; Kwon, Woo Joo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    This project is carrying out jointly with AECL to develop CANFLEX fuel which can enhance reactor safety, fuel economy and can be used with various fuel cycles (natural U, slightly enriched U, other advanced fuel). The final goal of this research is to load the CANFLEX fuel in commercial CANDU reactor for demonstration irradiation. The annual portion of research activities performed during this year are followings ; The detail design of CANFLEX-NU fuel was determined. Based on this design, various fabrication drawings and process specifications were revised. The seventeen CANFLEX-NU fuel bundles for reactivity test in ZED-2 and out-pile test, two CANFLEX-SEU fuel bundles for demo-irradiation in NRU were fabricated. Advanced tack welding machine was designed and sequence control software of automatic assembly welder was developed. The basic researches related to fabrication processes, such as weld evaluation by ECT, effect of additives in UO{sub 2}, thermal stabilities of Zr based metallic glasses, were curried out. 51 figs, 22 tabs, 42 refs. (Author).

  9. Carbide and Nitride Fuels for Advanced Burner Reactor

    International Nuclear Information System (INIS)

    The impacts of the mixed carbide and nitride fuels on the core performances and passive safety features of TRU burner were assessed and comapred with the metallic and oxide fuels. Targeting the potential design goals adopted in the Advanced Burner Reactor core concepts, the alternative TRU burner concepts were developed by loading carbide and nitride fuels. The neutron spectrum is softer than that of the metal core, but harder than that of the oxide core, and the core performance parameters such as fuel residence time, discharge burnup, flux level, etc are generally between the values of the metal and oxide cores. The margin to fuel melt was significantly increased because of the high thermal conductivity and high melting temperature, and hence there is an additional room to improve the thermal efficiency by increasing the operating temperature. The changed fuel composition affected the kinetics parameters and reactivity feedback coefficients, but the variations were minimal. The reduced core height decreases the sodium void worth, and the high thermal conductivity decreases the fuel temperature and Doppler constant. As a result, both carbide and nitride cores have favorable passive safety features without additional design fixes that are required in the oxide core concepts. (author)

  10. Development of Advanced Voloxidation Process for Treatment of Spent Fuel

    International Nuclear Information System (INIS)

    Data for evaluation of the effects of advanced voloxidation on pyroprocessing of spent oxide fuel with a determination for a path forward such was produced as follows: effect of particle size and particle structure on oxide reduction, assessment of decladding options for pyroprocessing, effect of removal timing of fission products, analysis of radioactivity and decay heat of advanced voloxidation process, proliferation resistance of advanced voloxidation process, Effect of advanced voloxidation process on shielding. Also, performance objectives for advanced voloxidation with respective to the down stream effects was established. The technology on design and manufacture of voloxidation and off gas treatment equipment was established. The possibility of fabrication of porous granule as a feed material for electro-reduction process was confirmed using rotary voloxidizer and SIMFUEL. The operational conditions for advanced voloxidation process consisting of 4 steps heat treatment was drawn to vaporize fission products and fabricate UO2 granule. The trapping test of Cs and Re(surrogate material of Tc) using newly developed filter were selectively separated at trapping efficiency of 99%, respectively. Data for oxidative decladding, vaporization rate of fission products, and particle size from experiment on voloxidation using spent fuel in ILN hot cell was acquisited including data of off gas trapping characteristics and verification of excellent performance of filter

  11. An analysis on the breeding capability and safety related parameters of advanced fast reactor fuels using recent cross-section set

    International Nuclear Information System (INIS)

    Highlights: • Breeding ratio of fast reactor fuels is computed with latest cross-section set. • Safety related parameters are also evaluated. • It is found that there are better prospects of utilization of thorium resources. • With large fast reactors, Th–233U fuel combination gives better B.G. -- Abstract: This study focuses on the evaluation of breeding capability as well as safety related neutronic parameters of advanced fast reactor fuels which comprises of fissile–fertile combination of metal, oxide, carbide and nitride, using the recent neutron cross-section set ENDF/B-VI.7. Sodium cooled fast breeder reactor similar to prototype Fast Breeder Reactor (PFBR) is used to evaluate the performance of various fuel types involving fissile isotopes of 233U and Pu and fertile isotopes of Th and 238U. The analysis is restricted to a comparison of neutronic parameters of a fresh core and does not take into account the effects of burnup and fission products. The breeding potential of the fuels are also compared with European cross-section set JEFF-3.1. The breeding ratio of advanced fuels evaluated with ENDF/B-VI.7 and JEFF-3.1 was found to be in good agreement. From this study, it is found that Th–233U combination for almost all fuel types with the present geometry and composition gives a lower breeding ratio value. Safety neutronic parameters such as effective delayed neutron fraction, Doppler defect and sodium void reactivity were also computed. In terms of breeding potential and safety neutronic parameters, the performance of Th–Pu system especially the metal fuel type can be a better option for future large fast reactors. The large negative Doppler feedback along with a negative sodium void reactivity for metal and hybrid combinations of Th–233U system makes it an attractive fuel cycle option even though there is a penalty over its breeding capability

  12. CANFLEX-RU fuel development programs as one option of advanced fuel cycles in Korea

    International Nuclear Information System (INIS)

    As one of the possible fuel cycles in Korea, RU (Recycled Uranium) fuel offers a very attractive alternative to the use of NU (Natural Uranium) and SEU in the CANDU reactors, because Korea is a unique country having both PWR and CANDU reactors. Korea can therefore exploit the natural synergism between the two reactor types to minimise overall waste production, and maximise energy derived from the fuel, by burning the spent fuel from its PWR reactors in CANDU reactors. Potential benefits can be derived from a number of stages in the fuel cycle: no enrichment required, no enrichment tails, direct conversion to UO2 lower sensitivity to 234U and 236U absorption in the CANDU reactor, expected lower cost relative to NU and SEU. These benefits all fit well with the PWR-CANDU fuel cycle synergy. RU arising from the reprocessing of European and Japanese oxide spent fuel by 2000 is projected to be approaching 25,000 te. The use of RU fuel in a CANDU-6 reactor should result in no serious radiological difficulties and no requirements for special precautions and should not require any new technologies for the fuel fabrication and handling. A KAERI's feasibility shows that the use of the CANFLEX bundle as the carrier for RU will be compatible with the reactor design, current safety and operational requirements, and there will be no significant fuel performance difference from the CANDU 37-element NU fuel bundle. Compared with the 37-element NU bundle, the RU fuel has significantly improved fuel cycle economics derived from increased burnups, a large reduction in fuel requirements and spent fuel arisings and the potential lower cost for RU material. There is the potential for annual fuel cost savings to be in the range of one-third to two-thirds, with enhanced operating margins using RU in the CANFLEX bundle design. These benefits provide the rationale for justifying R and D effort on the use of RU fuel for advanced fuel cycles in the CANDU reactors of Korea. The RU fuel

  13. Cycle update : advanced fuels and technologies for emissions reduction

    Energy Technology Data Exchange (ETDEWEB)

    Smallwood, G. [National Research Council of Canada, Ottawa, ON (Canada)

    2009-07-01

    This paper provided a summary of key achievements of the Program of Energy Research and Development advanced fuels and technologies for emissions reduction (AFTER) program over the funding cycle from fiscal year 2005/2006 to 2008/2009. The purpose of the paper was to inform interested parties of recent advances in knowledge and in science and technology capacities in a concise manner. The paper discussed the high level research and development themes of the AFTER program through the following 4 overarching questions: how could advanced fuels and internal combustion engine designs influence emissions; how could emissions be reduced through the use of engine hardware including aftertreatment devices; how do real-world duty cycles and advanced technology vehicles operating on Canadian fuels compare with existing technologies, models and estimates; and what are the health risks associated with transportation-related emissions. It was concluded that the main issues regarding the use of biodiesel blends in current technology diesel engines are the lack of consistency in product quality; shorter shelf life of biodiesel due to poorer oxidative stability; and a need to develop characterization methods for the final oxygenated product because most standard methods are developed for hydrocarbons and are therefore inadequate. 2 tabs., 13 figs.

  14. Fuel performance analysis of the Korea advanced nuclear fuel using ESCORE

    International Nuclear Information System (INIS)

    The Lead Use Assembly (LUA) of the Korea advanced nuclear fuel will be loaded in YGN4 cycle 7. The core cycle length is 16,248 MWD/MTU and 57,996 MWD/MTU is the maximum pin peak burnup. Fuel thermal and mechanical performance (i.e. maximum temperature, rod internal pressure, and cladding corrosion) evaluation is performed with ESCORE code which is developed by EPRI

  15. Design study and evaluation of advanced fuel fabrication systems for FBR fuel cycle

    International Nuclear Information System (INIS)

    The conceptual design study for advanced FBR fuel fabrication system has been performed for the purpose that the feature of small-scale fabrication system in the transition stage from LWR to FBR fuel cycle. On the small-scale of 50 ton heavy metal per year fabrication system, dry type fabrication systems have superior cost performance than the wet type, although waste amount is larger. (authors)

  16. Impact of advanced fuel cycle options on waste management policies

    International Nuclear Information System (INIS)

    OECD/NEA has performed a study on the impact of advanced fuel cycle options on waste management policies with 33 experts from 12 member countries, 1 non-member country and 2 international organizations. The study extends a series of previous ones on partitioning and transmutation (P and T) issues, focusing on the performance assessments for repositories of high-level waste (HLW) arising from advanced fuel cycles. This study covers a broader spectrum than previous studies, from present industrial practice to fully closed cycles via partially closed cycles (in terms of transuranic elements); 9 fuel cycle schemes and 4 variants. Elements of fuel cycles are considered primarily as sources of waste, the internal mass flows of each scheme being kept for the sake of mass conservation. The compositions, activities and heat loads of all waste flows are also tracked. Their impact is finally assessed on the waste repository concepts. The study result confirms the findings from the previous NEA studies on P and T on maximal reduction of the waste source term and maximal use of uranium resources. In advanced fuel cycle schemes the activity of the waste is reduced by burning first plutonium and then minor actinides and also the uranium consumption is reduced, as the fraction of fast reactors in the park is increased to 100%. The result of the repository performance assessments, analysing the effect of different HLW isotopic composition on repository performance and on repository capacity, shows that the maximum dose released to biosphere at any time in normal conditions remains, for all schemes and for all the repository concepts examined, well below accepted radiation protection thresholds. The major impact is on the detailed concept of the repositories, through heat load and waste volume. Advanced fuel cycles could allow a repository to cover waste produced from 5 to 20 times more electricity generation than PWR once-through cycle. Given the flexibility of the advanced fuel

  17. Advanced fuels for plutonium management in pressurized water reactors

    International Nuclear Information System (INIS)

    Several fuel concepts are under investigation at CEA with the aim of manage plutonium inventories in pressurized water reactors. This options range from the use of mature technologies like MOX adapted in the case of MOX-EUS (enriched uranium support) and COmbustible Recyclage A ILot (CORAIL) assemblies to more innovative technologies using IMF like DUPLEX and advanced plutonium assembly (APA). The plutonium burning performances reported to the electrical production go from 7 to 60 kg (TW h)-1. More detailed analysis covering economic, sustainability, reliability and safety aspects and their integration in the whole fuel cycle would allow identifying the best candidate

  18. Introducing advanced thorium-based fuel cycles in SA : an evolutionary approach / Maria Hendrina (Marina) du Toit

    OpenAIRE

    Du Toit, Maria Hendrina

    2013-01-01

    Past experience in several thorium fuelled research- and power reactors provides the basis and history of thorium. The material properties, fertile- and fissile isotope properties as well as the decay chain of thorium are discussed for purposes of evaluating thorium as a source of fuel. The different thorium-based fuel designs for PWR cores are discussed and resulting difficulties and solutions are outlined. The different options for each strategy are compared in terms of ad...

  19. Advanced CANDU reactors fuel analysis through optimal fuel management at approach to refuelling equilibrium

    International Nuclear Information System (INIS)

    The analysis of alternate CANDU fuels along with natural uranium-based fuel was carried out from the view point of optimal in-core fuel management at approach to refuelling equilibrium. The alternate fuels considered in the present work include thorium containing oxide mixtures (MOX), plutonium-based MOX, and Pressurised Water Reactor (PWR) spent fuel recycled in CANDU reactors (Direct Use of spent PWR fuel in CANDU (DUPIC)); these are compared with the usual natural UO2 fuel. The focus of the study is on the 'Approach to Refuelling Equilibrium' period which immediately follows the initial commissioning of the reactor. The in-core fuel management problem for this period is treated as an optimization problem in which the objective function is the refuelling frequency to be minimized by adjusting the following decision variables: the channel to be refuelled next, the time of the refuelling and the number of fresh fuel bundles to be inserted in the channel. Several constraints are also included in the optimisation problem which is solved using Perturbation Theory. Both the present 37-rod CANDU fuel bundle and the proposed CANFLEX bundle designs are part of this study. The results include the time to reach refuelling equilibrium from initial start-up of the reactor, the average discharge burnup, the average refuelling frequency and the average channel and bundle powers relative to natural UO2. The model was initially tested and the average discharge burnup for natural UO2 came within 2% of the industry accepted 199 MWh/kgHE. For this type of fuel, the optimization exercise predicted the savings of 43 bundles per full power year. In addition to producing average discharge burnups and other parameters for the advanced fuels investigated, the optimisation model also evidenced some problem areas like high power densities for fuels such as the DUPIC. Perturbation Theory has proven itself to be an accurate and valuable optimization tool in predicting the time between

  20. EDF advanced fuel management strategies for the next century

    International Nuclear Information System (INIS)

    The French nuclear fleet represents 57 PWRs in operation, accounting for 80 % of France's total electricity production. The performance achieved by EDF reactors, in terms of availability (82.6% in 1997) and good cost control, have allowed to improve the nuclear KWh cost by 2% since 1992. The implementation of longer fuel cycles on the 1300 MW reactors from 1996 has contributed to this improvement and, as competitiveness is one of the main challenges for EDF, improving core management strategies is still at the order of the day. With this aim, a thinking process has been initiated to evaluate the benefit brought by the use of a fuel assembly like ALLIANCE, the new fuel product developed by Framatome-Fragema and FCF (Framatome Cogema Fuels) in close cooperation with EDF. The considered product provides enhanced performance, particularly as regards discharge burnup (at least up to 70 GWd/t) and thermal-hydraulic and mechanical behaviour. Fuel management improvements rely on the expertise gained by Framatome through designing core management strategies in a wide range of operating conditions prevailing in nuclear reactors all over the world. It will however be taken into account the necessity for EDF to adopt a policy of stepwise change owing to the potential impact of a 'series effect' on its numerous units. The proposed paper will describe innovative fuel managements, achievable thanks to advanced fuel assembly performance, that are jointly investigated by EDF and Framatome. It includes the following optimization schemes: extending cycle length by using higher enrichments up to 5%, while keeping the same reload size (1/3 core for example for the 1300 MW reactors); decreasing reload size (from 1/3 to 1/4 core), while keeping the same cycle length, using more enriched (up to 5 %) fuel assemblies; reaching annual cycle, with maximization of fuel cycle cost optimization (1/5 core). Beyond such schemes, combinations of optimized loading patterns and neutronic features of

  1. Development of the advanced PHWR technology -Verification tests for CANDU advanced fuel-

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Jang Hwan; Suk, Hoh Chun; Jung, Moon Kee; Oh, Duk Joo; Park, Joo Hwan; Shim, Kee Sub; Jang, Suk Kyoo; Jung, Heung Joon; Park, Jin Suk; Jung, Seung Hoh; Jun, Ji Soo; Lee, Yung Wook; Jung, Chang Joon; Byun, Taek Sang; Park, Kwang Suk; Kim, Bok Deuk; Min, Kyung Hoh [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    This is the `94 annual report of the CANDU advanced fuel verification test project. This report describes the out-of pile hydraulic tests at CANDU-hot test loop for verification of CANFLEX fuel bundle. It is also describes the reactor thermal-hydraulic analysis for thermal margin and flow stability. The contents in this report are as follows; (1) Out-of pile hydraulic tests for verification of CANFLEX fuel bundle. (a) Pressure drop tests at reactor operation condition (b) Strength test during reload at static condition (c) Impact test during reload at impact load condition (d) Endurance test for verification of fuel integrity during life time (2) Reactor thermal-hydraulic analysis with CANFLEX fuel bundle. (a) Critical channel power sensitivity analysis (b) CANDU-6 channel flow analysis (c) Flow instability analysis. 61 figs, 29 tabs, 21 refs. (Author).

  2. Nuclear materials - fissile, fertile and dual-use structural materials involved in nuclear reactors

    International Nuclear Information System (INIS)

    The article presents a brief account of nuclear materials, with special emphasis on fissile, fertile and some important dual-use structural materials generally involved in nuclear reactors. The dual-use structural materials utilized in nuclear reactors have got important applications in both nuclear and non-nuclear fields. In the hostile environment, the important phenomena such as interactions between fission products and the surrounding elemental species, radiation-induced effects, corrosion, generation of gases, swelling and so forth, become increasingly complex and the performance of a nuclear reactor system thus becomes very much dependent on the physicochemical stability and nuclear compatibility of the dual-use structural materials used in fuel sub-assembly towards the fuel elements. In this context, the dual-use structural materials like stainless steel, zirconium alloys, etc. as claddings; water, liquid sodium or gases, etc. as coolants and water, boron, etc. as moderators, having good reliability and appropriate nuclear compatibility with the fuels, are of prime importance in reactor technology. In advanced designed reactors, development of novel fuels coupled with efficient dual-use structural materials may mitigate the challenges involved in optimizing the efficiency of the power reactors under specific experimental conditions. (author)

  3. Construction and engineering report for advanced nuclear fuel development facility

    International Nuclear Information System (INIS)

    The design and construction of the fuel technology development facility was aimed to accommodate general nuclear fuel research and development for the HANARO fuel fabrication and advanced fuel researches. 1. Building size and room function 1) Building total area : approx. 3,618m2, basement 1st floor, ground 3th floor 2) Room function : basement floor(machine room, electrical room, radioactive waste tank room), 1st floor(research reactor fuel fabrication facility, pyroprocess lab., metal fuel lab., nondestructive lab., pellet processing lab., access control room, sintering lab., etc), 2nd floor(thermal properties measurement lab., pellet characterization lab., powder analysis lab., microstructure analysis lab., etc), 3rd floor(AHU and ACU Room) 2. Special facility equipment 1) Environmental pollution protection equipment : ACU(2sets), 2) Emergency operating system : diesel generator(1set), 3) Nuclear material handle, storage and transport system : overhead crane(3sets), monorail hoist(1set), jib crane(2sets), tank(1set) 4) Air conditioning unit facility : AHU(3sets), packaged air conditioning unit(5sets), 5) Automatic control system and fire protection system : central control equipment(1set), lon device(1set), fire hose cabinet(3sets), fire pump(3sets) etc

  4. Construction and engineering report for advanced nuclear fuel development facility

    Energy Technology Data Exchange (ETDEWEB)

    Cho, S. W.; Park, J. S.; Kwon, S.J.; Lee, K. W.; Kim, I. J.; Yu, C. H

    2003-09-01

    The design and construction of the fuel technology development facility was aimed to accommodate general nuclear fuel research and development for the HANARO fuel fabrication and advanced fuel researches. 1. Building size and room function 1) Building total area : approx. 3,618m{sup 2}, basement 1st floor, ground 3th floor 2) Room function : basement floor(machine room, electrical room, radioactive waste tank room), 1st floor(research reactor fuel fabrication facility, pyroprocess lab., metal fuel lab., nondestructive lab., pellet processing lab., access control room, sintering lab., etc), 2nd floor(thermal properties measurement lab., pellet characterization lab., powder analysis lab., microstructure analysis lab., etc), 3rd floor(AHU and ACU Room) 2. Special facility equipment 1) Environmental pollution protection equipment : ACU(2sets), 2) Emergency operating system : diesel generator(1set), 3) Nuclear material handle, storage and transport system : overhead crane(3sets), monorail hoist(1set), jib crane(2sets), tank(1set) 4) Air conditioning unit facility : AHU(3sets), packaged air conditioning unit(5sets), 5) Automatic control system and fire protection system : central control equipment(1set), lon device(1set), fire hose cabinet(3sets), fire pump(3sets) etc.

  5. Methods for studying fuel management in advanced gas cooled reactors

    International Nuclear Information System (INIS)

    The methods used for studying fuel and absorber management problems in AGRs are described. The basis of the method is the use of ARGOSY lattice data in reactor calculations performed at successive time steps. These reactor calculations may be quite crude but for advanced design calculations a detailed channel-by-channel representation of the whole core is required. The main emphasis of the paper is in describing such an advanced approach - the ODYSSEUS-6 code. This code evaluates reactor power distributions as a function of time and uses the information to select refuelling moves and determine controller positions. (author)

  6. The US advanced fuel cycle program: Objectives and accomplishments

    International Nuclear Information System (INIS)

    Full text: The US Department of Energy has been running for approximately a decade an advanced fuel cycle program (currently named the Fuel Cycle R and D Program) devoted to lessening both the environmental burden of nuclear energy and the proliferation risk of accumulating Used Nuclear Fuel. Until very recently, this program was technically focused on achieving an optimized symbiosis between fuel cycle options on one hand, and the US geological repository on the other, with a relatively short term deployment focus. Following detailed technical analyses, this focus led to the selection of a limited set of technologies that were expected to both meet specific geology related criteria, and would be based on limited extrapolations of existing technologies. Recent developments in the US indicate that the Yucca Mountain repository might not be anymore the geology of reference; furthermore, the need for advanced fuel cycles has been postponed to the middle part of the century, with increased reliance on temporary storage of Used Nuclear Fuel in the interim. Consequently, the Fuel Cycle R and D Program is being redirected towards a science based, goal oriented focus, driven by the following three considerations: 1. the program is currently examining a broad set of options, including different geologic media and transmutation technologies, in order to understand their relationships and provide information for later decisions. 2. the R and D component of the program is focused on acquiring the basic understanding of key phenomena, defining the relevant challenges, and acquiring the basic tools necessary to resolve them. 3. the timeline of the program allows for a deployment of the successfully demonstrated technology in the 2040-2050 timeframe; this allows us to consider technologies that are not yet mature, but that might provide significant improvements in performances. The technical program is articulated along the following elements: - a systems integration task that

  7. The economics of advanced fuel cycles in CANDU (PHW) reactors

    International Nuclear Information System (INIS)

    The economic assessments of advanced fuel cycles performed within Ontario Hydro are collated and summarized. The results of the analyses are presented in a manner designed to provide a broad perspective of the economic issues regarding the advanced cycles. The enriched uranium fuel cycle is shown to be close to competitive at today's uranium prices, and its relative position vis-a-vis the natural uranium cycle will improve as uranium prices continue to rise. In the longer term, the plutonium-topped thorium cycle is identified as being the most economically desirable. It is suggested that this cycle may not be commercially attractive until the second or third decade of the next century. (auth)

  8. Advanced materials for alternative fuel capable directly fired heat engines

    Energy Technology Data Exchange (ETDEWEB)

    Fairbanks, J.W.; Stringer, J. (eds.)

    1979-12-01

    The first conference on advanced materials for alternative fuel capable directly fired heat engines was held at the Maine Maritime Academy, Castine, Maine. It was sponsored by the US Department of Energy, (Assistant Secretary for Fossil Energy) and the Electric Power Research Institute, (Division of Fossil Fuel and Advanced Systems). Forty-four papers from the proceedings have been entered into EDB and ERA and one also into EAPA; three had been entered previously from other sources. The papers are concerned with US DOE research programs in this area, coal gasification, coal liquefaction, gas turbines, fluidized-bed combustion and the materials used in these processes or equipments. The materials papers involve alloys, ceramics, coatings, cladding, etc., and the fabrication and materials listing of such materials and studies involving corrosion, erosion, deposition, etc. (LTN)

  9. Advanced modeling of oxy-fuel combustion of natural gas

    Energy Technology Data Exchange (ETDEWEB)

    Chungen Yin

    2011-01-15

    The main goal of this small-scale project is to investigate oxy-combustion of natural gas (NG) through advanced modeling, in which radiation, chemistry and mixing will be reasonably resolved. 1) A state-of-the-art review was given regarding the latest R and D achievements and status of oxy-fuel technology. The modeling and simulation status and achievements in the field of oxy-fuel combustion were also summarized; 2) A computer code in standard c++, using the exponential wide band model (EWBM) to evaluate the emissivity and absorptivity of any gas mixture at any condition, was developed and validated in detail against data in literature. A new, complete, and accurate WSGGM, applicable to both air-fuel and oxy-fuel combustion modeling and applicable to both gray and non-gray calculation, was successfully derived, by using the validated EWBM code as the reference mode. The new WSGGM was implemented in CFD modeling of two different oxy-fuel furnaces, through which its great, unique advantages over the currently most widely used WSGGM were demonstrated. 3) Chemical equilibrium calculations were performed for oxy-NG flame and air-NG flame, in which dissociation effects were considered to different degrees. Remarkable differences in oxy-fuel and air-fuel combustion were revealed, and main intermediate species that play key roles in oxy-fuel flames were identified. Different combustion mechanisms are compared, e.g., the most widely used 2-step global mechanism, refined 4-step global mechanism, a global mechanism developed for oxy-fuel using detailed chemical kinetic modeling (CHEMKIN) as reference. 4) Over 15 CFD simulations were done for oxy-NG combustion, in which radiation, chemistry, mixing, turbulence-chemistry interactions, and so on were thoroughly investigated. Among all the simulations, RANS combined with 2-step and refined 4-step mechanism, RANS combined with CHEMKIN-based new global mechanism for oxy-fuel modeling, and LES combined with different combustion

  10. A 252Cf based nondestructive assay system for fissile material

    International Nuclear Information System (INIS)

    A modulated 252Cf source assay system 'Shuffler' based on fast-or-thermal-neutron interrogation combined with delayed-neutron counting has been developed for the assay of fissile material. The 252Cf neutron source is repetitively transferred from the interrogation position to a shielded position while the delayed neutrons are counted in a high efficiency 3He neutron well-counter. For samples containing plutonium, this well-counter is also used in the passive coincidence mode to assay the effective 240Pu content. The design of an optimized neutron tailoring assembly for fast-neutron interrogation using a Monte Carlo Neutron Computer Code is described. The Shuffler system has been applied to the assay of fuel pellets, inventory samples, irradiated fuel and plutonium mixed-oxide fuel. The system can assay samples with fissile contents from a few milligrams up to several kilograms using thermal-neutron interrogation for the low mass samples and fast-neutron interrogation for the high mass samples. Samples containing 235U-238U, or 233U-Th, or UO2-PuO2 fuel mixtures have been assayed with the Shuffler system. (Auth.)

  11. Disposition scenarios and safeguardability of fissile materials under START Treaty

    International Nuclear Information System (INIS)

    Under the Strategic Arms Reduction Treaty (START-I) signed in 1991 and the Lisbon Protocol of 1992, a large inventory of fissile materials will be removed from the weapons fuel cycles of the United States and the Former Soviet Union (FSU). The Lisbon Protocol calls for Ukraine, Kazakstan, and Byelarus to become nonnuclear members of the treaty and for Russia to assume the responsibility of the treaty as a nuclear weapons state. In addition, the START-II Treaty, which was signed in 1993 by the United States and Russia, further reduces deployed nuclear warheads and adds to the inventory of excess special nuclear materials (SNM). Because storage of in-tact warheads has the potential for a open-quotes breakout,close quotes it would be desirable to dismantle the warheads and properly dispose of the SNMs under appropriate safeguards to prevent their reentry into the weapons fuel cycle. The SNM recovered from dismantled warheads can be disposed of in several ways, and the final choices may be up to the country having the title to the SNM. Current plans are to store them indefinitely, leaving serious safeguards concerns. Recognizing that the underlying objective of these treaties is to prevent the fissile materials from reentering the weapons fuel cycle, it is necessary to establish a verifiable disposal scheme that includes safeguards requirements. This paper identifies some realistic scenarios for the disposal of SNM from the weapons fuel cycle and examines the safeguardability of those scenarios

  12. Advanced coal-fueled industrial cogeneration gas turbine system

    Energy Technology Data Exchange (ETDEWEB)

    LeCren, R.T.; Cowell, L.H.; Galica, M.A.; Stephenson, M.D.; Wen, C.S.

    1991-07-01

    Advances in coal-fueled gas turbine technology over the past few years, together with recent DOE-METC sponsored studies, have served to provide new optimism that the problems demonstrated in the past can be economically resolved and that the coal-fueled gas turbine can ultimately be the preferred system in appropriate market application sectors. The objective of the Solar/METC program is to prove the technical, economic, and environmental feasibility of a coal-fired gas turbine for cogeneration applications through tests of a Centaur Type H engine system operated on coal fuel throughout the engine design operating range. The five-year program consists of three phases, namely: (1) system description; (2) component development; (3) prototype system verification. A successful conclusion to the program will initiate a continuation of the commercialization plan through extended field demonstration runs.

  13. The US Advanced Fuel Cycle Programme: Objectives and Accomplishments

    International Nuclear Information System (INIS)

    For approximately a decade, the United States Department of Energy has been conducting an advanced fuel cycle programme, presently named the Fuel Cycle R and D Program, devoted to lessening both the environmental burden of nuclear energy and the proliferation risk of accumulating used nuclear fuel. Currently, the programme is being redirected towards a science based, goal oriented focus with the objective of deploying successfully demonstrated technology in the 2040-2050 time frame. The present paper reports the key considerations of the science based research approach, the elements of the technical programme and the accomplishments in fast reactor research and development, the goal of which is to improve the primary issues that have inhibited fast reactor introduction in the past, namely, economics and safety. (author)

  14. Conjugate heat transfer simulations of advanced research reactor fuel

    International Nuclear Information System (INIS)

    Highlights: • Temperature predictions are enhanced by coupling heat transfer in solid and fluid zones. • Seven different cases are considered to observe trends in predicted temperature and pressure. • The seven cases consider high/medium/low power, flow, burnup, fuel material and geometry. • Simulations provide temperature predictions for performance/safety. Boiling is unlikely. • Simulations demonstrate that a candidate geometry can enhance performance/safety. - Abstract: The current work presents numerical simulations of coupled fluid flow and heat transfer of advanced U–Mo/Al and U–Mo/Mg research reactor fuels in support of performance and safety analyses. The objective of this study is to enhance predictions of the flow regime and fuel temperatures through high fidelity simulations that better capture various heat transfer pathways and with a more realistic geometric representation of the fuel assembly in comparison to previous efforts. Specifically, thermal conduction, convection and radiation mechanisms are conjugated between the solid and fluid regions. Also, a complete fuel element assembly is represented in three dimensional space, permitting fluid flow and heat transfer to be simulated across the entire domain. Seven case studies are examined that vary the coolant inlet conditions, specific power, and burnup to investigate the predicted changes in the pressure drop in the coolant and the fuel, clad and coolant temperatures. In addition, an alternate fuel geometry is considered with helical fins (replacing straight fins in the existing design) to investigate the relative changes in predicted fluid and solid temperatures. Numerical simulations predict that the clad temperature is sensitive to changes in the thermal boundary layer in the coolant, particularly in simultaneously developing flow regions, while the temperature in the fuel is anticipated to be unaffected. Finally, heat transfer between fluid and solid regions is enhanced with

  15. Economic and system aspects of CANDU advanced fuel cycles

    International Nuclear Information System (INIS)

    It is somewhat a paradox that Canada, which ranks as one of the world's leading uranium producers and has large economic uranium resources, should also have developed the CANDU reactor. This reactor is the most fuel efficient of all reactor types which are commercially available at the present time. The explanation of the paradox is that the design basis of the CANDU was established three decades ago when the full extent of Canadian uranium resources was unknown, and an early transition to recycle fuelling was anticipated as being necessary to sustain a growing power generation system. Consequently, the objectives of fuel efficiency and flexibility in using a variety of uranium, plutonium and thorium fuels were established at an early stage. One result of this is the ability to use the current design of CANDU in an advanced converter role with very little change in reactor design or operating procedures. As a result, in projections of future power costs, all major uncertainty is focused on fuel cycle parameters since the capital and operating costs are well defined by current commercial experience. The paper will examine the economic and resource characteristics of CANDU in an advanced converter role, both in terms of stand-alone technology and as a partner in a CANDU-light-water-reactor and in a CANDU-fast-breeder-reactor system. The use of results to establish cost targets to guide the current research and development program will be discussed, together with considerations of deployment strategy. (author)

  16. Systematic analysis of advanced fusion fuel in inertial fusion energy

    Science.gov (United States)

    Velarde, G.; Eliezer, S.; Henis, Z.; Piera, M.; Martinez-Val, J. M.

    1997-04-01

    Aneutronic fusion reactions can be considered as the cleanest way to exploit nuclear energy. However, these reactions present in general two main drawbacks.—very high temperatures are needed to reach relevant values of their cross sections—Moderate (and even low) energy yield per reaction. This value is still lower if measured in relation to the Z number of the reacting particles. It is already known that bremsstrahlung overruns the plasma reheating by fusion born charged-particles in most of the advanced fuels. This is for instance the case for proton-boron-11 fusion in a stoichiometric plasma and is also so in lithium isotopes fusion reactions. In this paper, the use of deuterium-tritium seeding is suggested to allow to reach higher burnup fractions of advanced fuels, starting at a lower ignition temperature. Of course, neutron production increases as DT contents does. Nevertheless, the ratio of neutron production to energy generation is much lower in DT-advanced fuel mixtures than in pure DT plasmas. One of the main findings of this work is that some natural resources (as D and Li-7) can be burned-up in a catalytic regime for tritium. In this case, neither external tritium breeding nor tritium storage are needed, because the tritium inventory after the fusion burst is the same as before it. The fusion reactor can thus operate on a pure recycling of a small tritium inventory.

  17. Utilization of spent PWR fuel-advanced nuclear fuel cycle of PWR/CANDU synergism

    Institute of Scientific and Technical Information of China (English)

    HUO Xiao-Dong; XIE Zhong-Sheng

    2004-01-01

    High neutron economy, on line refueling and channel design result in the unsurpassed fuel cycle flexibility and variety for CANDU reactors. According to the Chinese national conditions that China has both PWR and CANDU reactors and the closed cycle policy of reprocessing the spent PWR fuel is adopted, one of the advanced nuclear fuel cycles of PWR/CANDU synergism using the reprocessed uranium of spent PWR fuel in CANDU reactor is proposed, which will save the uranium resource (~22.5%), increase the energy output (~41%), decrease the quantity of spent fuels to be disposed (~2/3) and lower the cost of nuclear power. Because of the inherent flexibility of nuclear fuel cycle in CANDU reactor, and the low radiation level of recycled uranium(RU), which is acceptable for CANDU reactor fuel fabrication, the transition from the natural uranium to the RU can be completed without major modification of the reactor core structure and operation mode. It can be implemented in Qinshan Phase Ⅲ CANDU reactors with little or no requirement of big investment in new design. It can be expected that the reuse of recycled uranium of spent PWR fuel in CANDU reactor is a feasible and desirable strategy in China.

  18. Advanced fuel cycle on the basis of pyroelectrochemical process for irradiated fuel reprocessing and vibropacking technology

    International Nuclear Information System (INIS)

    For advanced nuclear fuel cycle in SSC RIAR there is developed the pyroelectrochemical process to reprocess irradiated fuel and produce granulated oxide fuel UO2, PuO2 or (U,Pu)O2 from chloride melts. The basic technological stage is the extraction of oxides as a crystal product with the methods either of the electrolysis (UO2 and UO2-PuO2) or of the precipitating crystalIization (PuO2). After treating the granulated fuel is ready for direct use to manufacture vibropacking fuel pins. Electrochemical model for (U,Pu)O2 coprecipitation is described. There are new processes being developed: electroprecipitation of mixed oxides - (U,Np)O2, (U,Pu,Np)O2, (U,Am)O2 and (U,Pu,Am)O2. Pyroelectrochemical production of mixed actinide oxides is used both for reprocessing spent fuel and for producing actinide fuel. Both the efficiency of pyroelectrochemical methods application for reprocessing nuclear fuel and of vibropac technology for plutonium recovery are estimated. (author)

  19. Impact of Nuclear Energy Futures on Advanced Fuel Cycle Options

    International Nuclear Information System (INIS)

    tripling market share by 2100 from the current 8.4% to 25%, equivalent to continuing the average market growth of last 50 years for an additional 100 years. Five primary spent fuel management strategies are assessed against each of the energy futures to determine the number of geological repositories needed and how the first repository would be used. The geological repository site at Yucca Mountain, Nevada, has the physical potential to accommodate all the spent fuel that will be generated by the current fleet of domestic commercial nuclear reactors, even with license extensions. If new nuclear plants are built in the future as replacements or additions, the United States will need to adopt spent fuel treatment to extend the life of the repository. Should a significant number of new nuclear plants be built, advanced fuel recycling will be needed to fully manage the spent fuel within a single repository. The analysis also considers the timeframe for most efficient implementation of new spent fuel management strategies. The mix of unprocessed spent fuel and processed high level waste in Yucca Mountain varies with each future and strategy. Either recycling must start before there is too much unprocessed waste emplaced or unprocessed waste will have to be retrieved later with corresponding costs. For each case, the latest date to implement reprocessing without subsequent retrieval is determined

  20. Fissile material containment efforts: an overview

    International Nuclear Information System (INIS)

    Fissile materials (plutonium and highly enriched uranium (HEU)) are the fundamental ingredients of all nuclear weapons and they also happen to be the most difficult and expensive part of a nuclear warhead to produce. For a comprehensive nuclear disarmament and non-proliferation regime to be established, it becomes essential that there is a global, verified ban on the production of fissile materials for nuclear explosives

  1. Simulations of Failure via Three-Dimensional Cracking in Fuel Cladding for Advanced Nuclear Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Lu, Hongbing [Univ. of Texas, Austin, TX (United States); Bukkapatnam, Satish; Harimkar, Sandip; Singh, Raman; Bardenhagen, Scott

    2014-01-09

    Enhancing performance of fuel cladding and duct alloys is a key means of increasing fuel burnup. This project will address the failure of fuel cladding via three-dimensional cracking models. Researchers will develop a simulation code for the failure of the fuel cladding and validate the code through experiments. The objective is to develop an algorithm to determine the failure of fuel cladding in the form of three-dimensional cracking due to prolonged exposure under varying conditions of pressure, temperature, chemical environment, and irradiation. This project encompasses the following tasks: 1. Simulate 3D crack initiation and growth under instantaneous and/or fatigue loads using a new variant of the material point method (MPM); 2. Simulate debonding of the materials in the crack path using cohesive elements, considering normal and shear traction separation laws; 3. Determine the crack propagation path, considering damage of the materials incorporated in the cohesive elements to allow the energy release rate to be minimized; 4. Simulate the three-dimensional fatigue crack growth as a function of loading histories; 5. Verify the simulation code by comparing results to theoretical and numerical studies available in the literature; 6. Conduct experiments to observe the crack path and surface profile in unused fuel cladding and validate against simulation results; and 7. Expand the adaptive mesh refinement infrastructure parallel processing environment to allow adaptive mesh refinement at the 3D crack fronts and adaptive mesh merging in the wake of cracks. Fuel cladding is made of materials such as stainless steels and ferritic steels with added alloying elements, which increase stability and durability under irradiation. As fuel cladding is subjected to water, chemicals, fission gas, pressure, high temperatures, and irradiation while in service, understanding performance is essential. In the fast fuel used in advanced burner reactors, simulations of the nuclear

  2. Simulations of Failure via Three-Dimensional Cracking in Fuel Cladding for Advanced Nuclear Fuels

    International Nuclear Information System (INIS)

    Enhancing performance of fuel cladding and duct alloys is a key means of increasing fuel burnup. This project will address the failure of fuel cladding via three-dimensional cracking models. Researchers will develop a simulation code for the failure of the fuel cladding and validate the code through experiments. The objective is to develop an algorithm to determine the failure of fuel cladding in the form of three-dimensional cracking due to prolonged exposure under varying conditions of pressure, temperature, chemical environment, and irradiation. This project encompasses the following tasks: 1. Simulate 3D crack initiation and growth under instantaneous and/or fatigue loads using a new variant of the material point method (MPM); 2. Simulate debonding of the materials in the crack path using cohesive elements, considering normal and shear traction separation laws; 3. Determine the crack propagation path, considering damage of the materials incorporated in the cohesive elements to allow the energy release rate to be minimized; 4. Simulate the three-dimensional fatigue crack growth as a function of loading histories; 5. Verify the simulation code by comparing results to theoretical and numerical studies available in the literature; 6. Conduct experiments to observe the crack path and surface profile in unused fuel cladding and validate against simulation results; and 7. Expand the adaptive mesh refinement infrastructure parallel processing environment to allow adaptive mesh refinement at the 3D crack fronts and adaptive mesh merging in the wake of cracks. Fuel cladding is made of materials such as stainless steels and ferritic steels with added alloying elements, which increase stability and durability under irradiation. As fuel cladding is subjected to water, chemicals, fission gas, pressure, high temperatures, and irradiation while in service, understanding performance is essential. In the fast fuel used in advanced burner reactors, simulations of the nuclear

  3. Development of Experimental Facilities for Advanced Spent Fuel Management Technology

    International Nuclear Information System (INIS)

    The advanced spent fuel management process(ACP), proposed to reduce the overall volume of the PWR spent fuel and improve safety and economy of the long-term storage of spent fuel, is under research and development. This technology convert spent fuels into pure metal-base uranium with removing the highly heat generating materials(Cs, Sr) efficiently and reducing of the decay heat, volume, and radioactivity from spent fuel by 1/4. In the next phase(2004∼2006), the demonstration of this technology will be carried out for verification of the ACP in a laboratory scale. For this demonstration, the hot cell facilities of α-γ type and auxiliary facilities are required essentially for safe handling of high radioactive materials. As the hot cell facilities for demonstration of the ACP, a existing hot cell of β-γ type will be refurbished to minimize construction expenditures of hot cell facility. In this study, the design requirements are established, and the process detail work flow was analysed for the optimum arrangement to ensure effective process operation in hot cell. And also, the basic and detail design of hot cell facility and process, and safety analysis was performed to secure conservative safety of hot cell facility and process

  4. Reduction of repository heat load using advanced fuel cycles

    International Nuclear Information System (INIS)

    With the geologic repository at Yucca Mountain already nearing capacity full before opening, advanced fuel cycles that introduce reprocessing, fast reactors, and temporary storage sites have the potential to allow the repository to support the current reactor fleet and future expansion. An uncertainty analysis methodology that combines Monte Carlo distribution sampling, reactor physics data simulation, and neural network interpolation methods enable investigation into the factor reduction of heat capacity by using the hybrid fuel cycle. Using a Super PRISM fast reactor with a conversion ratio of 0.75, burn ups reach up to 200 MWd/t that decrease the plutonium inventory by about 5 metric tons every 12 years. Using the long burn up allows the footprint of 1 single core loading of FR fuel to have an integral decay heat of about 2.5x105 MW*yr over a 1500 year period that replaces the footprint of about 6 full core loadings of LWR fuel for the number of years required to fuel the FR, which have an integral decay heat of about.3 MW*yr for the same time integral. This results in an increase of a factor of 4 in repository support capacity from implementing a single fast reactor in an equilibrium cycle. (authors)

  5. Ignition timing advance in the bi-fuel engine

    Directory of Open Access Journals (Sweden)

    Marek FLEKIEWICZ

    2009-01-01

    Full Text Available The influence of ignition timing on CNG combustion process has been presented in this paper. A 1.6 liter SI engine has been tested in the special program. For selected engine operating conditions, following data were acquired: in cylinder pressure, crank angle, fuel mass consumption and exhaust gases temperatures. For the timing advance correction varying between 0 to 15 deg crank angle, the internal temperature of combustion chamber, as well as the charge combustion ratio and ratio of heat release has been estimated. With the help of the mathematical model, emissions of NO, CO and CO2 were additionally estimated. Obtained results made it possible to compare the influence of ignition timing advance on natural gas combustion in the SI engine. The engine torque and in-cylinder pressure were used for determination of the optimum engine timing advance.

  6. Feasibility study on the development of advanced LWR fuel technology

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Youn Ho; Sohn, D. S.; Jeong, Y. H.; Song, K. W.; Song, K. N.; Chun, T. H.; Bang, J. G.; Bae, K. K.; Kim, D. H. and others

    1997-07-01

    Worldwide R and D trends related to core technology of LWR fuels and status of patents have been surveyed for the feasibility study. In addition, various fuel cycle schemes have been studied to establish the target performance parameters. For the development of cladding material, establishment of long-term research plan for alloy development and optimization of melting process and manufacturing technology were conducted. A work which could characterize the effect of sintering additives on the microstructure of UO{sub 2} pellet has been experimentally undertaken, and major sintering variables and their ranges have been found in the sintering process of UO{sub 2}-Gd{sub 2}O{sub 3} burnable absorber pellet. The analysis of state of the art technology related to flow mixing device for spacer grid and debris filtering device for bottom nozzle and the investigation of the physical phenomena related to CHF enhancement and the establishment of the data base for thermal-hydraulic performance tests has been done in this study. In addition, survey on the documents of the up-to-date PWR fuel assemblies developed by foreign vendors have been carried out to understand their R and D trends and establish the direction of R and D for these structural components. And, to set the performance target of the new fuel, to be developed, fuel burnup and economy under the extended fuel cycle length scheme were estimated. A preliminary study on the failure mechanism of CANDU fuel, key technology and advanced coating has been performed. (author). 190 refs., 31 tabs., 129 figs.

  7. Advanced Nuclear Fuel Cycle Transitions: Optimization, Modeling Choices, and Disruptions

    Science.gov (United States)

    Carlsen, Robert W.

    Many nuclear fuel cycle simulators have evolved over time to help understan the nuclear industry/ecosystem at a macroscopic level. Cyclus is one of th first fuel cycle simulators to accommodate larger-scale analysis with it liberal open-source licensing and first-class Linux support. Cyclus also ha features that uniquely enable investigating the effects of modeling choices o fuel cycle simulators and scenarios. This work is divided into thre experiments focusing on optimization, effects of modeling choices, and fue cycle uncertainty. Effective optimization techniques are developed for automatically determinin desirable facility deployment schedules with Cyclus. A novel method fo mapping optimization variables to deployment schedules is developed. Thi allows relationships between reactor types and scenario constraints to b represented implicitly in the variable definitions enabling the usage o optimizers lacking constraint support. It also prevents wasting computationa resources evaluating infeasible deployment schedules. Deployed power capacit over time and deployment of non-reactor facilities are also included a optimization variables There are many fuel cycle simulators built with different combinations o modeling choices. Comparing results between them is often difficult. Cyclus flexibility allows comparing effects of many such modeling choices. Reacto refueling cycle synchronization and inter-facility competition among othe effects are compared in four cases each using combinations of fleet of individually modeled reactors with 1-month or 3-month time steps. There are noticeable differences in results for the different cases. The larges differences occur during periods of constrained reactor fuel availability This and similar work can help improve the quality of fuel cycle analysi generally There is significant uncertainty associated deploying new nuclear technologie such as time-frames for technology availability and the cost of buildin advanced reactors

  8. Recovery Act: Advanced Direct Methanol Fuel Cell for Mobile Computing

    Energy Technology Data Exchange (ETDEWEB)

    Fletcher, James H. [University of North Florida; Cox, Philip [University of North Florida; Harrington, William J [University of North Florida; Campbell, Joseph L [University of North Florida

    2013-09-03

    ABSTRACT Project Title: Recovery Act: Advanced Direct Methanol Fuel Cell for Mobile Computing PROJECT OBJECTIVE The objective of the project was to advance portable fuel cell system technology towards the commercial targets of power density, energy density and lifetime. These targets were laid out in the DOE’s R&D roadmap to develop an advanced direct methanol fuel cell power supply that meets commercial entry requirements. Such a power supply will enable mobile computers to operate non-stop, unplugged from the wall power outlet, by using the high energy density of methanol fuel contained in a replaceable fuel cartridge. Specifically this project focused on balance-of-plant component integration and miniaturization, as well as extensive component, subassembly and integrated system durability and validation testing. This design has resulted in a pre-production power supply design and a prototype that meet the rigorous demands of consumer electronic applications. PROJECT TASKS The proposed work plan was designed to meet the project objectives, which corresponded directly with the objectives outlined in the Funding Opportunity Announcement: To engineer the fuel cell balance-of-plant and packaging to meet the needs of consumer electronic systems, specifically at power levels required for mobile computing. UNF used existing balance-of-plant component technologies developed under its current US Army CERDEC project, as well as a previous DOE project completed by PolyFuel, to further refine them to both miniaturize and integrate their functionality to increase the system power density and energy density. Benefits of UNF’s novel passive water recycling MEA (membrane electrode assembly) and the simplified system architecture it enabled formed the foundation of the design approach. The package design was hardened to address orientation independence, shock, vibration, and environmental requirements. Fuel cartridge and fuel subsystems were improved to ensure effective fuel

  9. Application of the Advanced Distillation Curve Method to Fuels for Advanced Combustion Engine Gasolines

    KAUST Repository

    Burger, Jessica L.

    2015-07-16

    © This article not subject to U.S. Copyright. Published 2015 by the American Chemical Society. Incremental but fundamental changes are currently being made to fuel composition and combustion strategies to diversify energy feedstocks, decrease pollution, and increase engine efficiency. The increase in parameter space (by having many variables in play simultaneously) makes it difficult at best to propose strategic changes to engine and fuel design by use of conventional build-and-test methodology. To make changes in the most time- and cost-effective manner, it is imperative that new computational tools and surrogate fuels are developed. Currently, sets of fuels are being characterized by industry groups, such as the Coordinating Research Council (CRC) and other entities, so that researchers in different laboratories have access to fuels with consistent properties. In this work, six gasolines (FACE A, C, F, G, I, and J) are characterized by the advanced distillation curve (ADC) method to determine the composition and enthalpy of combustion in various distillate volume fractions. Tracking the composition and enthalpy of distillate fractions provides valuable information for determining structure property relationships, and moreover, it provides the basis for the development of equations of state that can describe the thermodynamic properties of these complex mixtures and lead to development of surrogate fuels composed of major hydrocarbon classes found in target fuels.

  10. Update on fuel fabrication development and testing at Argonne National Laboratory

    International Nuclear Information System (INIS)

    In its effort to develop research reactor fuel with a high fissile loading, Argonne National Laboratory has continued its advanced fuel development efforts. Monolithic fuel, where the fuel is in the form of a single fuel foil, is being developed as the ultimate in fuel loading capacity. Work has been done on different monolithic fabrication methods that have resulted in process refinements. Effort is also underway to develop a uranium-molybdenum dispersion fuel plate that will be resistant to the irradiation shortcomings noted in previous tests. Alloying additions to the aluminum matrix are being investigated. These fuels are being fabricated for use in irradiation experiments scheduled for insertion in 2005. (author)

  11. Fissile material management strategies for sustainable nuclear energy. Proceedings of a technical meeting

    International Nuclear Information System (INIS)

    This publication is the proceedings of a technical meeting with the objective to identify fissile material management strategies for different nuclear fuel cycle options, to clarify issues and challenges existing in fissile material management and to seek possible solutions for these issues and challenges, focusing, in particular, on the sustainability of nuclear power in different fuel cycle options. It provides a comprehensive review of the relative merits of the different fuel cycle options, valuable technical information on the current status and future direction of the fuel cycle in Member States, and essential information for policy makers and those involved in strategic planning. It contains 35 individual papers relating to front and back end fuel cycle strategies as well as future sustainable fuel cycle technology options. They are indexed separately

  12. The fuel to clad heat transfer coefficient in advanced MX-type fuel pins

    International Nuclear Information System (INIS)

    Advanced fuels (mixed carbides, nitrides and carbonitrides) are characterised by a high thermal conductivity compared to that of oxide fuels (5 times greater) and their behaviour under irradiation (amount of swelling, fracture behaviour, restructuring) is far more sensitive to the design parameters and to the operating temperature than that of oxide fuels. The use of advanced fuels is therefore conditioned by the possibility of mastering the above phenomena, and the full exploitation of their favorable neutron characteristics depends upon a good understanding of the mutual relationships of the various parameters, which eventually affect the mechanical stability of the pin. By far the most important parameter is the radial temperature profile which controls the swelling of the fuel and the build-up of stress fields within the pin. Since the rate of fission gas swelling of these fuels is relatively large, a sufficient amount of free space has to be provided within the pin. This space originally appears as fabrication porosity and as fuel-to-clad clearance. Due to the large initial gap width and to the high fuel thermal conductivity, the range of the fuel operating temperatures is mainly determined by the fuel-to-clad heat transfer coefficient h, whose correct determination becomes one of the central points in modelling. During the many years of modelling activity in the field of oxide fuels, several theoretical models have been developed to calculate h, and a large amount of experimental data has been produced for the empirical adjustment of the parameters involved, so that the situation may be regarded as rather satisfactory. The analysis lead to the following conclusions. A quantitative comparison of experimental h-values with existing models for h requires rather sophisticated instrumented irradiation capsules, which permit the measurement of mechanical data (concerning fuel and clad) together with heat rating and temperatures. More and better well

  13. Feasibility study of advanced fuel burning nuclear reactors

    International Nuclear Information System (INIS)

    An investigation has been conducted to determine both physics, engineering and economic aspects of fusion power reactors based on magnetic confinement and on burning advanced fuels (AFs). DT burning Tokamaks are taken as reference concept. We show that the attractive features of advanced fuels, in particular of neutronlean proton-based AFs, can be combined, in appropriately designed AF reactors (high beta), with power densities comparable to or even higher than those achievable in DT Tokamaks. Moreover we identify physical requirements which would assure Q values well above unity. As an example a semi-open confinement scheme is analyzed based on a self-consistent plasma calculation. We find that a mirror, even if only ''semi-open'' as a result of strong diamagnetism, can barely be expected to achieve high Q values. Therefore confinement schemes such as compact tori, multipole surmacs etc. may be required to burn AFs. We conclude that the economics of AF reactors, as determined by the nuclear boiler power density, may be superior to that of DT-rectors if low recirculating power fractions can be obtained by appropriate plasma tayloring (high fractional transfer of fusion power to ions required). A more detailed investigation is suggested for proton-based fuel cycles. (orig.)

  14. Temperature Profile of the Solution Vessel of an Accelerator-Driven Subcritical Fissile Solution System

    International Nuclear Information System (INIS)

    Dynamic System Simulation (DSS) models of fissile solution systems have been developed and verified against a variety of historical configurations. DSS techniques have been applied specifically to subcritical accelerator-driven systems using fissile solution fuels of uranium. Initial DSS models were developed in DESIRE, a specialized simulation scripting language. In order to tailor the DSS models to specifically meet needs of system designers they were converted to a Visual Studio implementation, and one of these subsequently to National Instrument's LabVIEW for human factors engineering and operator training. Specific operational characteristics of subcritical accelerator-driven systems have been examined using a DSS model tailored to this particular class using fissile fuel.

  15. Temperature Profile of the Solution Vessel of an Accelerator-Driven Subcritical Fissile Solution System

    Energy Technology Data Exchange (ETDEWEB)

    Klein, Steven Karl [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Determan, John C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-09-14

    Dynamic System Simulation (DSS) models of fissile solution systems have been developed and verified against a variety of historical configurations. DSS techniques have been applied specifically to subcritical accelerator-driven systems using fissile solution fuels of uranium. Initial DSS models were developed in DESIRE, a specialized simulation scripting language. In order to tailor the DSS models to specifically meet needs of system designers they were converted to a Visual Studio implementation, and one of these subsequently to National Instrument’s LabVIEW for human factors engineering and operator training. Specific operational characteristics of subcritical accelerator-driven systems have been examined using a DSS model tailored to this particular class using fissile fuel.

  16. Development of Kinetic Mechanisms for Next-Generation Fuels and CFD Simulation of Advanced Combustion Engines

    Energy Technology Data Exchange (ETDEWEB)

    Pitz, William J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); McNenly, Matt J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Whitesides, Russell [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Mehl, Marco [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Killingsworth, Nick J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Westbrook, Charles K. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-12-17

    Predictive chemical kinetic models are needed to represent next-generation fuel components and their mixtures with conventional gasoline and diesel fuels. These kinetic models will allow the prediction of the effect of alternative fuel blends in CFD simulations of advanced spark-ignition and compression-ignition engines. Enabled by kinetic models, CFD simulations can be used to optimize fuel formulations for advanced combustion engines so that maximum engine efficiency, fossil fuel displacement goals, and low pollutant emission goals can be achieved.

  17. Prospects for advanced coal-fuelled fuel cell power plants

    International Nuclear Information System (INIS)

    As part of ECN's in-house R and D programmes on clean energy conversion systems with high efficiencies and low emissions, system assessment studies have been carried out on coal gasification power plants integrated with high-temperature fuel cells (IGFC). The studies also included the potential to reduce CO2 emissions, and to find possible ways for CO2 extraction and sequestration. The development of this new type of clean coal technology for large-scale power generation is still far off. A significant market share is not envisaged before the year 2015. To assess the future market potential of coal-fuelled fuel cell power plants, the promise of this fuel cell technology was assessed against the performance and the development of current state-of-the-art large-scale power generation systems, namely the pulverized coal-fired power plants and the integrated coal gasification combined cycle (IGCC) power plants. With the anticipated progress in gas turbine and gas clean-up technology, coal-fuelled fuel cell power plants will have to face severe competition from advanced IGCC power plants, despite their higher efficiency. (orig.)

  18. Radiation Monitoring System in Advanced Spent Fuel Conditioning Process Facility

    International Nuclear Information System (INIS)

    The Advanced spent fuel Conditioning Process is under development for effective management of spent fuel by converting UO2 into U-metal. For demonstration of this process, α-γ type new hot cell was built in the IMEF basement . To secure against radiation hazard, this facility needs radiation monitoring system which will observe the entire operating area before the hot cell and service area at back of it. This system consists of 7 parts; Area Monitor for γ-ray, Room Air Monitor for particulate and iodine in both area, Hot cell Monitor for hot cell inside high radiation and rear door interlock, Duct Monitor for particulate of outlet ventilation, Iodine Monitor for iodine of outlet duct, CCTV for watching workers and material movement, Server for management of whole monitoring system. After installation and test of this, radiation monitoring system will be expected to assist the successful ACP demonstration

  19. UF6 cylinder washing at ANF [Advanced Nuclear Fuels

    International Nuclear Information System (INIS)

    Advanced Nuclear Fuels (ANF) Corporation receives UF6 cylinders in its Richland, WA fuel fabrication plant from its utility customers. Following transfer of the UF6, designated cylinders are cleaned by washing using a specialized process which emphasizes personnel safety and low cost. The cleaning operation is conducted in accordance with the guidelines is conducted in accordance with the guidelines of ANSI Standard N14.1. Specialized equipment has been developed for carrying out this operation with a minimum of space and manpower. Equipment designs and process methods are selected to ensure that the cylinders meet cleanliness requirements while maintaining low personnel radiation exposures, avoiding introduction of organic material and achieving low applied labor. A photo tour is used to illustrate the equipment and methods used

  20. Advanced Neutron Source Reactor thermal analysis of fuel plate defects

    International Nuclear Information System (INIS)

    The Advanced Neutron Source Reactor (ANSR) is a research reactor designed to provide the highest continuous neutron beam intensity of any reactor in the world. The present technology for determining safe operations were developed for the High Flux Isotope Reactor (HFIR). These techniques are conservative and provide confidence in the safe operation of HFIR. However, the more intense requirements of ANSR necessitate the development of more accurate, but still conservative, techniques. This report details the development of a Local Analysis Technique (LAT) that provides an appropriate approach. Application of the LAT to two ANSR core designs are presented. New theories of the thermal and nuclear behavior of the U3Si2 fuel are utilized. The implications of lower fuel enrichment and of modifying the inspection procedures are also discussed. Development of the computer codes that enable the automate execution of the LAT is included

  1. Advanced Fuel Cycle Economic Tools, Algorithms, and Methodologies

    International Nuclear Information System (INIS)

    The Advanced Fuel Cycle Initiative (AFCI) Systems Analysis supports engineering economic analyses and trade-studies, and requires a requisite reference cost basis to support adequate analysis rigor. In this regard, the AFCI program has created a reference set of economic documentation. The documentation consists of the 'Advanced Fuel Cycle (AFC) Cost Basis' report (Shropshire, et al. 2007), 'AFCI Economic Analysis' report, and the 'AFCI Economic Tools, Algorithms, and Methodologies Report.' Together, these documents provide the reference cost basis, cost modeling basis, and methodologies needed to support AFCI economic analysis. The application of the reference cost data in the cost and econometric systems analysis models will be supported by this report. These methodologies include: the energy/environment/economic evaluation of nuclear technology penetration in the energy market - domestic and internationally - and impacts on AFCI facility deployment, uranium resource modeling to inform the front-end fuel cycle costs, facility first-of-a-kind to nth-of-a-kind learning with application to deployment of AFCI facilities, cost tradeoffs to meet nuclear non-proliferation requirements, and international nuclear facility supply/demand analysis. The economic analysis will be performed using two cost models. VISION.ECON will be used to evaluate and compare costs under dynamic conditions, consistent with the cases and analysis performed by the AFCI Systems Analysis team. Generation IV Excel Calculations of Nuclear Systems (G4-ECONS) will provide static (snapshot-in-time) cost analysis and will provide a check on the dynamic results. In future analysis, additional AFCI measures may be developed to show the value of AFCI in closing the fuel cycle. Comparisons can show AFCI in terms of reduced global proliferation (e.g., reduction in enrichment), greater sustainability through preservation of a natural resource (e.g., reduction in uranium ore depletion), value from weaning the

  2. Development of the advanced PHWR technology -Design and analysis of CANDU advanced fuel-

    Energy Technology Data Exchange (ETDEWEB)

    Suk, Hoh Chun; Shim, Kee Sub; Byun, Taek Sang; Park, Kwang Suk; Kang, Heui Yung; Kim, Bong Kee; Jung, Chang Joon; Lee, Yung Wook; Bae, Chang Joon; Kwon, Oh Sun; Oh, Duk Joo; Im, Hong Sik; Ohn, Myung Ryong; Lee, Kang Moon; Park, Joo Hwan; Lee, Eui Joon [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    This is the `94 annual report of the CANDU advanced fuel design and analysis project, and describes CANFLEX fuel design and mechanical integrity analysis, reactor physics analysis and safety analysis of the CANDU-6 with the CANFLEX-NU. The following is the R and D scope of this fiscal year : (1) Detail design of CANFLEX-NU and detail analysis on the fuel integrity, reactor physics and safety. (a) Detail design and mechanical integrity analysis of the bundle (b) CANDU-6 refueling simulation, and analysis on the Xe transients and adjuster system capability (c) Licensing strategy establishment and safety analysis for the CANFLEX-NU demonstration demonstration irradiation in a commercial CANDU-6. (2) Production and revision of CANFLEX-NU fuel design documents (a) Production and approval of CANFLEX-NU reference drawing, and revisions of fuel design manual and technical specifications (b) Production of draft physics design manual. (3) Basic research on CANFLEX-SEU fuel. 55 figs, 21 tabs, 45 refs. (Author).

  3. Advanced Diagnostics in Oxy-Fuel Combustion Processes

    DEFF Research Database (Denmark)

    Brix, Jacob; Toftegaard, Maja Bøg; Clausen, Sønnik;

    in the sampling equipment. The use of the IR technique for determination of particle temperatures, particle sizes, and number density proved reliable in both the swirl burner and the laboratory scale fixed bed reactor. When the technique was used in the swirl burner the subsequent data treatment was......This report sums up the findings in PSO-project 010069, “Advanced Diagnostics in Oxy- Fuel Combustion Processes”. Three areas of optic diagnostics are covered in this work: - FTIR measurements in a 30 kW swirl burner. - IR measurements in a 30 kW swirl burner. - IR measurements in a laboratory...

  4. High efficiency fuel cell/advanced turbine power cycles

    Energy Technology Data Exchange (ETDEWEB)

    Morehead, H. [Westinghouse Electric Corp., Orlando, FL (United States)

    1995-10-19

    An outline of the Westinghouse high-efficiency fuel cell/advanced turbine power cycle is presented. The following topics are discussed: The Westinghouse SOFC pilot manufacturing facility, cell scale-up plan, pressure effects on SOFC power and efficiency, sureCell versus conventional gas turbine plants, sureCell product line for distributed power applications, 20 MW pressurized-SOFC/gas turbine power plant, 10 MW SOFC/CT power plant, sureCell plant concept design requirements, and Westinghouse SOFC market entry.

  5. Economic potential of advanced fuel cycles in CANDU

    International Nuclear Information System (INIS)

    Advanced fuel cycles in CANDU offer the potential of a many-fold increase in energy yield over that which can be obtained from uranium resources using the current once-through natural uranium cycle. This paper examines the associated economics of alternative once-through and recycle fuelling. Results indicate that these cycles will limit the impact of higher uranium prices and offer the potential of a period of stable constant-dollar generating costs that are only approximately 20% higher than current levels

  6. Muon tomography algorithms for fissile nuclear materials detection

    International Nuclear Information System (INIS)

    Muon Tomography is a novel technique which exploits the multiple coulomb scattering of these particles for nondestructive inspection without the use of artificial radiation. It describes the concept and theory of cosmic-ray muon tomography, and discuss preliminary and advanced reconstruction algorithms (PoCA algorithm and MLS-EM algorithm), which take advantage of the scattering angle. Our algorithms are validated with simulation experimental demonstrations. Based upon the results from simulations, we conclude that muon radiography can be useful for detecting fissile nuclear materials. (authors)

  7. A worldwide program for controlling fissile material

    International Nuclear Information System (INIS)

    Fissile (fissionable) material, either plutonium or highly enriched uranium or both in combination, is an essential ingredient of any nuclear weapon. This is equally true of the first generation of pure fission weapons - such as those used on Hiroshima and Nagasaki - and more modern weapons which gain part of their yield from thermonuclear fusion. In the latter case, a fission explosion is required to ignite the fusion reaction. This paper proposes a worldwide programme based upon that linkage. Within this programme, however, the dominant element would be a cut-off in the production of fissile material for nuclear weapons. This is a subject to which various analysts have devoted attention over the years. Most notably, recent work by analysts at Princeton University in the USA has convincingly shown that an initial cut-off of production in the USA and the Soviet Union could be adequately verified. In recognition of this work, the approach taken here is to build upon the Princeton analysis, showing how the verification regime appropriate to an initial US-Soviet cut-off could be expanded to cover reductions in military fissile materials inventories in those two countries, production cut-offs and inventory reductions in other NWSs, and worldwide regulation of civilian programmes for fissile material production. The description of this ultimate fissile material control regime is couched primarily in terms of the criteria according to which the regime should be designed, and the nature of the institutions and agreements which should support it

  8. Design of the Advanced Gas Reactor Fuel Experiments for Irradiation in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover

    2005-10-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight particle fuel tests in the Advanced Test Reactor (ATR) located at the newly formed Idaho National Laboratory (INL) to support development of the next generation Very High Temperature Reactor (VHTR) in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments will be irradiated in an inert sweep gas atmosphere with on-line temperature monitoring and control combined with on-line fission product monitoring of the sweep gas. The final design phase has just been completed on the first experiment (AGR-1) in this series and the support systems and fission product monitoring system that will monitor and control the experiment during irradiation. This paper discusses the development of the experimental hardware and support system designs and the status of the experiment.

  9. Advanced fuel cycles: a rationale and strategy for adopting the low-enriched-uranium fuel cycle

    International Nuclear Information System (INIS)

    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)

  10. Advanced Fuels Campaign Light Water Reactor Accident Tolerant Fuel Performance Metrics

    Energy Technology Data Exchange (ETDEWEB)

    Brad Merrill; Melissa Teague; Robert Youngblood; Larry Ott; Kevin Robb; Michael Todosow; Chris Stanek; Mitchell Farmer; Michael Billone; Robert Montgomery; Nicholas Brown; Shannon Bragg-Sitton

    2014-02-01

    The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the United States’ nuclear industry. As a result, continual improvement of technology, including advanced materials and nuclear fuels, remains central to industry’s success. Decades of research combined with continual operation have produced steady advancements in technology and yielded an extensive base of data, experience, and knowledge on light water reactor (LWR) fuel performance under both normal and accident conditions. In 2011, following the Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex, enhancing the accident tolerance of LWRs became a topic of serious discussion. As a result of direction from the U.S. Congress, the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) initiated an Accident Tolerant Fuel (ATF) Development program. The complex multiphysics behavior of LWR nuclear fuel makes defining specific material or design improvements difficult; as such, establishing qualitative attributes is critical to guide the design and development of fuels and cladding with enhanced accident tolerance. This report summarizes a common set of technical evaluation metrics to aid in the optimization and down selection of candidate designs. As used herein, “metrics” describe a set of technical bases by which multiple concepts can be fairly evaluated against a common baseline and against one another. Furthermore, this report describes a proposed technical evaluation methodology that can be applied to assess the ability of each concept to meet performance and safety goals relative to the current UO2 – zirconium alloy system and relative to one another. The resultant ranked evaluation can then inform concept down-selection, such that the most promising accident tolerant fuel design option(s) can continue to be developed for lead test rod or lead test assembly

  11. Resonance self-indication as a method for estimating fissile elements in spent nuclear assemblies

    International Nuclear Information System (INIS)

    The integral version of the neutron resonance transmission analysis (NRTA) has been developed in order to solve the problem of estimating fissile elements in spent nuclear fuel. Two variants of the integral neutron resonance transmission analysis (the resonance self-indication (the average transmission by fission) and the average full transmission) have been described

  12. Achievements and prospects for advanced reactor design and fuel cycles

    International Nuclear Information System (INIS)

    The future of Nuclear Energy relies on the complementary optimization of reactors for NPPs and the associated nuclear fuel cycles. This is an apparent contradiction if we look in the so large effort made worldwide for developing advance reactors for power plants alone. The vision that focus the optimization effort in reactors and in the other side and separated in the associated fuel cycle jeopardizes the final results of an optimized nuclear system. The control of the primary source of energy is a key question and the technology involved and its control the main issue to be considered when the evaluation of advanced nuclear systems are under consideration. However the main reason of this situation is that reactors for NPP is still been costly, inefficient compared with other energy converters and increasingly complex to accomplish safety requirements. The maturity of nuclear technology and the present NPP are the background for the evolutionary concepts of reactors while the response to economy, safety, waste generation and management and proliferation resistance are the drivers for innovative concepts. Most traditional technology holders and NPP vendors have evolutionary LWR and HWR systems and participate directly or indirectly in innovative projects for future applications including fast reactors. EPR, AP 1000, KSNP, ABWR, WWER-600, ACR-700 and AHWR are examples of this fact. Example of continuous effort in fast reactors development are MONJU reactor, CEFR, FBTR and the emblematic Superphenix. Both reactors and nuclear fuel cycles should evolve throughout a breakthrough process if the energy demand mainly becomes large in developing countries. This may require a different approach that the one that drives the past 50 years mainly because the modules should be optimized for quite different electricity markets. Small and Medium Power Reactors like SMART, CAREM, IRIS, PBMR and HTGRs, enrichment processes optimized to be economics for small capacity production

  13. Objectives, Strategies, and Challenges for the Advanced Fuel Cycle Initiative

    International Nuclear Information System (INIS)

    This paper will summarize the objectives, strategies, and key chemical separation challenges for the Advanced Fuel Cycle Initiative (AFCI). The major objectives are as follows: Waste management--defer the need for a second geologic repository for a century or more, Proliferation resistance--be more resistant than the existing PUREX separation technology or uranium enrichment, Energy sustainability--turn waste management liabilities into energy source assets to ensure that uranium ore resources do not become a constraint on nuclear power, and Systematic, safe, and economic management of the entire fuel cycle. There are four major strategies for the disposal of civilian spent fuel: Once-through--direct disposal of all discharged nuclear fuel, Limited recycle--recycle transuranic elements once and then direct disposal, Continuous recycle--recycle transuranic elements repeatedly, and Sustained recycle--same as continuous except previously discarded depleted uranium is also recycled. The key chemical separation challenges stem from the fact that the components of spent nuclear fuel vary greatly in their influence on achieving program objectives. Most options separate uranium to reduce the weight and volume of waste and the number and cost of waste packages that require geologic disposal. Separated uranium can also be used as reactor fuel. Most options provide means to recycle transuranic (TRU) elements--plutonium (Pu), neptunium (Np), americium (Am), curium (Cm). Plutonium must be recycled to obtain repository, proliferation, and energy recovery benefits. U.S. non-proliferation policy forbids separation of plutonium by itself; therefore, one or more of the other transuranic elements must be kept with the plutonium; neptunium is considered the easiest option. Recycling neptunium also provides repository benefits. Americium recycling is also required to obtain repository benefits. At the present time, curium recycle provides relatively little benefit; indeed, recycling

  14. THE MISSION AND ACCOMPLISHMENTS FROM DOE’S FUEL CYCLE RESEARCH AND DEVELOPMENT (FCRD) ADVANCED FUELS CAMPAIGN

    Energy Technology Data Exchange (ETDEWEB)

    J. Carmack; L. Braase; F. Goldner

    2015-09-01

    The mission of the Advanced Fuels Campaign (AFC) is to perform Research, Development, and Demonstration (RD&D) activities for advanced fuel forms (including cladding) to enhance the performance and safety of the nation’s current and future reactors, enhance proliferation resistance of nuclear fuel, effectively utilize nuclear energy resources, and address the longer-term waste management challenges. This includes development of a state of the art Research and Development (R&D) infrastructure to support the use of a “goal oriented science based approach.” AFC uses a “goal oriented, science based approach” aimed at a fundamental understanding of fuel and cladding fabrication methods and performance under irradiation, enabling the pursuit of multiple fuel forms for future fuel cycle options. This approach includes fundamental experiments, theory, and advanced modeling and simulation. One of the most challenging aspects of AFC is the management, integration, and coordination of major R&D activities across multiple organizations. AFC interfaces and collaborates with Fuel Cycle Technologies (FCT) campaigns, universities, industry, various DOE programs and laboratories, federal agencies (e.g., Nuclear Regulatory Commission [NRC]), and international organizations. Key challenges are the development of fuel technologies to enable major increases in fuel performance (safety, reliability, power and burnup) beyond current technologies, and development of characterization methods and predictive fuel performance models to enable more efficient development and licensing of advanced fuels. Challenged with the research and development of fuels for two different reactor technology platforms, AFC targeted transmutation fuel development and focused ceramic fuel development for Advanced LWR Fuels.

  15. Development of Experimental Facilities for Advanced Spent Fuel Management Technology

    International Nuclear Information System (INIS)

    The Advanced spent fuel Conditioning Process Facility(ACPF) and hotcell system technologies were developed in this program for demonstrating safely and effectively the Advanced spent fuel Conditioning Process(ACP) on a lab-scale. With the analysis of work flow and characteristics of the process, ACP was successively demonstrated on a lab-scale experiments and the performance of process was evaluated. The hotcell system was comprehensively evaluated with those results and the design data for the engineering-scale demonstration was derived to propose the direction for the future research and development. The main items performed in this project were as follows. - The reconstruction of ACPF hotcell and demonstration of the ACP - The design and operation technologies for α-γ type nuclear hot cell facility - The overall evaluation of the performance, safety and operation ability of the hotcell system - The acquisition of the government licences for construction and operation and the IAEA licence for using nuclear materials The results of safety analysis and environmental effects assessment and performance data for ACPF had been used for acquiring the government licence for facility operation. The valuable experiences on pyroprocess facility design and operation knowledges would be applied to new Mock-up Facility being scheduled to be a previous stage facility of Integrated Pyroprocess Facility

  16. An advanced aqueous process for nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    To develop an advanced aqueous reprocessing process using a minimal organic solvent and compact equipment to separate U, Pu and long-lived minor actinides from spent nuclear fuels, we have developed a new type of anion exchanger (AR-01) and several novel extraction resins containing a chelating ligand such as CMPO, Cyanex-301 and R-BTP. A hot separation experiment for a spent BWR-fuel solution was carried out by an ion exchange column packed with AR-01. To enhance the separation factor between U and FPs, electrolytic reduction of UO22+ to U+4 was studied using a flow type electrolysis cell with carbon-fiber electrode. Separation behavior of Am(III) from simulated HLW by CMPO and R-BTP impregnated resins were investigated. On the basis of the experimental results, an advanced aqueous process which consists of anion exchange as main separation method, electrolytic reduction for reducing U(VI) to U(IV) and extraction chromatography for MA partitioning has been designed and evaluated preliminarily. (author)

  17. Advanced coal gasifier-fuel cell power plant systems design

    Science.gov (United States)

    Heller, M. E.

    1983-01-01

    Two advanced, high efficiency coal-fired power plants were designed, one utilizing a phosphoric acid fuel cell and one utilizing a molten carbonate fuel cell. Both incorporate a TRW Catalytic Hydrogen Process gasifier and regenerator. Both plants operate without an oxygen plant and without requiring water feed; they, instead, require makeup dolomite. Neither plant requires a shift converter; neither plant has heat exchangers operating above 1250 F. Both plants have attractive efficiencies and costs. While the molten carbonate version has a higher (52%) efficiency than the phosphoric acid version (48%), it also has a higher ($0.078/kWh versus $0.072/kWh) ten-year levelized cost of electricity. The phosphoric acid fuel cell power plant is probably feasible to build in the near term: questions about the TRW process need to be answered experimentally, such as weather it can operate on caking coals, and how effective the catalyzed carbon-dioxide acceptor will be at pilot scale, both in removing carbon dioxide and in removing sulfur from the gasifier.

  18. Advanced Coal-Fueled Gas Turbine Program. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Horner, M.W.; Ekstedt, E.E.; Gal, E.; Jackson, M.R.; Kimura, S.G.; Lavigne, R.G.; Lucas, C.; Rairden, J.R.; Sabla, P.E.; Savelli, J.F.; Slaughter, D.M.; Spiro, C.L.; Staub, F.W.

    1989-02-01

    The objective of the original Request for Proposal was to establish the technological bases necessary for the subsequent commercial development and deployment of advanced coal-fueled gas turbine power systems by the private sector. The offeror was to identify the specific application or applications, toward which his development efforts would be directed; define and substantiate the technical, economic, and environmental criteria for the selected application; and conduct such component design, development, integration, and tests as deemed necessary to fulfill this objective. Specifically, the offeror was to choose a system through which ingenious methods of grouping subcomponents into integrated systems accomplishes the following: (1) Preserve the inherent power density and performance advantages of gas turbine systems. (2) System must be capable of meeting or exceeding existing and expected environmental regulations for the proposed application. (3) System must offer a considerable improvement over coal-fueled systems which are commercial, have been demonstrated, or are being demonstrated. (4) System proposed must be an integrated gas turbine concept, i.e., all fuel conditioning, all expansion gas conditioning, or post-expansion gas cleaning, must be integrated into the gas turbine system.

  19. Advanced fuel cells for transportation applications. Final report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-02-10

    This Research and Development (R and D) contract was directed at developing an advanced technology compressor/expander for supplying compressed air to Proton Exchange Membrane (PEM) fuel cells in transportation applications. The objective of this project was to develop a low-cost high-efficiency long-life lubrication-free integrated compressor/expander utilizing scroll technology. The goal of this compressor/expander was to be capable of providing compressed air over the flow and pressure ranges required for the operation of 50 kW PEM fuel cells in transportation applications. The desired ranges of flow, pressure, and other performance parameters were outlined in a set of guidelines provided by DOE. The project consisted of the design, fabrication, and test of a prototype compressor/expander module. The scroll CEM development program summarized in this report has been very successful, demonstrating that scroll technology is a leading candidate for automotive fuel cell compressor/expanders. The objectives of the program are: develop an integrated scroll CEM; demonstrate efficiency and capacity goals; demonstrate manufacturability and cost goals; and evaluate operating envelope. In summary, while the scroll CEM program did not demonstrate a level of performance as high as the DOE guidelines in all cases, it did meet the overriding objectives of the program. A fully-integrated, low-cost CEM was developed that demonstrated high efficiency and reliable operation throughout the test program. 26 figs., 13 tabs.

  20. Performance evaluation of the Loviisa advanced type fuel rods

    International Nuclear Information System (INIS)

    The fuel vendor TVEL has supplied to Loviisa WWER-440 power plant six lead assemblies of an advanced type which have profiling of the fuel enrichment, demountability of the assembly and a reduced shroud wall thickness. The pool side examination programme of these assemblies is underway including visual inspections, diameter and length measurements between operation cycles, and end-of-life fission gas release measurements, determined from 85Kr activity in the plenum. Complementary evaluations and testing of models are done with the ENIGMA fuel performance code. The diameters of the corner rods have decreased to 30 μm during the first cycle and 40 to 70 μm after two cycles (with rod burnups of 24-30 MWd/kgU). The extent of creep-down is generally as expected, and agrees with the creep model adjusted for Russian Zr1%Nb cladding type and the Loviisa coolant and neutron flux conditions. The gap closure and reversed hoop strain are to be awaited during the third cycle so the new data will be an interesting validation exercise for the model and ENIGMA. Calculated temperatures stay low, and therefore low fission gas release fractions are anticipated as well

  1. The Adoption of Advanced Fuel Cycle Technology Under a Single Repository Policy

    International Nuclear Information System (INIS)

    Develops the tools to investigate the hypothesis that the savings in repository space associated with the implementation of advanced nuclear fuel cycles can result in sufficient cost savings to offset the higher costs of those fuel cycles.

  2. The Adoption of Advanced Fuel Cycle Technology Under a Single Repository Policy

    Energy Technology Data Exchange (ETDEWEB)

    Paul Wilson

    2009-11-02

    Develops the tools to investiage the hypothesis that the savings in repository space associated with the implementation of advanced nuclear fuel cycles can result in sufficient cost savings to offset the higher costs of those fuel cycles.

  3. Experiment Safety Assurance Package for Mixed Oxide Fuel Irradiation in an Average Power Position (I-24) in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    J. M . Ryskamp; R. C. Howard; R. C. Pedersen; S. T. Khericha

    1998-10-01

    The Fissile Material Disposition Program Light Water Reactor Mixed Oxide Fuel Irradiation Test Project Plan details a series of test irradiations designed to investigate the use of weapons-grade plutonium in MOX fuel for light water reactors (LWR) (Cowell 1996a, Cowell 1997a, Thoms 1997a). Commercial MOX fuel has been successfully used in overseas reactors for many years; however, weapons-derived test fuel contains small amounts of gallium (about 2 parts per million). A concern exists that the gallium may migrate out of the fuel and into the clad, inducing embrittlement. For preliminary out-of-pile experiments, Wilson (1997) states that intermetallic compound formation is the principal interaction mechanism between zircaloy cladding and gallium. This interaction is very limited by the low mass of gallium, so problems are not expected with the zircaloy cladding, but an in-pile experiment is needed to confirm the out-of-pile experiments. Ryskamp (1998) provides an overview of this experiment and its documentation. The purpose of this Experiment Safety Assurance Package (ESAP) is to demonstrate the safe irradiation and handling of the mixed uranium and plutonium oxide (MOX) Fuel Average Power Test (APT) experiment as required by Advanced Test Reactor (ATR) Technical Safety Requirement (TSR) 3.9.1 (LMITCO 1998). This ESAP addresses the specific operation of the MOX Fuel APT experiment with respect to the operating envelope for irradiation established by the Upgraded Final Safety Analysis Report (UFSAR) Lockheed Martin Idaho Technologies Company (LMITCO 1997a). Experiment handling activities are discussed herein.

  4. Modeling Constituent Redistribution in U-Pu-Zr Metallic Fuel Using the Advanced Fuel Performance Code BISON

    Energy Technology Data Exchange (ETDEWEB)

    Douglas Porter; Steve Hayes; Various

    2014-06-01

    The Advanced Fuels Campaign (AFC) metallic fuels currently being tested have higher zirconium and plutonium concentrations than those tested in the past in EBR reactors. Current metal fuel performance codes have limitations and deficiencies in predicting AFC fuel performance, particularly in the modeling of constituent distribution. No fully validated code exists due to sparse data and unknown modeling parameters. Our primary objective is to develop an initial analysis tool by incorporating state-of-the-art knowledge, constitutive models and properties of AFC metal fuels into the MOOSE/BISON (1) framework in order to analyze AFC metallic fuel tests.

  5. Studies of neutron methods for process control and criticality surveillance of fissile material processing facilities

    International Nuclear Information System (INIS)

    The development of radiochemical processes for fissile material processing and spent fuel handling need new control procedures enabling an improvement of plant throughput. This is strictly related to the implementation of continuous criticality control policy and developing reliable methods for monitoring the reactivity of radiochemical plant operations in presence of the process perturbations. Neutron methods seem to be applicable for fissile material control in some technological facilities. The measurement of epithermal neutron source multiplication with heuristic evaluation of measured data enables surveillance of anomalous reactivity enhancement leading to unsafe states. 80 refs., 47 figs., 33 tabs. (author)

  6. Advanced proton-exchange materials for energy efficient fuel cells.

    Energy Technology Data Exchange (ETDEWEB)

    Fujimoto, Cy H.; Grest, Gary Stephen; Hickner, Michael A.; Cornelius, Christopher James; Staiger, Chad Lynn; Hibbs, Michael R.

    2005-12-01

    The ''Advanced Proton-Exchange Materials for Energy Efficient Fuel Cells'' Laboratory Directed Research and Development (LDRD) project began in October 2002 and ended in September 2005. This LDRD was funded by the Energy Efficiency and Renewable Energy strategic business unit. The purpose of this LDRD was to initiate the fundamental research necessary for the development of a novel proton-exchange membranes (PEM) to overcome the material and performance limitations of the ''state of the art'' Nafion that is used in both hydrogen and methanol fuel cells. An atomistic modeling effort was added to this LDRD in order to establish a frame work between predicted morphology and observed PEM morphology in order to relate it to fuel cell performance. Significant progress was made in the area of PEM material design, development, and demonstration during this LDRD. A fundamental understanding involving the role of the structure of the PEM material as a function of sulfonic acid content, polymer topology, chemical composition, molecular weight, and electrode electrolyte ink development was demonstrated during this LDRD. PEM materials based upon random and block polyimides, polybenzimidazoles, and polyphenylenes were created and evaluated for improvements in proton conductivity, reduced swelling, reduced O{sub 2} and H{sub 2} permeability, and increased thermal stability. Results from this work reveal that the family of polyphenylenes potentially solves several technical challenges associated with obtaining a high temperature PEM membrane. Fuel cell relevant properties such as high proton conductivity (>120 mS/cm), good thermal stability, and mechanical robustness were demonstrated during this LDRD. This report summarizes the technical accomplishments and results of this LDRD.

  7. Underground autocatalytic-criticality potential and its implications to weapons fissile- material disposition

    International Nuclear Information System (INIS)

    Several options for weapons fissile-material disposition, such as once-through mixed- oxide (MOX) fuel in reactors or immobilisation in waste glass, would result in end products requiring geologic disposal. The criticality potential of the fissile end products containing U-235 and Pu-239 and the associated consequences in a geologic setting are important considerations for the final disposal of these materials. The possibility of underground criticality, and especially autocatalytic criticality, is affected by (1) groundwater leaking into a failed waste container, (2) preferential leaching of neutron absorbers or of fissile material from a failed container, and (3) preferential deposition of fissile material in the surrounding rock. Bowman and Venneri have pointed out that fissile material mixed with varying compositions of water and silica can undergo a nuclear chain reaction. Some configurations can become autocatalytically supercritical resulting in considerable energy release, terminated finally by disassembly. Some reviews rejected the Bowman and Venneri warning as implausible because of low probabilities of scenarios that could lead to such configurations. Sanchez et al. reported possible supercritical conditions in systems of Pu-SiO2-H2O and Pu-tuff-H2O but concluded that the probability of forming such combinations is extremely low. Kastenberg et al. studied the potential for autocatalytic criticality of plutonium or highly enriched uranium in the proposed Yucca Mountain geologic repository. They concluded that plutonium or uranium could, theoretically, become supercritical, but that such criticality is unlikely given the hydrology, geology and geochemistry of the Yucca Mountain site. These studies are not definitive. The possibility of criticality exists. Detailed mechanisms have not been sufficiently studied for clear conclusions on the probabilities of occurrence. More technical analysis is needed to understand the potential for underground autocatalytic

  8. State-of-the-art Report on Innovative Fuels for Advanced Nuclear Systems

    International Nuclear Information System (INIS)

    Development of innovative fuels such as homogeneous and heterogeneous fuels, ADS fuels, and oxide, metal, nitride and carbide fuels is an important stage in the implementation process of advanced nuclear systems. Several national and international R and D programmes are investigating minor actinide-bearing fuels due to their ability to help reduce the radiotoxicity of spent fuel and therefore decrease the burden on geological repositories. Minor actinides can be converted into a suitable fuel form for irradiation in reactor systems where they are transmuted into fission products with a significantly shorter half-life. This report compares recent studies of fuels containing minor actinides for use in advanced nuclear systems. The studies review different fuels for several types of advanced reactors by examining various technical issues associated with fabrication, characterisation, irradiation performance, design and safety criteria, as well as technical maturity. (authors)

  9. Shuffler instruments for the nondestructive assay of fissile materials

    International Nuclear Information System (INIS)

    A shuffler is a nondestructive assay instrument used to determine the fissile content of materials. It places an isotopic source of neutrons near the material to induce fissions, withdraws the source, and counts the delayed neutrons. The source is shuffled until a sufficient number of delayed neutrons have been counted. The shuffler technique is generally applied to difficult assay cases. The amount of material present may be very small (a few milligrams), and thus it does not spontaneously emit neutrons of consequence; the amount of material is also below an active well counter's level of sensitivity. On the other hand, the fissile amount may be fairly large, but the rate of spontaneously emitted neutrons may still be low (so a passive neutron count will not work) or the highest assay precision may be desired (favoring a shuffler over an active well counter) even if the material is inhomogeneous (making it difficult to interrogate with thermal neutrons). In all these cases, gamma-ray backgrounds, self- shielding, or matrix effects can make gamma-ray assays impractical. Materials ranging from highly radioactive spent-fuel assemblies to low-level waste drums have been assayed with shufflers, as have leached hulls, various process materials, scrap, and waste. This report presents a theoretical background for shufflers and describes techniques for practical applications. Procedures for assaying mixtures of fissile isotopes, inhomogeneous materials, and flowing liquids are discussed. It is shown how the precision and limits of detection of a shuffler can be calculated for a given neutron background rate. A section on data analysis gives a stepwise procedure for converting the measured counts into an assay value, including random, systematic, and total uncertainties. 31 refs

  10. Advanced Fuel Cycle Economic Tools, Algorithms, and Methodologies

    Energy Technology Data Exchange (ETDEWEB)

    David E. Shropshire

    2009-05-01

    The Advanced Fuel Cycle Initiative (AFCI) Systems Analysis supports engineering economic analyses and trade-studies, and requires a requisite reference cost basis to support adequate analysis rigor. In this regard, the AFCI program has created a reference set of economic documentation. The documentation consists of the “Advanced Fuel Cycle (AFC) Cost Basis” report (Shropshire, et al. 2007), “AFCI Economic Analysis” report, and the “AFCI Economic Tools, Algorithms, and Methodologies Report.” Together, these documents provide the reference cost basis, cost modeling basis, and methodologies needed to support AFCI economic analysis. The application of the reference cost data in the cost and econometric systems analysis models will be supported by this report. These methodologies include: the energy/environment/economic evaluation of nuclear technology penetration in the energy market—domestic and internationally—and impacts on AFCI facility deployment, uranium resource modeling to inform the front-end fuel cycle costs, facility first-of-a-kind to nth-of-a-kind learning with application to deployment of AFCI facilities, cost tradeoffs to meet nuclear non-proliferation requirements, and international nuclear facility supply/demand analysis. The economic analysis will be performed using two cost models. VISION.ECON will be used to evaluate and compare costs under dynamic conditions, consistent with the cases and analysis performed by the AFCI Systems Analysis team. Generation IV Excel Calculations of Nuclear Systems (G4-ECONS) will provide static (snapshot-in-time) cost analysis and will provide a check on the dynamic results. In future analysis, additional AFCI measures may be developed to show the value of AFCI in closing the fuel cycle. Comparisons can show AFCI in terms of reduced global proliferation (e.g., reduction in enrichment), greater sustainability through preservation of a natural resource (e.g., reduction in uranium ore depletion), value from

  11. Reactor physics ideas to design novel reactors with faster fissile growth

    International Nuclear Information System (INIS)

    There are several types of fission reactors operating in the world adopting generally the open fuel cycle which considers the naturally available fissile nuclide, viz., 235U. The accumulated discharged fuel is considered as waste in some countries. However the discharged fuel contains the precious man-made fissile plutonium which would provide the sole means of harnessing the nuclear energy from either depleted uranium or the natural thorium in future. It must be emphasized that the present day power reactors use just about 0.5% of the mined uranium and it would be imprudent to discard the rest of the mass as waste. It is therefore necessary to explore ways and means of exploiting the fertile mass which has the potential of providing the energy without the green house effects for millennia to come. This has to be done by innovating means of large scale fertile to fissile conversion and then using the man-made fissile material for sustenance as well as growth of fission nuclear power. This paper attempts to give a broad picture of the available options and the challenges in realizing the theoretical possibilities

  12. An Advanced Option for Sodium Cooled TRU Burner Loaded with Uranium-Free Fuels

    International Nuclear Information System (INIS)

    The sodium cooled fast reactors of this kind that are called burners are designed to have low conversion ratio by reducing fuel volume fraction or reducing neutron leakage or increasing neutron absorption. However, the typical SFR burners have a limited ability of TRU burning rate due to the fact that they use metallic or oxide fuels containing fertile nuclides such as 238U and 232Th and these fertile nuclides generate fissile nuclides through neutron capture even if they are designed to have low conversion ratio (e.g., 0.6). To further enhance the TRU burning rate, the removal of the fertile nuclides from the initial fuels is required and it will accelerate the reduction of TRUs that are accumulated in storages of LWR spent fuels. However, it has been well-known 4 that the removals of the fertile nuclides from the fuel degrade the inherent safety of the SFR burner cores through the significant decrease of the fuel Doppler effect, the increase of sodium void reactivity worth, and reduction of delayed neutron fraction. In this work, new option for the sodium cooled fast TRU burner cores loaded with fertile-free metallic fuels was proposed and the new cores were designed by using the suggested option. The cores were designed to enhance the inherent safety characteristics by using axially central absorber region and 6 or 12 ZrH1.8 moderator rods per fuel assembly. For each option, we considered two different types of fertile-free ternary metallic fuel (i.e., TRU-W-10Zr and TRU-Ni-10Zr). Also, we performed the BOR (Balance of Reactivity) analyses to show the self-controllability under ATWS as a measure of inherent safety. The core performance analysis showed that the new cores using axially central absorber region substantially improve the core performance parameters such as burnup reactivity swing and sodium void reactivity worth

  13. An Advanced Option for Sodium Cooled TRU Burner Loaded with Uranium-Free Fuels

    Energy Technology Data Exchange (ETDEWEB)

    You, WuSeung; Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of)

    2015-05-15

    The sodium cooled fast reactors of this kind that are called burners are designed to have low conversion ratio by reducing fuel volume fraction or reducing neutron leakage or increasing neutron absorption. However, the typical SFR burners have a limited ability of TRU burning rate due to the fact that they use metallic or oxide fuels containing fertile nuclides such as {sup 238}U and {sup 232}Th and these fertile nuclides generate fissile nuclides through neutron capture even if they are designed to have low conversion ratio (e.g., 0.6). To further enhance the TRU burning rate, the removal of the fertile nuclides from the initial fuels is required and it will accelerate the reduction of TRUs that are accumulated in storages of LWR spent fuels. However, it has been well-known 4 that the removals of the fertile nuclides from the fuel degrade the inherent safety of the SFR burner cores through the significant decrease of the fuel Doppler effect, the increase of sodium void reactivity worth, and reduction of delayed neutron fraction. In this work, new option for the sodium cooled fast TRU burner cores loaded with fertile-free metallic fuels was proposed and the new cores were designed by using the suggested option. The cores were designed to enhance the inherent safety characteristics by using axially central absorber region and 6 or 12 ZrH1.8 moderator rods per fuel assembly. For each option, we considered two different types of fertile-free ternary metallic fuel (i.e., TRU-W-10Zr and TRU-Ni-10Zr). Also, we performed the BOR (Balance of Reactivity) analyses to show the self-controllability under ATWS as a measure of inherent safety. The core performance analysis showed that the new cores using axially central absorber region substantially improve the core performance parameters such as burnup reactivity swing and sodium void reactivity worth.

  14. AECL programs in advanced systems research

    International Nuclear Information System (INIS)

    The AECL program in advanced systems research is directed in the long term to securing the option of obtaining fissile fuel by electronuclear breeding (accelerator breeder or fusion breeder) and to providing a basis from which AECL might move into stand alone fusion energy if warranted. In the short term the program is directed to reaping benefits from electronuclear technology. This report outlines the main activities and research facilities in both the long-term and short-term subprograms

  15. Canadian power reactor fuel

    International Nuclear Information System (INIS)

    The following subjects are covered: the basic CANDU fuel design, the history of the bundle design, the significant differences between CANDU and LWR fuel, bundle manufacture, fissile and structural materials and coolants used in the CANDU fuel program, fuel and material behaviour, and performance under irradiation, fuel physics and management, booster rods and reactivity mechanisms, fuel procurement, organization and industry, and fuel costs. (author)

  16. Advanced Diagnostics in Oxy-Fuel Combustion Processes

    OpenAIRE

    Brix, Jacob; Toftegaard, Maja Bøg; Clausen, Sønnik; Jensen, Anker Degn

    2012-01-01

    Denne rapport beskriver resultaterne, der er blevet opnået i PSO-projektet 010069, “Advanced Diagnostics in Oxy-Fuel Combustion Processes”. Tre områder af optisk diagnostik er inkluderet i rapporten: - FTIR målinger i en 30 kW swirlbrænder. - IR målinger i en 30 kW swirlbrænder. - IR målinger i en laboratorieskala fixed bed reaktor. Resultaterne, der blev opnået i swirlbrænderen, viser at FTIR teknikken er et værdifuldt værktøj til bestemmelse af gasfasetemperaturer. Når dens anvendelighed ev...

  17. Homogeneity survey of advanced spent fuel conditioning process hot cell

    International Nuclear Information System (INIS)

    The hot cell facility (ACPF) for research activities related to the advanced spent fuel conditioning process (ACP) is being constructed. The hot cell construction work will be finished in May, 2005. Hot cell is designed to permit safe handling of radioactive materials up to 1,385 TBq and to keep gamma and neutron dose-rate lower than the recommended ones. The dose-rate limit values following the Korean nuclear laws are 0.01 mSv/h at operation area and 0.15 mSv/h at maintenance area. The ACPF is a concrete structure with two rooms, and made its exterior walls of heavy concrete with density of 3.45 g/cm3 and the wall thickness is more than 90 cm

  18. Advanced Sphere-pac Fuel Design: Improving the Particle Fuel Performance with an Internally Cooled Pin

    International Nuclear Information System (INIS)

    Sphere-pac is a particle fuel type facilitating the remote production and being therefore designed for high minor actinide contents. The production of conventional pellet fuel contains many fabrication steps involving powder handling and mechanical processing. If a fuel is part of a combined light water and fast reactor cycle, it will most probably contain minor actinides and fission products, increasing the radiation dose of its reactants to a level, where a processing in hot-cells is necessary. In this case the conventional production route becomes difficult, as powder handling with the connected contamination risk in hot cells represents a major concern. Furthermore the maintenance of mechanical devices in hot cells is difficult and expensive. The suggested wet route by internal gelation is almost powder-less and requires only a solution mixing, drop generation plus heating and washing procedures. After those thermal treatment procedures follows which completes the production of the fuel particles. The paper describes some advances in the production method, plus new Sphere-pac designs for performance increase. (author)

  19. Advanced Fuels Campaign Light Water Reactor Accident Tolerant Fuel Performance Metrics Executive Summary

    Energy Technology Data Exchange (ETDEWEB)

    Shannon Bragg-Sitton

    2014-02-01

    Research and development (R&D) activities on advanced, higher performance Light Water Reactor (LWR) fuels have been ongoing for the last few years. Following the unfortunate March 2011 events at the Fukushima Nuclear Power Plant in Japan, the R&D shifted toward enhancing the accident tolerance of LWRs. Qualitative attributes for fuels with enhanced accident tolerance, such as improved reaction kinetics with steam resulting in slower hydrogen generation rate, provide guidance for the design and development of fuels and cladding with enhanced accident tolerance. A common set of technical metrics should be established to aid in the optimization and down selection of candidate designs on a more quantitative basis. “Metrics” describe a set of technical bases by which multiple concepts can be fairly evaluated against a common baseline and against one another. This report describes a proposed technical evaluation methodology that can be applied to evaluate the ability of each concept to meet performance and safety goals relative to the current UO2 – zirconium alloy system and relative to one another. The resultant ranked evaluation can then inform concept down-selection, such that the most promising accident tolerant fuel design option(s) can continue to be developed toward qualification.

  20. The advanced fuel cycle initiative: the future path for advanced spent fuel treatment and transmutation research in the United States

    International Nuclear Information System (INIS)

    The U. S. Department of Energy (DOE) has invested over USD 100 million in transmutation research and development over the past three years. The programme has evolved from an accelerator based transmutation programme to a multi-tier reactor and accelerator based programme. These changes have resulted in a significant re-focus of the research and development programme as well as a name change to reflect the new direction. The Advanced Accelerator Application (AAA) programme is now renamed the Advanced Fuel Cycle Initiative (AFCI). Research completed by the AAA programme in Fiscal Year 2002 points to a multi-phased AFCI Programme consisting of two elements that would be conducted in parallel as part of an integrated research effort: an intermediate-term technology element (AFCI Series One), which emphasises advanced technical enhancements to the current commercial nuclear power infrastructure; and a long term technology element (AFCI Series Two), which will require the introduction of next-generation nuclear energy systems to reduce the toxicity of nuclear waste. (author)

  1. Advanced operator interface design for CANDU-3 fuel handling system

    International Nuclear Information System (INIS)

    The Operator Interface for the CANDU 3 Fuel Handling (F/H) System incorporates several improvements over the existing designs. A functionally independent sit-down CRT (cathode-ray tube) based Control Console is provided for the Fuel Handling Operator in the Main Control Room. The Display System makes use of current technology and provides a user friendly operator interface. Regular and emergency control operations can be carried out from this control console. A stand-up control panel is provided as a back-up with limited functionality adequate to put the F/H System in a safe state in case of an unlikely non-availability of the Plant Display System or the F/H Control System'. The system design philosophy, hardware configuration and the advanced display system features are described in this paper The F/H Operator Interface System developed for CANDU 3 can be adapted to CANDU 9 as well as to the existing stations. (author)

  2. Advanced LWR Nuclear Fuel Cladding System Development Trade-Off Study

    Energy Technology Data Exchange (ETDEWEB)

    Kristine Barrett; Shannon Bragg-Sitton

    2012-09-01

    The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Research and Development (R&D) Pathway encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. To achieve significant operating improvements while remaining within safety boundaries, significant steps beyond incremental improvements in the current generation of nuclear fuel are required. Fundamental improvements are required in the areas of nuclear fuel composition, cladding integrity, and the fuel/cladding interaction to allow power uprates and increased fuel burn-up allowance while potentially improving safety margin through the adoption of an “accident tolerant” fuel system that would offer improved coping time under accident scenarios. With a development time of about 20 – 25 years, advanced fuel designs must be started today and proven in current reactors if future reactor designs are to be able to use them with confidence.

  3. Assessment of Startup Fuel Options for the GNEP Advanced Burner Reactor (ABR)

    Energy Technology Data Exchange (ETDEWEB)

    Jon Carmack (062056); Kemal O. Pasamehmetoglu (103171); David Alberstein

    2008-02-01

    The Global Nuclear Energy Program (GNEP) includes a program element for the development and construction of an advanced sodium cooled fast reactor to demonstrate the burning (transmutation) of significant quantities of minor actinides obtained from a separations process and fabricated into a transuranic bearing fuel assembly. To demonstrate and qualify transuranic (TRU) fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype is needed. The ABR would necessarily be started up using conventional metal alloy or oxide (U or U, Pu) fuel. Startup fuel is needed for the ABR for the first 2 to 4 core loads of fuel in the ABR. Following start up, a series of advanced TRU bearing fuel assemblies will be irradiated in qualification lead test assemblies in the ABR. There are multiple options for this startup fuel. This report provides a description of the possible startup fuel options as well as possible fabrication alternatives available to the program in the current domestic and international facilities and infrastructure.

  4. Japan's advanced reactor development and nuclear fuel policy

    International Nuclear Information System (INIS)

    That being the case, Japan has promoted development and utilization of nuclear energy supply structure in the face of its fragility of its energy supply structure in the face of its growing energy demand, but subject to the strict limitation of adherence to peaceful uses alone as stipulated in its Atomic Energy Basic Law. Furthermore, in order to make the most of limited uranium resources and at the same time solve the problem of appropriate treatment and disposal of reactivate waste from nuclear power generation, Japan has adopted nuclear fuel recycling, i. e. reprocessing of spent nuclear fuel for recovery of plutonium and other reusable components thereof for effective use as nuclear fuel, as one of the basic building blocks of its nuclear energy policy. As an advanced country in the field of peaceful uses of nuclear energy, Japan considers it important that it appropriately respond to growing demands for it to make an international contribution in that field, and that research and development and efforts to resolve common problems be based on international cooperation, and it intends to continue to play an international role in an active manner continue to play an international role in an active manner both in development and utilization of nuclear energy and in nuclear non-proliferation. In particular, concerning the Korean Peninsula Energy Development Organization (KEDO), Japan has high expectations that will function in such a way as to lead to relaxation of tension on the Korean such a way as to lead to relaxation of tension on the Korean Peninsula and more generally in Northeast Asia, and that its activities can be carried forward smoothly on the basis of cooperation among the countries concerned

  5. Enabling Advanced Modeling and Simulations for Fuel-Flexible Combustors

    Energy Technology Data Exchange (ETDEWEB)

    Heinz Pitsch

    2010-05-31

    The overall goal of the present project is to enable advanced modeling and simulations for the design and optimization of fuel-flexible turbine combustors. For this purpose we use a high-fidelity, extensively-tested large-eddy simulation (LES) code and state-of-the-art models for premixed/partially-premixed turbulent combustion developed in the PI's group. In the frame of the present project, these techniques are applied, assessed, and improved for hydrogen enriched premixed and partially premixed gas-turbine combustion. Our innovative approaches include a completely consistent description of flame propagation, a coupled progress variable/level set method to resolve the detailed flame structure, and incorporation of thermal-diffusion (non-unity Lewis number) effects. In addition, we have developed a general flamelet-type transformation holding in the limits of both non-premixed and premixed burning. As a result, a model for partially premixed combustion has been derived. The coupled progress variable/level method and the general flamelet tranformation were validated by LES of a lean-premixed low-swirl burner that has been studied experimentally at Lawrence Berkeley National Laboratory. The model is extended to include the non-unity Lewis number effects, which play a critical role in fuel-flexible combustor with high hydrogen content fuel. More specifically, a two-scalar model for lean hydrogen and hydrogen-enriched combustion is developed and validated against experimental and direct numerical simulation (DNS) data. Results are presented to emphasize the importance of non-unity Lewis number effects in the lean-premixed low-swirl burner of interest in this project. The proposed model gives improved results, which shows that the inclusion of the non-unity Lewis number effects is essential for accurate prediction of the lean-premixed low-swirl flame.

  6. Enabling Advanced Modeling and Simulations for Fuel-Flexible Combustors

    Energy Technology Data Exchange (ETDEWEB)

    Pitsch, Heinz

    2010-05-31

    The overall goal of the present project is to enable advanced modeling and simulations for the design and optimization of fuel-flexible turbine combustors. For this purpose we use a high fidelity, extensively-tested large-eddy simulation (LES) code and state-of-the-art models for premixed/partially-premixed turbulent combustion developed in the PI's group. In the frame of the present project, these techniques are applied, assessed, and improved for hydrogen enriched premixed and partially premixed gas-turbine combustion. Our innovative approaches include a completely consistent description of flame propagation; a coupled progress variable/level set method to resolve the detailed flame structure, and incorporation of thermal-diffusion (non-unity Lewis number) effects. In addition, we have developed a general flamelet-type transformation holding in the limits of both non-premixed and premixed burning. As a result, a model for partially premixed combustion has been derived. The coupled progress variable/level method and the general flamelet transformation were validated by LES of a lean-premixed low-swirl burner that has been studied experimentally at Lawrence Berkeley National Laboratory. The model is extended to include the non-unity Lewis number effects, which play a critical role in fuel-flexible combustor with high hydrogen content fuel. More specifically, a two-scalar model for lean hydrogen and hydrogen-enriched combustion is developed and validated against experimental and direct numerical simulation (DNS) data. Results are presented to emphasize the importance of non-unity Lewis number effects in the lean-premixed low-swirl burner of interest in this project. The proposed model gives improved results, which shows that the inclusion of the non-unity Lewis number effects is essential for accurate prediction of the lean-premixed low-swirl flame.

  7. Nuclear archaeology: Verifying declarations of fissile-material production

    International Nuclear Information System (INIS)

    Controlling the production of fissile material is an essential element of nonproliferation policy. Similarly, accounting for the past production of fissile material should be an important component of nuclear disarmament. This paper describes two promising techniques that make use of physical evidence at reactors and enrichment facilities to verify the past production of plutonium and highly enriched uranium. In the first technique, the concentrations of long-lived radionuclides in permanent components of the reactor core are used to estimate the neutron fluence in various regions of the reactor, and thereby verify declarations of plutonium production in the reactor. In the second technique, the ratio of the concentration of U-235 to that of U-234 in the tails is used to determine whether a given container of tails was used in the production of low- enriched uranium, which is suitable for reactor fuel, or highly enriched uranium, which can be used in nuclear weapons. Both techniques belong to the new field of open-quotes nuclear archaeology,close quotes in which the authors attempt to document past nuclear weapons activities and thereby lay a firm foundation for verifiable nuclear disarmament. 11 refs., 1 fig., 3 tabs

  8. The trafficking control of fissile and radioactive materials in Kazakhstan

    International Nuclear Information System (INIS)

    After disintegration of the USSR the problem of control for transfers of radioactive and fissile materials has getting more and more urgent in sovereign Kazakhstan. Fissile materials, which theoretically could be used to creation of nuclear weapons, were available at the following enterprises of Kazakhstan: The Institute of Nuclear Physics - National Nuclear Center - Almaty The Institute of Atomic Energy - National Nuclear Center - Kurchatov Mangyshlak Atomic Energy Combine - Aktau Ulbinskiy Metallurgical Plant - Ust-Kamenogorsk At last time mass-media have reported many incidents of a new type of theft. Individuals are detained, trying to sell fuel pellets made of enriched uranium, radioactive contaminated cables, or various components of technological purposes from the enterprises. A significant problem is the legal and illegal collection of ferrous and non-ferrous metals. This made at places of realization of nuclear tests (former Semipalatinsk nuclear test-site (STS), Azgir, Lyra and others) and at working and closed uranium mines. Nominally, the incidents can be divided into three groups: - heft of materials, hazardous in term of nonproliferation; - heft of materials, non-hazardous in terms of nonproliferation (radioactive scrap metal, sealed radioactive sources); - heft due to an inefficient system of physical protection (the tank for transportation of radioactive waste products at Institute of Atomic Energy of NNC RK)

  9. ASGARD - Advanced fuelS for Generation IV reActors: Reprocessing and Dissolution

    International Nuclear Information System (INIS)

    Conclusion: Thanks to its interdiciplinary nature ASGARD has created a common platform for many aspects of novel nuclear fuel cycles, 25% into the project everything is running according to plan with significant advances in most domains. The training and education scheme used in ASGARD has already been successfully implemented allowing young scientists in the field to present their results internationally and also visit other ASGARD labs. The future collaboration with e.g. SACESS and CINCH II will enable the creation of significant added value to the communities involved. More will come. We have only begun.....

  10. Science based integrated approach to advanced nuclear fuel development - vision, approach, and overview

    Energy Technology Data Exchange (ETDEWEB)

    Unal, Cetin [Los Alamos National Laboratory; Pasamehmetoglu, Kemal [IDAHO NATIONAL LAB; Carmack, Jon [IDAHO NATIONAL LAB

    2010-01-01

    Advancing the performance of Light Water Reactors, Advanced Nuclear Fuel Cycles, and Advanced Rcactors, such as the Next Generation Nuclear Power Plants, requires enhancing our fundamental understanding of fuel and materials behavior under irradiation. The capability to accurately model the nuclear fuel systems is critical. In order to understand specific aspects of the nuclear fuel, fully coupled fuel simulation codes are required to achieve licensing of specific nuclear fuel designs for operation. The backbone of these codes, models, and simulations is a fundamental understanding and predictive capability for simulating the phase and microstructural behavior of the nuclear fuel system materials and matrices. The purpose of this paper is to identify the modeling and simulation approach in order to deliver predictive tools for advanced fuels development. The coordination between experimental nuclear fuel design, development technical experts, and computational fuel modeling and simulation technical experts is a critical aspect of the approach and naturally leads to an integrated, goal-oriented science-based R & D approach and strengthens both the experimental and computational efforts. The Advanced Fuels Campaign (AFC) and Nuclear Energy Advanced Modeling and Simulation (NEAMS) Fuels Integrated Performance and Safety Code (IPSC) are working together to determine experimental data and modeling needs. The primary objective of the NEAMS fuels IPSC project is to deliver a coupled, three-dimensional, predictive computational platform for modeling the fabrication and both normal and abnormal operation of nuclear fuel pins and assemblies, applicable to both existing and future reactor fuel designs. The science based program is pursuing the development of an integrated multi-scale and multi-physics modeling and simulation platform for nuclear fuels. This overview paper discusses the vision, goals and approaches how to develop and implement the new approach.

  11. A contingency safe, responsible, economic, increased capacity spent nuclear fuel (SNF) advance fuel cycle

    International Nuclear Information System (INIS)

    The purpose of this paper is to have an Advanced Light Water (LWR) fuel cycle and an associated development program to provide a contingency plan to the current DOE effort to license once-through spent Light Water Reactor (LWR) fuel for disposition at Yucca Mountain (YM). The intent is to fully support the forthcoming June 2008 DOE submittal to the Nuclear Regulatory Commission (NRC) based upon the latest DOE draft DOE/EIS-0250F-SID dated October 2007 which shows that the latest DOE YM doses would readily satisfy the anticipated NRC and Environmental Protection Agency (EP) standards. The proposed Advance Fuel Cycle can offer potential resolution of obstacles that might arise during the NRC review and, particularly, during the final hearings process to be held in Nevada. Another reason for the proposed concept is that a substantial capacity growth of the YM repository will be necessary to accommodate the SNF of Advance Light Water Reactors (ALWRs) currently under consideration for United States (U.S.) electricity production (1) and the results of the recently issued study by the Electric Power Research Institute (EPRI) to reduce CO2 emissions (2). That study predicts that by 2030 U.S. nuclear power generation would grow by 64 Gigawatt electrical (GWe) and account for 25.5 percent of the overall U.S. electrical generation. The current annual SNF once-through fuel cycle accumulation would rise from 2000-2100 MT (Metric Tons) to about 3480 MT in 2030 and the total SNF inventory, would reach nearly 500,000 MT by 2100 if U. S. nuclear power continues to grow at 1.1 percent per year after 2030. That last projection does not account for any SNF reduction due to increased fuel burnup or any increased capacity needed 'to establish supply Global Nuclear Energy Partnership (GNEP,) arrangements among nations to provide nuclear fuel and taking back spent fuel for recycling without spreading enrichment and reprocessing technologies' (3). The anticipated capacity of 120 MT planned

  12. Generic Repository Concepts and Thermal Analysis for Advanced Fuel Cycles

    International Nuclear Information System (INIS)

    The current posture of the used nuclear fuel management program in the U.S. following termination of the Yucca Mountain Project, is to pursue research and development (R and D) of generic (i.e., non-site specific) technologies for storage, transportation and disposal. Disposal R and D is directed toward understanding and demonstrating the performance of reference geologic disposal concepts selected to represent the current state-of-the-art in geologic disposal. One of the principal constraints on waste packaging and emplacement in a geologic repository is management of the waste-generated heat. This paper describes the selection of reference disposal concepts, and thermal management strategies for waste from advanced fuel cycles. A geologic disposal concept for spent nuclear fuel (SNF) or high-level waste (HLW) consists of three components: waste inventory, geologic setting, and concept of operations. A set of reference geologic disposal concepts has been developed by the U.S. Department of Energy (DOE) Used Fuel Disposition Campaign, for crystalline rock, clay/shale, bedded salt, and deep borehole (crystalline basement) geologic settings. We performed thermal analysis of these concepts using waste inventory cases representing a range of advanced fuel cycles. Concepts of operation consisting of emplacement mode, repository layout, and engineered barrier descriptions, were selected based on international progress and previous experience in the U.S. repository program. All of the disposal concepts selected for this study use enclosed emplacement modes, whereby waste packages are in direct contact with encapsulating engineered or natural materials. The encapsulating materials (typically clay-based or rock salt) have low intrinsic permeability and plastic rheology that closes voids so that low permeability is maintained. Uniformly low permeability also contributes to chemically reducing conditions common in soft clay, shale, and salt formations. Enclosed modes are

  13. Development of advanced spent fuel management process. The fabrication and oxidation behavior of simulated metallized spent fuel

    International Nuclear Information System (INIS)

    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

  14. System study of CANDU/LWR synergy in advanced nuclear fuel cycles

    International Nuclear Information System (INIS)

    This report proposes a study that will evaluate the effects of advanced nuclear fuel cycles on resource utilisation, repository capacity, waste streams, economics, and proliferation resistance. The proposed fuel cycles are designed to exploit the unique synergy that exists between light water and CANDU reactors. Also, several fuel cycle simulation codes have been proposed to be used. (author)

  15. 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.

  16. Current and future research on passive safety devices and advanced fuels for innovative LWRs

    International Nuclear Information System (INIS)

    Presently an enrichment limit of 5% 235U is fixed in licenses all along the fuel cycle. Although working groups (e.g. DOE, WNA) are investigating incentives and prerequisites to increase the limit, it is widely expected that at least for the mid term (10-20 years) the limit will hold. For the EPR and SWR 1000 development the 5% enrichment limit was not a serious restriction, since for higher enrichment each fuel assembly has to be shimmed with fixed burnable absorbers anyway for criticality or reactivity reasons. Thus the fuel safety issues for these ALWR are the same as for currently operating LWR, essentially the influence of high burn-up on safety analyses (fuel response to transients e.g. RIA, LOCA). Operational behavior is of minor concern since the burn-up increase is slow enough to allow sufficient experience feedback. Worldwide the LWR fuel failure rates have not increased with burn-up. At present, and expected mid term, uranium prices are such that there is little incentive to look into other fissile materials like thorium. Nevertheless, on a small scale, R and D is ongoing to prepare a longer term alternative. Some R and D is also considered for the end of NPP lifetime fuel cycles, however there is little economic incentive for such (due to lifetime extension) rarely needed specific solutions. If nuclear fission energy is to be used in the real long term (beyond 50-100 years), then R and D on breeder reactor fuel will have to be resumed. (authors)

  17. Masters Study in Advanced Energy and Fuels Management

    Energy Technology Data Exchange (ETDEWEB)

    Mondal, Kanchan [Southern Illinois Univ., Carbondale, IL (United States)

    2014-12-08

    There are currently three key drivers for the US energy sector a) increasing energy demand and b) environmental stewardship in energy production for sustainability and c) general public and governmental desire for domestic resources. These drivers are also true for energy nation globally. As a result, this sector is rapidly diversifying to alternate sources that would supplement or replace fossil fuels. These changes have created a need for a highly trained workforce with a the understanding of both conventional and emerging energy resources and technology to lead and facilitate the reinvention of the US energy production, rational deployment of alternate energy technologies based on scientific and business criteria while invigorating the overall economy. In addition, the current trends focus on the the need of Science, Technology, Engineering and Math (STEM) graduate education to move beyond academia and be more responsive to the workforce needs of businesses and the industry. The SIUC PSM in Advanced Energy and Fuels Management (AEFM) program was developed in response to the industries stated need for employees who combine technical competencies and workforce skills similar to all PSM degree programs. The SIUC AEFM program was designed to provide the STEM graduates with advanced technical training in energy resources and technology while simultaneously equipping them with the business management skills required by professional employers in the energy sector. Technical training include core skills in energy resources, technology and management for both conventional and emerging energy technologies. Business skills training include financial, personnel and project management. A capstone internship is also built into the program to train students such that they are acclimatized to the real world scenarios in research laboratories, in energy companies and in government agencies. The current curriculum in the SIUC AEFM will help fill the need for training both recent

  18. Foresight Study on Advanced Conversion Technologies of Fossil Fuels

    International Nuclear Information System (INIS)

    The Observatorio de Prospectiva Tecnologica Industrial (OPTI) is a Foundation supported by the Ministry of Industry and Energy, (MINER) and has as main objective to provide a basic information and knowledge on technology evolution. This information will be accessible to the Administration and to the Companies and can be taking into account in planning and decision making of technology policies. Ciemat is member of OPTI and is the organism in charge of the actions in the Energy sector. CIEMAT has the responsibility on the realisation of the sector studies to get in three years (1998 to 2001) a foresight vision of the critical technology topics. The OPTI integrated strategic plan undertake the analysis of other seven technology sectors, with the same criteria on methodological aspects. Delphi method was used for the realization of the studies. It consisted of a survey conducted in two rounds using a questionnaire to check the experts opinion. The time frame of the studies was defined from 1999 to 2015. The study presented in this document has been performed by CIEMAT in the second stage of the OPTI activities. The main goal behind this study is to identify the advanced clean and efficient technologies for the conversion of fossil fuels to promote in our country. The questionnaire was addressed to 250 experts and the response rate was about the 37%, ratifying the final results. The spanish position and the barriers for the development of each technology has been determined and also the recommended measures to facilitate their performance in the future. This basic information is consider of main interest, taking in account the actual energetic situation with a foreseeable demand increase and fossil fuels dependence. (Author) 17 refs

  19. Use of advanced simulations in fuel performance codes

    International Nuclear Information System (INIS)

    The simulation of the cylindrical fuel rod behaviour in a reactor or a storage pool for spent fuel requires a fuel performance code. Such tool solves the equations for the heat transfer, the stresses and strains in fuel and cladding, the evolution of several isotopes and the behaviour of various fission products in the fuel rod. The main equations along with their limitations are briefly described. The current approaches adopted for overcoming these limitations and the perspectives are also outlined. (author)

  20. Irradiation tests of advanced plutonium-bearing fuels

    International Nuclear Information System (INIS)

    The capsule irradiation tests of uranium-plutonium mixed carbide and nitride fuels are under way in Japan Materials Testing Reactor (JMTR) for evaluating the fuel performance. The fuel pins containing these fuels, helium bonded to stainless steel cladding, have been irradiated in the NaK bonded capsules with a double metal containment in which thermocouples are installed. The present paper summarizes the results of the development in irradiation technology of these plutonium-bearing fuels utilizing JMTR. (author)

  1. Performance Evaluation of Metallic Dispersion Fuel for Advanced Research Reactors

    International Nuclear Information System (INIS)

    Uranium alloys with a high uranium density has been developed for high power research reactor fuel using low-enriched uranium (LEU). U-Mo alloys have been developed as candidate fuel material because of excellent irradiation behavior. Irradiation behavior of U-Mo/Al dispersion fuel has been investigated to develop high performance research reactor fuel as RERTR international research program. While plate-type and rod-type dispersion fuel elements are used for research reactors, HANARO uses rod-type dispersion fuel elements. PLATE code is developed by Argonne National Laboratory for the performance evaluation of plate-type dispersion fuel, but there is no counterpart for rod-type dispersion fuel. Especially, thermal conductivity of fuel meat decreases during the irradiation mainly because of interaction layer formation at the interface between the U-Mo fuel particle and Al matrix. The thermal conductivity of the interaction layer is not as high as the Al matrix. The growth of interaction layer is interactively affected by the temperature of fuel because it is associated with a diffusion reaction which is a thermally activated process. It is difficult to estimate the temperature profile during irradiation test due to the interdependency of fuel temperature and thermal conductivity changed by interaction layer growth. In this study, fuel performance of rod-type U-Mo/Al dispersion fuels during irradiation tests were estimated by considering the effect of interaction layer growth on the thermal conductivity of fuel meat

  2. The R and D program in support of advanced fuel cycles for CANDU

    International Nuclear Information System (INIS)

    Advanced fuel cycles for CANDU reactors are well on their way to being implemented. The first step is slightly enriched uranium (SEU) which is economical today. A new fuel bundle is seen as the vehicle for all fuels in CANDU. CANDU fuel fabricated from uranium recovered from fuel discharged from light-water reactors (LWR) is also economical today and readily achievable technically. Future fuel cycles would utilize plutonium recovered from light-water reactors or CANDUs and eventually thorium. R and D in support of these cycles focuses on those topics that require a high degree of confidence in their implementation such as fuel fabrication and defect-free performance to high burnup. Reactor physics codes and nuclear data for advanced fuel cycles will be validated against experiments. (author).

  3. The R and D program in support of advanced fuel cycles for CANDU

    International Nuclear Information System (INIS)

    Advanced fuel cycles for CANDU reactors are well on their way to being implemented. The first step is slightly enriched uranium (SEU) which is economical today. A new fuel bundle is seen as the vehicle for all fuels in CANDU. CANDU fuel fabricated from uranium recovered from fuel discharged from light-water reactors (LWR) is also economical today and readily achievable technically. Future fuel cycles would utilize plutonium recovered from light-water reactors or CANDU's and eventually thorium. R and D in support of these cycles focuses on those topics that require a high degree of confidence in their implementation such as fuel fabrication and defect-free performance to high burnup. Reactor physics codes and nuclear data for advanced fuel cycles will be validated against experiments. (author). 8 refs

  4. US-RERTR advanced fuel development plans: 1999

    International Nuclear Information System (INIS)

    Twelve fuel alloys were included in the very-high-density RERTR-1 and RERTR-2 microplate irradiation experiments. Experience gained fabrication and results from the post-irradiation examination of these fuels has allowed us to narrow the focus of our fuel development efforts in preparation for the next set of irradiation experiments. Specific technical problems in both the areas of fuel fabrication and irradiation performance remain to be addressed. Examples of these are powder fabrication, fuel phase gamma stability versus density, and fuel-matrix interaction. In order to more efficiently address metal alloy fuel performance issues work will continue on establishing a theoretical basis for alloy dispersion fuel irradiation performance. Plans to address these fuel development issues in the coming year be presented. (author)

  5. Thorium fueled high temperature gas cooled reactors. An assessment

    International Nuclear Information System (INIS)

    The use of thorium as a fertile fuel for the High Temperature Gas Cooled Reactor (HTR) instead of uranium has been reviewed. It has been concluded that the use of thorium might be beneficial to reduce the actinide waste production. To achieve a real advancement, the uranium of the spent fuel has to be recycled and the requested make-up fissile material for the fresh fuel has to be used in the form of highly-enriched uranium. A self-sustaining fuel cycle may be possible in the HTR of large core size, but this could reduce the inherent safety features of the design. (orig.)

  6. Advancement in fuel design and manufacturing technologies of NFI

    International Nuclear Information System (INIS)

    NFI has two works manufacturing PWR fuel, BWR fuel and other types of fuel with a total capacity of 465 ton-U/year and has employees approx. 500. NFI is appreciated as a unique company for supplying both PWR fuel and BWR fuel with excelent quality and carrying out R and D work covering a wide range of areas. NFI has had no leak problem except tow occasions both caused by baffle flow in PWRs. As to BWR fuel, approx. 1,000 bundles have been manufactured by NFI and used to date and not a single case of failure has occurred. (author)

  7. High-Level Functional and Operational Requirements for the Advanced Fuel Cycle Facility

    International Nuclear Information System (INIS)

    This document describes the principal functional and operational requirements for the proposed Advanced Fuel Cycle Facility (AFCF). The AFCF is intended to be the world's foremost facility for nuclear fuel cycle research, technology development, and demonstration. The facility will also support the near-term mission to develop and demonstrate technology in support of fuel cycle needs identified by industry, and the long-term mission to retain and retain U.S. leadership in fuel cycle operations. The AFCF is essential to demonstrate a more proliferation-resistant fuel cycle and make long-term improvements in fuel cycle effectiveness, performance and economy

  8. A review on the development of the advanced fuel fabrication technology

    International Nuclear Information System (INIS)

    In this state-of art report, the development status of the advanced nuclear fuel was investigated. The current fabrication technology for coated particle fuel and non-oxide fuel such as sol-gel technology, coating technology, and carbothermic reduction reaction has also been examined. In the view point of inherent safety and efficiency in the operation of power plant, the coated particle fuel will keep going on its reputation as nuclear fuel for a high temperature gas cooled reactor, and the nitride fuel is very prospective for the next liquid metal fast breeder reactor. 43 figs., 17 tabs., 96 refs. (Author)

  9. An advanced oxy-fuel power cycle with high efficiency

    Energy Technology Data Exchange (ETDEWEB)

    Gou, C.; Hong, H. [University of Science and Technology, Beijing (China). Mechanical School; Cai, R. [Chinese Academy of Sciences, Beijing (China). Institute of Engineering Thermophysics

    2006-07-01

    In this article, an innovative oxy-fuel power cycle is proposed as a promising CO{sub 2} emission mitigation solution. It includes two cases with different characteristics in the cycle configuration. Case 1 basically consists of a water steam Rankine cycle and a steam-CO{sub 2} recuperative-reheat cycle. Case 2 integrates some characteristics of Case 1 and a top Brayton cycle. The thermodynamic performances for the design conditions of these two cases were analysed using the advanced process simulator Aspen Plus and the results are given in detail. The corresponding exergy loss analyses were carried out to gain an understanding of the loss distribution. The MATIANT cycle, the CES cycle, and the GRAZ cycle were also evaluated as references. The results demonstrate that the proposed cycle has notable advantages in thermal efficiency, specific work, and technical feasibility compared with the reference cycles. For example, the thermal efficiency of Case 2 is 6.58 percentage points higher than that of the MATIANT cycle. (author)

  10. A Regional Advanced Fuel Cycle Partnership in Southeast Asia

    International Nuclear Information System (INIS)

    economic analysis was performed under the assumption that participating countries paid country housing each fuel cycle facility for services performed on materials burned in their reactors. The analysis revealed that the collaboration scenario would provide cost savings to every participant as compared to a once-through approach, excepting Japan. The cost borne by Japan reflects the construction of 31 GWe of fast burner reactor capacity to accomplish the consumption of transuranics produced by the participating countries. The cost premium for Japan relative to a once-through approach is about 10%. Therefore, equitable implementation of the strategy would require other participating nations to provide Japan with an additional service fee to support the cost burden associated with deployment of a large number of advanced reactors. [1] Nuclear Power in China, http://www.worldnuclear.org /info/inf63.html, September 2008. (authors)

  11. Recent Advances in Enzymatic Fuel Cells: Experiments and Modeling

    Directory of Open Access Journals (Sweden)

    Ivan Ivanov

    2010-04-01

    Full Text Available Enzymatic fuel cells convert the chemical energy of biofuels into electrical energy. Unlike traditional fuel cell types, which are mainly based on metal catalysts, the enzymatic fuel cells employ enzymes as catalysts. This fuel cell type can be used as an implantable power source for a variety of medical devices used in modern medicine to administer drugs, treat ailments and monitor bodily functions. Some advantages in comparison to conventional fuel cells include a simple fuel cell design and lower cost of the main fuel cell components, however they suffer from severe kinetic limitations mainly due to inefficiency in electron transfer between the enzyme and the electrode surface. In this review article, the major research activities concerned with the enzymatic fuel cells (anode and cathode development, system design, modeling by highlighting the current problems (low cell voltage, low current density, stability will be presented.

  12. On-Going Comparison of Advanced Fuel Cycle Options

    International Nuclear Information System (INIS)

    This paper summarizes the current comprehensive comparison of four major fuel cycle strategies: once-through, thermal recycle, thermal+fast recycle, fast recycle. It then proceeds to summarize comparison of the major technology options for the key elements of the fuel cycle that can implement each of the four strategies - separation processing, transmutation reactors, and fuels

  13. Recent advances in nuclear fuels technology for thermal reactors

    International Nuclear Information System (INIS)

    In today's competitive electrical generation, many nuclear power generators are lowering operating and fuel cycle costs by extending burnups, utilizing longer cycles, reducing outage duration, increasing peaking factors for more efficient fuel management; and by up rating to maximize energy output from the reactors. To better equip nuclear operators to meet these competitive challenges, Westinghouse has strategically aligned its goals to ensure that customer needs are met and that fuel supplied operates flawlessly. Westinghouse's fuel performance program implements design features and manufacturing processes to maximize margins to failure, specify bounds of reactor operation, and monitor critical operating parameters using BEACON software as well as specify and implement a robust Post Irradiation Examination (PIE) to obtain early feedback on fuel performance. Westinghouse's unwavering commitment to achieve flawless fuel performance and to innovate resulted in exceptional pressurized water reactor (PWR), boiling water reactor (BWR), and VVER fuel performance worldwide. This paper covers decades of continuous innovation in fuel design and manufacturing process which supports our outstanding fuel performance in all LWR fuel types. This paper also includes information about Westinghouse's state-of-the-art tools and methodologies utilized to improve fuel performance as well as recent developments in fuel cladding material. (author)

  14. Alternative Fuel and Advanced Technology Commercial Lawn Equipment (Spanish version); Clean Cities, Energy Efficiency & Renewable Energy (EERE)

    Energy Technology Data Exchange (ETDEWEB)

    Nelson, Erik

    2015-06-01

    Powering commercial lawn equipment with alternative fuels or advanced engine technology is an effective way to reduce U.S. dependence on petroleum, reduce harmful emissions, and lessen the environmental impacts of commercial lawn mowing. Numerous alternative fuel and fuel-efficient advanced technology mowers are available. Owners turn to these mowers because they may save on fuel and maintenance costs, extend mower life, reduce fuel spillage and fuel theft, and demonstrate their commitment to sustainability.

  15. Screening of advanced cladding materials and UN–U3Si5 fuel

    International Nuclear Information System (INIS)

    Highlights: • Screening methodology for advanced fuel and cladding. • Cladding candidates, except for silicon carbide, exhibit reactivity penalty versus zirconium alloy. • UN–U3Si5 fuels have the potential to exhibit reactor physics and fuel management performance similar to UO2. • Harder spectrum in the UN ceramic composite fuel increases transuranic build-up. • Fuel and cladding properties assumed in these assessments are preliminary. - Abstract: In the aftermath of Fukushima, a focus of the DOE-NE Advanced Fuels Campaign has been the development of advanced nuclear fuel and cladding options with the potential for improved performance in an accident. Uranium dioxide (UO2) fuels with various advanced cladding materials were analyzed to provide a reference for cladding performance impacts. For advanced cladding options with UO2 fuel, most of the cladding materials have some reactivity and discharge burn-up penalty (in GWd/t). Silicon carbide is one exception in that the reactor physics performance is predicted to be very similar to zirconium alloy cladding. Most candidate claddings performed similar to UO2–Zr fuel–cladding in terms of safety coefficients. The clear exception is that Mo-based materials were identified as potentially challenging from a reactor physics perspective due to high resonance absorption. This paper also includes evaluation of UN–U3Si5 fuels with Kanthal AF or APMT cladding. The objective of the U3Si5 phase in the UN–U3Si5 fuel concept is to shield the nitride phase from water. It was shown that UN–U3Si5 fuels with Kanthal AF or APMT cladding have similar reactor physics and fuel management performance over a wide parameter space of phase fractions when compared to UO2–Zr fuel–cladding. There will be a marginal penalty in discharge burn-up (in GWd/t) and the sensitivity to 14N content in UN ceramic composites is high. Analysis of the rim effect due to self-shielding in the fuel shows that the UN-based ceramic fuels

  16. Recent anode advances in solid oxide fuel cells

    Science.gov (United States)

    Sun, Chunwen; Stimming, Ulrich

    Solid oxide fuel cells (SOFCs) are electrochemical reactors that can directly convert the chemical energy of a fuel gas into electrical energy with high efficiency and in an environment-friendly way. The recent trends in the research of solid oxide fuel cells concern the use of available hydrocarbon fuels, such as natural gas. The most commonly used anode material Ni/YSZ cermet exhibits some disadvantages when hydrocarbons were used as fuels. Thus it is necessary to develop alternative anode materials which display mixed conductivity under fuel conditions. This article reviews the recent developments of anode in SOFCs with principal emphasis on the material aspects. In addition, the mechanism and kinetics of fuel oxidation reactions are also addressed. Various processes used for the cost-effective fabrication of anode have also been summarized. Finally, this review will be concluded with personal perspectives on the future research directions of this area.

  17. Feasibility and Safety Assessment for Advanced Reactor Concepts Using Vented Fuel

    International Nuclear Information System (INIS)

    Recent interest in fast reactor technology has led to renewed analysis of past reactor concepts such as Gas Fast Reactors and Sodium Fast Reactors. In an effort to make these reactors more economic, the fuel is required to stay in the reactor for extended periods of time; the longer the fuel stays within the core, the more fertile material is converted into usable fissile material. However, as burnup of the fuel-rod increases, so does the internal pressure buildup due to gaseous fission products. In order to reach the 30 year lifetime requirements of some reactor designs, the fuel pins must have a vented-type design to allow the buildup of fission products to escape. The present work aims to progress the understanding of the feasibility and safety issues related to gas reactors that incorporate vented fuel. The work was separated into three different work-scopes: 1. Quantitatively determine fission gas release from uranium carbide in a representative helium cooled fast reactor; 2. Model the fission gas behavior, transport, and collection in a Fission Product Vent System; and, 3. Perform a safety analysis of the Fission Product Vent System. Each task relied on results from the previous task, culminating in a limited scope Probabilistic Risk Assessment (PRA) of the Fission Product Vent System. Within each task, many key parameters lack the fidelity needed for comprehensive or accurate analysis. In the process of completing each task, the data or methods that were lacking were identified and compiled in a Gap Analysis included at the end of the report.

  18. Feasibility and Safety Assessment for Advanced Reactor Concepts Using Vented Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Klein, Andrew [Oregon State Univ., Corvallis, OR (United States). Nuclear Engineering and Radiation Health Physics; Matthews, Topher [Oregon State Univ., Corvallis, OR (United States); Lenhof, Renae [Oregon State Univ., Corvallis, OR (United States); Deason, Wesley [Oregon State Univ., Corvallis, OR (United States); Harter, Jackson [Oregon State Univ., Corvallis, OR (United States)

    2015-01-16

    Recent interest in fast reactor technology has led to renewed analysis of past reactor concepts such as Gas Fast Reactors and Sodium Fast Reactors. In an effort to make these reactors more economic, the fuel is required to stay in the reactor for extended periods of time; the longer the fuel stays within the core, the more fertile material is converted into usable fissile material. However, as burnup of the fuel-rod increases, so does the internal pressure buildup due to gaseous fission products. In order to reach the 30 year lifetime requirements of some reactor designs, the fuel pins must have a vented-type design to allow the buildup of fission products to escape. The present work aims to progress the understanding of the feasibility and safety issues related to gas reactors that incorporate vented fuel. The work was separated into three different work-scopes: 1. Quantitatively determine fission gas release from uranium carbide in a representative helium cooled fast reactor; 2. Model the fission gas behavior, transport, and collection in a Fission Product Vent System; and, 3. Perform a safety analysis of the Fission Product Vent System. Each task relied on results from the previous task, culminating in a limited scope Probabilistic Risk Assessment (PRA) of the Fission Product Vent System. Within each task, many key parameters lack the fidelity needed for comprehensive or accurate analysis. In the process of completing each task, the data or methods that were lacking were identified and compiled in a Gap Analysis included at the end of the report.

  19. Method and device for fabricating dispersion fuel comprising fission product collection spaces

    Energy Technology Data Exchange (ETDEWEB)

    Shaber, Eric L; Fielding, Randall S

    2015-05-05

    A method of fabricating a nuclear fuel comprising a fissile material, one or more hollow microballoons, a phenolic resin, and metal matrix. The fissile material, phenolic resin and the one or more hollow microballoons are combined. The combined fissile material, phenolic resin and the hollow microballoons are heated sufficiently to form at least some fissile material carbides creating a nuclear fuel particle. The resulting nuclear fuel particle comprises one or more fission product collection spaces. In a preferred embodiment, the fissile material, phenolic resin and the one or more hollow microballoons are combined by forming the fissile material into microspheres. The fissile material microspheres are then overcoated with the phenolic resin and microballoon. In another preferred embodiment, the fissile material, phenolic resin and the one or more hollow microballoons are combined by overcoating the microballoon with the fissile material, and phenolic resin.

  20. THE FISSILE MATERIAL TRANSPARENCY TECHNOLOGY DEMONSTRATION (FMTTD)

    International Nuclear Information System (INIS)

    The United States Department of Defense, Defense Threat Reduction Agency Cooperative Threat Reduction program is supporting the construction of a fissile material storage facility at Mayak in the Russian Federation. Up to 34 tons of weapon-grade plutonium will be stored in the facility to await disposition. In order to meet arms control and nonproliferation objectives, the U.S. Congress has requested assurances that the nuclear material stored at the Mayak facility is derived from dismantled nuclear weapons. The usual approach to identify the origin or state of radioactive materials is to measure the intensity and energy of neutron and gamma radiation emitted. However, the Russian Federation considers such details as isotopic composition and mass to be classified. The solution arrived at by a DOE multilaboratory team is to place the radioactive specimen, the gamma and neutron counters, and all the computational equipment behind an information barrier. In the Fissile Materials Transparency Technology Demonstration (FMTD), this equipment was configured and programmed to measure the following six attributes: isotopic ratio, threshold mass, absence of oxide, presence of plutonium, age, and symmetry. On August 16, 2000, at Los Alamos National Laboratory, a delegation of Russian officials observed the successful demonstration of this new technology (called an Attribute Measurement System with Information Barrier, or AMS/IB). The scientists were able to demonstrate without releasing classified information that the nuclear material sample being tested (a nuclear weapon pit) had the declared weapon-grade plutonium characteristics. Once fully developed, AMS/IB technology will protect sensitive information while providing the United States increased confidence that the mandated Russian fissile materials have been stored. Attribute measurement systems can play a role in a number of U.S.-Russian nuclear security regimes such as the Trilateral Initiative, the Plutonium

  1. A review of the prospects for fusion breeding of fissile material

    International Nuclear Information System (INIS)

    This report is the result of an eight month study by the AECL Fusion Status Study Group. The objectives of this study were to review the current status of fusion research, to evaluate the neutronic performance of various fusion-breeder systems, and to assess the economic and technological outlook for the fusion breeder as a source of fissile material to support CANDU reactors operating on the thorium fuel cycle

  2. Fuel fragmentation model advances using TEXAS-V

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, M.L.; El-Beshbeeshy, M.; Nilsuwankowsit, S.; Tang, J. [Wisconsin Univ., Madison, WI (United States). Dept. of Nuclear Engineering and Engineering Physics

    1998-01-01

    Because an energetic fuel-coolant interaction may be a safety hazard, experiments are being conducted to investigate the fuel-coolant mixing/quenching process (FARO) as well as the energetics of vapor explosion propagation for high temperature fuel melt simulants (KROTOS, WFCI, ZrEX). In both types of experiments, the dynamic breakup of the fuel is one of the key aspects that must be fundamentally understood to better estimate the magnitude of the mixing/quenching process or the explosion energetics. To aid our understanding the TEXAS fuel-coolant interaction computer model has been developed and is being used to analyze these experiments. Recently, the models for dynamic fuel fragmentation during the mixing and explosion phases of the FCI have been improved by further insights into these processes. The purpose of this paper is to describe these enhancements and to demonstrate their improvements by analysis of particular JRC FCI data. (author)

  3. Russian RERTR program: Advanced LEU fuel development for research reactors

    International Nuclear Information System (INIS)

    Pin-type fuel elements of new generation have been developed for conversion of Russian-built research reactors to LEU. This paper contains the main results of out-of-pile investigations during manufacturing of experimental batches of mini- and full size fuel elements for irradiation tests in 'MIR' (Dmitrovgrad) and VVR-M (Gatchina) reactor respectively. U-Mo LEU fuel granules have been manufactured both by centrifugal atomization and by hydride-dehydride process. Loading of mini-fuel elements made 4 and 6 g U/cm3, whereas full size ones were loaded to 5.3 g U/cm3. All fuel elements went through quality control of cladding thickness and uniformity of fuel granules distribution through the length of each one. (author)

  4. Experience with advanced driver fuels in EBR-II

    International Nuclear Information System (INIS)

    The Experimental Breeder Reactor II (EBR-II) is a complete nuclear power plant, incorporating a pool-type liquid-metal reactor (LMR) with a fuel-power thermal output of 62.5 MW and an electrical output of 20 MW. Initial criticality was in 1961, utilizing a metallic driver fuel design called the Mark-I. The fuel design has evolved over the last 30 yr, and significant progress has been made on improving performance. The first major innovations were incorporated into the Mark-II design, and burnup then increased dramatically. This design performed successfully, and fuel element lifetime was limited by subassembly hardware performance rather than the fuel element itself. Transient performance of the fuel was also acceptable and demonstrated the ability of EBR-II to survive severe upsets such as a loss of flow without scram. In the mid 1980s, with renewed interest in metallic fuels and Argonne's integral fast reactor (IFR) concept, the Mark-II design was used as the basis for new designs, the Mark-III and Mark-IV. In 1987, the Mark-III design began qualification testing to become a driver fuel for EBR-II. This was followed in 1989 by the Mark-IIIA and Mark-IV designs. The next fuel design, the Mark-V, is being planned to demonstrate the utilization of recycled fuel. The fuel cycle facility attached to EBR-II is being refurbished to produce pyroprocessed recycled fuel as part of the demonstration of the IFR

  5. CARA, new concept of advanced fuel element for HWR

    International Nuclear Information System (INIS)

    All Argentinean NPPs (2 in operation, 1 under construction), use heavy water as coolant and moderator. With very different reactor concepts (pressure Vessel and CANDU type designs), the fuel elements are completely different in its concepts too. Argentina produces both types of fuel elements at a manufacturing fuel element company, called CONUAR. The very different fuel element's designs produce a very complex economical behavior in this company, due to the low production scale. The competitiveness of the Argentinean electric system (Argentina has a market driven electric system) put another push towards to increase the economical competitiveness of the nuclear fuel cycle. At present, Argentina has a very active Slightly Enriched Uranium (SEU) Program for the pressure vessel HWR type, but without strong changes in the fuel concept itself. Then, the Atomic Energy Commission in Argentina (CNEA) has developed a new concept of fuel element, named CARA, trying to achieve very ambitious goals, and substantially improved the competitiveness of the nuclear option. The ambitious targets for CARA fuel element are compatibility (a single fuel element for all Argentinean's HWR) using a single diameter fuel rod, improve the security margins, increase the burnup and do not exceed the CANDU fabrication costs. In this paper, the CARA concept will be presented, in order to explained how to achieve all together these goals. The design attracted the interest of the nuclear power operator utility (NASA), and the fuel manufacturing company (CONUAR). Then a new Project is right now under planning with the cooperation of three parts (CNEA - NASA - CONUAR) in order to complete the whole development program in the shortest time, finishing in the commercial production of CARA fuel bundle. At the end of the this paper, future CARA development program will be described. (author)

  6. JAEA key facilities for global advanced fuel cycle R and D

    International Nuclear Information System (INIS)

    Advanced fuel cycle will be realized with the mid and long term R and D during the long-term transition period from LWR cycle to advanced reactor fuel cycle. Most of JAEA facilities have been utilized to establish the current LWR and FBR (Fast Breeder Reactor) fuel cycle by implementing evolutionary R and D. An assessment of today's state experimental facilities concerning the following research issues: reprocessing, Mox fuel fabrication, irradiation and post-irradiation examination, waste management and nuclear data measurement, is made. The revolutionary R and D requests new issues to be studied: the TRU multi-recycling, minor actinide recycling, the assessment of proliferation resistance and the assessment of cost reduction. To implement the revolutionary R and D for advanced fuel cycle, however, these facilities should be refurbished to install new machines and process equipment to provide more flexible testing parameters

  7. Technologies for Fissile Material Detection and Prevention of Fissile Material Introduction into International Shipping

    International Nuclear Information System (INIS)

    Prevention of the introduction of fissile materials into international shipping, and hence into a given country, is a complex problem. Some pieces of the solution to the puzzle are conceptually well defined, but lack definition of a technical pathway and/or operational implementation. Other elements are a little more fuzzy, and some elements are probably undefined at this point in time. This paper reviews the status of the more well-defined elements, and suggests needed additional measures to enhance the probability that fissile materials are not illicitly introduced into distant countries. International commerce proceeds through a number of steps from point of origin to final destination. Each step offers the possibility of a well-defined choke point to monitor and interdict the illicit shipment of fissile materials. However, because there are so many potential points and venues of entry into a large country such as the United States (e.g., air cargo, shipping containers, truck and rail transport, private vehicles, boats and planes, commercial passenger travel), it behooves the world to ensure that fissile material does not illicitly leave its point of origin

  8. Relevance of mesocatalytic hybrid reactors for accumulation of fissile nuclei and energy balance analysis

    Science.gov (United States)

    Chigrinov, S. E.; Kievitskaya, A. I.; Petlitskij, V. A.

    1993-05-01

    On the basis of the energy and angular spectra of particles emitted from the lateral surfaces of light element targets, the energy balance of a mesocatalytic hybrid reactor (MCHR) has been estimated, with the dependence upon fuel enrichment, type and volume fraction of coolant in the mesocatalytic and electronuclear channel blankets taken into account. It is shown that it is possible to generate a considerable amount of electric power in an MCHR due to burning up fissile nuclides in an MCHR blanket by choosing appropriate types of fuel composition and coolant. Despite some reduction of the fissile nuclide breeding ratio and of the number of nuclear reactors (NR) in the MCHR-NR system, the primary beam power gain is of the same magnitude as in the case of a natural uranium blanket with a hard neutron spectrum. A simplification in solving ecological, economic and safety problems in nuclear fuel reprocessing can be reached by burning the accumulated fissile nuclides directly in the MCHR blanket

  9. Advanced nuclear fuel cycles - Main challenges and strategic choices

    International Nuclear Information System (INIS)

    A graphical conceptual model of the uranium fuel cycles has been developed to capture the present, anticipated, and potential (future) nuclear fuel cycle elements. The once-through cycle and plutonium recycle in fast reactors represent two basic approaches that bound classical options for nuclear fuel cycles. Chief among these other options are mono-recycling of plutonium in thermal reactors and recycling of minor actinides in fast reactors. Mono-recycling of plutonium in thermal reactors offers modest savings in natural uranium, provides an alternative approach for present-day interim management of used fuel, and offers a potential bridging technology to development and deployment of future fuel cycles. In addition to breeder reactors' obvious fuel sustainability advantages, recycling of minor actinides in fast reactors offers an attractive concept for long-term management of the wastes, but its ultimate value is uncertain in view of the added complexity in doing so,. Ultimately, there are no simple choices for nuclear fuel cycle options, as the selection of a fuel cycle option must reflect strategic criteria and priorities that vary with national policy and market perspectives. For example, fuel cycle decision-making driven primarily by national strategic interests will likely favor energy security or proliferation resistance issues, whereas decisions driven primarily by commercial or market influences will focus on economic competitiveness

  10. Development of Advanced Hydrocarbon Fuels at Marshall Space Flight Center

    Science.gov (United States)

    Bai, S. D.; Dumbacher, P.; Cole, J. W.

    2002-01-01

    This was a small-scale, hot-fire test series to make initial measurements of performance differences of five new liquid fuels relative to rocket propellant-1 (RP-1). The program was part of a high-energy-density materials development at Marshall Space Flight Center (MSFC), and the fuels tested were quadricyclane, 1-7 octodiyne, AFRL-1, biclopropylidene, and competitive impulse noncarcinogenic hypergol (CINCH) (di-methyl-aminoethyl-azide). All tests were conducted at MSFC. The first four fuels were provided by the U.S. Air Force Research Laboratory (AFRL), Edwards Air Force Base, CA. The U.S. Army, Redstone Arsenal, Huntsville, AL, provided the CINCH. The data recorded in all hot-fire tests were used to calculate specific impulse and characteristic exhaust velocity for each fuel, then compared to RP-1 at the same conditions. This was not an exhaustive study, comparing each fuel to RP-1 at an array of mixture ratios, nor did it include important fuel parameters, such as fuel handling or long-term storage. The test hardware was designed for liquid oxygen (lox)/RP-1, then modified for gaseous oxygen/RP-1 to avoid two-phase lox at very small flow rates. All fuels were tested using the same thruster/injector combination designed for RP-1. The results of this test will be used to determine which fuels will be tested in future test programs.

  11. Development of an advanced 16x165 Westinghouse type PWR fuel assembly for Slovenia

    International Nuclear Information System (INIS)

    Industrias Nucleares do Brasil (INB), KEPCO Nuclear Fuel Company, Ltd. (KNFC), and Westinghouse Electric Company (Westinghouse) have jointly designed an advanced 16x16 Westinghouse type PWR fuel assembly. This advanced 16x16 Westinghouse type PWR fuel assembly, which will be implemented in both Kori Unit 2 (in Korea) and Angra Unit 1 (in Brazil) in January and March 2005, respectively, is an integral part of the utilities fuel management strategy. This same fuel design has also been developed for future use in Krsko Unit 1 (in Slovenia). In this paper we will describe the front-end nuclear fuel management activities utilized by the joint development team and describe how these activities played an integral part in defining the direction of the advanced 16x16 Westinghouse type PWR fuel assembly design. Additionally, this paper will describe how this design demonstrates improved margins under high duty plant operating conditions. The major reason for initiating this joint development program was to update the current 16x16 fuel assembly, which is also called 16STD. The current 16STD fuel assembly contains a non-optimized fuel rod diameter for the fuel rod pitch (i.e. 9.5 mm OD fuel rods at a 0.485 inch pitch), non-neutronic efficient components (i.e. Inconel Mid grids), no Intermediate Flow Mixer (IFM) grids, and other mechanical features. The advanced 16x16 fuel assembly is being designed for peak rod average burnups of up to 75 MWd/kgU and will use an optimized fuel rod diameter (i.e. 9.14 mm OD ZIRLOTM fuel rods), neutronic efficient components (i.e. ZIRLOTM Mid grids), ZIRLOTM Intermediate Flow Mixer (IFM) grids to improve Departure from Nucleate Boiling (DNB) margin, and many other mechanical features that improve design margins. Nuclear design activities in the areas of fuel cycle cost and fuel management were performed in parallel to the fuel assembly design efforts. As the change in reactivity due to the change in the fuel rod diameter influences directly the

  12. Advanced Composite Bipolar Plate for Unitized Regenerative Fuel Cell/Electrolyzer Systems Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Development of an advanced composite bipolar plate is proposed for a unitized regenerative fuel cell and electrolyzer system that operates on pure feed streams...

  13. Recent Advances in Fuel for Fast Reactors: Synthesis, Properties, Safety Performance

    International Nuclear Information System (INIS)

    This presentation provides an overview of advances made in fast reactor fuel safety research in the last decades. MOX fuels were selected for many fast reactors. The need to minimize waste through partitioning and transmutation strategies has seen new fuel forms evolve, including fertile and non-fertile targets to host the minor actinides. Despite the proven safety performance of MOX fuels and the progress made in nitride and carbide driver fuels and also in minor actinide oxide fuel research programmes, improvements in knowledge and understanding of the safety performance of these fuels can be made. Above all, breakthroughs in simulation and modelling need to be harnessed for dedicated experiments, leading to even more reliable and robust engineering codes for the qualification of all fast reactor fuels. (author)

  14. Advanced nuclear fuel study for the utilization of carbon-coated

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Myung Hyun [Kyung Hee Unviersity, Seoul (Korea)

    1998-03-01

    Advanced nuclear fuel design of carbon coated fuel particles(UCO fuel) was suggested to the current PWRs. Nuclear feasibility studying was forformed for the double heterogeneous UCO fuel by CASMO-3. UCO fuel showed nuclear feasibility when they were packed in the Ulchin3/4 fuel assembly. Nuclear safety was evaluated for the UCO fuel by FTC an dMTC, which had enough safety at operating condition. The average fuel temperature compared with conventional oxide fuel at hot full power condition was reduced by 150 deg K, which was caused by high conductivity of carbon matrix. A core design, used UCO fuel, was possible for same forformance with Ulchin3/4. But, UCO fuel enrichment exceed the PWR fuel enrichment limit 5w/o. Cycle length of UCO duel core was shortened by 90 EFPD satisfied with enrichment limit and thermal power. It is not good for using UCO fuel in PWRs in respect of fuel costs. (author). 19 refs., 71 figs., 25 tabs.

  15. Self shielding in cylindrical fissile sources in the APNea system

    Energy Technology Data Exchange (ETDEWEB)

    Hensley, D.

    1997-02-01

    In order for a source of fissile material to be useful as a calibration instrument, it is necessary to know not only how much fissile material is in the source but also what the effective fissile content is. Because uranium and plutonium absorb thermal neutrons so Efficiently, material in the center of a sample is shielded from the external thermal flux by the surface layers of the material. Differential dieaway measurements in the APNea System of five different sets of cylindrical fissile sources show the various self shielding effects that are routinely encountered. A method for calculating the self shielding effect is presented and its predictions are compared with the experimental results.

  16. Nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F

    2009-07-01

    nature of spent nuclear fuel, Anticipated evolution of fuel in dry storage, Anticipated evolution of fuel in deep geological disposal); Boiling-water reactor fuel (Similarities, and differences with PWR fuel, Axial and radial zoning, Rod and channel box sizes, Poisoning and reactivity control, Cladding specific characteristics, Trends in fuel evolution); 3 - Liquid-metal-cooled fast reactor fuel: Fast-neutron irradiation damage in structural materials (Fast-neutron-induced damage in metals, What materials should be used?); Fuels and targets for fast-reactor transmutation (Fast reactors: reactors affording the ability to carry out effective actinide transmutation, Recycling: homogeneous, or heterogeneous?); 4 - gas-cooled reactor fuel: Particle fuel (From the initial concept to the advanced TRISO particle concept, Kernel fabrication processes, Particle coating by chemical vapor deposition, Fuel element fabrication: particle compaction, Characterization of fuel particles, and elements, From HTR fuel to VHTR and GFR fuels: the GAIA facility at CEA/Cadarache); Irradiation behavior of particle fuels (Particle fuel: a variety of failure modes for a high-strength object, The amoeba effect, Fission product behavior, and diffusion in particle fuels); Mechanical modeling of particle fuel; Very-high-temperature reactor (VHTR) fuel; Gas-cooled fast reactor (GFR) fuel (The specifications for GFR fuel, GFR fissile material, First containment baffler materials, GFR fuel element concepts); 5 - Research reactor fuels (A considerable feedback from experience, Conversion of French reactors to low-enriched ({<=}20% U-235)U{sub 3}Si{sub 2} fuel, Conversion of all reactors: R and D requirements for high-performance reactors, An 'advanced' research reactor fuel: UMo, The startup fuel for the Jules Horowitz Reactor (JHR) will still be U{sub 3}Si{sub 2}-Al; 6 - An instrument for future fuel research: the Jules Horowitz Reactor (JHR): Fuel irradiation experiments in JHR, JHR: a flexible

  17. The U.S. Advanced Fuel Cycle Initiative: Development of separations technologies

    International Nuclear Information System (INIS)

    Spent nuclear fuel from 103 operating U.S. commercial nuclear power reactors is accumulating at a rate of about 2,000 metric tons per year. At this rate, the legislated capacity of the Yucca Mountain geologic repository (63,000 tons of commercial spent fuel) will be exceeded by 2015. Accordingly, the U.S. Department of Energy has instituted a new program, the Advanced Fuel Cycle Initiative, which is intended to provide the technologies necessary for the economical and environmentally sound processing of spent fuel. The goal of this technology development program is to preclude or significantly delay the need for a second geologic repository. Separations technologies are being developed that will support the processing of commercial spent fuel as well as the spent fuel arising from the operation of future advanced reactors

  18. The role of advanced calculation and simulation tools in the evolution of fuel

    International Nuclear Information System (INIS)

    This article is focused on the role of the advanced calculation/simulation tools on the development of the fuel designs as well as in the assessment of the effect of the changes in the operation. With this purpose, the article describes and shows some examples of the use by ENUSA of some of these tools in the fuel engineering. To conclude, the future on the evolution of the advanced tools is also presented. (Author)

  19. Advanced fuel developments for an industrial accelerator driven system prototype

    Energy Technology Data Exchange (ETDEWEB)

    Delage, Fabienne; Ottaviani, Jean Pierre [Commissariat a l' Energie Atomique CEA (France); Fernandez-Carretero, Asuncion; Staicu, Dragos [JRC-ITU (Germany); Boccaccini, Claudia-Matzerath; Chen, Xue-Nong; Mascheck, Werner; Rineiski, Andrei [Forschungszentrum Karlsruhe - FZK (Germany); D' Agata, Elio [JRC-IE (Netherlands); Klaassen, Frodo [NRG, PO Box 25, NL-1755 ZG Petten (Netherlands); Sobolev, Vitaly [SCK-CEN (Belgium); Wallenius, Janne [KTH Royal Institute of Technology (Sweden); Abram, T. [National Nuclear Laboratory - NNL (United Kingdom)

    2009-06-15

    Fuel to be used in an Accelerator Driven System (ADS) for transmutation in a fast spectrum, can be described as a highly innovative concept in comparison with fuels used in critical cores. ADS fuel is not fertile, so as to improve the transmutation performance. It necessarily contains a high concentration ({approx}50%) of minor actinides and plutonium. This unusual fuel composition results in high gamma and neutron emissions during its fabrication, as well as degraded core performance. So, an optimal ADS fuel is based on finding the best compromise between thermal, mechanical, chemical, neutronic and technological constraints. CERCER and CERMET composite fuels consisting of particles of (Pu,MA)O{sub 2} phases dispersed in a magnesia or molybdenum matrix are under investigation within the frame of the ongoing European Integrated Project EUROTRANS (European Research programme for Transmutation) which aims at performing a conceptual design of a 400 MWth transmuter: the European Facility for Industrial Transmutation (EFIT). Performances and safety of EFIT cores loaded with CERCER and CERMET fuels have been evaluated. Out-of-pile and in-pile experiments are carried out to gain knowledge on the properties and the behaviour of these fuels. The current paper gives an overview of the work progress. (authors)

  20. Completing the Design of the Advanced Gas Reactor Fuel Development and Qualification Experiments for Irradiation in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover

    2006-10-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the newly formed Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation.

  1. Completing the Design of the Advanced Gas Reactor Fuel Development and Qualification Experiments for Irradiation in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    The United States Department of Energy's Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the newly formed Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy's lead laboratory for nuclear energy development. The ATR is one of the world's premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation

  2. Efficiency improvement of nuclear power plant operation: the significant role of advanced nuclear fuel technologies

    International Nuclear Information System (INIS)

    Due to the increased liberalisation of the power markets, nuclear power generation is being exposed to high cost reduction pressure. In this paper we highlight the role of advanced nuclear fuel technologies to reduce the fuel cycle costs and therefore increase the efficiency of nuclear power plant operation. The key factor is a more efficient utilisation of the fuel and present developments at Siemens are consequently directed at (i) further increase of batch average burnup, (ii) improvement of fuel reliability, (iii) enlargement of fuel operation margins and (iv) improvement of methods for fuel design and core analysis. As a result, the nuclear fuel cycle costs for a typical LWR have been reduced during the past decades by about US$ 35 million per year. The estimated impact of further burnup increases on the fuel cycle costs is expected to be an additional saving of US$10 - 15 million per year. Due to the fact that the fuel will operate closer to design limits, a careful approach is required when introducing advanced fuel features in reload quantities. Trust and co-operation between the fuel vendors and the utilities is a prerequisite for the common success. (authors)

  3. Advanced fuel for fast breeder reactors: Fabrication and properties and their optimization

    International Nuclear Information System (INIS)

    The present design for FBR fuel rods includes usually MOX fuel pellets cladded into stainless steel tubes, together with UO2 axial blanket and stainless steel hexagonal wrappers. Mixed carbide, nitride and metallic fuels have been tested as alternative fuels in test reactors. Among others, the objectives to develop these alternative fuels are to gain a high breeding ratio, short doubling time and high linear ratings. Fuel rod and assembly designers are now concentrating on finding the combination of optimized fuel, cladding and wrapper materials which could result in improvement of fuel operational reliability under high burnups and load-follow mode of operation. The purpose of the meeting was to review the experience of advanced FBR fuel fabrication technology, its properties before, under and after irradiation, peculiarities of the back-end of the nuclear fuel cycle, and to outline future trends. As a result of the panel discussion, the recommendations on future Agency activities in the area of advanced FBR fuels were developed. A separate abstract was prepared for each of the 10 presentations of this meeting. Refs, figs and tabs

  4. Recent advances in monolithic solid oxide fuel cell development

    International Nuclear Information System (INIS)

    The Argonne Monolithic Solid Oxide Fuel Cell (MSOFC) is fabricated in a honey comb structure having alternate corrugated and flat layers similar to corrugated paperboard. This honeycomb structure, shown schematically in this paper is lightweight yet strong. The materials used to fabricate the MSOFC include yttria-stabilized zirconia as the electrolyte, alkaline earth-doped lanthanum chromite as the interconnection material (bipolar plate), strontium-doped lanthanum manganite as the cathode or air electrode, and yttria-stabilized zirconia-nickel cermet as the anode of fuel electrode. The high power densities of the MSOFC allow it to be used in many applications that would not be possible for other fuel cell designs. Some of these applications include lightweight, mobile power supplies as well as power sources for helicopters and aircraft, automobiles, space platforms, etc. The MSOFC is particularly suited for mobile power applications because of its ability to reform hydrocarbon fuels within the fuel channels

  5. Mechanical design and operating behaviour of advanced LWR fuel elements

    International Nuclear Information System (INIS)

    The development of fuel elements for pressurized and and boiling water reactors during the last years was marked by a reduction of the fuel cycle costs with security and reliability in operation remaining constant. The heightening of fuel discharge burnup and the improvement of neutron economy contributed essentially to that. The latter had been achieved by a reduction of the parasitic absorption within the fuel element and the leakage of neutrons of the reactor cores. These improvements could be obtained under complete observance of the safety-relevant requirements. Due to the change to fuel elements with a higher number of rods and correspondingly lower rod power it was even possible to raise the security margins partly. A survey of the state of experiences of Siemens/KWU is given. (orig./DG)

  6. Advanced anodes for high-temperature fuel cells

    DEFF Research Database (Denmark)

    Atkinson, A.; Barnett, S.; Gorte, R.J.;

    2004-01-01

    Fuel cells will undoubtedly find widespread use in this new millennium in the conversion of chemical to electrical energy, as they offer very high efficiencies and have unique scalability in electricity-generation applications. The solid-oxide fuel cell (SOFC) is one of the most exciting of these...... energy technologies; it is an all-ceramic device that operates at temperatures in the range 500-1,000degreesC. The SOFC offers certain advantages over lower temperature fuel cells, notably its ability to use carbon monoxide as a fuel rather than being poisoned by it, and the availability of high......-grade exhaust heat for combined heat and power, or combined cycle gas-turbine applications. Although cost is clearly the most important barrier to widespread SOFC implementation, perhaps the most important technical barriers currently being addressed relate to the electrodes, particularly the fuel electrode or...

  7. Impact of advanced fuel cycles on disposal of high-level radioactive waste

    International Nuclear Information System (INIS)

    During the last decade various international initiatives have been taken to investigate the role and feasibility of advanced nuclear energy systems. Such advanced nuclear systems aim at optimizing the utilisation of natural resources and minimizing the generation of long-lived radioactive waste. In the framework of the NEA project Impact of Advanced Nuclear Fuel Cycle Options on Waste Management Policies. SCK-CEN studied the impact of advanced fuel cycles on the geological disposal of high-level radioactive waste in a clay formation.The main objectives of the study are assessing the impact of advanced fuel cycles on the design of repositories for the disposal of high level radioactive waste and on their long-term radiological consequences

  8. ACSEPT, Toward the Future Demonstration of Advanced Fuel Treatments

    International Nuclear Information System (INIS)

    Actinide recycling by separation and transmutation is considered worldwide and particularly in several European countries as one of the most promising strategies to reduce the inventory of radioactive waste and to optimize the use of natural resources, thus contributing to making nuclear energy sustainable. In accordance with the Strategic Research Agenda (SRA) of the Sustainable Nuclear Energy Technology Platform (SNE-TP), the timelines of the FP7-EURATOM project ACSEPT (2008-2012) should allow the offering of technical solutions in terms of advanced closed fuel cycle technologies including the recycling of actinides and that may be reviewed by Governments, European utilities as well as Technology Providers at the time horizon 2012. By joining in its consortium 34 partners from 12 European countries plus Australia and Japan, ACSEPT is thus an essential contribution to the demonstration, in the long term, of the potential benefits of actinide recycling. To succeed, ACSEPT is organized into three technical domains: (i) Considering technically mature aqueous separation processes, ACSEPT works to optimize and select the most promising ones dedicated either to actinide partitioning or to grouped actinide separation. A substantial review was undertaken either to be sure that the right molecule families are being studied, or, on the contrary, to identify new candidates. After 18 months, results of the first hot tests should allow the validation of some process options. In addition, the first results on dissolution studies will be available as well as the progress in conversion techniques. (ii) Concerning pyrochemical separation processes, ACSEPT is focused on the enhancement of the two reference cores of process selected within EUROPART with specific attention to the exhaustive electrolysis in molten chloride (quantitative recovery of the actinides with the lowest amount of fission products) and to actinide back-extraction from an An-Al alloy. R and D efforts are also

  9. Advances in reprocessing technology to minimise nuclear fuel waste

    International Nuclear Information System (INIS)

    The responsible and effective management of nuclear wastes generated throughout the nuclear fuel cycle is the key element underpinning the current and future credibility of the industry. This paper presents an overview of the development of existing Purex reprocessing technology in the context of minimising waste streams arising from spent fuel reprocessing. These developments are presented in relation to BNFL's Thorp facility, designed for the reprocessing of oxide fuels. The paper proceeds to discuss potential opportunities for further waste reductions offered by radical reprocessing technologies, such as molten salts conditioning. (author)

  10. Update on Monitoring Technologies for International Safeguards and Fissile Material Verification

    International Nuclear Information System (INIS)

    Monitoring technologies are playing an increasingly important part in international safeguards and fissile material verification. The developments reduce the time an inspector must spend at a site while assuring continuity of knowledge. Monitoring technologies' continued development has produced new seal systems and integrated video surveillance advances under consideration for Trilateral Initiative use. This paper will present recent developments for monitoring systems at Embalse, Argentina, VNHEF, Sarov, Russian, and Savannah River Site, Aiken, South Carolina

  11. Administrative Co-ordination of Fissile Material Management and Accounting in the U.K.A.E.A

    International Nuclear Information System (INIS)

    The Authority are engaged as suppliers in fissile material production, distribution, recycle and reprocessing. As consumers, the Authority require fissile material for power reactors, a variety of prototypes, MTRs, zero-energy facilities and fuel development projects; and for other experimental and research purposes in laboratory quantities. Executive responsibility for these activities lies with the four Groups through which the Authority discharge these functions. It has been found useful to keep these activities under review in specialized inter-Group Committees, with a common secretariat. These Committees: (a) study all projects all proposals or work involving significant quantities of fissile material (plutonium and enriched uranium, other than natural U or U depleted in 235U) in the light of expected supplies over a number of years from all sources, including new production, scrap recovery and imports; and all uses including burn-up, losses and exports; (b) recommend the optimum allocation of specific amounts for approved purposes in relation to other calls upon available supplies, and having regard to the economic issues involved; (c) record and progress all approved allocations, and examine the nature, amount and purpose of all existing stockholdings in relation to current policies and objectives; (d) record and study all losses of fissile material during fabrication or other processing and the measures taken to reduce them; (e) assist in developing procedures and incentives to ensure that material is used economically and returned promptly. Each Group has considerable autonomy in its day-to-day use of fissile material. The administrative machinery described above provides a means by which the Authority’s scientists, engineers, accountants and administrators concerned with fissile material problems can operate collectively in a common frame of reference with a minimum of paperwork. The paper is illustrated with a simplified flowsheet of the main flows of

  12. Operational experiences in MOX fuel fabrication for the FUGEN advanced thermal reactor

    International Nuclear Information System (INIS)

    The Japan Nuclear Cycle Development Institute, JNC, has fabrication the MOX fuel for the Advanced Thermal Reactor, ATR, ''FUGEN'' in the Plutonium Fuel Fabrication Facility, PFFF, since 1974. For these 25 years, the MOX fuel fabrication has progressed in stable manner after overcoming several problems at the start up of FUGEN fuel fabrication. Through the experience, improvements on process equipment and conditions have been taken place to achieve efficient MOX fuel fabrication on an engineering scale as 10 tons MOX per year. Main features of current fabrication process are digested as one step blending with ball milling, pelletizing without granulation and sintering with batch type furnaces. This fabrication process has been demonstrated and confirmed to be applicable techniques for the MOX fuel fabrication on this scale. This paper discusses the FUGEN fuel fabrication focused on the MOX pellet fabrication with operational experiences and improvements to the process. (author)

  13. Study on advanced nuclear fuel cycle of PWR/CANDU synergism

    International Nuclear Information System (INIS)

    According to the concrete condition that China has both PWR and CANDU reactors, one of the advanced nuclear fuel cycle strategy of PWR/CANDU synergism ws proposed, i.e. the reprocessed uranium of spent PWR fuel was used in CANDU reactor, which will save the uranium resource, increase the energy output, decrease the quantity of spent fuels to be disposed and lower the cost of nuclear power. Because of the inherent flexibility of nuclear fuel cycle in CANDU reactor, the transition from the natural uranium to the recycled uranium (RU) can be completed without any changes of the structure of reactor core and operation mode. Furthermore, because of the low radiation level of RU, which is acceptable for CANDU reactor fuel fabrication, the present product line of fuel elements of CANDU reactor only need to be shielded slightly, also the conditions of transportation, operation and fuel management need not to be changed. Thus this strategy has significant practical and economical benefit

  14. Advanced Space Power Systems (ASPS): Regenerative Fuel Cells (RFC) Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The objective of the regenerative fuel cell project element is to develop power and energy storage technologies that enable new capabilities for future human space...

  15. Screening of advanced cladding materials and UN-U3Si5 fuel

    Science.gov (United States)

    Brown, Nicholas R.; Todosow, Michael; Cuadra, Arantxa

    2015-07-01

    In the aftermath of Fukushima, a focus of the DOE-NE Advanced Fuels Campaign has been the development of advanced nuclear fuel and cladding options with the potential for improved performance in an accident. Uranium dioxide (UO2) fuels with various advanced cladding materials were analyzed to provide a reference for cladding performance impacts. For advanced cladding options with UO2 fuel, most of the cladding materials have some reactivity and discharge burn-up penalty (in GWd/t). Silicon carbide is one exception in that the reactor physics performance is predicted to be very similar to zirconium alloy cladding. Most candidate claddings performed similar to UO2-Zr fuel-cladding in terms of safety coefficients. The clear exception is that Mo-based materials were identified as potentially challenging from a reactor physics perspective due to high resonance absorption. This paper also includes evaluation of UN-U3Si5 fuels with Kanthal AF or APMT cladding. The objective of the U3Si5 phase in the UN-U3Si5 fuel concept is to shield the nitride phase from water. It was shown that UN-U3Si5 fuels with Kanthal AF or APMT cladding have similar reactor physics and fuel management performance over a wide parameter space of phase fractions when compared to UO2-Zr fuel-cladding. There will be a marginal penalty in discharge burn-up (in GWd/t) and the sensitivity to 14N content in UN ceramic composites is high. Analysis of the rim effect due to self-shielding in the fuel shows that the UN-based ceramic fuels are not expected to have significantly different relative burn-up distributions at discharge relative to the UO2 reference fuel. However, the overall harder spectrum in the UN ceramic composite fuels increases transuranic build-up, which will increase long-term activity in a once-thru fuel cycle but is expected to be a significant advantage in a fuel cycle with continuous recycling of transuranic material. It is recognized that the fuel and cladding properties assumed in

  16. Slightly enriched uranium in CANDU: An economic first step towards advanced fuel cycles

    International Nuclear Information System (INIS)

    The natural-uranium fuelled Canada Deuterium-Uranium (CANDU) nuclear reactor system has proven to be a safe, reliable and economical producer of electricity for over a quarter of a century. The CANDU system, however, is not restricted to the use of natural-uranium fuel; a wide range of advanced fuel cycles can be accommodated. In the short term, slightly enriched uranium (SEU) is the most promising of these advanced fuel cycles. SEU offers a reduction in the total fuel cycle cost of between 25 and 50% relative to natural-uranium fuel. Uranium consumption is decreased by 30 to 40%. In addition the volume of spent fuel is reduced by a factor of two to three, depending on the enrichment selected. SEU also offers greater flexibility in the design of future CANDU reactors. A variety of fuel management options can be employed in CANDU with slightly enriched fuels. Fuel performance is expected to be good for the burnups of interest, but further fuel testing is planned and is currently in progress in order to confirm this. Programs in place at Atomic Energy of Canada Limited (AECL) will lead to the demonstration and introduction of slightly enriched uranium in CANDU. Ontario Hydro, a Canadian utility with twenty CANDUs operating or under construction, is considering a program which could lead to the implementation of SEU in its nuclear generating stations. (author). 30 refs, 7 figs

  17. Thermal-hydraulics analysis for advanced fuel to be used in Candu 600 nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Catana, Alexandru [RAAN, Institute for Nuclear Research, Str. Campului Nr. 1, Pitesti, Arges (Romania); Danila, Nicolae; Prisecaru, Ilie; Dupleac, Daniel [University POLITEHNICA of Bucharest (Romania)

    2008-07-01

    Two Candu 600 pressure tube nuclear reactors cover about 17% of Romania's electricity demand. These nuclear reactors are moderated/cooled with D{sub 2}O, fuelled on-power with Natural Uranium (NU) dioxide encapsulated in a standard (STD37) fuel bundle. High neutron economy is achieved using D{sub 2}O as moderator and coolant in separated systems. To reduce fuel cycle costs, programs were initiated in Canada, S.Korea, Argentina and Romania for the design and build new fuel bundles able to accommodate different fuel compositions. Candu core structure and modular fuel bundles, permits flexible fuel cycles. The main expected achievements are: reduced fuel cycle costs, increased discharge burn-up, plutonium and minor actinides management, thorium cycle, use of recycled PWR and in the same time waste minimization and operating cost reduction. These new fuel bundles are to be used in already operated Candu reactors. Advanced fuel bundle were proposed: CANFLEX bundle (Canada, S-Korea); the Romanian 'SEU43' bundle (Fig 1). In this paper thermal-hydraulic analysis in sub-channel approach is presented for SEU43. Comparisons with standard (STD37) fuel bundles are made using SEU-NU for NU fuel composition and SEU-0.96, for recycled uranium (RU) fuel with 0.96% U-235. Extended and comprehensive analysis must be made in order to assess the TH behaviour of SEU43. In this paper, considering STD37, SEU43-NU and SEU43-0.96 fuel bundles, main TH parameters were analysed: pressure drop, fuel highest temperatures, coolant density, critical heat flux. Differences between these fuel types are outlined. Benefits are: fuel costs reduction, spent fuel waste minimization, increase in competitiveness of nuclear power. Safety margins must be, at least, conserved. (authors)

  18. INITIAL IRRADIATION OF THE FIRST ADVANCED GAS REACTOR FUEL DEVELOPMENT AND QUALIFICATION EXPERIMENT IN THE ADVANCED TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover; David A. Petti

    2007-09-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation.

  19. Recent advances in thorium fuel cycles for CANDU reactors

    International Nuclear Information System (INIS)

    The once-through thorium fuel cycle in CANDU reactors provides an evolutionary approach to exploiting the energy potential of thorium. In the 'mixed bundle' strategy, the central 8 elements in a CANFLEX fuel bundle contain thoria, while the outermost 35 elements contain slightly enriched uranium (SEU). Detailed full-core fuel-management simulations have shown that this approach can be successfully implemented in existing CANDU reactors. Uranium requirements are lower than for the natural uranium fuel cycle. Further energy can be derived from the thorium by recycling the irradiated thoria fuel elements, containing 233U, as-is without any processing, into the center of a new mixed bundle. There are several examples of such 'demountable' bundles. Recycle of the central 8 thoria elements results in an additional burnup of 20 MW·d/kgHE from the thoria elements, for each recycle. The reactivity of these thoria elements remains remarkably constant over irradiation for each recycle. The natural uranium requirements for the mixed bundle (which includes the natural uranium feed required for the outer SEU fuel elements), without recycle, is about 10% lower than for the natural uranium fuel cycle. After the first recycle, the uranium requirements are -35% lower than for the natural uranium cycle, and remain fairly constant with further recycling (the total uranium requirement averaged over a number of cycles is 30% lower than a natural uranium fuelled CANDU reactor). This thorium cycle strategy is a cost-effective means of reducing uranium requirements, while producing a stockpile of valuable 233U, safeguarded in the spent fuel, that can be recovered in the future when predicated by economic or resource considerations. (author)

  20. Hydraulic test for non-instrumented capsule of advanced PWR fuel pellet

    International Nuclear Information System (INIS)

    This report presents the results of pressure drop test, vibration test and endurance test for Non-instrumented Capsule of Advanced PWR Fuel Pellet which were designed fabricated by KAERI. From the pressure drop test results, it is noted that the flow rate across the Non-instrumented Capsule of Advanced PWR Fuel Pellet corresponding to the pressure drop of 200 kPa is measured to be about 7.45 kg/sec. Vibration frequency for the Non-instrumented Capsule of Advanced PWR Fuel Pellet ranges from 13.0 to 32.3 Hz. RMS(Root Mean Square) displacement for the fuel rig is less than 11.6 μm, and the maximum displacement is less than 30.5 μm. The endurance test was carried out for 103 days and 17 hours

  1. 49 CFR 173.477 - Approval of packagings containing greater than 0.1 kg of non-fissile or fissile-excepted uranium...

    Science.gov (United States)

    2010-10-01

    ... kg of non-fissile or fissile-excepted uranium hexafluoride. 173.477 Section 173.477 Transportation... non-fissile or fissile-excepted uranium hexafluoride. (a) Each offeror of a package containing more than 0.1 kg of uranium hexafluoride must maintain on file for at least one year after the...

  2. A Path Forward to Advanced Nuclear Fuels: Spectroscopic Calorimetry of Nuclear Fuel Materials

    International Nuclear Information System (INIS)

    The goal is to relieve the shortage of thermodynamic and kinetic information concerning the stability of nuclear fuel alloys. Past studies of the ternary nuclear fuel UPuZr have demonstrated constituent redistribution when irradiated or with thermal treatment. Thermodynamic data is key to predicting the possibilities of effects such as constituent redistribution within the fuel rods and interaction with cladding materials

  3. Advanced multiphysics coupling for LWR fuel performance analysis

    International Nuclear Information System (INIS)

    Highlights: • Overviews the BISON nuclear fuel performance analysis code. • Discusses loose and tight coupling approaches applied to thermomechanics. • Outlines coupling to the DeCART neutronics code. • Demonstrates multiscale coupling using MARMOT. - Abstract: Even the most basic nuclear fuel analysis is a multiphysics undertaking, as a credible simulation must consider at a minimum coupled heat conduction and mechanical deformation. The need for more realistic fuel modeling under a variety of conditions invariably leads to a desire to include coupling between a more complete set of the physical phenomena influencing fuel behavior, including neutronics, thermal hydraulics, and mechanisms occurring at lower length scales. This paper covers current efforts toward coupled multiphysics LWR fuel modeling in three main areas. The first area covered in this paper concerns thermomechanical coupling. The interaction of these two physics, particularly related to the feedback effect associated with heat transfer and mechanical contact at the fuel/clad gap, provides numerous computational challenges. An outline is provided of an effective approach used to manage the nonlinearities associated with an evolving gap in BISON, a nuclear fuel performance application. A second type of multiphysics coupling described here is that of coupling neutronics with thermomechanical LWR fuel performance. DeCART, a high-fidelity core analysis program based on the method of characteristics, has been coupled to BISON. DeCART provides sub-pin level resolution of the multigroup neutron flux, with resonance treatment, during a depletion or a fast transient simulation. Two-way coupling between these codes was achieved by mapping fission rate density and fast neutron flux fields from DeCART to BISON and the temperature field from BISON to DeCART while employing a Picard iterative algorithm. Finally, the need for multiscale coupling is considered. Fission gas production and evolution

  4. Advanced reactors and associated fuel cycle facilities: safety and environmental impacts.

    Science.gov (United States)

    Hill, R N; Nutt, W M; Laidler, J J

    2011-01-01

    The safety and environmental impacts of new technology and fuel cycle approaches being considered in current U.S. nuclear research programs are contrasted to conventional technology options in this paper. Two advanced reactor technologies, the sodium-cooled fast reactor (SFR) and the very high temperature gas-cooled reactor (VHTR), are being developed. In general, the new reactor technologies exploit inherent features for enhanced safety performance. A key distinction of advanced fuel cycles is spent fuel recycle facilities and new waste forms. In this paper, the performance of existing fuel cycle facilities and applicable regulatory limits are reviewed. Technology options to improve recycle efficiency, restrict emissions, and/or improve safety are identified. For a closed fuel cycle, potential benefits in waste management are significant, and key waste form technology alternatives are described. PMID:21399407

  5. A Review of Thorium Utilization as an option for Advanced Fuel Cycle-Potential Option for Brazil in the Future

    International Nuclear Information System (INIS)

    Since the beginning of Nuclear Energy Development, Thorium was considered as a potential fuel, mainly due to the potential to produce fissile uranium 233. Several Th/U fuel cycles, using thermal and fast reactors were proposed, such as the Radkwoski once through fuel cycle for PWR and VVER, the thorium fuel cycles for CANDU Reactors, the utilization in Molten Salt Reactors, the utilization of thorium in thermal (AHWR), and fast reactors (FBTR) in India, and more recently in innovative reactors, mainly Accelerator Driven System, in a double strata fuel cycle. All these concepts besides the increase in natural nuclear resources are justified by non proliferation issues (plutonium constrain) and the waste radiological toxicity reduction. The paper intended to summarize these developments, with an emphasis in the Th/U double strata fuel cycle using ADS. Brazil has one of the biggest natural reserves of thorium, estimated in 1.2 millions of tons of ThO2, as will be reviewed in this paper, and therefore RandD programs would be of strategically national interest. In fact, in the past there was some projects to utilize Thorium in Reactors, as the ''Instinto/Toruna'' Project, in cooperation with France, to utilize Thorium in Pressurized Heavy Water Reactor, in the mid of sixties to mid of seventies, and the thorium utilization in PWR, in cooperation with German, from 1979-1988. The paper will review these initiatives in Brazil, and will propose to continue in Brazil activities related with Th/U fuel cycle

  6. Criteria and Evaluation for the Storage of Fissile Material in a Large and Varied Reactor Research and Development Programme

    International Nuclear Information System (INIS)

    Precise analysis of the neutron interaction between fissile material containers is possible through the application of Monte Carlo techniques such as the GEM code of the U.K.A.E.A. or O5R developed by Oak Ridge National Laboratory. While this is appropriate for well defined and inflexible arrays, many problems of practical materials storage do not require this rigor, nor are many materials storage configurations sufficiently well defined to permit full advantage to be derived from such treatment. An analysis which is amenable to slide rule calculation has been found sufficient for most of the problems that arise in a laboratory which has a large inventory of fissile material in the many forms required for a large, extensive, and varied reactor research and development programme (including fuels and materials development). This presentation is directed toward the nuclear safety specialist who must, with limited support facilities, derive criteria for the safe storage of fissile material without undue economic penalty. (author)

  7. An implicit solution framework for reactor fuel performance simulation

    International Nuclear Information System (INIS)

    The simulation of nuclear reactor fuel performance involves complex thermomechanical processes between fuel pellets, made of fissile material, and the protective cladding that surrounds the pellets. An important design goal for a fuel is to maximize the life of the cladding thereby allowing the fuel to remain in the reactor for a longer period of time to achieve higher degrees of burnup. This presentation examines various mathematical and computational issues that impact the modeling of the thermomechanical response of reactor fuel, and are thus important to the development of INL's fuel performance analysis code, BISON. The code employs advanced methods for solving coupled partial differential equation systems that describe multidimensional fuel thermomechanics, heat generation, and transport within the fuel

  8. Advanced light and heavy water reactors for improved fuel utilization

    International Nuclear Information System (INIS)

    On 26-29 November 1984 the Agency convened at its Headquarters in Vienna the Technical Committee and Workshop on Advanced Light and Heavy Water Reactor Technology in order to provide an opportunity to review and discuss the current status and recent development in the lay-out and design of advanced water reactor and to identify areas in which additional research and development are needed. The meeting was attended by 45 participants from 16 nations and 2 international organizations presenting 25 papers. The Conference presentations were divided into sessions devoted to the following topics: Advanced light water reactor programmes (6 papers); Advanced light water design, technology and physics (12 papers); Advanced heavy water reactors (7 papers). A separate abstract was prepared for each of these papers

  9. Measurements of inventories with mixed fissile materials

    Energy Technology Data Exchange (ETDEWEB)

    Rinard, P.M.; Krick, M.S.; Kelley, T.; Schneider, C.M. [and others

    1997-11-01

    An inventory with a large number of diverse items containing mixtures of uranium and plutonium has been measured with two nondestructive assay (NDA) instruments used in four modes. A segmented gamma scanner (SGS) was used to find the number of cans and the positions of the fissile materials by scanning each item in front of a transmissions source; at each position, uranium and plutonium isotopics were measured with the passive gamma rays emitted. A shuffler was then used in both the passive and active modes to measure the masses of the two elements. The measured masses for the inventory items were generally in agreement with the declared values, but anomalies were identified for a small fraction of the inventory.

  10. Fission dynamics with systems of intermediate fissility

    Indian Academy of Sciences (India)

    E Vardaci; A Di Nitto; P N Nadtochy; A Brondi; G La Rana; R Moro; M Cinausero; G Prete; N Gelli; E M Kozulin; G N Knyazheva; I M Itkis

    2015-08-01

    A 4 light charged particle spectrometer, called 8 LP, is in operation at the Laboratori Nazionali di Legnaro, Italy, for studying reaction mechanisms in low-energy heavy-ion reactions. Besides about 300 telescopes to detect light charged particles, the spectrometer is also equipped with an anular PPAC system to detect evaporation residues and a two-arm time-of-flight spectrometer to detect fission fragments. The spectrometer has been used in several fission dynamics studies using as a probe light charged particles in the fission and evaporation residues (ER) channels. This paper proposes a journey within some open questions about the fission dynamics and a review of the main results concerning nuclear dissipation and fission time-scale obtained from several of these studies. In particular, the advantages of using systems of intermediate fissility will be discussed.

  11. Comparative sodium void effects for different advanced liquid metal reactor fuel and core designs

    International Nuclear Information System (INIS)

    An analysis of metal-, oxide-, and nitride-fueled advanced liquid metal reactor cores was performed to investigate the calculated differences in sodium void reactivity, and to determine the relationship between sodium void reactivity and burnup reactivity swing using the three fuel types. The results of this analysis indicate that nitride fuel has the least positive sodium void reactivity for any given burnup reactivity swing. Thus, it appears that a good design compromise between transient overpower and loss of flow response is obtained using nitride fuel. Additional studies were made to understand these and other nitride advantages. (author)

  12. SunLine Transit Agency Advanced Technology Fuel Cell Bus Evaluation: First Results Report

    Energy Technology Data Exchange (ETDEWEB)

    Eudy, L.; Chandler, K.

    2011-03-01

    This report describes operations at SunLine Transit Agency for their newest prototype fuel cell bus and five compressed natural gas (CNG) buses. In May 2010, SunLine began operating its sixth-generation hydrogen fueled bus, an Advanced Technology (AT) fuel cell bus that incorporates the latest design improvements to reduce weight and increase reliability and performance. The agency is collaborating with the U.S. Department of Energy's (DOE) National Renewable Energy Laboratory (NREL) to evaluate the bus in revenue service. This report provides the early data results and implementation experience of the AT fuel cell bus since it was placed in service.

  13. Evolution of nuclear fuels

    International Nuclear Information System (INIS)

    Nuclear fuel is the primary energy source for sustaining the nuclear fission chain reactions in a reactor. The fuels in the reactor cores are exposed to highly aggressive environment and varieties of advanced fuel materials with improved nuclear properties are continuously being developed to have optimum performance in the existing core conditions. Fabrications of varieties of nuclear fuels used in diverse forms of reactors are mainly based on two naturally occurring nuclear source elements, uranium as fissile 235U and fertile 238U, and thorium as fertile 232Th species. The two metals in the forms of alloys with specific elements, ceramic oxides like MOX and ceramic non-oxide as mixed carbide and nitride with suitable nuclear properties like higher metal density, thermal conductivity, etc. are used as fuels in different reactor designs. In addition, efficiency of various advanced fuels in the forms of dispersion, molten salt and other types are also under investigations. The countries which have large deposits of thorium but limited reserves of uranium, are trying to give special impetus on the development of thorium-based fuels for both thermal and fast reactors in harnessing nuclear energy for peaceful uses of atomic energy. (author)

  14. Advanced diagnostics in oxy-fuel combustion processes

    Energy Technology Data Exchange (ETDEWEB)

    Brix, J.; Clausen, Soennik; Degn Jensen, A. (Technical Univ. of Denmark. CHEC Research Centre, Kgs. Lyngby (Denmark)); Boeg Toftegaard, M. (DONG Energy Power, Hvidovre (Denmark))

    2012-07-01

    This report sums up the findings in PSO-project 010069, ''Advanced Diagnostics in Oxy-Fuel Combustion Processes''. Three areas of optic diagnostics are covered in this work: - FTIR measurements in a 30 kW swirl burner. - IR measurements in a 30 kW swirl burner. - IR measurements in a laboratory scale fixed bed reactor. The results obtained in the swirl burner have proved the FTIR method as a valuable technique for gas phase temperature measurements. When its efficacy is evaluated against traditional thermocouple measurements, two cases, with and without probe beam stop, must however be treated separately. When the FTIR probe is operated with the purpose of gas phase concentration measurements the probe needs to operate with a beam stop mounted in front of it. With this beam stop in place it was shown that the measured gas phase temperature was affected by cooling, induced by the cooled beam stop. Hence, for a more accurate determination of gas phase temperatures the probe needed to operate without the beam stop. When this was the case, the FTIR probe showed superior to traditional temperature measurements using a thermocouple as it could measure the fast temperature fluctuations. With the beam stop in place the efficacy of the FTIR probe for gas temperature determination was comparable to the use of a traditional thermocouple. The evaluation of the FTIR technique regarding estimation of gas phase concentrations of H{sub 2}O, CO{sub 2} and CO showed that the method is reliable though it cannot be stated as particularly accurate. The accuracy of the method is dependent on the similarity of the reference emission spectra of the gases with those obtained in the experiments, as the transmittance intensity is not a linear function of concentration. The length of the optical path also affects the steadiness of the measurements. The length of the optical path is difficult to adjust on the small scales that are the focus of this work. However

  15. Fissile and non-fissile element separation in concrete radioactive waste drums using the SIMPHONIE method

    International Nuclear Information System (INIS)

    The simultaneous photon and neutron interrogation experiment (SIMPHONIE) method, applied to radioactive waste drum characterization, has already been treated in [F. Jallu, A. Lyoussi, C. Passard, E. Payan, H. Recroix, G. Nurdin, A. Buisson, J. Allano, Nucl. Instr. and Meth. B 170 (2000) 489]. First experimental results carried out with U and Pu bare samples were presented, that showed the feasibility of quantifying fissile (235U, 239,241Pu, ...) and non-fissile (234,236,238U, 238,240Pu, ...) elements separately in only one measurement, using both active neutron interrogation and induced photofission interrogation techniques simultaneously. This paper presents new experimental results carried out with U samples embedded in a concrete matrix. These results have been obtained using the DGA/ETCA MiniLinatron pulsed linear electron accelerator located at Arcueil, France. Mass detection limits of less than 2 g of matter have been obtained with the preliminary setup used in these experiments

  16. Fissile material disposition program: Screening of alternate immobilization candidates for disposition of surplus fissile materials

    International Nuclear Information System (INIS)

    With the end of the Cold War, the world faces for the first time the need to dismantle vast numbers of ''excess'' nuclear weapons and dispose of the fissile materials they contain, together with fissile residues in the weapons production complex left over from the production of these weapons. If recently agreed US and Russian reductions are fully implemented, tens of thousands of nuclear weapons, containing a hundred tons or more of plutonium and hundreds of tonnes* of highly enriched uranium (HEU), will no longer be needed worldwide for military purposes. These two materials are the essential ingredients of nuclear weapons, and limits on access to them are the primary technical barrier to prospective proliferants who might desire to acquire a nuclear weapons capability. Theoretically, several kilograms of plutonium, or several times that amount of HEU, is sufficient to make a nuclear explosive device. Therefore, these materials will continue to be a potential threat to humanity for as long as they exist

  17. Fissile material disposition program: Screening of alternate immobilization candidates for disposition of surplus fissile materials

    Energy Technology Data Exchange (ETDEWEB)

    Gray, L.W.

    1996-01-08

    With the end of the Cold War, the world faces for the first time the need to dismantle vast numbers of ``excess`` nuclear weapons and dispose of the fissile materials they contain, together with fissile residues in the weapons production complex left over from the production of these weapons. If recently agreed US and Russian reductions are fully implemented, tens of thousands of nuclear weapons, containing a hundred tons or more of plutonium and hundreds of tonnes* of highly enriched uranium (HEU), will no longer be needed worldwide for military purposes. These two materials are the essential ingredients of nuclear weapons, and limits on access to them are the primary technical barrier to prospective proliferants who might desire to acquire a nuclear weapons capability. Theoretically, several kilograms of plutonium, or several times that amount of HEU, is sufficient to make a nuclear explosive device. Therefore, these materials will continue to be a potential threat to humanity for as long as they exist.

  18. Fuel cycle flexibility in Advanced Heavy Water Reactor (AHWR) with the use of Th-LEU fuel

    International Nuclear Information System (INIS)

    The Advanced Heavy Water Reactor (AHWR) is being designed for large scale commercial utilization of thorium (Th) and integrated technological demonstration of the thorium cycle in India. The AHWR is a 920 MW(th), vertical pressure tube type cooled by boiling light water and moderated by heavy water. Heat removal through natural circulation and on-line fuelling are some of the salient features of AHWR design. The physics design of AHWR offers considerable flexibility to accommodate different kinds of fuel cycles. Our recent efforts have been directed towards a case study for the use of Th-LEU fuel cycle in a once-through mode. The discharged Uranium from Th-LEU cycle has proliferation resistant characteristics. This paper gives the initial core, fuel cycle characteristics and online refueling strategy of Th-LEU fuel in AHWR. (author)

  19. Advances in nuclear fuel cycle materials and concepts. Vol. 1

    International Nuclear Information System (INIS)

    This presentation gives an overview of the new trends in the materials used in various steps of the nuclear fuel cycle. This will cover fuels for various types of reactors (PWRs, HTRs, ... etc.) cladding materials, control rod materials, reactor structural materials, as well as materials used in the back end of the fuel cycle. Problems associated with corrosion of fuel cladding materials as well as those in control rod materials (B4 C swelling...etc.), and approaches for combating these influences are reviewed. For the case of reactor pressure vessel materials issues related to the influences of alloy composition, design approaches including the use of more forged parts and minimizing, as for as possible, longitudinal welds especially in the central region, are discussed. Furthermore the application of techniques for recovery of pre-irradiation mechanical properties of PVS components is also covered. New candidate materials for the construction of high level waste containers including modified types of stainless steel (high Ni and high MO), nickel-base alloys and titanium alloys are also detailed. Finally, nuclear fuel cycle concepts involving plutonium and actinides recycling shall be reviewed. 28 figs., 6 tabs

  20. Assessments of Uncertainties in Advanced Equilibrium Fuel Cycles

    International Nuclear Information System (INIS)

    Approximately 15 billion dollars has been paid into the Nuclear Waste Fund, about 55,000 metric tons of spent fuel resides at reactor and storage sites, and the time when Yucca Mountain will be available for disposal of spent fuel is at best uncertain. Results from a substantial number of studies show that space requirements for geologic disposal of radioactive waste can be reduced by about a factor of 50 by selective removal of a few radioactive isotopes from reprocessed spent fuel. Although the issues associated with alternative recycle options are well understood by many who have studied various fuel cycle options, the consequences of various choices should also be well understood. Uncertainty permeates every aspect of the fuel cycle and characterization of these parameters is required to simplify models and identify important contributors to the overall uncertainty. Importance analysis on various parameters for a single reactor identifies these key contributors so that they may be integrated into a larger model such as DANESS. With major uncertainties minimized from a technical perspective, future analyses will look at the economic costs associated with each. (authors)

  1. Gasoline Ultra Efficient Fuel Vehicle with Advanced Low Temperature Combustion

    Energy Technology Data Exchange (ETDEWEB)

    Confer, Keith

    2014-09-30

    The objective of this program was to develop, implement and demonstrate fuel consumption reduction technologies which are focused on reduction of friction and parasitic losses and on the improvement of thermal efficiency from in-cylinder combustion. The program was executed in two phases. The conclusion of each phase was marked by an on-vehicle technology demonstration. Phase I concentrated on short term goals to achieve technologies to reduce friction and parasitic losses. The duration of Phase I was approximately two years and the target fuel economy improvement over the baseline was 20% for the Phase I demonstration. Phase II was focused on the development and demonstration of a breakthrough low temperature combustion process called Gasoline Direct- Injection Compression Ignition (GDCI). The duration of Phase II was approximately four years and the targeted fuel economy improvement was 35% over the baseline for the Phase II demonstration vehicle. The targeted tailpipe emissions for this demonstration were Tier 2 Bin 2 emissions standards.

  2. Advanced nuclear fuel cycle. Optimization by recycling instructive elements

    International Nuclear Information System (INIS)

    Rare-metals and rare-earths produced by fission reaction of uranium 235 in nuclear reactors and consequently contained in spent fuels are considered as potential resources for strategic material in many fields of recent industry. The report consists of several contributed papers concerning with possible utility of such fission products as ruthenium, rhodium, palladium, technetium, and neodymium, and with their recovery and separation from spent fuels as well as possible utilization of actinides and long-lived radioactive elements as radiation sources. To conclude, the present report proposes a new national strategy study to reorient the present scheme of reprocessing of spent fuels and radioactive waste disposal from a new perspective. (S. Ohno)

  3. Advances in PEM fuel cells with CFD techniques

    Energy Technology Data Exchange (ETDEWEB)

    Robalinho, Eric; Cunha, Edgar Ferrari da; Zararya, Ahmed; Linardi, Marcelo [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)], Email: eric@ipen.br; Cekinski, Efrain [Instituto de Pesquisas Tecnologicas (IPT), Sao Paulo, SP (Brazil)

    2010-07-01

    This paper presents some applications of computational fluid dynamics techniques in the optimization of Proton Exchange Membrane Fuel Cell (PEMFC) designs. The results concern: modeling of gas distribution channels, the study for both porous anode and cathode and the three-dimensional modeling of a partial geometry layer containing catalytic Gas Diffusion Layers (GDL) and membrane. Numerical results of the simulations of graphite plates flow channels, using ethanol as fuel, are also presented. Some experimental results are compared to the corresponding numerical ones for several cases, demonstrating the importance and usefulness of this computational tool. (author)

  4. Recycling : The advanced fuel cycle for existing reactors

    International Nuclear Information System (INIS)

    In 1993, the Installed capacity of the world's 427 nuclear power plants was over 335 GWe. Additional plants representing 67 GWe were under construction or on order. Taking construction schedules into consideration, their start-up will stretch out over a period of ten years. Nuclear power will therefore increase by 20% at best in ten years, transiting into a relatively modest 2% average annual growth rate. Of these units, about 80% are light water reactors, whether PWR, BWR, or WER. All of these reactors utilize enriched uranium oxide fuel clad with zirconium alloy. From a fuel perspective, these reactors form a pretty homogeneous group. During reactor residence, energy is supplied by fission of three-fourths of the Initial uranium 235, but also by plutonium fission, which is formed in the fuel as soon as it is Irradiated. The plutonium supplies 40% of the generated power. When the fuel is unloaded, it consists of four elements : fission products and structural materials, such as cladding and end-fittings, which are the reel waste, and residual plutonium and uranium, which are energy materials that can be recycled in accordance with French legislation applicable to both non-nuclear and nuclear industries : 'the purpose of this law is to... make use of waste by reusing, recycling or otherwise obtaining reusable material or energy from.'. The nuclear power industry has entered a phase in which most of its capital-intensive projects are behind it. Now, It must depose Itself to ensuring the competitiveness of nuclear energy compared to other sources of power generation, while protecting the environment and respecting safety regulations. Significant gains have been achieved by improving fuel performance : optimization of fuel design, utilization of less neutron-absorbent materials, and increases in fuel burn-up have made it possible to increase the amount of energy derived from one kilogram of natural uranium by more than 50%. Recycling of the fuel in light water reactor

  5. Microwave Processing of Simulated Advanced Nuclear Fuel Pellets

    International Nuclear Information System (INIS)

    Throughout the three-year project funded by the Department of Energy (DOE) and lead by Virginia Tech (VT), project tasks were modified by consensus to fit the changing needs of the DOE with respect to developing new inert matrix fuel processing techniques. The focus throughout the project was on the use of microwave energy to sinter fully stabilized zirconia pellets using microwave energy and to evaluate the effectiveness of techniques that were developed. Additionally, the research team was to propose fundamental concepts as to processing radioactive fuels based on the effectiveness of the microwave process in sintering the simulated matrix material.

  6. Validation of BWR advanced core and fuel nuclear designs with power reactor measurements

    International Nuclear Information System (INIS)

    Power reactor measurements have been important in validating the reliability, performance characteristics and economics of BWR advanced core and fuel designs. Such measurements go beyond the data obtainable from normal reactor operation and provide detailed benchmark data necessary to verify design and licensing computer design and simulation models. In some cases, such as in the validation of the performance of zirconium barrier pellet-cladding-interaction (PCI) resistant cladding, the BWR power reactor measurements have subjected the advanced fuel design to operating conditions more severe than normal operating conditions, thereby providing nuclear-thermal-mechanical-corrosion performance data for accelerated or extended conditions of operation. In some cases destructive measurements have been carried out on BWR power reactor fuel to provide microscopic and macroscopic data of importance in validating design and licensing analysis methods. There is not uniform agreement among core and fuel designers on the needs for special power reactor core and fuel measurements for validation of advanced designs. The General Electric approach has been to error on the side of extensive, detailed measurements so as to assure reliable performance licensing and economic design and predictive capability. This paper is a summary of some of the validative power reactor measurements that have been carried out on advanced BWR core and fuel designs. Some comparisons of predictions with the data are summarized

  7. University-industry alliance to advance fuel cell discoveries

    OpenAIRE

    Trulove, Susan

    2004-01-01

    Virginia universities will partner with Battelle and other industry partners, Virginia's Center for Innovative Technology, and Los Alamos National Laboratory, to build upon Virginia Tech discoveries in a program called "Bridging the Gap Between New Materials, Fuel Cell Devices and Products."

  8. Performance of advanced high-temperature fuels for nuclear propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Stark, W.A.; Butt, D.P.; Storms, E.K.; Wallace, T.C. [Los Alamos National Lab., NM (United States)

    1994-12-31

    Nuclear propulsion using hydrogen has been demonstrated to operate at nearly twice the performance level of today`s chemical rockets. However, higher temperatures lead to a variety of degradations that compromise safety and longevity. Foremost among these is the melting of the propulsion reactor fuel. The melting behaviour of the U-Zr-C and U-Nb-C systems have been evaluated.

  9. Analysis of Advanced Fuel Assemblies and Core Designs for the Current and Next Generations of LWRs

    International Nuclear Information System (INIS)

    The objective of the project is to design and analyze advanced fuel assemblies for use in current and future light water reactors and to assess their ability to reduce the inventory of transuranic elements, while preserving operational safety. The reprocessing of spent nuclear fuel can delay or avoid the need for a second geological repository in the US. Current light water reactor fuel assembly designs under investigation could reduce the plutonium inventory of reprocessed fuel. Nevertheless, these designs are not effective in stabilizing or reducing the inventory of minor actinides. In the course of this project, we developed and analyzed advanced fuel assembly designs with improved thermal transmutation capability regarding transuranic elements and especially minor actinides. These designs will be intended for use in thermal spectrum (e.g., current and future fleet of light water reactors in the US). We investigated various fuel types, namely high burn-up advanced mixed oxides and inert matrix fuels, in various geometrical designs that are compliant with the core internals of current and future light water reactors. Neutronic/thermal hydraulic effects were included. Transmutation efficiency and safety parameters were used to rank and down-select the various designs.

  10. Modified-open fuel cycle performance with breed-and-burn advanced reactor concepts

    International Nuclear Information System (INIS)

    Recent advances in fast reactor designs enable significant increase in the uranium utilization in an advanced fuel cycle. The category of fast reactors, collectively termed breed-and-burn reactor concepts, can use a large amount of depleted uranium as fuel without requiring enrichment with the exception of the initial core critical loading. Among those advanced concepts, some are foreseen to operate within a once-through fuel cycle such as the Traveling Wave Reactor, CANDLE reactor or Ultra-Long Life Fast Reactor, while others are intended to operate within a modified-open fuel cycle, such as the Breed-and-Burn reactor and the Energy Multiplier Module. This study assesses and compares the performance of the latter category of breed-and-burn reactors at equilibrium state. It is found that the two reactor concepts operating within a modified-open fuel cycle can significantly improve the sustainability and security of the nuclear fuel cycle by decreasing the uranium resources and enrichment requirements even further than the breed-and-burn core concepts operating within the once-through fuel cycle. Their waste characteristics per unit of energy are also found to be favorable, compared to that of currently operating PWRs. However, a number of feasibility issues need to be addressed in order to enable deployment of these breed-and-burn reactor concepts. (author)

  11. Analysis of Advanced Fuel Assemblies and Core Designs for the Current and Next Generations of LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Ragusa, Jean; Vierow, Karen

    2011-09-01

    The objective of the project is to design and analyze advanced fuel assemblies for use in current and future light water reactors and to assess their ability to reduce the inventory of transuranic elements, while preserving operational safety. The reprocessing of spent nuclear fuel can delay or avoid the need for a second geological repository in the US. Current light water reactor fuel assembly designs under investigation could reduce the plutonium inventory of reprocessed fuel. Nevertheless, these designs are not effective in stabilizing or reducing the inventory of minor actinides. In the course of this project, we developed and analyzed advanced fuel assembly designs with improved thermal transmutation capability regarding transuranic elements and especially minor actinides. These designs will be intended for use in thermal spectrum (e.g., current and future fleet of light water reactors in the US). We investigated various fuel types, namely high burn-up advanced mixed oxides and inert matrix fuels, in various geometrical designs that are compliant with the core internals of current and future light water reactors. Neutronic/thermal hydraulic effects were included. Transmutation efficiency and safety parameters were used to rank and down-select the various designs.

  12. Advanced high throughput MOX fuel fabrication technology and sustainable development

    International Nuclear Information System (INIS)

    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

  13. Advanced Silicon Carbide from Molecular Engineering and Actinide Fuels

    International Nuclear Information System (INIS)

    In the frame of nuclear fuels studies for generation IV, carbides or oxycarbides assemblies are one of the engaged material for high temperature reactors. The design of the fuels is not yet defined but some structures are actually considered with SiC as matrix for the actinide fuel. In this work we have studied the synthesis of a multi-scale structure controlled SiC matrix using molecular silicon organometallic precursors. The aim of this work was to develop a way to obtain multi-scale SiC matrix material which could be engineered to fit in any fuel structure defined for generation IV fuels. The control of this multi-scale structure was done using several simulation methods specific of the low temperature solution synthesis of the precursor. In a first step, we have focused our effort on the synthesis of the SiC material. A first level of template was successfully done by the use of solid silica 500 nm balls. A second level of template was studied by the use of meso-porous silica, structured at a 50 nm level. At least, supra-molecular simulation in non aqueous media was considered with the difficulty to build a molecular assembly (inverse micelles). In a second step, we have functionalized the primary silane phase with actinide complexing agent in order to blend directly the actinide inside this primary phase in a controlled way. During these studies, a new one pot synthesis route to obtain the functionalized primary silane phase was developed. (authors)

  14. Further evaluations of the toxicity of irradiated advanced heavy water reactor fuels.

    Science.gov (United States)

    Edwards, Geoffrey W R; Priest, Nicholas D

    2014-11-01

    The neutron economy and online refueling capability of heavy water moderated reactors enable them to use many different fuel types, such as low enriched uranium, plutonium mixed with uranium, or plutonium and/or U mixed with thorium, in addition to their traditional natural uranium fuel. However, the toxicity and radiological protection methods for fuels other than natural uranium are not well established. A previous paper by the current authors compared the composition and toxicity of irradiated natural uranium to that of three potential advanced heavy water fuels not containing plutonium, and this work uses the same method to compare irradiated natural uranium to three other fuels that do contain plutonium in their initial composition. All three of the new fuels are assumed to incorporate plutonium isotopes characteristic of those that would be recovered from light water reactor fuel via reprocessing. The first fuel investigated is a homogeneous thorium-plutonium fuel designed for a once-through fuel cycle without reprocessing. The second fuel is a heterogeneous thorium-plutonium-U bundle, with graded enrichments of U in different parts of a single fuel assembly. This fuel is assumed to be part of a recycling scenario in which U from previously irradiated fuel is recovered. The third fuel is one in which plutonium and Am are mixed with natural uranium. Each of these fuels, because of the presence of plutonium in the initial composition, is determined to be considerably more radiotoxic than is standard natural uranium. Canadian nuclear safety regulations require that techniques be available for the measurement of 1 mSv of committed effective dose after exposure to irradiated fuel. For natural uranium fuel, the isotope Pu is a significant contributor to the committed effective dose after exposure, and thermal ionization mass spectrometry is sensitive enough that the amount of Pu excreted in urine is sufficient to estimate internal doses, from all isotopes, as low

  15. A blueprint for GNEP advanced burner reactor startup fuel fabrication facility

    International Nuclear Information System (INIS)

    Research highlights: → This article discusses use of WG-plutonium as the startup fuel for Advanced Burner Reactor. → The presence of gallium in WG fuel may compromise the fuel integrity. → There is no facility exists to remove gallium from plutonium except at laboratory scale. → This article discusses the processes and issues associated with the gallium removal. → The article provides realistic scenario to all stack-holders involved in designing and operating ABR. - Abstract: The purpose of this article is to identify the requirements and issues associated with design of GNEP Advanced Burner Reactor Fuel Facility. The report was prepared in support of providing data for preparation of a NEPA Environmental Impact Statement in support the U.S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP). One of the GNEP objectives was to reduce the inventory of long lived actinide from the light water reactor (LWR) spent fuel. The LWR spent fuel contains Plutonium (Pu)-239 and other transuranics (TRU) such as Americium-241. One of the options is to transmute or burn these actinides in fast neutron spectra as well as generate the electricity. A sodium-cooled Advanced Recycling Reactor (ARR) concept was proposed to achieve this goal. However, fuel with relatively high TRU content has not been used in the fast reactor. To demonstrate the utilization of TRU fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype of ARR was proposed, which would necessarily be started up using weapons grade (WG) Pu fuel. The WG Pu is distinguished by relatively highest proportions of Pu-239 and lesser amount of other actinides. The WG Pu was assumed to be used as the startup fuel along with TRU fuel in lead test assemblies. Because such fuel is not currently being produced in the US, a new facility (or new capability in an existing facility) was being considered for fabrication of WG Pu fuel for the ABR. It was estimated that the facility will provide the

  16. Irradiation of the First Advanced Gas Reactor Fuel Development and Qualification Experiment in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover; David A. Petti

    2008-10-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation. The design of the first experiment (designated AGR-1) was completed in 2005, and the fabrication and assembly of the test train as well as the support systems and fission product monitoring system that monitor and control the experiment during irradiation were completed in September 2006. The experiment was inserted in the ATR in December 2006, and is serving as a shakedown test of the multi-capsule experiment design that will be used in the subsequent irradiations as well as a test of the early variants of the fuel produced under this program. The experiment test train as well as the monitoring, control, and data collection systems are discussed and the status of the experiment is provided.

  17. Storage tank with internal neutron absorbers for fissile materials

    International Nuclear Information System (INIS)

    The advantages of the new storage tank design with internal neutron absorbers for fissile materials compared to the traditional slab tank design are as follows: The available volume of this vessel design has a value of more than 80%. The tank does not require neutron absorbers to be spaced between adjacent tanks as with the case with slab tanks. This results in a saving in cost and space. The tank is smaller in height and breadth and thus requires less room space for the same amount of storage capacity. Costs are reduced in manufacturing material and room space. The outer surface area of the tank is roughly 7 times smaller than the outer surface area of the slab tank and in less area for heat transfer in case of fire - improved safety. The tank manufacturing will require less welding points to be done and the type of welding required can be automated by machinery. The welding method required, is similar to the welding method that is used to weld the fuel rods for conventional HWR nuclear reactors and thus the technology is known, automated and proven. The tube bundle can be easily removed and more space is available inside the tank for inspection and maintenance of the vessel due to the circular geometry. The vessel design is similar to pressure vessel design which is a proven and well documented method.

  18. PHWR advanced fuel R and D for the 21st Century in Korea

    International Nuclear Information System (INIS)

    As the first CANDU-PHWR nuclear power plant in Korea, Wolsong Unit 1 has been successfully operated since 1983. The CANDU installed electric-generation capacity was about 50 % of the installed electric-generation capacity of nuclear power plants in Korea in 1983 but then decreased to less than 10 % of the total installed nuclear electric-generation capacity by 1996. This CANDU installed electric-generation capacity has recovered to about 20 % of the total installed nuclear electric-generation capacity in 1999, because Wolsong Units 2, 3 and 4 have been placed into commercial operation in 1997, 1998 and 1999, respectively. This indicates that CANDU reactors are not the majority of nuclear power plants in Korea. Since the period of the late 1970s, nuclear fuel design and fabrication technologies have been engaged as one of the important R and D activities in Korea. As one of the early R and D activities leading to nuclear power industrialization in Korea, the project to develop the design and fabrication technology of CANDU-6 37-element fuel had been successfully carried out from 1981 to 1987 by KAERI. Just after the successful completion of the 37-element fuel R and D, KAERI has developed a CANDU-6 advanced fuel. The key targets of the development program are safety enhancement, reduction of spent fuel volume, and economic improvements, using the inherent characteristics and advantages of CANDU technology. The CANFLEX and DUPIC R and D programs have been conducted under Korea's Nuclear Energy R and D Project as national mid- and long-term programs since 1992. As the second of the CANDU R and D products in Korea, the CANFLEX-NU fuel has been jointly developed by KAERI and AECL.The fuel has demonstrated its irradiation performance in a Canadian commercial power reactor, Pt. Lepreau Generating Station since 1998 September. The RU(SEU) and DUPIC fuels are expected to be developed continuously until about the year 2010 for their use in CANDU reactors. Beside these fuel

  19. Fissile material measurements using the differential die-away self interrogation technique

    Energy Technology Data Exchange (ETDEWEB)

    Schear, Melissa A [Los Alamos National Laboratory; Menlove, Howard O [Los Alamos National Laboratory; Tobin, Stephen J [Los Alamos National Laboratory; Evans, Louise G [Los Alamos National Laboratory; Lee, S Y [Los Alamos National Laboratory

    2010-01-01

    Currently, there is substantial research effort focused on quantifying plutonium (Pu) mass in spent fuel using non-destructive assay (NDA) techniques. Of the several techniques being investigated for this purpose, Differential Die-Away Self-Interrogation (DDSI) is a recently proposed, neutron-based NDA technique capable of quantifying the total fissile content in an assembly. Unlike the conventional Differential Die-Away (DDA) technique, DOSI does not require an external neutron source for sample interrogation, but rather, uses the spontaneous fission neutrons originating from {sup 244}Cm within the spent fuel for self-interrogation. The essence of the technique lies in the time separation between the detection of spontaneous fission neutrons from {sup 244}Cm and the detection of induced fission neutrons at a later time. The DDSI detector design imposes this time separation by optimizing the die-away times ({tau}) of the detector and sample interrogation regions to obtain an early and late neutron distribution respectively. The ratio of the count rates in the late gate to the early gate for singles, doubles, and triples is directly proportional to the fissile content present in the sample, which has already been demonstrated for simplified fuel cases using the Monte Carlo N-Particle eXtended (MCNPX) code. The current work applies the DDSI concept to more complex samples, specifically spent Pressurized Water Reactor (PWR) assemblies with varying isotopics resulting from a range of initial enrichment, bumup, and cooling time. We assess the feasibility of using the late gate to early gate ratio as a reliable indicator of overall fissile mass for a range of assemblies by defining a {sup 239}Pu effective mass which indicates the mass of {sup 239}Pu that would yield the same DDSI signal as the combined mass of major fissile isotopes present in the sample. This work is important for assessing the individual capability of the DDSI instrument in quantifying fissile mass in

  20. The processes for fissile material flux management: theory vs practice

    International Nuclear Information System (INIS)

    Bel-V has carried out a thorough control of some licensees in order to check if the official accountability of fissile material matched the experimental fissile material fluxes in the processes. The following approach was adopted: 1) Understand how the licensee has kept count of the official accountability over the years (it must be noted that the past method used was evaluated and recognized by the regulator); 2) Observe and get an in-depth understanding of the processes carried out by the licensee; 3) Compare these processes with the information (yields, chemical reactions, byproducts formation...) published in scientific literature; 4) Localize parts of the installation or equipment where an accumulation of fissile material could take place; 5) Ask the licensee for a sample analysis in parts of the equipment where the localization study showed a risk of accumulation or loss of fissile material; 6) Compare the analysis results with the official accountability values; 7) If applicable, suggest any process modification in order to improve the recovery of fissile material; 8) In the future, eventually estimate the potential necessary corrections of the official accountability of fissile material (administrative task). Bel-V has not detected any major problems that necessitate immediate action. However, this study has identified parts of equipment in some facilities where the accumulation rate of fissile material is higher than expected. On the other hand, experimental analyzes have excluded some places where fissile material was supposed to be prone to accumulation. Finally the licensees and Bel-V have localized places where chemical side reactions or very slow kinetic reactions have altered the molecular form of the fissile material. The article is followed by the slides of the presentation

  1. Advances in the generation of a new emulsified fuel

    Energy Technology Data Exchange (ETDEWEB)

    Chavez, A. [Technical Consultancy, Energy Plus UC, Huitzilac, Morelos (Mexico); Ramirez, M. [Instituto Mexicano del Petroleo, Programa de Aseguramiento de Hidrocarburos, Mexico, D.F. (Mexico); Medina, E. [Universidad Nacional Autonoma de Mexico, Departamento de Termofluidos, Facultad de Ingenieria, Mexico, D.F. (Mexico); Bolado, R.; Mora, J. [Instituto Mexicano del Petroleo, Laboratorio de Combustion, Veracruz (Mexico)

    2011-08-15

    The development of a new emulsified fuel is described, from the conceptual idea to the semi-industrial tests of the final product. The starting point was the necessity to lower the particulate matter (PM) emissions produced by the combustion of more than 200 MBD of heavy fuel oil (HFO) used for electric power conversion. The major component of HFO is a vacuum residue of the oil refining process mixed with light cycle oils to make it pumpable. An alternative to handle and burn the high viscosity residue (solid at room temperature) is by converting it in an oil-in-water emulsion. The best emulsions resulted of 70% residue in 30% water, Sauter Mean Diameter of 10-20 {mu}m and a stability of more than 90 days. Spray burning tests of the emulsion against HFO in a semi-industrial 500 kW furnace showed a reduction in PM emissions of 24-36%. (orig.)

  2. Advanced ECU Software Development Method for Fuel Cell Systems

    Institute of Scientific and Technical Information of China (English)

    TIAN Shuo; LIU Yuan; XIA Wenchuan; LI Jianqiu; YANG Minggao

    2005-01-01

    The electronic control unit (ECU) in electrical powered hybrid and fuel cell vehicles is exceedingly complex. Rapid prototyping control is used to reduce development time and eliminate errors during software development. This paper describes a high-efficiency development method and a flexible tool chain suitable for various applications in automotive engineering. The control algorithm can be deployed directly from a Matlab/Simulink/Stateflow environment into the ECU hardware together with an OSEK real-time operating system (RTOS). The system has been successfully used to develop a 20-kW fuel cell system ECU based on a Motorola PowerPC 555 (MPC555) microcontroller. The total software development time is greatly reduced and the code quality and reliability are greatly enhanced.

  3. Recent advances in Carbon Nanotube based Enzymatic Fuel Cells

    Directory of Open Access Journals (Sweden)

    Serge eCosnier

    2014-10-01

    Full Text Available This review summarizes recent trends in the field of enzymatic fuel cells. Thanks to the high specificity of enzymes, biofuel cells can generate electrical energy by oxidation of a targeted fuel (sugars, alcohols or hydrogen at the anode and reduction of oxidants (O2, H2O2 at the cathode in complex media. The combination of carbon nanotubes, enzymes and redox mediators was widely exploited to develop biofuel cells since the electrons, involved in the bio-electrocatalytic processes, can be efficiently transferred from or to an external circuit. Original approaches to construct electron transfer based CNT-bioelectrodes and impressive biofuel cell performances are reported as well as biomedical applications.

  4. Advances in fuel channel technology for CANDU reactors

    International Nuclear Information System (INIS)

    The components of the CANDU fuel channels are being developed to have service lives of over 30 years with large margins of safety. Information from research programs and the examination of components removed from reactors has enable improvements to be made to pressure tubes, spacers, calandria tubes and end fittings. Improvements have also been made to the channel design to facilitate planned retubing. (author). 22 refs., 5 tabs., 31 figs

  5. Hybrid fuel cell bus demonstration: advanced technology moves bus forward

    International Nuclear Information System (INIS)

    The Province of Manitoba has been actively pursuing hydrogen since 2001 as one part of a portfolio of renewable energy alternatives. Six priority hydrogen actions have been underway covering a variety of opportunities, including two recently completed major transit bus and refueling demonstrations. A brief overview of Manitoba's activities on hydrogen will be provided, emphasizing the lessons learned from recent projects such as the hydrogen Hybrid Fuel Cell Bus demonstration, and in particular implications for the research community. (author)

  6. Variants of closing the nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Andrianova, E. A., E-mail: Andrianova-EA@nrcki.ru; Davidenko, V. D.; Tsibulskiy, V. F.; Tsibulskiy, S. V. [National Research Center Kurchatov Institute (Russian Federation)

    2015-12-15

    Influence of the nuclear energy structure, the conditions of fuel burnup, and accumulation of new fissile isotopes from the raw isotopes on the main parameters of a closed fuel cycle is considered. The effects of the breeding ratio, the cooling time of the spent fuel in the external fuel cycle, and the separation of the breeding area and the fissile isotope burning area on the parameters of the fuel cycle are analyzed.

  7. Variants of closing the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Influence of the nuclear energy structure, the conditions of fuel burnup, and accumulation of new fissile isotopes from the raw isotopes on the main parameters of a closed fuel cycle is considered. The effects of the breeding ratio, the cooling time of the spent fuel in the external fuel cycle, and the separation of the breeding area and the fissile isotope burning area on the parameters of the fuel cycle are analyzed

  8. Screening of advanced cladding materials and UN–U{sub 3}Si{sub 5} fuel

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Nicholas R., E-mail: nbrown@bnl.gov; Todosow, Michael; Cuadra, Arantxa

    2015-07-15

    Highlights: • Screening methodology for advanced fuel and cladding. • Cladding candidates, except for silicon carbide, exhibit reactivity penalty versus zirconium alloy. • UN–U{sub 3}Si{sub 5} fuels have the potential to exhibit reactor physics and fuel management performance similar to UO{sub 2}. • Harder spectrum in the UN ceramic composite fuel increases transuranic build-up. • Fuel and cladding properties assumed in these assessments are preliminary. - Abstract: In the aftermath of Fukushima, a focus of the DOE-NE Advanced Fuels Campaign has been the development of advanced nuclear fuel and cladding options with the potential for improved performance in an accident. Uranium dioxide (UO{sub 2}) fuels with various advanced cladding materials were analyzed to provide a reference for cladding performance impacts. For advanced cladding options with UO{sub 2} fuel, most of the cladding materials have some reactivity and discharge burn-up penalty (in GWd/t). Silicon carbide is one exception in that the reactor physics performance is predicted to be very similar to zirconium alloy cladding. Most candidate claddings performed similar to UO{sub 2}–Zr fuel–cladding in terms of safety coefficients. The clear exception is that Mo-based materials were identified as potentially challenging from a reactor physics perspective due to high resonance absorption. This paper also includes evaluation of UN–U{sub 3}Si{sub 5} fuels with Kanthal AF or APMT cladding. The objective of the U{sub 3}Si{sub 5} phase in the UN–U{sub 3}Si{sub 5} fuel concept is to shield the nitride phase from water. It was shown that UN–U{sub 3}Si{sub 5} fuels with Kanthal AF or APMT cladding have similar reactor physics and fuel management performance over a wide parameter space of phase fractions when compared to UO{sub 2}–Zr fuel–cladding. There will be a marginal penalty in discharge burn-up (in GWd/t) and the sensitivity to {sup 14}N content in UN ceramic composites is high

  9. γ-ray self-absorption of cylindrical fissile material

    Institute of Scientific and Technical Information of China (English)

    HUANG Yong-Yi; CHENG Yi-Ying; TIAN Dong-Feng; LU Fu-Quan; YANG Fu-Jia

    2005-01-01

    The self-absorption of γ-ray emitted from cylindrical fissile materials, such as 235U and 239Pu, does not possess spherical symmetry. The analytical formulae of self-absorption for γ-ray throughout the cylinder have been obtained. The intensity of γ-ray is a function of γ-ray outgoing directions and cylindrical configurations, accordingly one can acquire the information about geometrical configuration of cylindrical fissile materials through multi-location measurements. Further more, the method is given in this article. The result can be applied to the fissile material safeguard, such as nuclear monitoring and verifying.

  10. Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    D. E. Shropshire

    2009-01-01

    The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program’s understanding of the cost drivers that will determine nuclear power’s cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-iradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

  11. Radio-toxicity of spent fuel of the advanced heavy water reactor.

    Science.gov (United States)

    Anand, S; Singh, K D S; Sharma, V K

    2010-01-01

    The Advanced Heavy Water Reactor (AHWR) is a new power reactor concept being developed at Bhabha Atomic Research Centre, Mumbai. The reactor retains many desirable features of the existing Pressurised Heavy Water Reactor (PHWR), while incorporating new, advanced safety features. The reactor aims to utilise the vast thorium resources available in India. The reactor core will use plutonium as the make-up fuel, while breeding (233)U in situ. On account of this unique combination of fuel materials, the operational characteristics of the fuel as determined by its radioactivity, decay heat and radio-toxicity are being viewed with great interest. Radio-toxicity of the spent fuel is a measure of potential radiological hazard to the members of the public and also important from the ecological point of view. The radio-toxicity of the AHWR fuel is extremely high to start with, being approximately 10(4) times that of the fresh natural U fuel used in a PHWR, and continues to remain relatively high during operation and subsequent cooling. A unique feature of this fuel is the peak observed in its radio-toxicity at approximately 10(5) y of decay cooling. The delayed increase in fuel toxicity has been traced primarily to a build-up of (229)Th, (230)Th and (226)Ra. This phenomenon has been observed earlier for thorium-based fuels and is confirmed for the AHWR fuel. This paper presents radio-toxicity data for AHWR spent fuel up to a period of 10(6) y and the results are compared with the radio-toxicity of PHWR. PMID:19776247

  12. Validation of the COBRA code for dry out power calculation in CANDU type advanced fuels

    International Nuclear Information System (INIS)

    Stern Laboratories perform a full scale CHF testing of the CANFLEX bundle under AECL request. This experiment is modeled with the COBRA IV HW code to verify it's capacity for the dry out power calculation . Good results were obtained: errors below 10 % with respect to all data measured and 1 % for standard operating conditions in CANDU reactors range . This calculations were repeated for the CNEA advanced fuel CARA obtaining the same performance as the CANFLEX fuel. (author)

  13. Executive summary: advanced-fuel fusion systems, the D-3He satellite approach

    International Nuclear Information System (INIS)

    An evaluation was made of the potential advantages and feasibility of fusion power plants designed to employ near-term non-D--T fuels such as deuterium and D-3He. The following topics are discussed here: (1) cost studies and net-energy analysis, (2) D-3He Bumpy Torus satellite, (3) exploratory studies of a D-3He field-reversed mirror satellite, (4) preliminary advanced fuel pellet studies, and (5) 3He neutral beam injector

  14. Advanced oxidation-resistant iron-based alloys for LWR fuel cladding

    Science.gov (United States)

    Terrani, K. A.; Zinkle, S. J.; Snead, L. L.

    2014-05-01

    Application of advanced oxidation-resistant iron alloys as light water reactor fuel cladding is proposed. The motivations are based on specific limitations associated with zirconium alloys, currently used as fuel cladding, under design-basis and beyond-design-basis accident scenarios. Using a simplified methodology, gains in safety margins under severe accidents upon transition to advanced oxidation-resistant iron alloys as fuel cladding are showcased. Oxidation behavior, mechanical properties, and irradiation effects of advanced iron alloys are briefly reviewed and compared to zirconium alloys as well as historic austenitic stainless steel cladding materials. Neutronic characteristics of iron-alloy-clad fuel bundles are determined and fed into a simple economic model to estimate the impact on nuclear electricity production cost. Prior experience with steel cladding is combined with the current understanding of the mechanical properties and irradiation behavior of advanced iron alloys to identify a combination of cladding thickness reduction and fuel enrichment increase (∼0.5%) as an efficient route to offset any penalties in cycle length, due to higher neutron absorption in the iron alloy cladding, with modest impact on the economics.

  15. The nuclear fuel cycle with advanced reactor systems - analysis of its economic fundamentals and possibilities

    International Nuclear Information System (INIS)

    The purpose of the study is to analyse the nuclear fuel cycle of alternative advanced reactor systems with respect to their different mass flows of nuclear fuel and to judge the economic feasibility of these advanced nuclear technologies using a specific fuel cycle model. It is the particular importance of this subject that many technical, physical, political and economic coherences are combined in a very complex manner. A detailed description of the problem is given in the introductional chapter 1. The following chapter 2 gives a sufficient survey of the different techniques and technical facilities of the nuclear fuel cycles in question. Part 3 includes an investigation of logical coherences between typical fuel cycle mass flows which consequently leads to a mathematical model. This model is described in part 4. Chapter 5 then deals with the application of this model by the quantitative estimation and valuation of the economic differences between the conventional and advanced nuclear technology. In the final part of this study the influence of a very important parameter in this context, the price of plutonium, is discussed with respect to the time of introduction of the advanced reactor technology under economic conditions. (orig.)

  16. The advanced fuel cycle facility (AFCF) role in the global nuclear energy partnership

    International Nuclear Information System (INIS)

    The Global Nuclear Energy Partnership (GNEP), launched in February, 2006, proposes to introduce used nuclear fuel recycling in the United States with improved proliferation-resistance and a more effective waste management approach. This program is evaluating ways to close the fuel cycle in a manner that builds on recent laboratory breakthroughs in U.S. national laboratories and draws on international and industry partnerships. Central to moving this advanced fuel recycling technology from the laboratory to commercial implementation is a flexible research, development and demonstration facility, called the Advanced Fuel Cycle Facility (AFCF). The AFCF was introduced as one of three projects under GNEP and will provide the U.S. with the capabilities to evaluate technologies that separate used fuel into reusable material and waste in a proliferation-resistant manner. The separations technology demonstration capability is coupled with a remote transmutation fuel fabrication demonstration capability in an integrated manner that demonstrates advanced safeguard technologies. This presentation will discuss the key technical and programmatic features of AFCF and their support of the GNEP objectives. (authors)

  17. Endurance test for non-instrumented capsule of advanced PWR fuel pellet (test procedure)

    International Nuclear Information System (INIS)

    This test procedure details the test loop, test method, and test procedure for pressure drop, vibration and endurance test of Non-instrumented Capsule of Advanced PWR Fuel Pellet. From the pressure drop test, the hydraulic design requirements of the capsule are verified. HANARO limit condition is checked and the compatibility with HANARO core is verified. From flow induced vibration test vibration frequency, vibration displacement are investigated. The wear of Non-instrumented Capsule of Advanced PWR Fuel Pellet is investigated through endurance test, and these data are used to evaluate the expected wear of during maximum resident time of Non-instrumented Capsule

  18. Opportunities for mixed oxide fuel testing in the advanced test reactor to support plutonium disposition

    International Nuclear Information System (INIS)

    Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. These issues include the following: (1) MOX fuel fabrication process verification; (2) Whether and how to use burnable poisons to depress MOX fuel initial reactivity, which is higher than that of urania; (3) The effects of WGPu isotopic composition; (4) The feasibility of loading MOX fuel with plutonia content up to 7% by weight; (5) The effects of americium and gallium in WGPu; (6) Fission gas release from MOX fuel pellets made from WGPu; (7) Fuel/cladding gap closure; (8) The effects of power cycling and off-normal events on fuel integrity; (9) Development of radial distributions of burnup and fission products; (10) Power spiking near the interfaces of MOX and urania fuel assemblies; and (11) Fuel performance code validation. The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory possesses many advantages for performing tests to resolve most of the issues identified above. We have performed calculations to show that the use of hafnium shrouds can produce spectrum adjustments that will bring the flux spectrum in ATR test loops into a good approximation to the spectrum anticipated in a commercial LWR containing MOX fuel while allowing operation of the test fuel assemblies near their optimum values of linear heat generation rate. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. The requirements for planning and implementing a test program in the ATR have been identified. The facilities at Argonne National Laboratory-West can meet all potential needs for pre- and post-irradiation examination that might arise in a MOX fuel qualification program

  19. Systems Analysis of an Advanced Nuclear Fuel Cycle Based on a Modified UREX+3c Process

    International Nuclear Information System (INIS)

    The research described in this report was performed under a grant from the U.S. Department of Energy (DOE) to describe and compare the merits of two advanced alternative nuclear fuel cycles -- named by this study as the 'UREX+3c fuel cycle' and the 'Alternative Fuel Cycle' (AFC). Both fuel cycles were assumed to support 100 1,000 MWe light water reactor (LWR) nuclear power plants operating over the period 2020 through 2100, and the fast reactors (FRs) necessary to burn the plutonium and minor actinides generated by the LWRs. Reprocessing in both fuel cycles is assumed to be based on the UREX+3c process reported in earlier work by the DOE. Conceptually, the UREX+3c process provides nearly complete separation of the various components of spent nuclear fuel in order to enable recycle of reusable nuclear materials, and the storage, conversion, transmutation and/or disposal of other recovered components. Output of the process contains substantially all of the plutonium, which is recovered as a 5:1 uranium/plutonium mixture, in order to discourage plutonium diversion. Mixed oxide (MOX) fuel for recycle in LWRs is made using this 5:1 U/Pu mixture plus appropriate makeup uranium. A second process output contains all of the recovered uranium except the uranium in the 5:1 U/Pu mixture. The several other process outputs are various waste streams, including a stream of minor actinides that are stored until they are consumed in future FRs. For this study, the UREX+3c fuel cycle is assumed to recycle only the 5:1 U/Pu mixture to be used in LWR MOX fuel and to use depleted uranium (tails) for the makeup uranium. This fuel cycle is assumed not to use the recovered uranium output stream but to discard it instead. On the other hand, the AFC is assumed to recycle both the 5:1 U/Pu mixture and all of the recovered uranium. In this case, the recovered uranium is reenriched with the level of enrichment being determined by the amount of recovered plutonium and the combined amount of the

  20. Systems Analysis of an Advanced Nuclear Fuel Cycle Based on a Modified UREX+3c Process

    Energy Technology Data Exchange (ETDEWEB)

    E. R. Johnson; R. E. Best

    2009-12-28

    The research described in this report was performed under a grant from the U.S. Department of Energy (DOE) to describe and compare the merits of two advanced alternative nuclear fuel cycles -- named by this study as the “UREX+3c fuel cycle” and the “Alternative Fuel Cycle” (AFC). Both fuel cycles were assumed to support 100 1,000 MWe light water reactor (LWR) nuclear power plants operating over the period 2020 through 2100, and the fast reactors (FRs) necessary to burn the plutonium and minor actinides generated by the LWRs. Reprocessing in both fuel cycles is assumed to be based on the UREX+3c process reported in earlier work by the DOE. Conceptually, the UREX+3c process provides nearly complete separation of the various components of spent nuclear fuel in order to enable recycle of reusable nuclear materials, and the storage, conversion, transmutation and/or disposal of other recovered components. Output of the process contains substantially all of the plutonium, which is recovered as a 5:1 uranium/plutonium mixture, in order to discourage plutonium diversion. Mixed oxide (MOX) fuel for recycle in LWRs is made using this 5:1 U/Pu mixture plus appropriate makeup uranium. A second process output contains all of the recovered uranium except the uranium in the 5:1 U/Pu mixture. The several other process outputs are various waste streams, including a stream of minor actinides that are stored until they are consumed in future FRs. For this study, the UREX+3c fuel cycle is assumed to recycle only the 5:1 U/Pu mixture to be used in LWR MOX fuel and to use depleted uranium (tails) for the makeup uranium. This fuel cycle is assumed not to use the recovered uranium output stream but to discard it instead. On the other hand, the AFC is assumed to recycle both the 5:1 U/Pu mixture and all of the recovered uranium. In this case, the recovered uranium is reenriched with the level of enrichment being determined by the amount of recovered plutonium and the combined amount

  1. Advanced Automotive Fuels Research, Development, and Commercialization Cluster (OH)

    Energy Technology Data Exchange (ETDEWEB)

    Linkous, Clovis; Hripko, Michael; Abraham, Martin; Balendiran, Ganesaratnam; Hunter, Allen; Lovelace-Cameron, Sherri; Mette, Howard; Price, Douglas; Walker, Gary; Wang, Ruigang

    2013-08-31

    Technical aspects of producing alternative fuels that may eventually supplement or replace conventional the petroleum-derived fuels that are presently used in vehicular transportation have been investigated. The work was centered around three projects: 1) deriving butanol as a fuel additive from bacterial action on sugars produced from decomposition of aqueous suspensions of wood cellulose under elevated temperature and pressure; 2) using highly ordered, openly structured molecules known as metal-organic framework (MOF) compounds as adsorbents for gas separations in fuel processing operations; and 3) developing a photocatalytic membrane for solar-driven water decomposition to generate pure hydrogen fuel. Several departments within the STEM College at YSU contributed to the effort: Chemistry, Biology, and Chemical Engineering. In the butanol project, sawdust was blended with water at variable pH and temperature (150 – 250{degrees}C), and heated inside a pressure vessel for specified periods of time. Analysis of the extracts showed a wide variety of compounds, including simple sugars that bacteria are known to thrive upon. Samples of the cellulose hydrolysate were fed to colonies of Clostridium beijerinckii, which are known to convert sugars to a mixture of compounds, principally butanol. While the bacteria were active toward additions of pure sugar solutions, the cellulose extract appeared to inhibit butanol production, and furthermore encouraged the Clostridium to become dormant. Proteomic analysis showed that the bacteria had changed their genetic code to where it was becoming sporulated, i.e., the bacteria were trying to go dormant. This finding may be an opportunity, as it may be possible to genetically engineer bacteria that resist the butanol-driven triggering mechanism to stop further fuel production. Another way of handling the cellulosic hydrolysates was to simply add the enzymes responsible for butanol synthesis to the hydrolytic extract ex-vivo. These

  2. Study on process basic requirements of experimental facility of advanced spent fuel management process

    International Nuclear Information System (INIS)

    The advanced spent fuel management process, which was proposed to reduce the overall volume of the PWR spent fuel and improve safety and economy of the long-term storage of spent fuel, is under research and development. Hot cell facilities of α-γ type and inert atmosphere are required essentially for safe hot test and verification of this process. In this study, design basic data are established, and these data include process flow, process condition and yields, mass and radioactivity balance of radionuclides, process safety considerations, etc. And also, these data will be utilized for basic and detail design of hot cell facility, secured conservative safety and effective operability

  3. Fuel Effects on Ignition and Their Impact on Advanced Combustion Engines (Poster)

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, J.; Li, H.; Neill, S.

    2006-08-01

    The objective of this report is to develop a pathway to use easily measured ignition properties as metrics for characterizing fuels in advanced combustion engine research--correlate IQT{trademark} measured parameters with engine data. In HCCL engines, ignition timing depends on the reaction rates throughout compression stroke: need to understand sensitivity to T, P, and [O{sub 2}]; need to rank fuels based on more than one set of conditions; and need to understand how fuel composition (molecular species) affect ignition properties.

  4. SunLine Transit Agency Advanced Technology Fuel Cell Bus Evaluation: Fourth Results Report

    Energy Technology Data Exchange (ETDEWEB)

    Eudy, L.; Chandler, K.

    2013-01-01

    SunLine Transit Agency, which provides public transit services to the Coachella Valley area of California, has demonstrated hydrogen and fuel cell bus technologies for more than 10 years. In May 2010, SunLine began demonstrating the advanced technology (AT) fuel cell bus with a hybrid electric propulsion system, fuel cell power system, and lithium-based hybrid batteries. This report describes operations at SunLine for the AT fuel cell bus and five compressed natural gas buses. The U.S. Department of Energy's National Renewable Energy Laboratory (NREL) is working with SunLine to evaluate the bus in real-world service to document the results and help determine the progress toward technology readiness. NREL has previously published three reports documenting the operation of the fuel cell bus in service. This report provides a summary of the results with a focus on the bus operation from February 2012 through November 2012.

  5. Advanced evaluation of spent fuel in long term wet storage in Slovakia

    International Nuclear Information System (INIS)

    Experiences with an advanced spent nuclear fuel management in Slovakia are presented in this paper. The evaluation and monitoring procedures are based on practices at the Slovak wet interim spent fuel storage facility in NPP Jaslovske Bohunice. Since 1999, leak testing of WWER-440 fuel assemblies are provided by special leak tightness detection system 'Sipping in Pool' delivered by Framatome-ANP with external heating for the precise defects determination. In 2006 a new inspection stand 'SVYP-440' for the monitoring of spent nuclear fuel condition was inserted. This stand has the possibility to open WWER-440 fuel assemblies and examine fuel elements. Optimal ways of spent fuel disposal and monitoring of nuclear fuel condition were designed. With appropriate approach of conservativeness, new factor for specifying of spent fuel leak tightness is introduced in the paper. By using computer simulations (based on SCALE 4.4a code) for fission products creation and measurements by system 'Sipping in Pool', the limit values of leak tightness were established. The mean value of the leak tightness coefficient is expected to be around 10-10

  6. GOVERNOR ELECTRONICS FOR DIESEL ENGINES : High availability platform for real-time control and advanced fuel efficiency algorithms

    OpenAIRE

    Holmström, Johnny

    2013-01-01

    Fossil fuel is a rare commodity and the combustion of this fuel results in negative environmental effects. This paper evaluates and validates the electronics needed to run intelligent algorithms to lower the fuel consumption for commercial vessels. This is done by integrating advanced fuel saving functions into an electronic device that controls the fuel injection of large diesel engines, as known as a diesel engine governor. The control system is classified as a safety critical system. This ...

  7. Feasible advanced fuel cycle options for CANDU reactors in the Republic of Korea

    International Nuclear Information System (INIS)

    Taking into account the view points on nuclear safety, nuclear waste, non-proliferation and economics from the public, international environment, and utilities, the SEU/RU and DUPIC fuel cycles would be feasible options of advanced fuel cycles for CANDU-PHWRs in the Republic of Korea in the mid- and long-terms, respectively. Comparing with NU fuel, 0.9 % or 1.2 % SEU fuel would increase fuel burnup and hence reduce the spent fuel arisings by a factor of 2 or 3, and also could reduce CANDU fuel cycle costs by 20 to 30%. RU offers similar benefits as 0.9% SEU and is very attractive due to the significantly improved fuel cycle economics, substantially increased burnups, large reduction in fuel requirements as well as in spent fuel arisings. For RU use in a CANDU reactor, re-enrichment is not required. There are 25,000 tes RU produced from reprocessing operations in Europe and Japan, which would theoretically provide sufficient fuel for 500 CANDU 6 reactor-years of operation. According to the physics, thermal-hydraulic and thermal-mechanical assessments of CANFLEX-0.9% RU fuel for a CANDU-6 reactor, the fuel could be introduced into the reactor in a straight-forward fashion. A series of assessments of CANFLEX-DUPIC physics on the compatibility of the fuel design in the existing CANDU 6 reactors has shown that the poisoning of the central element of DUPIC with, for example, natural dysprosium, reduces the void reactivity of the fuel, and that a 2 bundle shift refuelling scheme would be the most appropriate in-core fuel management scheme for a CANDU-6 reactor. The average discharge burnup is ∼15 MWd/kgHE. Although these results have shown promising results for the DUPIC fuel cycle, more in-depth studies are required in the areas of ROP system, large LOCA safety analyses, and so on. The recycling fuel cycles of RU and DUPIC for CANDU are expected to achieve the environmental 3R's (Reduce, Reuse, Recycle) as applied to global energy use in the short- and long

  8. Summary of session 1: 'Advanced fuel cycles and reactor concepts'

    International Nuclear Information System (INIS)

    Full text: During the opening session of the Scientific Forum presentations were made by the IAEA Director General, Mohamed ElBaradei, Mr. Carlo Rubbia from Italy, and by the Chairman of the Scientific Forum, Mr. B. Bigot from France. Mr. A. Kakodkar from India was the moderator of the session The audience included some 180 participants. Four keynote speakers from the UK, Argentina, Japan and India as well as one panelist from the USA contributed to the first session. The highlights can be summarized as follows: Nuclear energy as an emission free energy source is indispensable for sustainable development. The importance of continuous R and D in support of innovative reactors and fuel cycles was stressed. The overall goal for these technologies includes better uranium resource utilization and improved waste management strategies. Moreover, the development of accelerator-driven systems for transmutation and energy production is regarded as an important long-term option. In this respect the closure of the fuel cycle with fast reactors is considered to be essential. In meeting these objectives, a focus on economics, proliferation resistance, and safety is paramount. The transition to innovative nuclear energy systems from current systems must be gradual via a combination of evolutionary and innovative technologies. (author)

  9. Scope and verification of a Fissile Material (Cutoff) Treaty

    Energy Technology Data Exchange (ETDEWEB)

    Hippel, Frank N. von [Program on Science and Global Security, Princeton University, 221 Nassau St., 2nd floor, Princeton, NJ 08542 (United States)

    2014-05-09

    A Fissile Material Cutoff Treaty (FMCT) would ban the production of fissile material - in practice highly-enriched uranium and separated plutonium - for weapons. It has been supported by strong majorities in the United Nations. After it comes into force, newly produced fissile materials could only be produced under international - most likely International Atomic Energy Agency - monitoring. Many non-weapon states argue that the treaty should also place under safeguards pre-existing stocks of fissile material in civilian use or declared excess for weapons so as to make nuclear-weapons reductions irreversible. This paper discusses the scope of the FMCT, the ability to detect clandestine production and verification challenges in the nuclear-weapons states.

  10. Operation experience of the advanced fuel assemblies at Unit 1 of Volgodonsk NPP within four fuel cycles

    International Nuclear Information System (INIS)

    The first commissioning of Volgodonsk NPP Unit 1 with standard reactor WWER-1000 (project V-320) was in 2001. The reactor core, starting from the first fuel charge, was arranged completely with Advanced Fuel Assemblies (AFAs). In this way, it is possible to obtain the experience in startup and operation of the core, completely arranged with AFAs, and also to get a possibility of performing the comprehensive check for justification of newly commissioned units and justification of design solutions accepted in the design of reactor core for Taiwan NPP, Bushehr NPP and Kudankulam NPP. The first fuel charge of the Volgodonsk NPP Unit 1 is a reference and unified for Tiawan NPP (V-428), Bushehr NPP (V-446), Kudankulam NPP(V-412) with small differences caused by design features of RP V-320. The first core charge of Unit 1 of Volgodonsk NPP was arranged of 163 AFAs, comprising 61 CPS ARs and 42 BAR bundles. The subsequent fuel charges were arranged of AFAs with gadolinium oxide integrated into fuel instead of BAR. By 2005 the results of operation of the core at Unit 1 of Volgodonsk NPP during four fuel cycles showed that AFA is sufficiently reliable and serviceable. The activity of the primary coolant of the Volgodonsk NPP is at stable low level. During the whole time of the core operation of the Volgodonsk NPP Unit 1 no leaky AFAs were revealed. The modifications of the internals, made during pre-operational work, are reasonable and effective to provide for fuel mechanical stability in the course of operation. The modifications, made in AFA structure during operation of the Volgodonsk NPP Unit 1, are aimed at improving the service and operational reliability of its components. Correctness of the solutions taken is confirmed by AFAs operation experience both at the Volgodonsk NPP, and at other operating Russian NPPs

  11. Use of freeze-casting in advanced burner reactor fuel design

    Energy Technology Data Exchange (ETDEWEB)

    Lang, A. L.; Yablinsky, C. A.; Allen, T. R. [Dept. of Engineering Physics, Univ. of Wisconsin Madison, 1500 Engineering Drive, Madison, WI 53711 (United States); Burger, J.; Hunger, P. M.; Wegst, U. G. K. [Thayer School of Engineering, Dartmouth College, 8000 Cummings Hall, Hanover, NH 03755 (United States)

    2012-07-01

    This paper will detail the modeling of a fast reactor with fuel pins created using a freeze-casting process. Freeze-casting is a method of creating an inert scaffold within a fuel pin. The scaffold is created using a directional solidification process and results in open porosity for emplacement of fuel, with pores ranging in size from 300 microns to 500 microns in diameter. These pores allow multiple fuel types and enrichments to be loaded into one fuel pin. Also, each pore could be filled with varying amounts of fuel to allow for the specific volume of fission gases created by that fuel type. Currently fast reactors, including advanced burner reactors (ABR's), are not economically feasible due to the high cost of operating the reactors and of reprocessing the fuel. However, if the fuel could be very precisely placed, such as within a freeze-cast scaffold, this could increase fuel performance and result in a valid design with a much lower cost per megawatt. In addition to competitive costs, freeze-cast fuel would also allow for selective breeding or burning of actinides within specific locations in fast reactors. For example, fast flux peak locations could be utilized on a minute scale to target specific actinides for transmutation. Freeze-cast fuel is extremely flexible and has great potential in a variety of applications. This paper performs initial modeling of freeze-cast fuel, with the generic fast reactor parameters for this model based on EBR-II. The core has an assumed power of 62.5 MWt. The neutronics code used was Monte Carlo N-Particle (MCNP5) transport code. Uniform pore sizes were used in increments of 100 microns. Two different freeze-cast scaffold materials were used: ceramic (MgO-ZrO{sub 2}) and steel (SS316L). Separate models were needed for each material because the freeze-cast ceramic and metal scaffolds have different structural characteristics and overall porosities. Basic criticality results were compiled for the various models

  12. Use of freeze-casting in advanced burner reactor fuel design

    International Nuclear Information System (INIS)

    This paper will detail the modeling of a fast reactor with fuel pins created using a freeze-casting process. Freeze-casting is a method of creating an inert scaffold within a fuel pin. The scaffold is created using a directional solidification process and results in open porosity for emplacement of fuel, with pores ranging in size from 300 microns to 500 microns in diameter. These pores allow multiple fuel types and enrichments to be loaded into one fuel pin. Also, each pore could be filled with varying amounts of fuel to allow for the specific volume of fission gases created by that fuel type. Currently fast reactors, including advanced burner reactors (ABR's), are not economically feasible due to the high cost of operating the reactors and of reprocessing the fuel. However, if the fuel could be very precisely placed, such as within a freeze-cast scaffold, this could increase fuel performance and result in a valid design with a much lower cost per megawatt. In addition to competitive costs, freeze-cast fuel would also allow for selective breeding or burning of actinides within specific locations in fast reactors. For example, fast flux peak locations could be utilized on a minute scale to target specific actinides for transmutation. Freeze-cast fuel is extremely flexible and has great potential in a variety of applications. This paper performs initial modeling of freeze-cast fuel, with the generic fast reactor parameters for this model based on EBR-II. The core has an assumed power of 62.5 MWt. The neutronics code used was Monte Carlo N-Particle (MCNP5) transport code. Uniform pore sizes were used in increments of 100 microns. Two different freeze-cast scaffold materials were used: ceramic (MgO-ZrO2) and steel (SS316L). Separate models were needed for each material because the freeze-cast ceramic and metal scaffolds have different structural characteristics and overall porosities. Basic criticality results were compiled for the various models. Preliminary results

  13. Proliferation resistance of advanced sustainable nuclear fuel cycles

    International Nuclear Information System (INIS)

    Intrinsic and extrinsic proliferation barriers of a pyro-process-based nuclear fuel cycle are discussed. While technical characteristics of the process raise new challenges for safeguards, others naturally facilitate the implementation of more integrated schemes for unattended continuous monitoring. In particular, the concept of operations accountability and model-assisted methods are revisited. While traditional safeguards constructs, such as material control and accountability, place greater emphasis on input/output characterization of nuclear processes, a model- based discrete event accountability approach could explicitly verify not only facility use but also internal operational dynamics. Under the proposed remote integral safeguards approach, transparency can be achieved efficiently, without divulging competitive or national security sensitive information. (author)

  14. Advanced Materials for PEM-Based Fuel Cell Systems

    Energy Technology Data Exchange (ETDEWEB)

    James E. McGrath

    2005-10-26

    Proton exchange membrane fuel cells (PEMFCs) are quickly becoming attractive alternative energy sources for transportation, stationary power, and small electronics due to the increasing cost and environmental hazards of traditional fossil fuels. Two main classes of PEMFCs include hydrogen/air or hydrogen/oxygen fuel cells and direct methanol fuel cells (DMFCs). The current benchmark membrane for both types of PEMFCs is Nafion, a perfluorinated sulfonated copolymer made by DuPont. Nafion copolymers exhibit good thermal and chemical stability, as well as very high proton conductivity under hydrated conditions at temperatures below 80 °C. However, application of these membranes is limited due to their high methanol permeability and loss of conductivity at high temperatures and low relative humidities. These deficiencies have led to the search for improved materials for proton exchange membranes. Potential PEMs should have good thermal, hydrolytic, and oxidative stability, high proton conductivity, selective permeability, and mechanical durability over long periods of time. Poly(arylene ether)s, polyimides, polybenzimidazoles, and polyphenylenes are among the most widely investigated candidates for PEMs. Poly(arylene ether)s are a promising class of proton exchange membranes due to their excellent thermal and chemical stability and high glass transition temperatures. High proton conductivity can be achieved through post-sulfonation of poly(arylene ether) materials, but this most often results in very high water sorption or even water solubility. Our research has shown that directly polymerized poly(arylene ether) copolymers show important advantages over traditional post-sulfonated systems and also address the concerns with Nafion membranes. These properties were evaluated and correlated with morphology, structure-property relationships, and

  15. Advanced Materials for PEM-Based Fuel Cell Systems

    Energy Technology Data Exchange (ETDEWEB)

    James E. McGrath; Donald G. Baird; Michael von Spakovsky

    2005-10-26

    Proton exchange membrane fuel cells (PEMFCs) are quickly becoming attractive alternative energy sources for transportation, stationary power, and small electronics due to the increasing cost and environmental hazards of traditional fossil fuels. Two main classes of PEMFCs include hydrogen/air or hydrogen/oxygen fuel cells and direct methanol fuel cells (DMFCs). The current benchmark membrane for both types of PEMFCs is Nafion, a perfluorinated sulfonated copolymer made by DuPont. Nafion copolymers exhibit good thermal and chemical stability, as well as very high proton conductivity under hydrated conditions at temperatures below 80 degrees C. However, application of these membranes is limited due to their high methanol permeability and loss of conductivity at high temperatures and low relative humidities. These deficiencies have led to the search for improved materials for proton exchange membranes. Potential PEMs should have good thermal, hydrolytic, and oxidative stability, high proton conductivity, selective permeability, and mechanical durability over long periods of time. Poly(arylene ether)s, polyimides, polybenzimidazoles, and polyphenylenes are among the most widely investigated candidates for PEMs. Poly(arylene ether)s are a promising class of proton exchange membranes due to their excellent thermal and chemical stability and high glass transition temperatures. High proton conductivity can be achieved through post-sulfonation of poly(arylene ether) materials, but this most often results in very high water sorption or even water solubility. Our research has shown that directly polymerized poly(arylene ether) copolymers show important advantages over traditional post-sulfonated systems and also address the concerns with Nafion membranes. These properties were evaluated and correlated with morphology, structure-property relationships, and states of water in the membranes. Further improvements in properties were achieved through incorporation of inorganic

  16. CANDU advanced fuel cycles: key to energy sustainability

    International Nuclear Information System (INIS)

    In the fast-growing economies of the Pacific Basin region, sustainability is an important requisite for new energy development. Many countries in this region have seen, and continue to see, very large increases in energy and electricity demand. The investment in any nuclear technology is large. Countries making that investment want to ensure that the technology can be sustained and that it can evolve in an ever-changing environment. Three key aspects in ensuring a sustainable energy future are: technological sustainability; economic sustainability; and environmental sustainability (including resource utilization). The fuel-cycle flexibility of the CANDU reactor provides a ready path to sustainable energy development in both the short and the long term. (author). 23 refs

  17. Ambient Laboratory Coater for Advanced Gas Reactor Fuel Development

    Energy Technology Data Exchange (ETDEWEB)

    Duane D. Bruns; Robert M. Counce; Irma D. Lima Rojas

    2010-06-09

    this research is targeted at developing improved experimentally-based scaling relationships for the hydrodynamics of shallow, gas-spouted beds of dense particles. The work is motivated by the need to more effctively scale up shallow spouted beds used in processes such as in the coating of nuclear fuel particles where precise control of solids and gas circulation is critically important. Experimental results reported here are for a 50 mm diameter spouted bed containing two different types of bed solids (alumina and zirconia) at different static bed depths and fluidized by air and helium. Measurements of multiple local average pressures, inlet gas pressure fluctuations, and spout height were used to characterize the bed hydrodynamics for each operating condition. Follow-on studies are planned that include additional variations in bed size, particle properties, and fluidizing gas. The ultimate objective is to identify the most important non-dimensional hydrodynamic scaling groups and possible spouted-bed design correlations based on these groups.

  18. ADVANCED HYDROGEN TRANSPORT MEMBRANES FOR VISION 21 FOSSIL FUEL PLANTS

    International Nuclear Information System (INIS)

    Eltron Research Inc., and team members CoorsTek, McDermott Technology, inc., Sued Chemie, Argonne National Laboratory, and Oak Ridge National Laboratory are developing an environmentally benign, inexpensive, and efficient method for separating hydrogen from gas mixtures produced during industrial processes, such as coal gasification. This project was motivated by the National Energy Technology Laboratory (NETL) Vision 21 initiative which seeks to economically eliminate environmental concerns associated with the use of fossil fuels. This objective is being pursued using dense membranes based in part on Eltron-patented ceramic materials with a demonstrated ability for proton and electron conduction. The technical goals are being addressed by modifying single-phase and composite membrane composition and microstructure to maximize proton and electron conductivity without loss of material stability. Ultimately, these materials must enable hydrogen separation at practical rates under ambient and high-pressure conditions, without deactivation in the presence of feedstream components such as carbon dioxide, water, and sulfur

  19. Magnetic separation - Advanced nanotechnology for future nuclear fuel recycle

    International Nuclear Information System (INIS)

    The unique properties of magnetic nanoparticles (MNPs), such as their extremely small size and high surface area to volume ratio, provide better kinetics for the adsorption of metal ions from aqueous solutions. In this work, we demonstrated the separation of minor actinides using complex conjugates of MNPs with diethylenetriamine-pentaacetic acid (DTPA) chelator. The sorption results show the strong affinity of DTPA towards Am (III) and Pu (IV) by extracting 97% and 80% of actinides, respectively. It is shown that the extraction process is highly dependent on the pH of the solution. If these long-term heat generating actinides can be efficiently removed from the used fuel raffinates, the volume of material that can be placed in a given amount of repository space can be significantly increased. (authors)

  20. Ambient Laboratory Coater for Advanced Gas Reactor Fuel Development

    International Nuclear Information System (INIS)

    This research is targeted at developing improved experimentally-based scaling relationships for the hydrodynamics of shallow, gas-spouted beds of dense particles. The work is motivated by the need to more effctively scale up shallow spouted beds used in processes such as in the coating of nuclear fuel particles where precise control of solids and gas circulation is critically important. Experimental results reported here are for a 50 mm diameter spouted bed containing two different types of bed solids (alumina and zirconia) at different static bed depths and fluidized by air and helium. Measurements of multiple local average pressures, inlet gas pressure fluctuations, and spout height were used to characterize the bed hydrodynamics for each operating condition. Follow-on studies are planned that include additional variations in bed size, particle properties, and fluidizing gas. The ultimate objective is to identify the most important non-dimensional hydrodynamic scaling groups and possible spouted-bed design correlations based on these groups.