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Sample records for th-233u based msr

  1. Cost-based optimizations of power density and target-blanket modularity for 232Th/233U-based ADEP

    International Nuclear Information System (INIS)

    Krakowski, R.A.

    1995-01-01

    A cost-based parametric systems model is developed for an Accelerator-Driven Energy Production (ADEP) system based on a 232 Th/ 233 U fuel cycle and a molten-salt (LiF/BeF 2 /ThF 3 ) fluid-fuel primary system. Simplified neutron-balance, accelerator, reactor-core, chemical-processing, and balance-of-plant models are combined parametrically with a simplified costing model. The main focus of this model is to examine trade offs related to fission power density, reactor-core modularity, 233 U breeding rate, and fission product transmutation capacity

  2. Th/U-233 multi-recycle in PWRs

    International Nuclear Information System (INIS)

    Yun, D.; Kim, T.K.; Taiwo, T.A.

    2010-01-01

    Th/U-233 multirecycle. Recent studies done internationally and in the U.S. are briefly summarized. Additionally, the previous U.S. thorium breeder experiment in the Shippingport reactor is briefly discussed. The objective of this work and the reactor design issues associated with multirecycle of Th/U-233 are discussed in Section 3. The approaches required to achieve a sustainable system are discussed and evaluated. Homogeneous assembly modeling results are presented in this section. In Section 4, a 17-by-17 heterogeneous assembly design has been selected and evaluated, based on its positive attributes for sustainable Th/U-233 multirecycle. A feasibility study is briefly discussed at the end of this section followed by recommendations for future activities. Section 5 discusses the attributes of the 17-by-17 heterogeneous assembly design. The material mass flow data and fuel cycle impact data are reported in this section. Discussions on the fuel cycle implications of thorium fuel utilization are provided in Section 6. This includes information on fuel sources, fuel manufacturing, fuel reprocessing, and re-fabrication. The conclusions of the study are provided in Section 7.

  3. Th/U-233 multi-recycle in PWRs.

    Energy Technology Data Exchange (ETDEWEB)

    Yun, D.; Kim, T. K.; Taiwo, T. A.; Nuclear Engineering Division

    2010-09-07

    Th/U-233 multirecycle. Recent studies done internationally and in the U.S. are briefly summarized. Additionally, the previous U.S. thorium breeder experiment in the Shippingport reactor is briefly discussed. The objective of this work and the reactor design issues associated with multirecycle of Th/U-233 are discussed in Section 3. The approaches required to achieve a sustainable system are discussed and evaluated. Homogeneous assembly modeling results are presented in this section. In Section 4, a 17-by-17 heterogeneous assembly design has been selected and evaluated, based on its positive attributes for sustainable Th/U-233 multirecycle. A feasibility study is briefly discussed at the end of this section followed by recommendations for future activities. Section 5 discusses the attributes of the 17-by-17 heterogeneous assembly design. The material mass flow data and fuel cycle impact data are reported in this section. Discussions on the fuel cycle implications of thorium fuel utilization are provided in Section 6. This includes information on fuel sources, fuel manufacturing, fuel reprocessing, and re-fabrication. The conclusions of the study are provided in Section 7.

  4. Computational Approach in Determination of 233U and 233Th Fermi Energy

    International Nuclear Information System (INIS)

    Kurniadi, R.; Perkasa, Y. S.; Waris, A.

    2010-01-01

    There are several methods to get Fermi energy such as hermit polynomial expansion and Wigner-Kirkwood expansion, these are analytical method. In this paper will be discussed numerical approach of calculating Fermi energy of 233 Th and 233 U nuclei. Our work demonstrates the simple technique of determining Fermi energy.

  5. High conversion Th-U{sup 233} fuel assembly for current generation of PWRs

    Energy Technology Data Exchange (ETDEWEB)

    Baldova, D.; Fridman, E. [Reactor Safety Div., Helmholtz-Zentrum Dresden-Rossendorf, POB 510119, Dresden, 01314 (Germany)

    2012-07-01

    This paper presents a preliminary design of a high conversion Th-U{sup 233} fuel assembly applicable for current generation of Pressurized Water Reactor (PWRs). The considered fuel assembly has a typical 17 x 17 PWR lattice. However in order to increase the conversion of Th{sup 232} to U{sup 233}, the assembly was subdivided into the two regions called seed and blanket. The central seed region has a higher than blanket U{sup 233} content and acts as a neutron source for the peripheral blanket region. The latest acts as a U{sup 233} breeder. While the seed fuel pins have a standard dimensions the blanket fuel radius was increased in order to reduce the moderation and to facilitate the resonance neutron absorption in blanket Th{sup 232}. The U{sup 233} content in the seed and blanket regions was optimized to achieve maximal initial to discharged fissile inventory ratio (FIR) taking into account the target fuel cycle length of 12 months with 3-batch reloading scheme. In this study the neutronic calculations were performed on the fuel assembly level using Helios deterministic lattice transport code. The fuel cycle length and the core k{sub eff} were estimated by applying the Non Linear Reactivity Model. The applicability of the HELIOS code for the analysis of the Th-based high conversion designs was confirmed with the help of continuous-energy Monte-Carlo code SERPENT. The results of optimization studies show that for the heterogeneous seed and blanket (SB) fuel assembly the FIR of about 0.95 can be achieved. (authors)

  6. Neutron inelastic-scattering cross sections of 232Th, 233U, 235U, 238U, 239Pu and 240Pu

    International Nuclear Information System (INIS)

    Smith, A.B.; Guenther, P.T.

    1982-01-01

    Differential-neutron-emission cross sections of 232 Th, 233 U, 235 U, 238 U, 239 Pu and 240 Pu are measured between approx. = 1.0 and 3.5 MeV with the angle and magnitude detail needed to provide angle-integrated emission cross sections to approx. 232 Th, 233 U, 235 U and 238 U inelastic-scattering values, poor agreement is observed for 240 Pu, and a serious discrepancy exists in the case of 239 Pu

  7. Preliminary Study for Radioactivity Evaluation of MSR compared with LWR

    International Nuclear Information System (INIS)

    Lee, Geun Hyeong; Kim, Hee Reyoung

    2014-01-01

    LWR uses fuel as 235 U and fissile material as solid (enriched uranium). Those cannot control its component artificially and hard to change fuel frequently. Therefore this fuel remains as much as possible. That makes risk of high radiation leakage because of long neutron irradiation time. On the other hand, MSR (Molten Salt Reactor) uses fuel as thorium-uranium; fissile 233 U when 232 Th absorbs one neutron, and fissile material as liquid (molten salt). It has plenty of benefits respect to radioactive safety. It leads nuclear fuel dump when accident happens, diminishes basic fission substances' radiation and even the cost (Th exist 3∼4 times more on the earth compared with natural uranium). Source term is much lower than conventional LWR in order to processing time. Radiation exposure from volatile fission products in severe accidents is thought to be negligible due to the continuous removal mechanism. The generation of high level radioactive wastes from MSR is estimated to be much smaller than that of conventional LWR because of its less converting probability of thorium to minor actinides. It was thought the fundamental approach to MSR would make it possible to realize the safety of reactor when considering the severe accidents affecting on nuclear power plants due to natural disaster

  8. Preliminary Study for Radioactivity Evaluation of MSR compared with LWR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Geun Hyeong; Kim, Hee Reyoung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2014-05-15

    LWR uses fuel as {sup 235}U and fissile material as solid (enriched uranium). Those cannot control its component artificially and hard to change fuel frequently. Therefore this fuel remains as much as possible. That makes risk of high radiation leakage because of long neutron irradiation time. On the other hand, MSR (Molten Salt Reactor) uses fuel as thorium-uranium; fissile {sup 233}U when {sup 232}Th absorbs one neutron, and fissile material as liquid (molten salt). It has plenty of benefits respect to radioactive safety. It leads nuclear fuel dump when accident happens, diminishes basic fission substances' radiation and even the cost (Th exist 3∼4 times more on the earth compared with natural uranium). Source term is much lower than conventional LWR in order to processing time. Radiation exposure from volatile fission products in severe accidents is thought to be negligible due to the continuous removal mechanism. The generation of high level radioactive wastes from MSR is estimated to be much smaller than that of conventional LWR because of its less converting probability of thorium to minor actinides. It was thought the fundamental approach to MSR would make it possible to realize the safety of reactor when considering the severe accidents affecting on nuclear power plants due to natural disaster.

  9. Evaluation of cross sections of Th-232 and U-233

    International Nuclear Information System (INIS)

    Dias, A.M.

    1978-01-01

    The cross sections in multigroups of Th-232 and U-233 are evaluated by comparison of theoretical results and experimental data obtained through experiments on the fast reactors IBR-I, EBR-II, BR-I and AETR. The deviation between calculated values and experimental results is about 10%. They are therefore satisfatory for neutronic calculations [pt

  10. Resonance Region Covariance Analysis Method and New Covariance Data for Th-232, U-233, U-235, U-238, and Pu-239

    International Nuclear Information System (INIS)

    Leal, Luiz C.; Arbanas, Goran; Derrien, Herve; Wiarda, Dorothea

    2008-01-01

    Resonance-parameter covariance matrix (RPCM) evaluations in the resolved resonance region were done for 232Th, 233U, 235U, 238U, and 239Pu using the computer code SAMMY. The retroactive approach of the code SAMMY was used to generate the RPCMs for 233U, 235U. RPCMs for 232Th, 238U and 239Pu were generated together with the resonance parameter evaluations. The RPCMs were then converted in the ENDF format using the FILE32 representation. Alternatively, for computer storage reasons, the FILE32 was converted in the FILE33 cross section covariance matrix (CSCM). Both representations were processed using the computer code PUFF-IV. This paper describes the procedures used to generate the RPCM with SAMMY.

  11. Fuel utilization improvement in PWRs using the denatured 233U-Th cycle

    International Nuclear Information System (INIS)

    Jones, H.M.; Schwenk, G.A.; Toops, E.C.; Yotinen, V.O.

    1980-06-01

    A number of changes in PWR core design and/or operating strategy were evaluated to assess the fuel utilization improvement achievable by their implementation in a PWR using thorium-based fuel and operating in a recycle mode. The reference PWR for this study was identical to the B and W Standard Plant except that the fuel pellets were of denatured ( 233 U/ 238 U-Th)O 2 . An initial scoping study identified the three most promising improvement concepts as (1) a very tight lattice, (2) thorium blankets, and (3) ThO 2 rods placed in available guide tubes. A conceptual core design incorporating these changes was then developed, and the fuel utilization of this modified design was compared with that of the reference case

  12. Contribution to the study of {sup 233}U production with MOX-ThPu fuel in PWR reactor. Transition scenarios towards Th/{sup 233}U iso-generating concepts in thermal spectrum. Development of the MURE fuel evolution code; Contribution a l'etude de la production d'{sup 233}U en combustible MOX-ThPu en reacteur a eau sous pression. Scenarios de transition vers des concepts isogenerateurs Th/{sup 233}U en spectre thermique. Developpement du code MURE d'evolution du combustible

    Energy Technology Data Exchange (ETDEWEB)

    Michel-Sendis, F

    2006-12-15

    If nuclear power is to provide a significant fraction of the growing world energy demand, only through the breeding concept can the development of sustainable nuclear energy become a reality. The study of such a transition, from present-day nuclear technologies to future breeding concepts is therefore pertinent. Among these future concepts, those using the thorium cycle Th/U-233 in a thermal neutron spectrum are of particular interest; molten-salt type thermal reactors would allow for breeding while requiring comparatively low initial inventories of U-233. The upstream production of U-233 can be obtained through the use of thorium-plutonium mixed oxide fuel in present-day light water reactors. This work presents, firstly, the development of the MURE evolution code system, a C++ object-oriented code that allows the study, through Monte Carlo (M.C.) simulation, of nuclear reactors and the evolution of their fuel under neutron irradiation. The M.C. methods are well-suited for the study of any reactor, whether it'd be an existing reactor using a new kind of fuel or a future concept altogether, the simulation is only dependent on nuclear data. Exact and complex geometries can be simulated and continuous energy particle transport is performed. MURE is an interface with MCNP, the well-known and validated transport code, that allows, among other functionalities, to simulate constant power and constant reactivity evolutions. Secondly, the study of MOX ThPu fuel in a conventional light water reactor (REP) is presented; it explores different plutonium concentrations and isotopic qualities in order to evaluate their safety characteristics. Simulation of their evolution allows us to quantify the production of U-233 at the end of burnup. Last, different french scenarios validating a possible transition towards a park of thermal Th/U-233 breeders, are presented. In these scenarios, U-233 is produced in ThPu moxed light water reactors. (author)

  13. Contribution to the study of {sup 233}U production with MOX-ThPu fuel in PWR reactor. Transition scenarios towards Th/{sup 233}U iso-generating concepts in thermal spectrum. Development of the MURE fuel evolution code; Contribution a l'etude de la production d'{sup 233}U en combustible MOX-ThPu en reacteur a eau sous pression. Scenarios de transition vers des concepts isogenerateurs Th/{sup 233}U en spectre thermique. Developpement du code MURE d'evolution du combustible

    Energy Technology Data Exchange (ETDEWEB)

    Michel-Sendis, F

    2006-12-15

    If nuclear power is to provide a significant fraction of the growing world energy demand, only through the breeding concept can the development of sustainable nuclear energy become a reality. The study of such a transition, from present-day nuclear technologies to future breeding concepts is therefore pertinent. Among these future concepts, those using the thorium cycle Th/U-233 in a thermal neutron spectrum are of particular interest; molten-salt type thermal reactors would allow for breeding while requiring comparatively low initial inventories of U-233. The upstream production of U-233 can be obtained through the use of thorium-plutonium mixed oxide fuel in present-day light water reactors. This work presents, firstly, the development of the MURE evolution code system, a C++ object-oriented code that allows the study, through Monte Carlo (M.C.) simulation, of nuclear reactors and the evolution of their fuel under neutron irradiation. The M.C. methods are well-suited for the study of any reactor, whether it'd be an existing reactor using a new kind of fuel or a future concept altogether, the simulation is only dependent on nuclear data. Exact and complex geometries can be simulated and continuous energy particle transport is performed. MURE is an interface with MCNP, the well-known and validated transport code, that allows, among other functionalities, to simulate constant power and constant reactivity evolutions. Secondly, the study of MOX ThPu fuel in a conventional light water reactor (REP) is presented; it explores different plutonium concentrations and isotopic qualities in order to evaluate their safety characteristics. Simulation of their evolution allows us to quantify the production of U-233 at the end of burnup. Last, different french scenarios validating a possible transition towards a park of thermal Th/U-233 breeders, are presented. In these scenarios, U-233 is produced in ThPu moxed light water reactors. (author)

  14. The status of 232Th and 233U for CENDL-3.0

    International Nuclear Information System (INIS)

    Liu Ping

    2003-01-01

    The new version CENDL-3.0 of China: Evaluated nuclear data library has been updated, and contains about 200 nuclides. Among them, the data of following nuclides have been newly evaluated or reevaluated: fissile nuclides 15, structure materials 18, light nuclides 3, fission products 116. The 232 Th and 233 U are newly evaluated

  15. A LMFBR for thorium utilization and for the U233/Th fuel rods specification

    International Nuclear Information System (INIS)

    Ishiguro, Y.; Dias, A.F.

    1982-01-01

    The use of U 233 /Th as fuel in the middle part of LMFBR core and the Pu/U in the external part of the core, are proposed. The basic neutronic and safety characteristics and the specifications of fuel rods to be used in the internal core, are presented. (E.G.) [pt

  16. Study on Utilization of Super Grade Plutonium in Molten Salt Reactor FUJI-U3 using CITATION Code

    Science.gov (United States)

    Wulandari, Cici; Waris, Abdul; Pramuditya, Syeilendra; Asril, Pramutadi AM; Novitrian

    2017-07-01

    FUJI-U3 type of Molten Salt Reactor (MSR) has a unique design since it consists of three core regions in order to avoid the replacement of graphite as moderator. MSR uses floride as a nuclear fuel salt with the most popular chemical composition is LiF-BeF2-ThF4-233UF4. ThF4 and 233UF4 are the fertile and fissile materials, respectively. On the other hand, LiF and BeF2 working as both fuel and heat transfer medium. In this study, the super grade plutonium will be utilized as substitution of 233U since plutonium is easier to be obtained compared to 233U as main fuel. Neutronics calculation was performed by using PIJ and CITATION modules of SRAC 2002 code with JENDL 3.2 as nuclear data library.

  17. Compilation of criticality data involving thorium or 233U and light water moderation

    Energy Technology Data Exchange (ETDEWEB)

    Gore, B.F.

    1978-07-01

    The literature has been searched for criticality data for light water moderated systems which contain thorium or /sup 233/U, and data found are compiled herein. They are from critical experiments, extrapolations, and exponential experiments performed with homogeneous solutions and metal spheres of /sup 233/U; with lattices of fuel rods containing highly enriched /sup 235/UO/sub 2/ - ThO/sub 2/ and /sup 233/UO/sub 2/ - ThO/sub 2/; and with arrays of cyclinders of /sup 233/U solutions. The extent of existing criticality data has been compared with that necessary to implement a thorium-based fuel cycle. No experiments have been performed with any solutions containing thorium. Neither do data exist for homogeneous /sup 233/U systems with H/U < 34, except for solid metal systems. Arrays of solution cylinders up to 3 x 3 x 3 have been studied. Data for solutions containing fixed or soluble poisons are very limited. All critical lattices using /sup 233/UO/sub 2/ - ThO/sub 2/ fuels (LWBR program) were zoned radially, and in most cases axially also. Only lattice experiments using /sup 235/UO/sub 2/ - ThO/sub 2/ fuels have been performed using a single fuel rod type. Critical lattices of /sup 235/UO/sub 2/ - ThO/sub 2/ rods poisoned with boron have been measured, but only exponential experiments have been performed using boron-poisoned lattices of /sup 233/UO/sub 2/ - ThO/sub 2/ rods. No criticality data exist for denatured fuels (containing significant amounts of /sup 238/U) in either solution or lattice configurations.

  18. Monte Carlo analyses of simple U233 O2-ThO2 and U235 O2-ThO2 lattices with ENDF/B-IV data (AWBA development program)

    International Nuclear Information System (INIS)

    Hardy, J. Jr.; Ullo, J.J.

    1980-09-01

    A number of water-moderated Th-U235 and Th-U233 lattice integral experiments were analyzed in a consistent manner, with ENDF/B-IV data and detailed Monte Carlo methods. These experiments provide a consistent test of the nuclear data. The ENDF/B-IV data are found to perform reasonably well. Adequate agreement is found with integral measurements of thorium capture. Calculated K/sub eff/ values show a generally coherent pattern which is consistent with K/sub eff/ results obtained for homogeneous aqueous critical assemblies. Harder prompt fission spectra for U233 and U235 can correct the principal discrepancy observed with ENDF/B-IV, a bias trend in K/sub eff/ attributed to an underprediction of leakage

  19. Development of Indian cross section data files for Th-232 and U-233 and integral validation studies

    International Nuclear Information System (INIS)

    Ganesan, S.

    1988-01-01

    This paper presents an overview of the tasks performed towards the development of Indian cross section data files for Th-232 and U-233. Discrepancies in various neutron induced reaction cross sections in various available evaluated data files have been obtained by processing the basic data into multigroup form and intercomparison of the latter. Interesting results of integral validation studies for capture, fission and (n,2n) cross sections for Th-232 by analyses of selected integral measurements are presented. In the resonance range, energy regions where significant differences in the calculated self-shielding factors for Th-232 occur have been identified by a comparison of self-shielded multigroup cross sections derived from two recent evaluated data files, viz., ENDF/B-V (Rev.2) and JENDL-2, for several dilutions and temperatures. For U-233, the three different basic data files ENDF/B-IV, JENDL-2 and ENDL-84 were intercompared. Interesting observations on the predictional capability of these files for the criticality of the spherical metal U-233 system are given. The current status of Indian data file is presented. (author) 62 ref

  20. Nuclear reactor for breeding 233U

    International Nuclear Information System (INIS)

    Bohanan, C.S.; Jones, D.H.; Raab, H.F. Jr.; Radkowsky, A.

    1976-01-01

    A light-water-cooled nuclear reactor capable of breeding 233 U for use in a light-water breeder reactor includes physically separated regions containing 235 U fissile material and 238 U fertile material and 232 Th fertile material and 239 Pu fissile material, if available. Preferably the 235 U fissile material and 238 U fertile material are contained in longitudinally movable seed regions and the 239 Pu fissile material and 232 Th fertile material are contained in blanket regions surrounding the seed regions. 1 claim, 5 figures

  1. Collective and single-particle excitations in the heavy deformable nuclei 234U, 233U, 231Th, 230Pa and 232Pa

    International Nuclear Information System (INIS)

    Kotthaus, Tanja

    2010-01-01

    In this thesis five heavy deformed isotopes from the mass region A≥230, namely 234 U, 233 U, 231 Th, 230 Pa and 232 Pa, were investigated by means of deuteron-induced neutron transfer reactions. The even-even isotope 234 U has been studied with the 4π-γ-spectrometer MINIBALL at the Cologne Tandem accelerator. Excited nuclei in the isotope 234 U were produced using the reaction 235 U(d,t) at a beam energy of 11 MeV. The target thickness was 3.5 mg/cm 2 . The analysis of the γγ-coincidence data yielded a reinterpretation of the level scheme in 12 cases. Considering its decay characteristics, the 4 + state at an excitation energy of 1886.7 keV is a potential candidate for a two-phonon vibrational state. The isotopes 233 U, 231 Th, 230 Pa and 232 Pa were investigated at the Munich Q3D spectrometer. For each isotope an angular distribution with angles between 5 and 45 were measured. In all four cases the energy of the polarized deuteron beam (vector polarization of 80%) was 22 MeV. As targets 234 U (160 μg/cm 2 ), 230 Th (140 μg/cm 2 ) and 231 Pa (140 μg/cm 2 ) were used. The experimental angular distributions were compared to results of DWBA calculations. For the odd isotope 233 U spin and parity for 33 states are assigned and in the other odd isotope 231 Th 22 assignments are made. The excitation spectra of the two odd-odd isotopes 230 Pa and 232 Pa were investigated for the first time. For the isotope 230 Pa 63 states below an excitation energy of 1.5 MeV are identified. Based on the new experimental data the Nilsson configuration of the ground state is either 1/2[530] p -5/2[633] n or 1/2[530] p +3/2[631] n . In addition 12 rotational bands are proposed and from this six values for the GM splitting energy are deduced as well as two new values for the Newby shift. In the other odd-odd isotope 232 Pa 40 states below an excitation energy of 850 keV are observed and suggestions for the groundstate band and its GM partner are made. From this one GM splitting

  2. Measurement of the fission cross section induced by fast neutrons of the {sup 232}Th/{sup 233}U nuclei within the innovating fuel cycles framework; Mesure de la section efficace de fission induite par neutrons rapides des noyaux {sup 232}Th/{sup 233}U dans le cadre des cycles de combustible innovants

    Energy Technology Data Exchange (ETDEWEB)

    Grosjean, C

    2005-03-15

    The thorium-U{sup 233} fuel cycle might provided safer and cleaner nuclear energy than the present Uranium/Pu fuelled reactors. Over the last 10 years, a vast campaign of measurements has been initiated to bring the precision of neutron data for the key nuclei (Th{sup 232}, Pa{sup 233} and U{sup 233}) at the level of those for the U-Pu cycle. This is the framework of these measurements, the energy dependent neutron induced fission cross section of Th{sup 232} and U{sup 233} has been measured from 1 to 7 MeV with a target accuracy lesser than 5 per cent. These measurements imply the accurate determination of the fission rate, the number of the target nuclei as well as the incident neutron flux impinging on the target, the latter has been obtained using the elastic scattering (n,p). The cross section of which is very well known in a large neutron energy domain ({approx} 0,5 % from 1 eV to 50 MeV) compared to the U{sup 235}(n,f) reaction. This technique has been applied for the first time to the Th{sup 232}(n,f) and U{sup 233}(n,f) cases. A Hauser-Feshbach statistical model has been developed. It consists of describing the different decay channels of the compound nucleus U{sup 234} from 0,01 to 10 MeV neutron energy. The parameters of this model were adjusted in order to reproduce the measured fission cross section of U{sup 233}. From these parameters, the cross sections from the following reactions could be extracted: inelastic scattering U{sup 233}(n,n'), radiative capture U{sup 233}(n,{gamma}) and U{sup 233}(n,2n). These cross sections are still difficult to measure by direct neutron reactions. The calculated values have allowed us to fill the lack of experimental data for the major fissile nucleus of the thorium cycle. (author)

  3. Measurement of neutron-induced fission cross-sections of Th232, U238, U233 and Np237 relative to U235 from 1 MeV to 200 MeV

    Energy Technology Data Exchange (ETDEWEB)

    Shcherbakov, O.A.; Laptev, A.B.; Petrov, G.A. [Petersburg Nuclear Physics Inst., Gatchina, Leningrad district (Russian Federation); Fomichev, A.V.; Donets, A.Y.; Osetrov, O.I.

    1998-11-01

    The measurements of neutron-induced cross-section ratios for Th232, U238, U233 and Np237 relative to U235 have been carried out in the energy range from 1 MeV up to 200 MeV using the neutron time-of-flight spectrometer GNEIS based on 1 GeV proton synchrocyclotron. Below 20 MeV, the results of present measurements are roughly in agreement with evaluated data though there are some discrepances to be resolved. (author)

  4. Mass dependence of azimuthal asymmetry in the fission of 232Th and 233,235,236,238U by polarized photons

    International Nuclear Information System (INIS)

    Denyak, V.V.; Khvastunov, V.M.; Paschuk, S.A.; Schelin, H.R.

    2013-01-01

    Fission of the even-even nuclei 232 Th, 236,238 U and even-odd nuclei 233,235 U by linearly polarized photons has been studied at excitation energies in the region of a giant dipole resonance. The performed investigations unambiguously showed the existence of the fragment mass dependence of the cross section azimuthal asymmetry in the photofission of 236 U and 238 U. In addition, the obtained results provided the first evidence for the possible difference between the asymmetry values in asymmetric and symmetric mass distribution regions in the case of 236 U. The measured cross section azimuthal asymmetry of the fission of 232 Th does not show any fragment mass dependence. In the even-odd nuclei 233 U and 235 U the difference between the far-asymmetric and other mass distribution regions was also observed but with the statistical uncertainty not small enough for definitive conclusion. (orig.)

  5. The fission cross sections of 230Th, 232Th, 233U, 234U, 236U, 238U, 237Np, 239Pu and 242Pu relative 235U at 14.74 MeV neutron energy

    International Nuclear Information System (INIS)

    Meadows, J.W.

    1986-12-01

    The measurement of the fission cross section ratios of nine isotopes relative to 235 U at an average neutron energy of 14.74 MeV is described with particular attention to the determination of corrections and to sources of error. The results are compared to ENDF/B-V and to other measurements of the past decade. The ratio of the neutron induced fission cross section for these isotopes to the fission cross section for 235 U are: 230 Th - 0.290 +- 1.9%; 232 Th - 0.191 +- 1.9%; 233 U - 1.132 +- 0.7%; 234 U - 0.998 +- 1.0%; 236 U - 0.791 +- 1.1%; 238 U - 0.587 +- 1.1%; 237 Np - 1.060 +- 1.4%; 239 Pu - 1.152 +- 1.1%; 242 Pu - 0.967 +- 1.0%. 40 refs., 11 tabs., 9 figs

  6. Status of thorium cycle nuclear data evaluations: Comparison of cross-section line shapes of JENDL-3 and ENDF-B-VI files for 230Th, 232Th, 231Pa, 233Pa, 232U, 233U and 234U

    International Nuclear Information System (INIS)

    Ganesan, S.; McLaughlin, P.K.

    1992-02-01

    Since 1990, one of the most interesting developments in the field of nuclear data for nuclear technology applications is that several new evaluated data files have been finalized and made available to the International Atomic Energy Agency (IAEA) for distribution to its Member States. Improved evaluated nuclear data libraries such as ENDF/B-VI from the United States and JENDL-3 from Japan were developed over a period of 10-15 years. This report is not an evaluation of the evaluations. The report as presented here gives a first look at the cross section line shapes of the isotopes that are important to the thorium fuel cycle derived from the two recently evaluated data files: JENDL-3 and ENDF/B-VI. The basic evaluated data files JENDL-3 and ENDF/B-VI were point-processed successfully using the codes LINEAR and RECENT. The point data were multigrouped in three different group structures using the GROUPIE code. Graphs of intercomparisons of cross section line shapes of JENDL-3 and ENDF/B-VI are presented in this paper for the following isotopes of major interest to studies of the thorium fuel cycle: 230 Th, 232 Th, 231 Pa, 233 Pa, 232 U, 233 U and 234 U. Comparisons between JENDL-3 and ENDF/B-VI which were performed at the point and group levels show large discrepancies in various cross sections. We conclude this report with a general remark that it is necessary to perform sensitivity studies to assess the impacts of the discrepancies between the two different sets of data on calculated reactor design and safety parameters of specific reactor systems and, based on the results of such sensitivity studies, to undertake new tasks of evaluations. (author). 2 refs, 245 figs, 8 tabs

  7. Sources of neutronics data involving thorium of 233U and light water moderation

    International Nuclear Information System (INIS)

    Davenport, L.C.

    1978-11-01

    A literature search has been conducted to locate sources of neutronics data for light water moderated systems which contain thorium and/or uranium-233. It is concluded that insufficient data is currently available to validate neutronics design methods for licensing the 233 UO 2 -ThO 2 fuel cycle in light water reactors. A summary of the neutronics data sources found is reported in this document. These sources include critical and exponential experiments with lattices of fuel rods containing 233 U + Th or 235 U + Th. A few experiments using homogeneous aqueous solutions of 233 UO 2 (NO 3 ) 2 or 233 UO 2 F 2 are also included. The only critical lattice data using both 233 U and Th came from the LWBR program. All these experiments were zoned radially and in most cases axially also. Geometrically clean lattice critical data were measured for the CETR and TUPE programs. Both series used 235 UO 2 -ThO 2 pellets. A series of 21 exponential experiments using 3% 233 UO 2 - 97% ThO 2 fuel vibratory compacted to 92% of theoretical density in Zircaloy-2 tubing was performed at BNL using both unpoisoned and boric acid poisoned H 2 O moderator. For completeness, homogeneous systems are listed in which basic neutronics data have been measured. However, it is expected that most data concerning homogeneous systems will be applied to criticality safety problems rather than neutronics methods validation

  8. Development of automation and remotisation systems for fabrication of (Th-233U)O2 MOX fuel for AHWR

    International Nuclear Information System (INIS)

    Saraswat, Anupam; Danny, K.M.; Chakraborty, S.; Somayajulu, P.S.; Kumar, Arun; Mittal, R.; Prasad, R.S.; Mahule, K.N.; Panda, S.; Jayarajan, K.

    2011-01-01

    To meet the ever increasing power requirement of India, country is planning to utilize its large thorium reserves for the third stage of nuclear power program based on Thorium-Uranium 233 fuel in A.H.W.R. Although there are many advantages of (Th- 233 U)O 2 fuel cycle, presence of radiological hazards due to the presence of 1000-2000 ppm level of 232 U in the 233 U fuel and inertness of ThO 2 makes handling and fabrication of fuel difficult. The associated high alpha and gamma activity demands high level of automation and remote handling in alpha tight hot cells. To demonstrate automation and remotisation in (Th- 233 U)O 2 fuel fabrication, a mock up facility is being set up at BARC. This facility shall develop automation systems required for remote fuel fabrication in a simulated hot cell environment. There are many innovative schemes and systems being developed like integrated powder pellet system, remote viewing system for hot cell application etc. Low visibility inside the hot cell has always been a problem for the operator. To overcome this problem a remote viewing system has been developed by which entire hot cell area can be scanned with the use of a joystick and the display can be seen on a LCD monitor. The viewing system is made up of radiation resistant optics which can work even in high gamma fields. It consists of objective end assembly which is used to scan the hot cell area with the help of prism doublets and drive mechanism for capturing full 360 deg solid angle view. There is a Galilean telescope and focusing system used for focusing images of distant objects. Drive mechanism can be controlled by the joystick available to the operator. System has a high resolution CCD display and camera which gives a clear display of objects lying inside the hot cell area. Integrated powder pellet system is being developed for fabrication of MOX pellets from feed powder. This will be automated system which will take input in the form of MOX powder and convert it

  9. Bi-stability in accelerator driven 233U breeders

    International Nuclear Information System (INIS)

    Ghosh, Biplab; Degweker, S.B.

    2011-01-01

    Research on Accelerator Driven Systems (ADSs) is being carried out around the world primarily with the objective of waste transmutation. Presently, the volume of waste in India is small and therefore there is little incentive to develop ADS based waste transmutation technology immediately. On the other hand, the indigenous U availability is limited and hence there is a strong incentive for breeding. Moreover the large Th deposits in the country provide a clear incentive to develop Th related technologies. Th has the additional advantage that it produces very little trans-uranic waste. While Pu fuelled fast reactors using advanced metallic fuel can have high breeding ratios due to the hard spectrum in such reactors, Th fuelled critical reactors can at best be self sustaining or marginal breeders. A possible way to improve the breeding of Th fueled reactors is to use an external neutron source as is done in ADSs. ADSs can not only give improved breeding but also permit greater flexibility in type of fuel that may be used and have the potential to considerably simplify the Th fuel cycle as in the case of the Th burner. In this paper we study various issues associated with breeding in ADSs such as the energy economics of breeding in ADSs using various types of neutron sources and the effect of the reactor spectrum and the discharge fluence (or irradiation time) of the fuel on the breeding performance. We show that even with non-fissioning, non-power- producing targets such as Pb or LBE it is possible to choose the fuel irradiation time so that the breeder produces sufficient power to drive the accelerator and export the balance to the grid, without significantly diminishing the 233 U breeding rate. By increasing the discharge fluence (irradiation time) it is possible to increase the power. However, the 233 U production rate falls off rapidly to about half its maximum value. This is the Th burner region. As the equations governing the breeding process are non

  10. Using {sup 233}U-doped crystals to access the few-eV isomeric transition in {sup 229}Th

    Energy Technology Data Exchange (ETDEWEB)

    Stellmer, Simon; Schreitl, Matthias; Kazakov, Georgy A.; Sterba, Johannes H.; Schumm, Thorsten [Vienna Center for Quantum Science and Technology (VCQ) and Atominstitut, TU Wien, Vienna (Austria)

    2016-07-01

    The isotope {sup 229}Th possesses an exceptionally low-lying isomeric state at an energy of only a few eV. While direct laser excitation of the isomer is a tantalizing future prospect, the stage is not yet set for nuclear laser spectroscopy: too little is known about the energy, lifetime, and internal conversion pathways of the isomer. Alternative routes to populate the isomer are needed for further investigations. We use the alpha decay {sup 233}U →{sup 229g,m}Th to populate the isomer with a probability of 2%. The {sup 233}U is embedded into VUV-transparent crystals, as the isomer transition is expected around 160 nm. The wavelength of the gamma ray, emitted upon de-excitation of the isomer into the ground state, is measured with a spectrometer. Calculations show that the isomer emission is not obscured by radioluminescence of the crystal. We report on the current status of the experiment.

  11. Mass dependence of azimuthal asymmetry in the fission of {sup 232}Th and {sup 233,235,236,238}U by polarized photons

    Energy Technology Data Exchange (ETDEWEB)

    Denyak, V.V. [National Science Center ' ' Kharkov Institute of Physics and Technology' ' , Kharkiv (Ukraine); Pele Pequeno Principe Research Institute, Curitiba (Brazil); Khvastunov, V.M. [National Science Center ' ' Kharkov Institute of Physics and Technology' ' , Kharkiv (Ukraine); Paschuk, S.A. [Federal University of Technology - Parana, Curitiba (Brazil); Schelin, H.R. [Federal University of Technology - Parana, Curitiba (Brazil); Pele Pequeno Principe Research Institute, Curitiba (Brazil)

    2013-04-15

    Fission of the even-even nuclei {sup 232}Th, {sup 236,238}U and even-odd nuclei {sup 233,235}U by linearly polarized photons has been studied at excitation energies in the region of a giant dipole resonance. The performed investigations unambiguously showed the existence of the fragment mass dependence of the cross section azimuthal asymmetry in the photofission of {sup 236}U and {sup 238}U. In addition, the obtained results provided the first evidence for the possible difference between the asymmetry values in asymmetric and symmetric mass distribution regions in the case of {sup 236}U. The measured cross section azimuthal asymmetry of the fission of {sup 232}Th does not show any fragment mass dependence. In the even-odd nuclei {sup 233}U and {sup 235}U the difference between the far-asymmetric and other mass distribution regions was also observed but with the statistical uncertainty not small enough for definitive conclusion. (orig.)

  12. Thermal Stabilization of 233UO2, 233UO3, and 233U3O8

    International Nuclear Information System (INIS)

    Thein, S.M.; Bereolos, P.J.

    2000-01-01

    This report identifies an appropriate thermal stabilization temperature for 233 U oxides. The temperature is chosen principally on the basis of eliminating moisture and other residual volatiles. This report supports the U. S. Department of Energy (DOE) Standard for safe storage of 233 U (DOE 2000), written as part of the response to Recommendation 97-1 of the Defense Nuclear Facilities Safety Board (DNFSB), addressing safe storage of 233 U

  13. A general overview of generation IV molten salt reactor (MSR) and the use of thorium as fuel

    Energy Technology Data Exchange (ETDEWEB)

    Yamaguchi, Carlos H.; Stefani, Giovanni L.; Santos, Thiago A., E-mail: carlos.yamaguchi@usp.br, E-mail: giovanni.stefani@ipen.br, E-mail: thiago.santos@ufabc.edu.br [Universidade de Sao Paulo (USP), SP (Brazil). Instituto de Fisica; Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Universidade Federal do ABC (CECS/UFABC), Santo Andre, SP (Brazil). Centro de Engenharia, Modelagem e Ciencias Sociais Aplicadas

    2017-07-01

    The molten salt reactors (MSRs) make use of fluoride salt as primary cooler, at low pressure. Although considered a generation IV reactor, your concept isn't new, since in the 1960 years the Oak Ridge National Laboratory created a little prototype of 8MWt. Over the 20{sup th} century, other countries, like UK, Japan, Russia, China and France also did research in the area, especially with the use of thorium as fuel. This goes with the fact that Brazil possess the biggest reserve of thorium in the world. In the center of nuclear engineering at IPEN is being created a study group connected to thorium reactors, which purpose is to investigate reactors using thorium to produce {sup 233}U and tailing burn, thus making the MSR using thorium as fuel, an object of study. This present work searches to do a general summary about the researches of MSR's, having as focus the utilization of thorium with the goal being to show it's efficiency and utilization is doable. (author)

  14. Determination of the extraction efficiency for {sup 233}U source α-recoil ions from the MLL buffer-gas stopping cell

    Energy Technology Data Exchange (ETDEWEB)

    Wense, Lars v.d.; Seiferle, Benedict; Thirolf, Peter G. [Ludwig-Maximilians-Universitaet Muenchen, Garching (Germany); Laatiaoui, Mustapha [GSI Helmholtzzentrum fuer Schwerionenforschung GmbH, Darmstadt (Germany); Helmholtz Institut Mainz, Mainz (Germany)

    2015-03-01

    Following the α decay of {sup 233}U, {sup 229}Th recoil ions are shown to be extracted in a significant amount from the MLL buffer-gas stopping cell. The produced recoil ions and subsequent daughter nuclei are mass purified with the help of a customized quadrupole mass spectrometer. The combined extraction and mass purification efficiency for {sup 229}Th{sup 3+} is determined via MCP-based measurements and via the direct detection of the {sup 229}Th α decay. A large value of (10±2)% for the combined extraction and mass purification efficiency of {sup 229}Th{sup 3+} is obtained at a mass resolution of about 1u/e. In addition to {sup 229}Th, also other α-recoil ions of the {sup 233,} {sup 232}U decay chains are addressed. (orig.)

  15. Preserving high-purity 233U

    International Nuclear Information System (INIS)

    Krichinsky, Alan; Giaquinto, Joe; Canaan, Doug

    2016-01-01

    The MARC X Conference hosted a workshop for the scientific community to communicate needs for high-purity 233 U and its by-products in order to preserve critical items otherwise slated for downblending and disposal. Currently, only small portions of the U.S. holdings of separated 233 U are being preserved. However, many additional kilograms of 233 U (>97 % pure) still are destined to be disposed, and it is unlikely that this material will ever be replaced due to a lack of operating production capability. Summaries of information conveyed at the workshop and feedback obtained from the scientific community are presented herein. (author)

  16. Comparison of potential radiological impacts of 233U and 239Pu fuel cycles

    International Nuclear Information System (INIS)

    Meyer, H.R.; Little, C.A.; Witherspoon, J.P.; Till, J.E.

    1979-01-01

    Nuclear fuel cycles utilizing 233 U are currently the subject of considerable interest in the United States. This paper focuses on the identification of significant differences between the off-site radiological hazards posed by 232 Th/ 233 U (Th/U) and 238 U/ 239 Pu (U/Pu) fuel cycles, and represents a portion of our involvement in the Nonproliferation Alternative Systems Assessment Program (NASAP), to be used in support of the International Fuel Cycle Evaluation (INFCE). The major contributors to radiological dose are likely to be uranium mining and milling (58.5% of total fuel cycle dose), reprocessing (33.9%), and light-water reactor power generation (7.3%). The remainder of the cycle, including enrichment processes, fuel fabrication, transportation, and waste management, contributes only 0.3% to total estimated fuel cycle dose

  17. Coulomb effects in isobaric cold fission from reactions 233U(nth,f), 235U(nth,f),239Pu(nth,f) and 252Cf(sf)

    International Nuclear Information System (INIS)

    Montoya, Modesto

    2013-01-01

    The Coulomb effect hypothesis, formerly used to interpret fluctuations in the curve of maximal total kinetic energy as a function of light fragment mass in reactions 233 U(n th ,f), 235 U(n th ,f) and 239 Pu(n th ,f), is confirmed in high kinetic energy as well as in low excitation energy windows, respectively. Data from reactions 233 U(n th ,f), 235 U(n th ,f), 239 Pu(n th ,f) and 252 Cf(sf) show that, between two isobaric fragmentations with similar Q-values, the more asymmetric charge split reaches the higher value of total kinetic energy. Moreover, in isobaric charge splits with different Q-values, similar preference for asymmetrical fragmentations is observed in low excitation energy windows. (author).

  18. PERFORMA NEUTRONIK BAHAN BAKAR LiF-BeF2-ThF4-UF4 PADA SMALL MOBILE-MOLTEN SALT REACTOR

    Directory of Open Access Journals (Sweden)

    S. N. Rokhman

    2015-04-01

    Full Text Available Telah dilakukan analisis terhadap performa neutronik bahan bakar garam lebur LiF-BeF2-ThF4-UF4 pada Small Mobile-Molten Salt Reactor (SM-MSR. Penyesuaian konfigurasi teras dan temperatur operasi harus dilakukan untuk penggunaan bahan bakar baru tersebut agar mencapai keff > 1 dan CR (conversion ratio > 1 pada fraksi 0,5% 233U, 20% 232Th, 28% Li, 51,5% Be. Setelah didapat nilai keff ≈ 1 dan CR ≈ 1, dilakukan analisis pengaruh perubahan Th terhadap Be dan Be terhadap Li yang terlihat dalam perubahan parameter keff dan CR. Setelah itu fraksi 233U divariasi antara 0,5–0,46% untuk memperoleh keff > 1 dan CR > 1. Dalam perhitungan koefisien reaktifitas temperatur (αT, temperatur teras dinaikkan sebesar +25K dan +50K., dan untuk koefisien reaktifitas void (αV, densitas bahan bakar dikurangi hingga 90%. Hasil perhitungan menunjukkan bahwa pengurangan Th terhadap Be menyebabkan penurunan nilai CR dan naiknya keff akibat berkurangnya material fertil. Sebaliknya penambahan Be terhadap Li mengakibatkan terjadi kenaikan nilai keff dan menurunkan CR, akibat laju serapan Li lebih besar dari Be. Pada 5 (lima fraksi 233U dalam rentang 0,5–0,49%, hasil perhitungan keff dan CR masing-masing bervariasi dalam rentang 1,00001 - 1,00327 dan 1,00016 - 1,00731. Untuk faktor puncak daya (PPF, hasil perhitungan memberikan nilai dalam rentang 2,4311 -2,4714. Sedangkan untuk parameter keselamatan, koefisien reaktivitas temperatur (αT dan reaktivitas void (αV masingmasing bernilai negatif dalam rentang 4,972×10-5 - 5,909×10-5 dan 2,596×10-2- 2,8287×10-2 ∆k/k/K. Dapat disimpulkan bahwa teras SM-MSR memberikan nilai negatif di kedua koefisien reaktivitas tersebut untuk setiap fraksi,, sehingga memenuhi kriteria keselamatan dan keselamatan melekat. Kata kunci: SM-MSR (small mobile-molten salt reactor, bahan bakar LiF-BeF2-ThF4-UF4, keselamatan melekat, koefisien reaktivitas temperatur, koefisien reaktivitas void   The analysis of neutronic performance has

  19. Consultants' Meeting on Review Benchmarking of Nuclear Data for the Th/U Fuel Cycle. Summary Report

    International Nuclear Information System (INIS)

    Capote Noy, R.

    2011-02-01

    A summary is given of the Consultants' Meeting (CM) on Review and Benchmarking of Nuclear Data for the Th/U Fuel Cycle. An IAEA Coordinated Research Project (CRP) on 'Nuclear Data for Th/U Fuel Cycle' was concluded in 2005. The CRP activities resulted in new evaluated nuclear data files for 232 Th, 231 , 233 Pa (later adopted for the ENDF/B-VII.0 library) and improvements to existing evaluations for 232 , 233 , 234 , 236 U. Available nuclear data evaluations for 230 - 232 Th, 231,233 Pa and 232 , 233 , 234 U were reviewed including ROSFOND2010, CENDL-3.1, JENDL-4, JEFF-3.1.1, MINSKACT, and ENDF/B-VII.0 libraries. Benchmark results of available evaluations for 232 Th and 233 U were also discussed. (author)

  20. The fission cross sections of /sup 230/Th, /sup 232/Th, /sup 233/U, /sup 234/U, /sup 236/U, /sup 238/U, /sup 237/Np, /sup 239/Pu and /sup 242/Pu relative /sup 235/U at 14. 74 MeV neutron energy

    Energy Technology Data Exchange (ETDEWEB)

    Meadows, J.W.

    1986-12-01

    The measurement of the fission cross section ratios of nine isotopes relative to /sup 235/U at an average neutron energy of 14.74 MeV is described with particular attention to the determination of corrections and to sources of error. The results are compared to ENDF/B-V and to other measurements of the past decade. The ratio of the neutron induced fission cross section for these isotopes to the fission cross section for /sup 235/U are: /sup 230/Th - 0.290 +- 1.9%; /sup 232/Th - 0.191 +- 1.9%; /sup 233/U - 1.132 +- 0.7%; /sup 234/U - 0.998 +- 1.0%; /sup 236/U - 0.791 +- 1.1%; /sup 238/U - 0.587 +- 1.1%; /sup 237/Np - 1.060 +- 1.4%; /sup 239/Pu - 1.152 +- 1.1%; /sup 242/Pu - 0.967 +- 1.0%. 40 refs., 11 tabs., 9 figs.

  1. Measurement of 233U/234U ratios in contaminated groundwater using alpha spectrometry

    International Nuclear Information System (INIS)

    Harrison, Jennifer J.; Payne, Timothy E.; Wilsher, Kerry L.; Thiruvoth, Sangeeth; Child, David P.; Johansen, Mathew P.; Hotchkis, Michael A.C.

    2016-01-01

    The uranium isotope 233 U is not usually observed in alpha spectra from environmental samples due to its low natural and fallout abundance. It may be present in samples from sites in the vicinity of nuclear operations such as reactors or fuel reprocessing facilities, radioactive waste disposal sites or sites affected by clandestine nuclear operations. On an alpha spectrum, the two most abundant alpha emissions of 233 U (4.784 MeV, 13.2%; and 4.824 MeV, 84.3%) will overlap with the 234 U doublet peak (4.722 MeV, 28.4%; and 4.775 MeV, 71.4%), if present, resulting in a combined 233+234 U multiplet. A technique for quantifying both 233 U and 234 U from alpha spectra was investigated. A series of groundwater samples were measured both by accelerator mass spectrometry (AMS) to determine 233 U/ 234 U atom and activity ratios and by alpha spectrometry in order to establish a reliable 233 U estimation technique using alpha spectra. The Genie™ 2000 Alpha Analysis and Interactive Peak Fitting (IPF) software packages were used and it was found that IPF with identification of three peaks ( 234 U minor, combined 234 U major and 233 U minor, and 233 U major) followed by interference correction on the combined peak and a weighted average activity calculation gave satisfactory agreement with the AMS data across the 233 U/ 234 U activity ratio range (0.1–20) and 233 U activity range (2–300 mBq) investigated. Correlation between the AMS 233 U and alpha spectrometry 233 U was r 2  = 0.996 (n = 10). - Highlights: • Describes a technique for deconvoluting the combined 233 U and 234 U multiplet in alpha spectra. • Enables 233 U and 234 U activities and 233 U/ 234 U ratios to be quantified without requiring additional analysis and measurement. • Applicable to an environmental matrix (groundwater) using standard alpha spectrometry counting equipment, operation and set-up.

  2. Multiplicity and energy of neutrons from {sup 233}U(n{sub th},f) fission fragments

    Energy Technology Data Exchange (ETDEWEB)

    Nishio, Katsuhisa; Kimura, Itsuro; Nakagome, Yoshihiro [Kyoto Univ. (Japan)

    1998-03-01

    The correlation between fission fragments and prompt neutrons from the reaction {sup 233}U(n{sub th},f) was measured with improved accuracy. The results determined the neutron multiplicity and emission energy as a function of fragment mass and total kinetic energy. The average energy as a function of fragment mass followed a nearly symmetric distribution centered about the equal mass-split and formed a remarkable contrast with the saw-tooth distribution of the average neutron multiplicity. The neutron multiplicity from the specified fragment decreases linearly with total kinetic energy, and the slope of multiplicity with kinetic energy had the minimum value at about 130 u. The level density parameter versus mass determined from the neutron data showed a saw-tooth structure with the pronounced minimum at about 128 and generally followed the formula by Gilbert and Cameron, suggesting that the neutron emission process was very much affected by the shell-effect of the fission fragment. (author)

  3. 233U Assay A Neutron NDA System

    International Nuclear Information System (INIS)

    Hensley, D.C.; Lucero, A.J.; Pierce, L.

    1998-01-01

    The assay of highly enriched 233 U material presents some unique challenges. Techniques which apply to the assay of materials of Pu or enriched 235 U do not convert easily over to the assay of 233 U. A specialized neutron assay device is being fabricated to exploit the singles neutron signal, the weak correlated neutron signal, and an active correlated signal. These pieces of information when combined with γ ray isotopics information should give a good overall determination of 233 U material now stored in bldg. 3019 at the Oak Ridge National Laboratory

  4. 233U Assay A Neutron NDA System

    Energy Technology Data Exchange (ETDEWEB)

    Hensley, D.C.; Lucero, A.J.; Pierce, L.

    1998-11-17

    The assay of highly enriched {sup 233}U material presents some unique challenges. Techniques which apply to the assay of materials of Pu or enriched {sup 235}U do not convert easily over to the assay of {sup 233}U. A specialized neutron assay device is being fabricated to exploit the singles neutron signal, the weak correlated neutron signal, and an active correlated signal. These pieces of information when combined with {gamma} ray isotopics information should give a good overall determination of {sup 233}U material now stored in bldg. 3019 at the Oak Ridge National Laboratory.

  5. An optimized symbiotic fusion and molten-salt fission reactor system

    International Nuclear Information System (INIS)

    Blinkin, V.L.; Novikov, V.M.

    A symbiotic fusion-fission reactor system which breeds nuclear fuel is discussed. In the blanket of the controlled thermonuclear reactor (CTR) uranium-233 is generated from thorium, which circulates in the form of ThF 4 mixed with molten sodium and beryllium fluorides. The molten-salt fission reactor (MSR) burns up the uranium-233 and generates tritium for the fusion reactor from lithium, which circulates in the form of LiF mixed with BeF 2 and 233 UF 4 through the MSR core. With a CTR-MSR thermal power ratio of 1:11 the system can produce electrical energy and breed fuel with a doubling time of 4-5 years. The system has the following special features: (1) Fuel reprocessing is much simpler and cheaper than for contemporary fission reactors; reprocessing consists simply in continuous removal of 233 U from the salt circulating in the CTR blanket by the fluorination method and removal of xenon from the MSR fuel salt by gas scavenging; the MSR fuel salt is periodically exchanged for fresh salt and the 233 U is then removed from it; (2) Tritium is produced in the fission reactor, which is a much simpler system than the fusion reactor; (3) The CTR blanket is almost ''clean''; no tritium is produced in it and fission fragment activity does not exceed the activity induced in the structural materials; (4) Almost all the thorium introduced into the CTR blanket can be used for producing 233 U

  6. Fabrication routes for Thorium and Uranium233 based AHWR fuel

    International Nuclear Information System (INIS)

    Danny, K.M.; Saraswat, Anupam; Chakraborty, S.; Somayajulu, P.S.; Kumar, Arun

    2011-01-01

    India's economic growth is on a fast growth track. The growth in population and economy is creating huge demand for energy which has to be met with environmentally benign technologies. Nuclear Energy is best suited to meet this demand without causing undue environmental impact. Considering the large thorium reserves in India, the future nuclear power program will be based on Thorium- Uranium 233 fuel cycle. The major characteristic of thorium as the fuel of future comes from its superior fuel utilization. 233 U produced in a reactor is always contaminated with 232 U. This 232 U undergoes a decay to produce 228 Th and it is followed by decay chain including 212 Bi and 208 Tl. Both 212 Bi and 208 Tl are hard gamma emitters ranging from 0.6 MeV-1.6 MeV and 2.6 MeV respectively, which necessitates its handling in hot cell. The average concentration of 232 U is expected to exceed 1000 ppm after a burn-up of 24,000 MWD/t. Work related to developing the fuel fabrication technology including automation and remotization needed for 233 U based fuels is in progress. Various process for fuel fabrication have been developed i.e. Coated Agglomerate Pelletisation (CAP), impregnation technique (Pellet/Gel), Sol Gel Micro-sphere Pelletisation (SGMP) apart from Powder to Pellet (POP) route. This paper describes each process with respect to its advantages, disadvantages and its amenability to automation and remotisation. (author)

  7. Delayed Fission Gamma-ray Characteristics of Th-232 U-233 U-235 U-238 and Pu-239

    Energy Technology Data Exchange (ETDEWEB)

    Lane, Taylor [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Parma, Edward J. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

    2015-08-01

    Delayed fission gamma-rays play an important role in determining the time dependent ioniz- ing dose for experiments in the central irradiation cavity of the Annular Core Research Reactor (ACRR). Delayed gamma-rays are produced from both fission product decay and from acti- vation of materials in the core, such as cladding and support structures. Knowing both the delayed gamma-ray emission rate and the time-dependent gamma-ray energy spectrum is nec- essary in order to properly determine the dose contributions from delayed fission gamma-rays. This information is especially important when attempting to deconvolute the time-dependent neutron, prompt gamma-ray, and delayed gamma-ray contribution to the response of a diamond photo-conducting diode (PCD) or fission chamber in time frames of milliseconds to seconds following a reactor pulse. This work focused on investigating delayed gamma-ray character- istics produced from fission products from thermal, fast, and high energy fission of Th-232, U-233, U-235, U-238, and Pu-239. This work uses a modified version of CINDER2008, a transmutation code developed at Los Alamos National Laboratory, to model time and energy dependent photon characteristics due to fission. This modified code adds the capability to track photon-induced transmutations, photo-fission, and the subsequent radiation caused by fission products due to photo-fission. The data is compared against previous work done with SNL- modified CINDER2008 [ 1 ] and experimental data [ 2 , 3 ] and other published literature, includ- ing ENDF/B-VII.1 [ 4 ]. The ability to produce a high-fidelity (7,428 group) energy-dependent photon fluence at various times post-fission can improve the delayed photon characterization for radiation effects tests at research reactors, as well as other applications.

  8. Fast and thermal data testing of 233U critical assemblies

    International Nuclear Information System (INIS)

    Wright, R.Q.; Jordan, W.C.; Leal, L.C.

    1999-01-01

    Many sources have been used to obtain 233 U benchmark descriptions. Unfortunately, some of these are not reliable since a thorough and complete benchmark evaluation often has not been done. For 24 yr a principal source for 233 U benchmarks has been the Cross Section Evaluation Working Group (CSEWG) Benchmark Specifications. The CSEWG specifications included only two fast benchmarks and three thermal benchmarks. The thermal benchmarks were H 2 O-moderated thorium-oxide exponential lattices. Since the thorium-oxide lattices were exponential experiments, they have not been widely used. CSEWG has also used the 233 U Oak Ridge National Laboratory (ORNL) spheres for many years. One advantage of the CSEWG fast benchmarks, JEZEBEL-23 and FLATTOP-23, is that experiments were done for central-reaction-rate ratios. These reaction-rate ratios provide very valuable information to data testers and evaluators that would not otherwise be available. In recent years the International Handbook of Evaluated Criticality Safety Benchmark Experiments has, in general, been a very useful and reliable source. The Handbook does not include central-reaction-rate ratio experiments, however. A new set of 233 U benchmark experiments has been added to the most recent release of the Handbook, U233-SOL-THERM-004. These are paraffin-reflected cylinders of 233 U uranyl-nitrate solutions. Unfortunately, the estimated benchmark uncertainties are on the order of 0.9 to 1.0% in k eff . Benchmark testing has been done for some of these U233-SOL-THERM-004 experiments. The authors have also discovered that the benchmark specifications for the Thomas uranyl-nitrate experiments given in Ref. 5 are incorrect. One problem with the Ref. 5 specifications is that the excess acid was not included. As part of this work, the authors developed revised specifications that include an excess acid correlation based on information from the experimental logbook

  9. Initial ORNL site assessment report on the storage of 233U

    International Nuclear Information System (INIS)

    Bereolos, P.J.; Yong, L.K.; Sadlowe, A.R.; Ramey, D.W.; Krichinsky, A.M.

    1998-03-01

    The 233 U storage facility at ORNL is Building 3019. The inventory stored in Building 3019 consists of 426.5 kg of 233 U contained in 1,387.1 kg of total uranium. The inventory is primarily in the form of uranium oxides; however, uranium metal and other compounds are also stored. Over 99% of the inventory is contained in 1,007 packages stored in tube vaults within the facility. A tank of thorium nitrate solution, the P-24 Tank, contains 0.13 kg of 233 U in ∼ 4,000 gal. of solution. The facility is receiving additional 233 U for storage from the remediation of the Molten Salt Reactor Experiment (MSRE) at ORNL. Consolidation of material from sites with small holdings is also adding to the 233 U inventory. Additionally, small quantities ( 233 U are in other research facilities at ORNL. A risk assessment process was chosen to evaluate the stored material and packages based on available package records. The risk scenario was considered the failure of a package (or a group of similar packages) in the Building 3019 inventory. The probability of such a failure depends on packaging factors such as the age and material of construction of the containers. The consequence of such a failure depends on the amount and form of the material within the packages. One thousand seven packages were categorized with this methodology resulting in 859 low-risk packages, 147 medium-risk packages, and 1 high-risk package. This initial assessment also documents the status of the evaluation of the Building 3019 and its systems for safe storage of 233 U. The final assessment report for ORNL storage of 233 U is scheduled for June 1999. The report will document the facility assessments, the specific package inspection plan, and the results of initial package inspections

  10. A preliminary simulation of an ADS using MCNPX for U233 production

    International Nuclear Information System (INIS)

    Barros, Graiciany P.; Pereira, Claubia; Veloso, Maria A.F.; Costa, Antonella L.

    2009-01-01

    The code MCNPX is used to evaluate a simplified model of a hybrid system regenerator accelerator (ADS - Accelerator Driven Subcritical), for energy and 233 U production. The concept consists of coupling a high-energy particle accelerator with a sub-critical reactor core, using 232 ThO 2 + 233 UO 2 as initial composition. In this work, the spallation source definition used in MCNPX is a point source of 1000 MeV. The system consists in three coaxial cylinders. The internal cylinder is the spallation target that is a thick natural Pb. The intermediate cylinder is the core, composed by the mixture of fuel ( 232 ThO 2 + 9.5% 233 UO 2 ) and Pb coolant; and lead, as reflector, composes the external cylinder. The goal is to begin studies to evaluate the regenerator blanket composition when submitted to a neutron flux during a time step. The effective multiplication coefficient of the system and the variation of the composition of the regenerative layer are analyzed. The preliminary results show the possibility of utilization of this system. (author)

  11. Measurement of the $^{233}$U neutron capture cross section at the n_TOF facility at CERN

    CERN Document Server

    Carrapiço, Carlos; Berthoumieux, Eric; Gonçalves, Isabel; Gunsing, Frank

    2012-12-12

    The Thorium-Uranium (Th-U) fuel cycle has been envisaged as an alternative to the Uranium-Plutonium (U-Pu) fuel cycle for electricity generation using nuclear power reactors. Indeed, thorium can be used as a nuclear fuel, and several studies and R&D programs seem to provide evidence on the sustainability of the Th-U fuel cycle, due to (i) the natural abundance of Thorium, (ii) the improved proliferation resistance offered by the Th-U fuel cycle relative to the U-Pu fuel cycle, (iii) the better neutronics performance of the Th-U fuel cycle throughout the whole neutron energy range compared to the U-Pu fuel cycle, (iv) the lower radiotoxicity of the generated spent fuel in reactors with Th-U fuel cycle and, consequently (v) better economics and public acceptance of the reactors operated using the Th-U fuel cycle compared to those using the U-Pu fuel cycle (prior to the Generation IV nuclear reactors). In a nuclear reactor operated using the Th-U fuel cycle, $^{233}$U is a key nuclide governing the neutr...

  12. Spectral shift controlled reactors, denatured U-233/thorium cycle

    International Nuclear Information System (INIS)

    1978-05-01

    This paper presents technical and economic data on the SSCR which may be of use in the International Fuel Cycle Evaluation Program to intercompare alternative nuclear systems. Included in this paper are data on the denatured U-233/thorium cycle. This cycle shows a proliferation advantage over more classical thorium fuel cycle (e.g., highly-enriched U-235/thorium or plutonium/thorium) due to the elimination of chemically-separable, concentrated fissile material from unirradiated nuclear fuel. The U-233 is denatured by mixing with depleted uranium to a concentration no greater than 12 w/o. An exogenous source of U-233 is assumed in this paper, since U-233 does not occur in nature and only a limited supply has been produced to date for research and development work

  13. Interim assessment of the denatured 233U fuel cycle: feasibility and nonproliferation characteristics

    International Nuclear Information System (INIS)

    Abbott, L.S.; Bartine, D.E.; Burns, T.J.

    1979-12-01

    A fuel cycle that employs 233 U denatured with 238 U and mixed with thorium fertile material is examined with respect to its proliferation-resistance characteristics and its technical and economic feasibility. The rationale for considering the denatured 233 U fuel cycle is presented, and the impact of the denatured fuel on the performance of Light-Water Reactors, Spectral-Shift-Controlled Reactors, Gas-Cooled Reactors, Heavy-Water Reactors, and Fast Breeder Reactors is discussed. The scope of the R, D and D programs to commercialize these reactors and their associated fuel cycles is also summarized and the resource requirements and economics of denatured 233 U cycles are compared to those of the conventional Pu/U cycle. In addition, several nuclear power systems that employ denatured 233 U fuel and are based on the energy center concept are evaluated

  14. A compact multi-plate fission chamber for the simultaneous measurement of 233U capture and fission cross-sections

    Directory of Open Access Journals (Sweden)

    Bacak M.

    2017-01-01

    Full Text Available 233U plays the essential role of fissile nucleus in the Th-U fuel cycle. A particularity of 233U is its small neutron capture cross-section which is about one order of magnitude lower than the fission cross-section on average. Therefore, the accuracy in the measurement of the 233U capture cross-section essentially relies on efficient capture-fission discrimination thus a combined setup of fission and γ-detectors is needed. At CERN n_TOF the Total Absorption Calorimeter (TAC coupled with compact fission detectors is used. Previously used MicroMegas (MGAS detectors showed significant γ-background issues above 100 eV coming from the copper mesh. A new measurement campaign of the 233U capture cross-section and alpha ratio is planned at the CERN n_TOF facility. For this measurement, a novel cylindrical multi ionization cell chamber was developed in order to provide a compact solution for 14 active targets read out by 8 anodes. Due to the high specific activity of 233U fast timing properties are required and achieved with the use of customized electronics and the very fast ionizing gas CF4 together with a high electric field strength. This paper describes the new fission chamber and the results of the first tests with neutrons at GELINA proving that it is suitable for the 233U measurement.

  15. Plutonium and U-233 mines

    International Nuclear Information System (INIS)

    Milgram, M.S.

    1983-08-01

    A comparison is made among second generation reactor systems fuelled primarily with fissile plutonium and/or U-233 in uranium or thorium. This material is obtained from irradiated fuel from first generation CANDU reactors fuelled by natural or enriched uranium and thorium. Except for plutonium-thorium reactors, second generation reactors demand similar amounts of reprocessing throughput, but the most efficient plutonium burning systems require a large prior allocation of uranium. Second generation reactors fuelled by U-233 make more efficient use of resources and lead to more flexible fuelling strategies, but require development of first generation once-through thorium cycles and early demonstration of the commercial viability of thorium fuel reprocessing. No early implementation of reprocessing technology is required for these cycles

  16. Within the framework of the new fuel cycle 232Th/233U, determination of the 233Pa(n.γ) radiative capture cross section for neutron energies ranging between 0 and 1 MeV

    International Nuclear Information System (INIS)

    Boyer, S.

    2004-10-01

    The Thorium cycle Th 232 /U 233 may face brilliant perspectives through advanced concepts like molten salt reactors or accelerator driven systems but it lacks accurate nuclear data concerning some nuclei. Pa 233 is one of these nuclei, its high activity makes the direct measurement of its radiative neutron capture cross-section almost impossible. This difficulty has been evaded by considering the transfer reaction Th 232 (He 3 ,p)Pa 234 * in which the Pa 234 nucleus is produced in various excited states according to the amount of energy available in the reaction. The first chapter deals with the thorium cycle and its assets to contribute to the quenching of the fast growing world energy demand. The second chapter gives a detailed description of the experimental setting. A scintillation detector based on deuterated benzene (C 6 D 6 ) has been used to counter gamma ray cascades. The third chapter is dedicated to data analysis. In the last chapter we compare our experimental results with ENDF and JENDL data and with computed values from 2 statistical models in the 0-1 MeV neutron energy range. Our results disagree clearly with evaluated data: our values are always above ENDF and JENDL data but tend to near computed values. We have also perform the measurement of the radiative neutron cross-section of Pa 231 for a 110 keV neutron: σ(n,γ) 2.00 ± 0.14 barn. (A.C.)

  17. Water-Reflected 233U Uranyl Nitrate Solutions in Simple Geometry

    International Nuclear Information System (INIS)

    Elam, K.R.

    2001-01-01

    A number of critical experiments involving 233 U were performed in the Oak Ridge National Laboratory Building 9213 Critical Experiments Facility during the years 1952 and 1953. These experiments, reported in Reference 1, were directed toward determining bounding values for the minimum critical mass, minimum critical volume, and maximum safe pipe size of water-moderated solutions of 233 U. Additional information on the critical experiments was found in the experimental logbooks. Two experiments utilizing uranyl nitrate (UO 2 (NO 3 ) 2 ) solutions in simple geometry are evaluated in this report. Experiment 37 is in a 10.4-inch diameter sphere, and Experiment 39 is in a 10-inch diameter cylinder. The 233 U concentration ranges from 49 to 62 g 233 U/l. Both experiments were reflected by at least 6 inches of water in all directions. Paraffin-reflected uranyl nitrate experiments, also reported in Reference 1, are evaluated elsewhere. Experiments with smaller paraffin reflected 5-, 6-, and 7.5-inch diameter cylinders are evaluated in U233-SOL-THERM-004. Experiments with paraffin reflected 8-, 8.5-, 9-, 10-, and 12-inch diameter cylinders are evaluated in U233-SOL-THERM-002. Later experiments with highly-enriched 235 U uranyl fluoride solution in the same 10.4-inch diameter sphere are reported in HEU-SOL-THERM-010. Both experiments were judged acceptable for use as criticality-safety benchmark experiments

  18. Interim assessment of the denatured 233U fuel cycle: feasibility and nonproliferation characteristics

    International Nuclear Information System (INIS)

    Abbott, L.S.; Bartine, D.E.; Burns, T.J.

    1978-12-01

    A fuel cycle that employs 233 U denatured with 238 U and mixed with thorium fertile material is examined with respect to its proliferation-resistance characteristics and its technical and economic feasibility. The rationale for considering the denatured 233 U fuel cycle is presented, and the impact of the denatured fuel on the performance of Light-Water Reactors, Spectral-Shift-Controlled Reactors, Gas-Cooled Reactors, Heavy-Water Reactors, and Fast Breeder Reactors is discussed. The scope of the R, D and D programs to commercialize these reactors and their associated fuel cycles is also summarized and the resource requirements and economics of denatured 233 U cycles are compared to those of the conventional Pu/U cycle. In addition, several nuclear power systems that employ denatured 233 U fuel and are based on the energy center concept are evaluated. Under this concept, dispersed power reactors fueled with denatured or low-enriched uranium fuel are supported by secure energy centers in which sensitive activities of the nuclear cycle are performed. These activities include 233 U production by Pu-fueled transmuters (thermal or fast reactors) and reprocessing. A summary chapter presents the most significant conclusions from the study and recommends areas for future work

  19. Experimental 233U nondestructive assay with a random driver

    International Nuclear Information System (INIS)

    Goris, P.

    1979-01-01

    Nondestructive assay (NDA) of 233 U in quantities up to 15 grams containing 7 ppM 232 U age 2 years was investigated with a random driver. A passive singles counting technique showed a reproducibility within 0.2% at the 95% confidence level. This technique would be applicable throughout a process in which all of the 233 U had the same 232 U content at the same age. Where the 232 U content varies, determination of 233 U fissile content would require active NDA. Active coincidence counting utilizing a 238 Pu, Li neutron source and a plastic scintillator detector system showed a reproducibility limit within 15% at the 95% confidence limit. The active technique was found to be very dependent on the detector system resolving time in order to make proper random coincidence corrections associated with the high gamma activity from the 232 U decay chain

  20. Optimization of the binary breeder reactor. VIII annular core fueled with 233U - 238U and Pu-238U

    International Nuclear Information System (INIS)

    Nascimento, J.A. do; Ishiguro, Y.

    1988-04-01

    First cycle burnup characteristics of a 1200 MWe binary breeder reactor with annular core fueled with metallic 233 U- 238 U-Zr, Pu- 238 U-Zr and Th in the blankets have been analysed. The Doppler effect is small as expected in a metal fueled fast reactor. The sodium void reactivity is, in general, smaller than in metal fueled homogeneous fast reactors of 1 m core height. The estimates of the required and available control rod worths show a large shutdown margin throughout the operational cycle. There are flexibilities in the blanket fueling and well balanced breeding in the two cycles, uranium and thorium, with doubling times of about 20 years are possible. (author) [pt

  1. Studies on the recovery of 233U from phosphate containing aqueous waste using DBDECMP as extractant

    International Nuclear Information System (INIS)

    Sagar, V.B.; Oak, M.S.; Pawar, S.M.; Sivaramakrishnan, C.K.; Patil, S.K.

    1990-01-01

    A method for the recovery and purification of 233 U from phosphate containing analytical waste is developed. Extraction studies with Di-butyl N,N-diethylcarbamoylmethylphosphonate (DBDECMP) in xylene were carried out to explore the feasibility of separation and purification of 233 U from such wastes. Based on the data obtained, optimum conditions for the recovery of 233 U are suggested. (author) 11 refs.; 1 fig.; 3 tabs

  2. Neutron data evaluation of 233U

    International Nuclear Information System (INIS)

    Maslov, V.M.; Tetereva, N.A.; Kagalenko, A.B.; Kornilov, N.V.; Baba, M.; Hasegawa, A.

    2003-08-01

    Consistent evaluation of 233 U measured data base is performed. Hauser-Feshbach- Moldauer theory, coupled channel model and double-humped fission barrier model are employed. Total, differential scattering, fission and (n,xn) data are calculated using fission cross section data description as a major constraint. The direct excitation of ground state is calculated within rigid rotator model. Average resonance parameters are provided, which reproduce evaluated cross sections in the range of 0.6-40.5 keV. (author)

  3. Within the framework of the new fuel cycle {sup 232}Th/{sup 233}U, determination of the {sup 233}Pa(n.{gamma}) radiative capture cross section for neutron energies ranging between 0 and 1 MeV; Dans le cadre du nouveau cycle de combustible {sup 232}Th/{sup 233}U, determination de la section efficace de capture radiative {sup 233}Pa(n,{gamma}) pour des energies de neutrons comprises entre 0 et 1 MeV

    Energy Technology Data Exchange (ETDEWEB)

    Boyer, S

    2004-10-15

    The Thorium cycle Th{sup 232}/U{sup 233} may face brilliant perspectives through advanced concepts like molten salt reactors or accelerator driven systems but it lacks accurate nuclear data concerning some nuclei. Pa{sup 233} is one of these nuclei, its high activity makes the direct measurement of its radiative neutron capture cross-section almost impossible. This difficulty has been evaded by considering the transfer reaction Th{sup 232}(He{sup 3},p)Pa{sup 234}* in which the Pa{sup 234} nucleus is produced in various excited states according to the amount of energy available in the reaction. The first chapter deals with the thorium cycle and its assets to contribute to the quenching of the fast growing world energy demand. The second chapter gives a detailed description of the experimental setting. A scintillation detector based on deuterated benzene (C{sub 6}D{sub 6}) has been used to counter gamma ray cascades. The third chapter is dedicated to data analysis. In the last chapter we compare our experimental results with ENDF and JENDL data and with computed values from 2 statistical models in the 0-1 MeV neutron energy range. Our results disagree clearly with evaluated data: our values are always above ENDF and JENDL data but tend to near computed values. We have also perform the measurement of the radiative neutron cross-section of Pa{sup 231} for a 110 keV neutron: {sigma}(n,{gamma}) 2.00 {+-} 0.14 barn. (A.C.)

  4. Criticality safety validation: Simple geometry, single unit 233U systems

    International Nuclear Information System (INIS)

    Putman, V.L.

    1997-06-01

    Typically used LMITCO criticality safety computational methods are evaluated for suitability when applied to INEEL 233 U systems which reasonably can be modeled as simple-geometry, single-unit systems. Sixty-seven critical experiments of uranium highly enriched in 233 U, including 57 aqueous solution, thermal-energy systems and 10 metal, fast-energy systems, were modeled. These experiments include 41 cylindrical and 26 spherical cores, and 41 reflected and 26 unreflected systems. No experiments were found for intermediate-neutron-energy ranges, or with interstitial non-hydrogenous materials typical of waste systems, mixed 233 U and plutonium, or reflectors such as steel, lead, or concrete. No simple geometry experiments were found with cubic or annular cores, or approximating infinite sea systems. Calculations were performed with various tools and methodologies. Nine cross-section libraries, based on ENDF/B-IV, -V, or -VI.2, or on Hansen-Roach source data, were used with cross-section processing methods of MCNP or SCALE. The k eff calculations were performed with neutral-particle transport and Monte Carlo methods of criticality codes DANT, MCNP 4A, and KENO Va

  5. Chemical and spectrochemical production analysis of ThO2 and 233UO2-ThO2 pellets for the light water breeder reactor core for Shippingport (LWBR development program)

    International Nuclear Information System (INIS)

    Bukowski, J.F.; Hollis, E.D.

    1975-06-01

    The Bettis Atomic Power Laboratory has utilized wet chemical, emission spectrochemical, and mass spectrometric analytical techniques for the production analysis of the ThO 2 and 233 UO 2 -ThO 2 (1 to 6 wt percent 233 UO 2 ) pellets for the Light Water Breeder Reactor (LWBR) core for Shippingport. Proof of the fuel breeding concept necessitates measurement of precise and accurate chemical characterization of all fuel pellets before core life. Chemistry's efforts toward this goal are presented in three main sections: (1) general discussions relating the chemical requirements for ThO 2 and 233 UO 2 -ThO 2 core materials to the analytical capabilities, (2) technical discussions of the chemical and instrumental technology applied for the analysis of aluminum, boron, calcium, carbon, chloride plus bromide, chromium, cobalt, copper, dysprosium, europium, fluoride, gadolinium, iron, magnesium, manganese, mercury, molybdenum, nickel, nitrogen, samarium, silicon, titanium, vanadium, thorium, and uranium (total, trace, and uranium VI), and (3) a formal presentation of the analytical procedures as applied to the LWBR Development Program. (U.S.)

  6. Interim assessment of the denatured /sup 233/U fuel cycle: feasibility and nonproliferation characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Abbott, L.S.; Bartine, D.E.; Burns, T.J. (eds.)

    1978-12-01

    A fuel cycle that employs /sup 233/U denatured with /sup 238/U and mixed with thorium fertile material is examined with respect to its proliferation-resistance characteristics and its technical and economic feasibility. The rationale for considering the denatured /sup 233/U fuel cycle is presented, and the impact of the denatured fuel on the performance of Light-Water Reactors, Spectral-Shift-Controlled Reactors, Gas-Cooled Reactors, Heavy-Water Reactors, and Fast Breeder Reactors is discussed. The scope of the R, D and D programs to commercialize these reactors and their associated fuel cycles is also summarized and the resource requirements and economics of denatured /sup 233/U cycles are compared to those of the conventional Pu/U cycle. In addition, several nuclear power systems that employ denatured /sup 233/U fuel and are based on the energy center concept are evaluated. Under this concept, dispersed power reactors fueled with denatured or low-enriched uranium fuel are supported by secure energy centers in which sensitive activities of the nuclear cycle are performed. These activities include /sup 233/U production by Pu-fueled transmuters (thermal or fast reactors) and reprocessing. A summary chapter presents the most significant conclusions from the study and recommends areas for future work.

  7. Fuel cycle cost analysis on molten-salt reactors

    International Nuclear Information System (INIS)

    Shimazu, Yoichiro

    1976-01-01

    An evaluation is made of the fuel cycle costs for molten-salt reactors (MSR's), developed at Oak Ridge National Laboratory. Eight combinations of conditions affecting fuel cycle costs are compared, covering 233 U-Th, 235 U-Th and 239 Pu-Th fuels, with and without on-site continuous fuel reprocessing. The resulting fuel cycle costs range from 0.61 to 1.18 mill/kWh. A discussion is also given on the practicability of these fuel cycles. The calculations indicate that somewhat lower fuel cycle costs can be expected from reactor operation in converter mode on 235 U make-up with fuel reprocessed in batches every 10 years to avoid fission product precipitation, than from operation as 233 U-Th breeder with continuous reprocessing. (auth.)

  8. Isotopic dilution requirements for 233U criticality safety in processing and disposal facilities

    International Nuclear Information System (INIS)

    Elam, K.R.; Forsberg, C.W.; Hopper, C.M.; Wright, R.Q.

    1997-11-01

    The disposal of excess 233 U as waste is being considered. Because 233 U is a fissile material, one of the key requirements for processing 233 U to a final waste form and disposing of it is to avoid nuclear criticality. For many processing and disposal options, isotopic dilution is the most feasible and preferred option to avoid nuclear criticality. Isotopic dilution is dilution of fissile 233 U with nonfissile 238 U. The use of isotopic dilution removes any need to control nuclear criticality in process or disposal facilities through geometry or chemical composition. Isotopic dilution allows the use of existing waste management facilities, that are not designed for significant quantities of fissile materials, to be used for processing and disposing of 233 U. The amount of isotopic dilution required to reduce criticality concerns to reasonable levels was determined in this study to be ∼ 0.66 wt% 233 U. The numerical calculations used to define this limit consisted of a homogeneous system of silicon dioxide (SiO 2 ), water (H 2 O), 233 U, and depleted uranium (DU) in which the ratio of each component was varied to determine the conditions of maximum nuclear reactivity. About 188 parts of DU (0.2 wt% 235 U) are required to dilute 1 part of 233 U to this limit in a water-moderated system with no SiO 2 present. Thus, for the US inventory of 233 U, several hundred metric tons of DU would be required for isotopic dilution

  9. Disposition of Uranium -233 (sup 233U) in Plutonium Metal and Oxide at the Rocky Flats Environmental Technology Site

    International Nuclear Information System (INIS)

    Freiboth, Cameron J.; Gibbs, Frank E.

    2000-01-01

    This report documents the position that the concentration of Uranium-233 ( 233 U) in plutonium metal and oxide currently stored at the DOE Rocky Flats Environmental Technology Site (RFETS) is well below the maximum permissible stabilization, packaging, shipping and storage limits. The 233 U stabilization, packaging and storage limit is 0.5 weight percent (wt%), which is also the shipping limit maximum. These two plutonium products (metal and oxide) are scheduled for processing through the Building 371 Plutonium Stabilization and Packaging System (PuSPS). This justification is supported by written technical reports, personnel interviews, and nuclear material inventories, as compiled in the ''History of Uranium-233 ( 233 U) Processing at the Rocky Flats Plant In Support of the RFETS Acceptable Knowledge Program'' RS-090-056, April 1, 1999. Relevant data from this report is summarized for application to the PuSPS metal and oxide processing campaigns

  10. Criticality considerations for 233U fuels in an HTGR fuel refabrication facility

    International Nuclear Information System (INIS)

    McNeany, S.R.; Jenkins, J.D.

    1978-01-01

    Eleven 233 U solution critical assemblies spanning an H/ 233 U ratio range of 40 to 2000 and a bare metal 233 U assembly have been calculated with the ENDF/B-IV and Hansen-Roach cross sections. The results from these calculations are compared with the experimental results and with each other. An increasing disagreement between calculations with ENDF/B and Hansen-Roach data with decreasing H/ 233 U ratio was observed, indicative of large differences in their intermediate energy cross sections. The Hansen-Roach cross sections appeared to give reasonably good agreement with experiments over the whole range; whereas the ENDF/B calculations yielded high values for k/sub eff/ on assemblies of low moderation. It is concluded that serious problems exist in the ENDF/B-IV representation of the 233 U cross sections in the intermediate energy range and that further evaluation of this nuclide is warranted. In addition, it is recommended that an experimental program be undertaken to obtain 233 U criticality data at low H/ 233 U ratios for verification of generalized criticality safety guidelines. Part II of this report presents the results of criticality calculations on specific pieces of equipment required for HTGR fuel refabrication. In particular, fuel particle storage hoppers and resin carbonization furnaces are criticality safe up to 22.9 cm (9.0 in.) in diameter providing water or other hydrogenous moderators are excluded. In addition, no criticality problems arise due to accumulation of particles in the off-gas scrubber reservoirs provided reasonable administrative controls are exercised

  11. Efficiency of an LBE spallation target in an accelerator-driven molten salt subcritical reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bak, Sang-In [Sungkyunkwan University, Suwon (Korea, Republic of); Hong, Seung-Woo [Sungkyunkwan University, Suwon (Korea, Republic of); Kadi, Yacine [CERN, Geneva (Switzerland)

    2016-10-15

    An Accelerator-Driven System (ADS) combined with a subcritical Molten Salt Reactor (MSR) is a type of hybrid reactor originally designed to breed uranium from thorium or to incinerate long-lived minor actinides in nuclear wastes. In an MSR, the salt material is used not only as a nuclear fuel but also as a primary coolant. In addition, this material is used as a target for inducing spallation neutrons in most AD-MSR concepts. A high energy proton beam impinges on a heavy metal target to induce spallation reactions and produces neutrons. Accordingly, a reliable proton accelerator is needed to feed the source neutrons. As ADSs have been criticized for requiring high power accelerators, minimization of beam power is an important aspect of ADS design. A primary concern associated with ADS development is stable high-power accelerators. We therefore studied the neutron source efficiencies of an AD-MSR involving chloride fuels by including a Pb-Bi eutectic (LBE) spallation target. The proton source efficiency and the accelerator beam power required have been studied for an AD-MSR. Adoption of an LBE spallation target induces an increase in proton source efficiencies in comparison to the case without a spallation target. Thus the presence of an efficient spallation target is useful in the reduction of the beam power of an accelerator. Almost 33 % of the beam power can be reduced in comparison to the case without the target for NaCl-Th/{sup 233}U fuel, and about 16 % for NaCl-U/TRU fuel. The beam power amplifications increase by 1.5 times for NaCl-Th/{sup 233}U and 1.2 times for NaCl-U/TRU in comparison with the no target AD-MSR.

  12. Criticality safety validation: Simple geometry, single unit {sup 233}U systems

    Energy Technology Data Exchange (ETDEWEB)

    Putman, V.L.

    1997-06-01

    Typically used LMITCO criticality safety computational methods are evaluated for suitability when applied to INEEL {sup 233}U systems which reasonably can be modeled as simple-geometry, single-unit systems. Sixty-seven critical experiments of uranium highly enriched in {sup 233}U, including 57 aqueous solution, thermal-energy systems and 10 metal, fast-energy systems, were modeled. These experiments include 41 cylindrical and 26 spherical cores, and 41 reflected and 26 unreflected systems. No experiments were found for intermediate-neutron-energy ranges, or with interstitial non-hydrogenous materials typical of waste systems, mixed {sup 233}U and plutonium, or reflectors such as steel, lead, or concrete. No simple geometry experiments were found with cubic or annular cores, or approximating infinite sea systems. Calculations were performed with various tools and methodologies. Nine cross-section libraries, based on ENDF/B-IV, -V, or -VI.2, or on Hansen-Roach source data, were used with cross-section processing methods of MCNP or SCALE. The k{sub eff} calculations were performed with neutral-particle transport and Monte Carlo methods of criticality codes DANT, MCNP 4A, and KENO Va.

  13. Fission cross-section measurements on 233U and minor actinides at the CERN n-TOF facility

    International Nuclear Information System (INIS)

    Calviani, M.; Cennini, P.; Chiaveri, E.; Dahlfors, M.; Ferrari, A.; Herrera-Martinez, A.; Kadi, Y.; Sarchiapone, L.; Vlachoudis, V.; Colonna, N.; Terlizzi, R.; Abbondanno, U.; Marrone, S.; Belloni, F.; Fujii, K.; Moreau, C.; Aerts, G.; Andriamonje, S.; Berthoumieux, E.; Dridi, W.; Gunsing, F.; Pancin, J.; Perrot, L.; Plukis, A.; Alvarez, H.; Duran, I.; Paradela, C.; Alvarez-Velarde, F.; Cano-Ott, D.; Embid-Sesura, M.; Gonzalez-Romero, E.; Guerrero, C.; Martinez, T.; Vincente, M. C.; Andrzejewski, J.; Assimakopoulos, P.; Audouin, L.; David, S.; Ferrant, L.; Stephan, C.; Tassan-Got, L.; Badurek, G.; Jericha, E.; Leeb, H.; Oberhummer, H.; Pigni, M. T.; Baumann, P.; Kerveno, M.; Lukic, S.; Rudolf, G.; Becvar, F.; Calvino, F.; Capote, R.; Carrapico, C.; Chepel, V.; Ferreira-Marques, R.; Goncalves, I.; Lindote, A.; Lopes, I.; Neves, F.; Cortes, G.; Poch, A.; Pretel, C.; Couture, A.; Cox, J.; O'Brien, S.; Wiescher, M.; Dillmann, I.; Heil, M.; Kaeppeler, F.; Mosconi, M.; Plag, R.; Walter, S.; Wisshak, K.; Domingo-Pardo, C.; Eleftheriadis, C.; Furman, W.; Goverdovski, A.; Gramegna, F.; Mastinu, P.; Praena, J.; Haas, B.; Haight, R.; Igashira, M.; Karadimos, D.; Karamanis, D.; Ketlerov, V.; Koehler, P.; Konovalov, V.; Kossionides, E.; Krticka, M.; Lampoudis, C.; Lozano, M.; Marganiec, J.; Massimi, C.; Mengoni, A.; Milazzo, P. M.; Papachristodoulou, C.; Papadopoulos, C.; Patronis, N.; Pavlik, A.; Pavlopoulos, P.; Plompen, A.; Quesada, J.; Rauscher, T.; Reifarth, R.; Rubbia, C.; Rullhusen, P.; Salgado, J.; Santos, C.; Savvidis, I.; Tagliente, G.; Tain, J. L.; Tavora, L.; Vannini, G.; Vaz, P.; Ventura, A.; Villamarin, D.; Vlastou, R.; Voss, F.

    2010-01-01

    Neutron-induced fission cross-sections of minor actinides have been measured at the white neutron source n-TOF at CERN, Geneva. The studied isotopes include 233 U, interesting for Th/U based nuclear fuel cycles, 241, 243 Am and 245 Cm, relevant for transmutation and waste reduction studies in new generation fast reactors (Gen-IV) or Accelerator Driven Systems. The measurements take advantage of the unique features of the n-TOF facility, namely the wide energy range, the high instantaneous neutron flux and the low background. Results for the involved isotopes are reported from ∼30 meV to around 1 MeV neutron energy. The measurements have been performed with a dedicated Fission Ionization Chamber (FIC), relative to the standard cross-section of the 235 U fission reaction, measured simultaneously with the same detector. Results are here reported. (authors)

  14. Preserving Ultra-Pure Uranium-233

    International Nuclear Information System (INIS)

    Krichinsky, Alan M.; Goldberg, Steven A.; Hutcheon, Ian D.

    2011-01-01

    Uranium-233 ( 233 U) is a synthetic isotope of uranium formed under reactor conditions during neutron capture by natural thorium ( 232 Th). At high purities, this synthetic isotope serves as a crucial reference material for accurately quantifying and characterizing uranium-bearing materials assays and isotopic distributions for domestic and international nuclear safeguards. Separated, high purity 233 U is stored in vaults at Oak Ridge National Laboratory (ORNL). These materials represent a broad spectrum of 233 U from the standpoint of isotopic purity - the purest being crucial for precise analyses in safeguarding uranium. All 233 U at ORNL is currently scheduled to be disposed of by down-blending with depleted uranium beginning in 2015. This will reduce safety concerns and security costs associated with storage. Down-blending this material will permanently destroy its potential value as a certified reference material for use in uranium analyses. Furthermore, no credible options exist for replacing 233 U due to the lack of operating production capability and the high cost of restarting currently shut down capabilities. A study was commissioned to determine the need for preserving high-purity 233 U. This study looked at the current supply and the historical and continuing domestic need for this crucial isotope. It examined the gap in supplies and uses to meet domestic needs and extrapolated them in the context of international safeguards and security activities - superimposed on the recognition that existing supplies are being depleted while candidate replacement material is being prepared for disposal. This study found that the total worldwide need by this projection is at least 850 g of certified 233 U reference material over the next 50 years. This amount also includes a strategic reserve. To meet this need, 18 individual items totaling 959 g of 233 U were identified as candidates for establishing a lasting supply of certified reference materials (CRM), all

  15. Production of 232,233Pa in 6Li+232Th Collisions in the Classical Trajectory Approach

    International Nuclear Information System (INIS)

    Aleshin, V.P.

    2000-01-01

    The semiclassical model of nuclear reactions with loosely bound projectiles (V.P. Aleshin, B.I. Sidorenko, Acta Phys. Pol. B29, 325 (1998)) is refined and compared with experimental data of Rama Rao et al. on the excitation function for the production of 232,233 Pa in 6 Li+ 232 Th collisions at E = 30-50 MeV. The main contribution to the production of 232 Pa is the 2 neutron emission from excited states of 234 Pa formed in the ( 6 Li,α) reaction. The main source of 233 Pa is the ( 6 Li,αp) reaction followed by γ transitions from excited states of 233 Th to 233 Th (g.s.) which transforms to 233 Pa through β - decay. The ground state of 6 Li regarded as a combination of n+p+α is modeled with the K=2, l x =l y =0 hyperspherical function. The calculation underpredicts the excitation function of 232 Pa by a factor of 0.6 and overpredicts the excitation function of 233 Pa by a factor of 2.3, on the average. With the more realistic wave function of 6 Li both factors are expected to be closer to 1. (author)

  16. Material control and accountability aspects of safeguards for the USA 233U/Th fuel recycle plant

    International Nuclear Information System (INIS)

    Carpenter, J.A. Jr.; McNeany, S.R.; Angelini, P.; Holder, N.D.; Abraham, L.

    1978-01-01

    The materials control and accountability aspects of the reprocessing and refabrication of a conceptual large-scale HTGR fuel recycle plant have been discussed. Two fuel cycles were considered. The traditional highly enriched uranium cycle uses an initial or makeup fuel element with a fissile enrichment of 93% 235 U. The more recent medium enriched uranium cycle uses initial or makeup fuel elements with a fissile enrichment less than 20% 235 U. In both cases, 233 U bred from the fertile thorium is recycled. Materials control and accountability in the plant will be by means of a real-time accountability method. Accountability data will be derived from monitoring of total material mass through the processes and a system of numerous assays, both destructive and nondestructive

  17. Performance of LMFBR fuel pins with (Pu,Th)O/sub 2-x/ and UO2

    International Nuclear Information System (INIS)

    Lawrence, L.A.

    1983-09-01

    The irradiation performance of (Pu,Th)O/sub 2-x/ and UO 2 fueled pins for breeder reactor application were compared to the extensive performance data base for the (U,Pu)O/sub 2-x/ fuel system. Th-Pu and 238 U- 233 U based fuel systems were candidate fuel fertile/fissile isotopic combinations for development of alternatives to the current LMFBR fuel cycle. Initial screening tests were conducted in the EBR-II to obtain comparative performance data because of the limited experience with these fuel systems. In some cases, 235 U was used as a substitute for 233 U because of the difficulties in fabrication of available 233 U due to its high gamma ray emission rate

  18. Recovery of 233U from irradiated J rods - operating experience

    International Nuclear Information System (INIS)

    Lakshmanan, K.; Natarajan, D.; Muthukumar, M.; Halder, Surajit; Jayakrishnan, G.; Selvarasan, M.; Kuppusamy, V.; Ganesan, V.; Vijayakumar, V.

    2000-01-01

    The first campaign of reprocessing was completed in 1988 and 233 U was delivered for fabrication of fuel for KAMINI. After revamping the facility, the second campaign was started in Dec 1998 and has processed some of the high density thoria rods from CIRUS successfully. Currently the campaign is in progress and it is planned to process the irradiated thorium rods from Dhruva. Interim 23 process selective to 233 U is adopted for separation of uranium from thorium and fission products

  19. Thorium Molten Salt Nuclear Energy Synergetic System (THORIMS-NES)

    International Nuclear Information System (INIS)

    Yoshioka, Ritsuo; Mitachi, Koshi

    2013-01-01

    The authors have been promoting nuclear energy technology based on thorium molten salt as Thorium Molten Salt Nuclear Energy Synergetic System (THORIMS-NES). This system is a combination of fission power reactor of Molten Salt Reactor (MSR), and Accelerator Molten Salt Breeder (AMSB) for production of fissile 233 U with connecting chemical processing facility. In this paper, concept of THORIMS-NES, advantages of thorium and molten salt recent MSR design results such as FUJI-U3 using 233 U fuel, FUJI-Pu, large sized super-FUJI, pilot plant miniFUJI, AMSB, and chemical processing facility are described. (author)

  20. Assessment of the available {sup 233}U cross-section evaluations in the calculation of critical benchmark experiments

    Energy Technology Data Exchange (ETDEWEB)

    Leal, L.C.; Wright, R.Q.

    1996-10-01

    In this report we investigate the adequacy of the available {sup 233}U cross-section data for calculation of experimental critical systems. The {sup 233}U evaluations provided in two evaluated nuclear data libraries, the U.S. Data Bank [ENDF/B (Evaluated Nuclear Data Files)] and the Japanese Data Bank [JENDL (Japanese Evaluated Nuclear Data Library)] are examined. Calculations were performed for six thermal and ten fast experimental critical systems using the S{sub n} transport XSDRNPM code. To verify the performance of the {sup 233}U cross-section data for nuclear criticality safety application in which the neutron energy spectrum is predominantly in the epithermal energy range, calculations of four numerical benchmark systems with energy spectra in the intermediate energy range were done. These calculations serve only as an indication of the difference in calculated results that may be expected when the two {sup 233}U cross-section evaluations are used for problems with neutron spectra in the intermediate energy range. Additionally, comparisons of experimental and calculated central fission rate ratios were also made. The study has suggested that an ad hoc {sup 233}U evaluation based on the JENDL library provides better overall results for both fast and thermal experimental critical systems.

  1. Assessment of the Available (Sup 233)U Cross Sections Evaluations in the Calculation of Critical Benchmark Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Leal, L.C.

    1993-01-01

    In this report we investigate the adequacy of the available {sup 233}U cross-section data for calculation of experimental critical systems. The {sup 233}U evaluations provided in two evaluated nuclear data libraries, the U. S. Data Bank [ENDF/B (Evaluated Nuclear Data Files)] and the Japanese Data Bank [JENDL (Japanese Evaluated Nuclear Data Library)] are examined. Calculations were performed for six thermal and ten fast experimental critical systems using the Sn transport XSDRNPM code. To verify the performance of the {sup 233}U cross-section data for nuclear criticality safety application in which the neutron energy spectrum is predominantly in the epithermal energy range, calculations of four numerical benchmark systems with energy spectra in the intermediate energy range were done. These calculations serve only as an indication of the difference in calculated results that may be expected when the two {sup 233}U cross-section evaluations are used for problems with neutron spectra in the intermediate energy range. Additionally, comparisons of experimental and calculated central fission rate ratios were also made. The study has suggested that an ad hoc {sup 233}U evaluation based on the JENDL library provides better overall results for both fast and thermal experimental critical systems.

  2. Comment on "Zircon U-Th-Pb dating using LA-ICP-MS: Simultaneous U-Pb and U-Th dating on 0.1 Ma Toya Tephra, Japan" by Hisatoshi Ito

    Science.gov (United States)

    Guillong, M.; Schmitt, A. K.; Bachmann, O.

    2015-04-01

    Laser ablation inductively coupled plasma mass spectrometry (LA-ICP-MS) of eight zircon reference materials and synthetic zircon-hafnon end-members indicate that corrections for abundance sensitivity and molecular zirconium sesquioxide ions (Zr2O3+) are critical for reliable determination of 230Th abundances in zircon. Other polyatomic interferences in the mass range 223-233 amu are insignificant. When corrected for abundance sensitivity and interferences, activity ratios of (230Th)/(238U) for the zircon reference materials we used average 1.001 ± 0.010 (1σ error; mean square of weighted deviates MSWD = 1.45; n = 8). This includes the 91500 and Plešovice zircons, which were deemed unsuitable for calibration of (230Th)/(238U) by Ito (2014). Uranium series zircon ages generated by LA-ICP-MS without mitigating (e.g., by high mass resolution) or correcting for abundance sensitivity and molecular interferences on 230Th such as those presented by Ito (2014) are potentially unreliable.

  3. Th and U fuel photofission study by NTD for AD-MSR subcritical assembly

    Energy Technology Data Exchange (ETDEWEB)

    Sajo-Bohus, Laszlo; Greaves, Eduardo D.; Barros, Haydn; Pino, Felix; Barrera, Maria T.; Farina, Fulvio [Universidad Simón Bolívar, Nuclear Physics Laboratory, Apdo 89000, Caracas 1080A (Venezuela, Bolivarian Republic of); Davila, Jesus [Física Médica C. A. and Universidad Central de Venezuela, Caracas (Venezuela, Bolivarian Republic of)

    2015-07-23

    During the last decade a considerable effort has been devoted for developing energy generating systems based on advanced nuclear technology within the design concepts of GEN-IV. Thorium base fuel systems such as accelerator driven nuclear reactors are one of the often mentioned attractive and affordable options. Several radiotherapy linear accelerators are on the market and due to their reliability, they could be employed as drivers for subcritical liquid fuel assemblies. Bremsstrahlung photons with energies above 5.5MeV, induce (γ,n) and (e,e’n) reactions in the W-target. Resulting gamma radiation and photo or fission neutrons may be absorbed in target materials such as thorium and uranium isotopes to induce sustained fission or nuclear transmutation in waste radioactive materials. Relevant photo driven and photo-fission reaction cross sections are important for actinides {sup 232}Th, {sup 238}U and {sup 237}Np in the radiotherapy machines energy range of 10-20 MV. In this study we employ passive nuclear track detectors (NTD) to determine fission rates and neutron production rates with the aim to establish the feasibility for gamma and photo-neutron driven subcritical assemblies. To cope with these objectives a 20 MV radiotherapy machine has been employed with a mixed fuel target. Results will support further development for a subcritical assembly employing a thorium containing liquid fuel. It is expected that acquired technological knowledge will contribute to the Venezuelan nuclear energy program.

  4. Th and U fuel photofission study by NTD for AD-MSR subcritical assembly

    Science.gov (United States)

    Sajo-Bohus, Laszlo; Greaves, Eduardo D.; Davila, Jesus; Barros, Haydn; Pino, Felix; Barrera, Maria T.; Farina, Fulvio

    2015-07-01

    During the last decade a considerable effort has been devoted for developing energy generating systems based on advanced nuclear technology within the design concepts of GEN-IV. Thorium base fuel systems such as accelerator driven nuclear reactors are one of the often mentioned attractive and affordable options. Several radiotherapy linear accelerators are on the market and due to their reliability, they could be employed as drivers for subcritical liquid fuel assemblies. Bremsstrahlung photons with energies above 5.5MeV, induce (γ,n) and (e,e'n) reactions in the W-target. Resulting gamma radiation and photo or fission neutrons may be absorbed in target materials such as thorium and uranium isotopes to induce sustained fission or nuclear transmutation in waste radioactive materials. Relevant photo driven and photo-fission reaction cross sections are important for actinides 232Th, 238U and 237Np in the radiotherapy machines energy range of 10-20 MV. In this study we employ passive nuclear track detectors (NTD) to determine fission rates and neutron production rates with the aim to establish the feasibility for gamma and photo-neutron driven subcritical assemblies. To cope with these objectives a 20 MV radiotherapy machine has been employed with a mixed fuel target. Results will support further development for a subcritical assembly employing a thorium containing liquid fuel. It is expected that acquired technological knowledge will contribute to the Venezuelan nuclear energy program.

  5. System design description of forced-convection molten-salt corrosion loops MSR-FCL-3 and MSR-FCL-4

    International Nuclear Information System (INIS)

    Huntley, W.R.; Silverman, M.D.

    1976-11-01

    Molten-salt corrosion loops MSR-FCL-3 and MSR-FCL-4 are high-temperature test facilities designed to evaluate corrosion and mass transfer of modified Hastelloy N alloys for future use in Molten-Salt Breeder Reactors. Salt is circulated by a centrifugal sump pump to evaluate material compatibility with LiF-BeF 2 -ThF 4 -UF 4 fuel salt at velocities up to 6 m/s (20 fps) and at salt temperatures from 566 to 705 0 C (1050 to 1300 0 F). The report presents the design description of the various components and systems that make up each corrosion facility, such as the salt pump, corrosion specimens, salt piping, main heaters, salt coolers, salt sampling equipment, and helium cover-gas system, etc. The electrical systems and instrumentation and controls are described, and operational procedures, system limitations, and maintenance philosophy are discussed

  6. Process technology for the molten-salt reactor 233U--Th cycle

    International Nuclear Information System (INIS)

    Hightower, J.R. Jr.

    1975-01-01

    After a brief description of the design features of the molten-salt breeder reactor, fuel processing for removal of 233 Pa and fission products is examined. Some recent developments in processing technology are discussed

  7. Remote fabrication of (Th, {sup 233}U)O{sub 2} pellet-type fuels for CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Feraday, M A

    1981-05-15

    Thorium fuels enriched with {sup 233}U must be fabricated in shielded cells because of high gamma and alpha activity. A conceptual design of a remotely operated plant to produce gamma-active pellet fuels has been made. The plant consists of eight fabrication canyons, two repair canyons, and several miscellaneous cells. Process equipment is modular, easily disconnected, and mounted on plates for easy removal. Equipment consists of a combination of robotics, hard automation, and conventional process equipment. The plant is operated from a central control room with the assistance of a sophisticated computer-based control and information system. Many of the automated process steps are preprogrammed on the control computer and executed on demand by the supervising operator. The technology to build such a plant exists today but needs to be adapted to the needs of the recycle fuel industry. (author)

  8. MSR Founders Narrative and Content Analysis of Scholarly Papers

    DEFF Research Database (Denmark)

    Tackney, Charles T.; Chappell, Stacie F.; Sato, Toyoko

    2017-01-01

    This is a founders’ narrative and research paper content analysis of the first 15 years of the Management Spirituality and Religion Interest Group (MSR) of the Academy of Management. Based on archival data and founder interviews, our inquiry recounts how the early collaborators established......: a corpus epitomizing MSR research and practice. The combined study is a benchmark of founding and institutionalization for current and potential MSR members. By tracing the research trends MSR has taken in light of the founding aspirations, we illuminate the distinctive values, tensions, and meanings...

  9. Standardization of thermal and epithermal INAA methods for simultaneous determination of U and Th in mixed oxide samples

    International Nuclear Information System (INIS)

    Acharya, R.; Pujari, P.K.; Chandra, Ruma

    2010-01-01

    Full text: Uranium and thorium are important fuel materials for nuclear power program. In recent years utilization of thoria based fuel has assumed significance due to higher energy requirements. Thorium based mixed oxide is the proposed fuel for Advanced Heavy Water Reactors (AHWR). In this respect, studies are carried out through preparation of natural U and Th mixed oxides by powder metallurgical route, wherein composition of U and Th is specific and requires strict control in terms their contents and homogeneity in the mixture. Stringent chemical quality control necessitates compositional characterization of the fuel material i.e. accurate and precise determination of U and Th. A suitable method which does not need any chemical dissolution and yields high precision results with minima sample handling is desirable. Instrumental neutron activation analysis (INAA) using reactor neutron is the technique of choice. In view of this, INAA methods namely thermal lNAA (TNAA) (utilizing whole reactor neutrons) and epithermal INAA (ENAA) (utilizing epicadmium neutrons) were standardized for the determination of U and Th in presence of each other in mixed oxide samples. In the present work pneumatic carrier facility (PCF) of Dhruva reactor and self-serve facility of CIRUS reactor were used for TNAA and ENAA respectively. Standards, synthetic samples and mixed oxide samples prepared in cellulose matrix, were irradiated for 1 minute at PCF of Dhruva reactor and for 1 hour at CIRUS reactor under cadmium cover (0.5 mm). Radioactive assay was carried out using 40% relative efficiency HPGe detector. Peak areas under the full energy peaks were evaluated by peak fit method using the PHAST software. Both activation and daughter products of U ( 239 U, 74.6 keV and 239 Np, 277 keV) and Th ( 233 Th, 86 keV and 233 Pa, 312 keV) were used for their concentration determination. The method was validated by analyzing synthetic mixed oxide samples (6-48%U-Th mixed oxide). The % deviations

  10. U-233 fuelled low critical mass solution reactor experiment PURNIMA II

    International Nuclear Information System (INIS)

    Srinivasan, M.; Chandramoleshwar, K.; Pasupathy, C.S.; Rasheed, K.K.; Subba Rao, K.

    1987-01-01

    A homogeneous U-233 uranyl nitrate solution fuelled BeO reflected, low critical mass reactor has been built at the Bhabha Atomic Research Centre, India. Christened PURNIMA II, the reactor was used for the study of the variation of critical mass as a function of fuel solution concentration to determine the minimum critical mass achievable for this geometry. Other experiments performed include the determination of temperature coefficient of reactivity, study of time behaviour of photoneutrons produced due to interaction between decaying U-233 fission product gammas and the beryllium reflector and reactor noise measurements. Besides being the only operational U-233 fuelled reactor at present, PURNIMA II also has the distinction of having attained the lowest critical mass of 397 g of fissile fuel for any operating reactor at the current time. The paper briefly describes the facility and gives an account of the experiments performed and results achieved. (author)

  11. The molten salt reactors (MSR) pyro chemistry and fuel cycle for innovative nuclear systems

    International Nuclear Information System (INIS)

    Brossard, Ph.; Garzenne, C.; Mouney, H.

    2002-01-01

    In the frame of the studies on next generation nuclear systems, and especially for the molten salt reactors and for the integrated fuel cycle (as IFR), the fuel cycle constraints must be taken into account in the preliminary studies of the system to improve the cycle and reactor optimisation. Among the purposes for next generation nuclear systems, sustainability and waste (radio-toxicity and mass) management are important goals. These goals imply reprocessing and recycling strategies. The objectives of this workshop are to present and to share the different strategies and scenarios, the needs based on these scenarios, the experimental facilities available today or in the future and their capabilities, the needs for demonstration. It aims at: identifying the needs for fuel cycle based on solid fuel or liquid fuel, and especially, the on-line reprocessing or clean up for the molten salt reactors; assessing the state-of-the-art on the pyro-chemistry applied to solid fuel and to present the research activities; assessing the state-of-the-art on liquid fuels (or others), and to present the research activities; expressing the R and D programs for pyro-chemistry, molten salt, and also to propose innovative processes; and proposing some joint activities in the frame of GEDEON and PRACTIS programs. This document brings together the transparencies of 18 contributions dealing with: scenario studies with AMSTER concept (Scenarios, MSR, breeders (Th) and burners); fuel cycle for innovative systems; current reprocessing of spent nuclear fuel (SNF) in molten salts (review of pyro-chemistry processes (non nuclear and nuclear)); high temperature NMR spectroscopies in molten salts; reductive extraction of An from molten fluorides (salt - liquid metal extraction); electrochemistry characterisation; characterisation with physical methods - extraction coefficient and kinetics; electrolytic extraction; dissolution-precipitation of plutonium in the eutectic LiCl-KCl (dissolution and

  12. Integral benchmark test of JENDL-4.0 for U-233 systems with ICSBEP handbook

    International Nuclear Information System (INIS)

    Kuwagaki, Kazuki; Nagaya, Yasunobu

    2017-03-01

    The integral benchmark test of JENDL-4.0 for U-233 systems using the continuous-energy Monte Carlo code MVP was conducted. The previous benchmark test was performed only for U-233 thermal solution and fast metallic systems in the ICSBEP handbook. In this study, MVP input files were prepared for uninvestigated benchmark problems in the handbook including compound thermal systems (mainly lattice systems) and integral benchmark test was performed. The prediction accuracy of JENDL-4.0 was evaluated for effective multiplication factors (k eff 's) of the U-233 systems. As a result, a trend of underestimation was observed for all the categories of U-233 systems. In the benchmark test of ENDF/B-VII.1 for U-233 systems with the ICSBEP handbook, it is reported that a decreasing trend of calculated k eff values in association with a parameter ATFF (Above-Thermal Fission Fraction) is observed. The ATFF values were also calculated in this benchmark test of JENDL-4.0 and the same trend as ENDF/B-VII.1 was observed. A CD-ROM is attached as an appendix. (J.P.N.)

  13. MSR (Pu converters) and MSBRs in commercial nuclear power stations

    International Nuclear Information System (INIS)

    Reichle, L.F.C.

    1977-01-01

    Molten Salt Reactors are likely to be the best way to achieve lowest-cost, safe, reliable and environmentally compatible commercial nuclear power in the early 1990's. This conclusion is based on work performed by the industrial members of the U.S. Molten Salt Group and by the Oak Ridge National Laboratory - both of whom are in general agreement on the status and prospects of Molten Salt Reactor technology. The MSBR Development Program is a 14-year program comprised of a 250 MWe MSTR (Molten Salt Test Reactor), a 1,000 MWe MSBR (Molten Salt Breeder Reactor) Demonstration Plant, and related development work. Plutonium from LWRs will fuel MSR (Pu Converters) which, in turn, will produce U-233 to fuel MSBRs. Because of the low inventory of fissile material in MSRs, a given amount of Pu will start-up many more MSRs than LMFBRs or GCFRs. MSRs can be expected to produce energy at a cost that will be competitive with LWRs before LMFBRs or GCFRs. They will use less uranium and require less enrichment. They will have a much lower development cost. They have the potential of producing high-temperature process heat. MSRs use a fluid fuel, and therefore eliminate the high cost of fuel fabrication. They have on-stream refueling and high thermal efficiency. They will have construction costs comparable to LWRs. MSRs have relative safety and environmental advantages, such as no possibility of a LOCA, low inventory of fissile material, continuous removal of fission products and on-site storage of spent fuel wastes

  14. Review of thorium-U233 cycle thermal reactor benchmark studies (AWBA Development Program)

    International Nuclear Information System (INIS)

    Ullo, J.J.; Hardy, J. Jr.; Steen, N.M.

    1980-03-01

    A survey is made of existing integral experiments for U233 systems and thorium-uranium based fuel systems. The aim is to understand to what extent they give a consistent test of ENDF/B-IV nuclear data. A principal result is that ENDF/B-IV leads to an underprediction of neutron leakage. Results from testing alternate thorium data sets are presented. For one evaluation due to Leonard, the results depict a possible growing discrepancy between measured integral parameters such as rho 02 and I 232 and the differential data, which underpredicts these parameters. Sensitivities to other nuclear data components, notably the fission neutron spectrum, were determined. A new harder U233 spectrum significantly reduces a bias trend in K/sub eff/ vs leakage

  15. Calculated critical parameters in simple geometries for oxide and nitrate water mixtures of U-233, U-235 and Pu-239 with thorium. Final report

    International Nuclear Information System (INIS)

    Converse, W.E.; Bierman, S.R.

    1979-11-01

    Calculations have been performed on water mixtures of oxides and nitrates of 233 U, 235 U, and 239 Pu with chemically similar thorium compounds to determine critical dimensions for simple geometries (sphere, cylinder, and slab). Uranium enrichments calculated were 100%, 20%, 10%, and 5%; plutonium calculations assumed 100% 239 Pu. Thorium to uranium or plutonium weight ratios (Th: U or Pu) calculated were 0, 1, 4, and 8. Both bare and full water reflection conditions were calculated. The results of the calculations are plotted showing a critical dimension versus the uranium or plutonium concentration. Plots of K-infinity and material buckling for each material type are also shown

  16. Scoping studies of 233U breeding fusion fission hybrid

    International Nuclear Information System (INIS)

    Maniscalco, J.A.; Hansen, L.F.; Allen, W.O.

    1978-05-01

    Neutronic calculations have been carried out in order to design a laser fusion driven hybrid blanket which maximizes 233 U production per unit of thermal energy (greater than or equal to 1 kg/MW/sub T/-year) with acceptable fusion energy multiplication (M/sub F/ approximately 4). Two hybrid blankets, a thorium and a uranium-thorium blanket, are discussed in detail and their performance is evaluated by incorporating them into an existing hybrid design (the LLL/Bechtel design). The overall performance of the two laser fusion driven 233 U producers is discussed and estimates are given of (1) the number of equivalent thermal power fission reactors (LWR, HWR, SSCR and HTGR) that these fusion breeders can fuel, (2) their capital cost, and (3) the cost of electricity in the combined fusion breder-converter reactor scenario

  17. Scoping studies of 233U breeding fusion fission hybrid

    International Nuclear Information System (INIS)

    Maniscalco, J.A.; Hansen, L.F.; Allen, W.O.

    1978-01-01

    Neutronic calculations have been carried out in order to design a laser fusion driven hybrid blanket which maximizes 233 U production per unit of thermal energy (greater than or equal to 1 kg/MW/sub T/-year) with acceptable fusion energy multiplication (M/sub F/ approx. 4). Two hybrid blankets, a thorium and a uranium--thorium blanket, are discussed in detail and their performance is evaluated by incorporating them into an existing hybrid design (the LLL/Bechtel design). The overall performance of the two laser fusion driven 233 U producers is discussed and estimates are given of (1) the number of equivalent thermal power fission reactors (LWR, HWR, SSCR and HTGR) that these fusion breeders can fuel, (2) their capital cost, and (3) the cost of electricity in the combined fusion breeder-converter reactor scenario

  18. Study of the isotopic exchange associated with ionic exchange for the radiochemical separation of 233-Th

    International Nuclear Information System (INIS)

    Sepulveda Munita, C.J.A.

    1983-01-01

    The isotopic ion exchange procedure is applied in order to establish an analytical method for the determination of thorium by means of the 233 Th activity, when the presence of interfering elements does not allow a direct non-destructive activation analysis. The separation is based on the retention of 233 Th by a thorium saturated resin, due to the isotopic exchange effect, and subsequent elution of the interfering radioisotopes with a solution of thorium in diluted hydrochloric acid. The interfering elements were those which either present a great affinity for the resin or emit gamma rays with energies close to that of 233 Th (86.6 KeV), when a NaI(Tl) detector is used to obtain the gama-ray spectra of the irradiated samples. The equilibrium time for the thorium isotopic ion exchange and the distribution coefficients for the interfering elements were determined by using Bio-Rad AG 50W resins (100-200 mesh), with 4% to 8% of divinylbenzene. The best separation conditions were established in terms of the thorium and hydrochloric acid concentrations in the solution, the resin cross-linking degree, and the solution flow through the resin. The analytical method was applied to the determination of thorium in samples of ammonium diuranate as well in standard rock samples from the United States Geological Survey. The sensitivity, precision and accuracy of the method are also discussed. (Author) [pt

  19. Excited levels of Pa-233; Niveles excitados del Pa-233

    Energy Technology Data Exchange (ETDEWEB)

    Vara Cuadrado, J M

    1969-07-01

    A study of Pa-233 excited levels from the alpha decay of Np-237 and from beta decay of Th-233 has been performed. The alpha decay spectrum was measured with a semiconductor spectrometer of 18 keV effective resolution (FWHM). Over 13 new lines were identified. The gamma ray spectra of Np-237 and Th-233 were obtained with a Ge-Li detector low and medium range energy lines, and with Si-Li detector for the low energy region. A continuous purification method of Np-237 from its comparatively short-lived daughter Pa-233 was applied. A high number of new lines were identified in both spectra. The gamma-gamma coincidence spectra were obtained with INa(T{sub 1}) detectors. (Author) 54 refs.

  20. Final Oak Ridge National Laboratory Site Assessment Report on the Storage of 233U

    International Nuclear Information System (INIS)

    Bereolos, P.J.; Yong, L.K.

    1999-01-01

    This assessment characterizes the 233 U inventories and storage facility at Oak Ridge National Laboratory (ORNL). This assessment is a commitment in the U.S. Department of Energy (DOE) Implementation Plan (IP), ''Safe Storage of Uranium-233,'' in response to the Defense Nuclear Facilities Safety Board's Recommendation 97-1

  1. Uranium Age Determination by Measuring the 230Th / 234U Ratio

    International Nuclear Information System (INIS)

    LAMONT, STEPHEN P.

    2004-01-01

    A radiochemical isotope dilution mass spectrometry method has been developed to determine the age of uranium materials. The amount of 230Th activity, the first progeny of 234U, that had grown into a small uranium metal sample was used to determine the elapsed time since the material was last radiochemically purified. To preserve the sample, only a small amount of oxidized uranium was removed from the surface of the sample and dissolved. Aliquots of the dissolved sample were spiked with 233U tracer and radiochemically purified by anion-exchange chromatography. The 234U isotopic concentration was then determined by thermal ionization mass spectrometry. Additional aliquots of the sample were spiked with 229Th tracer, and the thorium was purified using two sequential anion-exchange chromatography separations. The isotopic concentrations of 230Th and 232Th were determined by TIMS. The lack of any 232Th confirmed the assumption that all thorium was removed from the uranium sample at the time of purification. The 230Th and 234U mass concentrations were converted to activities and the 230Th/234U ratio for the sample was calculated. The experimental 230Th/234U ratio showed the uranium in this sample was radiochemically purified in about 1945. Isotope dilution thermal ionization mass spectrometry has sufficient sensitivity to determine the age of 100 samples of uranium. This method could certainly be employed as a nuclear forensic method to determine the age of small quantities of uranium metal or salts. Accurate determination of the ultra-trace 230Th radiochemically separated from the uranium is possible due to the use of 229Th as an isotope dilution tracer. The precision in the experimental age of the uranium could be improved by making additional replicate measurements of the 230Th/234U isotopic ratio or using a larger initial sample

  2. 233U breeding in accelerator-driven sub-critical fast reactor

    International Nuclear Information System (INIS)

    Yang Yongwei; An Yu

    1999-01-01

    Accelerator-driven Sub-critical Fast Reactor (ADFR) is chosen as fissile-material-breeding reactor. (U-Pu)O x is chosen as fuel in the core and ThO 2 as fertile material in the blanket zone to breed 233 U. Molten lead is chosen as coolant because of its better neutronic and chemical characteristics over sodium. The program system used for neutronics study consists of: LAHET, for the simulation of the interaction between the proton with medium energy and the nuclei of the target; MCNP4A, for the simulation of neutron transport with energy below 20 MeV in the sub-critical reactor; CONNECT1, for the processing of some tallies provided by the output of MCNP4A in order to prepare micro-cross sections for elements used for burnup calculation; ORIGEN2, used for multi-region burnup calculation; CONNECT2, for the processing of atom densities of some elements provided in the output of ORIGEN2 in order to prepare input to LAHET calculation for next time step. The calculated results show that the proposed case is feasible for breeding fissile material considering the criticality safety, power density, burnup, etc

  3. Evaluation of temperature coefficients of reactivity for 233U--thorium fueled HTGR lattices. Final report

    International Nuclear Information System (INIS)

    Newman, D.F.; Leonard, B.R. Jr.; Trapp, T.J.; Gore, B.F.; Kottwitz, D.A.; Thompson, J.K.; Purcell, W.L.; Stewart, K.B.

    1977-05-01

    A comparison of calculated and measured neutron multiplication factors as a function of temperature was made for three graphite-moderated lattices in the High Temperature Lattice Test Reactor (HTLTR) using 233 UO 2 --ThO 2 fuels in varying amounts and configurations. Correlation of neutronic analysis methods and cross section data with the experimental measurements forms the basis for assessing the accuracy of the methods and data and developing confidence in the ability to predict the temperature coefficient of reactivity for various High Temperature Gas-Cooled Reactor (HTGR) conditions in which 233 U and thorium are present in the fuel. The calculated values of k/sub infinity/(T) were correlated with measured values using two least-squares-fitted correlation coefficients: (1) a normalization factor, and (2) a temperature coefficient bias factor. These correlations indicate the existence of a negative (nonconservative) bias in temperature coefficients of reactivity calculated using ENDF/B-IV cross section data

  4. Amster: a molten-salt reactor concept generating its own 233U and incinerating transuranium elements

    International Nuclear Information System (INIS)

    Lecarpentier, D.; Garzenne, C.; Vergnes, J.; Mouney, H.; Delpech, M.

    2002-01-01

    In the coming century, sustainable development of atomic energy will require the development of new types of reactors able to exceed the limits of the existing reactor types, be it in terms of optimum use of natural fuel resources, reduction in the production of long-lived radioactive waste, or economic competitiveness. Of the various candidates with the potential to meet these needs, molten-salt reactors are particularly attractive, in the light of the benefits they offer, arising from two fundamental features: - A liquid fuel does away with the constraints inherent in solid fuel, leading to a drastic simplification of the fuel cycle, in particular making in possible to carry out on-line pyrochemical reprocessing; - Thorium cycle and thermal spectrum breeding. The MSBR concept proposed by ORNL in the 1970's thus gave a breeding factor of 1.06, with a doubling time of about 25 years. However, given the tight neutron balance of the thorium cycle (the η of 233 U is about 2.3), MSBR performance is only possible if there are strict constraints set on the in-line reprocessing unit: all the 233 Pa must be removed from the core so that it can decay on the 233 U in no more than about ten days (or at least 15 tonnes of salt to be extracted from the core daily), and the absorbing fission products, in particular the rare earths, must be extracted in about fifty days. With the AMSTER MSR concept, which we initially developed for incinerating transuranium elements, we looked to reduce the mass of salt to be reprocessed in order to minimise the size and complexity of the reprocessing unit coupled to the reactor, and the quantity of transuranium elements sent for disposal, as this is directly proportional to the mass of salt reprocessed for extraction of the fission products. Given that breeding was not an absolute necessity, because the reactor can be started by incinerating the transuranium elements from the spent fuel assemblies of current reactors, or if necessary by loading

  5. Neutron multipilication factors as a function of temperature: a comparison of calculated and measured values for lattices using 233UO2-ThO2 fuel in graphite

    International Nuclear Information System (INIS)

    Newman, D.F.; Gore, B.F.

    1978-01-01

    Neutron multiplication factors calculated as a function of temperature for three graphite-moderated 233 UO 2 -ThO 2 -fueled lattices are correlated with the values measured for these lattices in the high-temperature lattice test reactor (HTLTR). The correlation analysis is accomplished by fitting calculated values of k/sub infinity/(T) to the measured values using two least-squares-fitted correlation coefficients: (a) a normalization factor and (b) a temperature coefficient bias factor. These correlations indicate the existence of a negative (nonconservative) bias in temperature coefficients of reactivity calculated using ENDF/B-IV cross-section data. Use of an alternate cross-section data set for thorium, which has a smaller resonance integral than ENDF/B-IV data, improved the agreement between calculated and measured temperature coefficients of reactivity for the three experimental lattices. The results of the correlations are used to estimate the bias in the temperature coefficient of reactivity calculated for a lattice typical of fresh 233 U recycle fuel for a high-temperature gas-cooled reactor (HTGR). This extrapolation to a lattice having a heavier fissile loading than the experimental lattices is accomplished using a sensitivity analysis of the estimated bias to alternate thorium cross-section data used in calculations of k/sub infinity/(T). The envelope of uncertainty expected to contain the actual values for the temperature coefficient of the reactivity for the 233 U-fueled HTGR lattice studied remains negative at 1600 K (1327 0 C). Although a broader base of experimental data with improved accuracy is always desirable, the existing data base provided by the HTLTR experiments is judged to be adequate for the verification of neutronic calculations for the HTGR containing 233 U fuel at its current state of development

  6. Measurements of neutron induced capture and fission reactions on $^{233}$ U (EAR1)

    CERN Multimedia

    The $^{233}$U plays the essential role of ssile nucleus in the Th-U fuel cycle, which has been proposed as a safer and cleaner alternative to the U-Pu fuel cycle. Considered the scarce data available to assess the capture cross section, a measurement was proposed and successfully performed at the n_TOF facility at CERN using the 4$\\pi$ Total Absorp- tion Calorimeter (TAC). The measurement was extremely dicult due to the need to accurately distinguish between capture and fission $\\gamma$-rays without any additional discrim-ination tool and the measured capture cross section showed a signicant disagreement in magnitude when compared with the ENDF/B-VII.1 library despite the agreement in shape. We propose a new measurement that is aimed at providing a higher level of dis-crimination between competing nuclear reactions, to extend the neutron energy range and to obtain more precise and accurate data, thus fullling the demands of the "NEA High Priority Nuclear Data Request List". The setup is envisaged as a combin...

  7. Investigation of neutronic behavior in a CANDU reactor with different (Am, Th, {sup 235}U)O{sub 2} fuel matrixes

    Energy Technology Data Exchange (ETDEWEB)

    Gholamzadeh, Z. [Talca Univ. (Chile). Dept. of Physics; Feghhi, S.A.H. [Shahid Beheshti Univ., Tehran (Iran, Islamic Republic of). Dept. of Radiation Application

    2014-11-15

    Recently thorium-based fuel matrixes are taken into consideration for nuclear waste incineration because of thorium proliferation resistance feature moreover its breeding or convertor ability in both thermal and fast reactors. In this work, neutronic influences of adding Am to (Th-{sup 235}U)O{sub 2} on effective delayed neutron fraction, reactivity coefficients and burn up of a fed CANDU core has been studied using MCNPX 2.6.0 computational code. Different atom fractions of Am have been introduced in the fuel matrix to evaluate its effects on neutronic parameters of the modeled core. The computational data show that adding 2% atom fraction of Am to thorium-based fuel matrix won't noticeably change reactivity coefficients in comparison with the fuel matrix containing 1% atom fraction of Am. The use of 2% atom fraction of Am resulted in a higher delayed neutron fraction. According to the obtained data, 32.85 GWd burn up of the higher Americium-containing fuel matrix resulted in 55.2%, 26.5%, 41.9% and 2.14% depletion of {sup 241}Am, {sup 243}Am, {sup 235}U and {sup 232}Th respectively. 132.8 kg of {sup 233}U fissile element is produced after the burn up time and the nuclear core multiplication factor increases in rate of 2390 pcm. The less americium-containing fuel matrix resulted in higher depletion of {sup 241/243}Am, {sup 235}U and {sup 232}Th while the nuclear core effective multiplication factor increases in rate of 5630 pcm after the burn up time with 9.8 kg additional {sup 233}U production.

  8. Radionuclides release possibility analysis of MSR at various accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Choong Wie; Kim, Hee Reyoung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    There are some accidents which go beyond our expectation such as Fukushima Daiichi nuclear disaster and amounts of radionuclides release to environment, so more effort and research are conducted to prevent it. MSR (Molten Salt Reactor) is one of GEN-IV reactor types, and its coolant and fuel are mixtures of molten salt. MSR has a schematic like figure 1 and it has different features with the solid fuel reactor, but most important and interesting feature of MSR is its many safety systems. For example, MSR has a large negative void coefficient. Even though power increases, the reactor slows down soon. Radionuclides release possibility of MSR was analyzed at various accident conditions including Chernobyl and Fukushima ones. The MSR was understood to prevent the severe accident by the negative reactivity coefficient and the absence of explosive material such as water at the Chernobyl disaster condition. It was expected to contain fuel salts in the reactor building and not to release radionuclides into environment even if the primary system could be ruptured or broken and fuel salts would be leaked at the Fukushima Daiichi nuclear disaster condition of earthquake and tsunami. The MSR, which would not lead to the severe accident and therefore prevents the fuel release to the environment at many expected scenarios, was thought to have priority in the aspect of accidents. A quantitative analysis and a further research are needed to evaluate the possibility of radionuclide release to the environment at the various accident conditions based on the simple comparison of the safety feature between MSR and solid fuel reactor.

  9. Simultaneous measurement of neutrons and fission fragments of thermal neutron fission of U-233

    International Nuclear Information System (INIS)

    Itsuro Kimura; Katsuhisa Nishio; Yoshihiro Nakagome

    2000-01-01

    The multiplicity and the energy of prompt neutrons from the fragments for 233 U(n th , f) were measured as functions of fragment mass and total kinetic energy. Average neutron energy against the fragment mass showed a nearly symmetric distribution about the half mass division with two valleys at 98 and 145 u. The slope of the neutron multiplicity with total kinetic energy depended on the fragment mass and showed the minimum at about 130 u. The obtained neutron data were applied to determine the total excitation energy of the system, and the resulting value in the typical asymmetric fission lied between 22 and 25 MeV. The excitation energy agreed with that determined by subtracting the total kinetic energy from the Q-value within 1 MeV, thus satisfied the energy conservation. In the symmetric fission, where the mass yield was drastically suppresses, the total excitation energy is significantly large and reaches to about 40 MeV, suggesting that fragment pairs are preferentially formed in a compact configuration at the scission point [ru

  10. Interaction between U and Th on their uptake, distribution, and toxicity in V S. alfredii based on the phytoremediation of U and Th.

    Science.gov (United States)

    Huang, Zhenling; Tang, Siqun; Zhang, Lu; Ma, Lijian; Ding, Songdong; Du, Liang; Zhang, Dong; Jin, Yongdong; Wang, Ruibing; Huang, Chao; Xia, Chuanqin

    2017-01-01

    Variant Sedum alfredii Hance (V S. alfredii) could simultaneously take up U and Th from water with the highest concentrations recorded as 1.84 × 10 4 and 6.72 × 10 3  mg/kg in the roots, respectively. Th stimulated U uptake by V S. alfredii roots at Th 10 (10 μM of Th), however, the opposite was observed at Th 100 (100 μM of Th). A similar result was found in the effect of U on the uptake of Th by V S. alfredii. Subcellular fractionation studies of V S. alfredii indicated that U and Th were mainly stored in cell wall fraction, and much less was found in organelle and soluble fractions. Chemical form examination results showed that water-soluble U and Th were the predominant chemical forms in this plant. Addition of the other radionuclide in aqueous solutions altered the concentration and percentage of U or Th in cell wall fraction and in water-soluble form, resulting in the change of the uptake capacity of U or Th by V S. alfredii roots. Comparing with single U or Th treatment, the plant cells revealed more swollen chloroplasts and enhanced thickening in cell walls under the U 100  + Th 100 treatment, as observed by TEM. Those results collectively displayed that V S. alfredii may be utilized as a potential plant to simultaneously remove U and Th from aqueous solutions (rhizofiltration).

  11. Measurement of the Neutron Capture Cross Sections of $^{233}$U, $^{237}$Np, $^{240,242}$Pu, $^{241,243}$Am and $^{245}$Cm with a Total Absorption Calorimeter at n_TOF

    CERN Multimedia

    Beer, H; Wiescher, M; Cox, J; Rapp, W; Embid, M; Dababneh, S

    2002-01-01

    Accurate and reliable neutron capture cross section data for actinides are necessary for the poper design, safety regulation and precise performance assessment of transmutation devices such as Fast Critical Reactors or Accelerator Driven Systems (ADS). The goal of this proposal is the measurement of the neutron capture cross sections of $^{233}$U, $^{237}$Np, $^{240,242}$Pu, $^{241,243}$Am and $^{245}$Cm at n_TOF with an accuracy of 5~\\%. $^{233}$U plays an essential role in the Th fuel cycle, which has been proposed as a safer and cleaner alternative to the U fuel cycle. The capture cross sections of $^{237}$Np,$^{240,242}$Pu, $^{241,243}$Am and $^{245}$Cm play a key role in the design and optimization of a strategy for the Nuclear Waste Transmutation. A high accuracy can be achieved at n_TOF in such measurements due to a combination of features unique in the world: high instantaneous neutron fluence and excellent energy resolution of the facility, innovative Data Acquisition System based on flash ADCs and t...

  12. Recovery of thorium along with uranium 233 from Thorex waste solution employing Chitosan

    International Nuclear Information System (INIS)

    Priya, S.; Reghuram, D.; Kumaraguru, K.; Vijayan, K.; Jambunathan, U.

    2003-01-01

    The low level waste solution, generated from Thorex process during the processing of U 233 , contains thorium along with traces of Th 228 and U 233 . Chitosan, a natural bio-polymer derived from Chitin, was earlier used to recover the uranium and americium. The studies were extended to find out its thorium sorption characteristics. Chitosan exhibited very good absorption of thorium (350 mg/g). Chitosan was equilibrated directly with the low level waste solution at different pH after adjusting its pH, for 60 minutes with a Chitosan to aqueous ratio of 1:100 and the raffinates were filtered and analysed. The results showed more than 99% of thorium and U 233 could be recovered by Chitosan between pH 4 and 5. Loaded thorium and uranium could be eluted from the Chitosan by 1M HNO 3 quantitatively. (author)

  13. Modular Stellarator Fusion Reactor (MSR) concept

    International Nuclear Information System (INIS)

    Miller, R.L.; Krakowski, R.A.

    1981-01-01

    A preliminary conceptual study has been made of the Modulator Stellarator Reactor (MSR) as a stedy-state, ignited, DT-fueled, magnetic fusion reactor. The MSR concept combines the physics of classic stellarator confinement with an innovative, modular-coil design. Parametric tradeoff calculations are described, leading to the selection of an interim design point for a 4.8-GWt plant based on Alcator transport scaling and an average beta value of 0.04 in an l = 2 system with a plasma aspect ratio of 11. Neither an economic analysis nor a detailed conceptual engineering design is presented here, as the primary intent of this scoping study is the elucidation of key physics tradeoffs, constraints, and uncertainties for the ultimate power-reactor embodiment

  14. Excited levels of Pa-233

    International Nuclear Information System (INIS)

    Vara Cuadrado, J. M.

    1969-01-01

    A study of Pa-233 excited levels from the alpha decay of Np-237 and from beta decay of Th-233 has been performed. The alpha decay spectrum was measured with a semiconductor spectrometer of 18 keV effective resolution (FWHM). Over 13 new lines were identified. The gamma ray spectra of Np-237 and Th-233 were obtained with a Ge-Li detector low and medium range energy lines, and with Si-Li detector for the low energy region. A continuous purification method of Np-237 from its comparatively short-lived daughter Pa-233 was applied. A high number of new lines were identified in both spectra. The gamma-gamma coincidence spectra were obtained with INa(T 1 ) detectors. (Author) 54 refs

  15. Study of Advanced Reactor Mixed Oxide Fuel Production of (U,Th)O2

    International Nuclear Information System (INIS)

    Busron-Masduki; Damunir; Pristi-Hartati; R-Sukarsono; Bangun-Wasito

    2000-01-01

    The high price and starting scarcity of reserved of oil drive the people to drill the alternative nuclear energy. Accelerator-driven Transmutation Waste (ATW) is a prospective technology to solve the problem of used fuel waste, to reduce the anxiety of long term disposal waste, to increase the public acceptance of nuclear energy enter into the third millennium. The future of large nuclear energy appears in many-branched industry will depend on the capability to generate relatively low priced fuel on the basis of commercial nuclear energy. Utilization of uranium-233 -thorium cycle insures long-term fuel supply, makes the nuclear energy production more flexible and enables the self-provision regime to be realized in future. Flowsheet of mixed oxide fuel production for advanced reactor of (U,Th)O 2 is a combination of existing manufacturing equipment and quality assurance program from commercial LWR and HTR. The front-end of flowsheet using sol-gel process. The external sol-gel process is chosen due to simple equipment can anticipate refabrication of U-233 which always contains a few hundred ppm of U-232 and its gamma-emitting daughters, besides yielding smaller waste. The decision to choose external sol-gel process encourages to develop External Gelation Thorium (EGT). In order to get higher density and relatively low compaction pressures (i.e. for advanced LWR) adopted flowsheet EGT is developed to be Sol-Gel Microsphere Pelletization (SGMP). Using the optimal parameters, SGMP become established flowsheet for producing mixed oxide fuel of (U,Th)O 2 for advanced reactor. (author)

  16. Subcritical multiplication measurements with a BeO reflected 233U uranyl nitrate solution system

    International Nuclear Information System (INIS)

    Job, P.K.; Srinivasan, M.; Nargundkar, V.R.; Chandramoleshwar, K.; Pasupathy, C.S.; Das, S.; Mayankutty, P.C.

    1978-01-01

    A series of subcritical multiplication measurements were carried out in PURNIMA with 233 U uranyl nitrate solution contained in all 11 x 11 cm 2 square sectional tank and reflected by 30 cm thickness of BeO on all sides. The objective of these experiments was to determine the 'Minimum critical mass' of the system in rectangular parellelopiped geometry. The rectangular aluminium core tank was attached to the bottom of an alpha tight glove box. BeO reflector was arranged below the glove box outside the core tank. The system multiplication was measured as a function of solution concentration and core volume by means of neutron detectors placed outside the assembly. The extrapolated critical mass was obtained through conventional inverse counts plot. The maximum amount of 233 U used was 120 gms. The rectangular geometry was estimated to be 235 +- 10 gms, in the concentration range of 80 to 120 gms/litre of 233 U. The experimental set up, procedure adopted, method of analysis and the details of the results are described. (author)

  17. Fission Product Yields of 233U, 235U, 238U and 239Pu in Fields of Thermal Neutrons, Fission Neutrons and 14.7-MeV Neutrons

    Science.gov (United States)

    Laurec, J.; Adam, A.; de Bruyne, T.; Bauge, E.; Granier, T.; Aupiais, J.; Bersillon, O.; Le Petit, G.; Authier, N.; Casoli, P.

    2010-12-01

    The yields of more than fifteen fission products have been carefully measured using radiochemical techniques, for 235U(n,f), 239Pu(n,f) in a thermal spectrum, for 233U(n,f), 235U(n,f), and 239Pu(n,f) reactions in a fission neutron spectrum, and for 233U(n,f), 235U(n,f), 238U(n,f), and 239Pu(n,f) for 14.7 MeV monoenergetic neutrons. Irradiations were performed at the EL3 reactor, at the Caliban and Prospero critical assemblies, and at the Lancelot electrostatic accelerator in CEA-Valduc. Fissions were counted in thin deposits using fission ionization chambers. The number of fission products of each species were measured by gamma spectrometry of co-located thick deposits.

  18. Characterization and densification studies on ThO{sub 2}-UO{sub 2} pellets derived from ThO{sub 2} and U{sub 3}O{sub 8} powders

    Energy Technology Data Exchange (ETDEWEB)

    Kutty, T.R.G. [Radiometallurgy Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)]. E-mail: tkutty@magnum.barc.ernet.in; Hegde, P.V. [Radiometallurgy Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Khan, K.B. [Radiometallurgy Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Jarvis, T. [Radiometallurgy Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Sengupta, A.K. [Radiometallurgy Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Majumdar, S. [Radiometallurgy Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Kamath, H.S. [Nuclear Fuels Group, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)

    2004-12-01

    ThO{sub 2} containing around 2-3% {sup 233}UO{sub 2} is the proposed fuel for the forthcoming Indian Advanced Heavy Water Reactor (AHWR). This fuel is prepared by powder metallurgy technique using ThO{sub 2} and U{sub 3}O{sub 8} powders as the starting material. The densification behaviour of the fuel was evaluated using a high temperature dilatometer in four different atmospheres Ar, Ar-8%H{sub 2}, CO{sub 2} and air. Air was found to be the best medium for sintering among them. For Ar and Ar-8%H{sub 2} atmospheres, the former gave a slightly higher densification. Thermogravimetric studies carried out on ThO{sub 2}-2%U{sub 3}O{sub 8} granules in air showed a continuous decrease in weight up to 1500 deg. C. The effectiveness of U{sub 3}O{sub 8} in enhancing the sintering of ThO{sub 2} has been established.

  19. Analysis of Hydrogen Generation and Accumulation in U-233 Tube Vaults

    International Nuclear Information System (INIS)

    Ally, M.R.; Willis, K.J.

    1999-01-01

    The purpose of the 233 U Safe Storage Program is to enhance the safe storage of 233 U-bearing materials. This report describes the work done at the Oak Ridge National Laboratory's Radiochemical Development Facility (RDF) to address questions related to possible hydrogen generation and accumulation in 233 U tube vaults. The objective of this effort was to verify assumptions in the mathematical model used to estimate the hydrogen content of the gaseous atmosphere that possibly could occur inside the tube vaults in Building 3019 and to evaluate proposed measures for mitigating any hydrogen concerns. A mathematical model was developed using conservative assumptions to evaluate possible hydrogen generation and accumulation in the tube vaults. The model concluded that an equilibrium concentration would be established below the lower flammability limit (LFL) of 4.1% hydrogen. The major assumptions used in the model that were validated are as follows: (1) The shield plug does not form a seal with the tube vault wall, thus allowing the hydrogen gas to diffuse past the shield plug to the upper section of the tube vault. (2) The tube vault end-cap leaks sufficiently to allow air to be drawn into the tube vault by the off-gas system, thereby purging hydrogen from the upper section of the tube vault. (3) Any hydrogen gas generated completely mixes with the other gases present in the lower section of the tube vault and does not stratify beneath the shield plug. (4) The diffusion coefficient determined from the literature for constant diffusion of hydrogen in air is valid. The coefficient is corrected for temperatures from 0 to 25 C. Another assumption used in the model, that hydrogen generated by radiolytic decomposition of hydrogen-bearing materials (e.g., moisture and plastic) leaks from the cans under steady-state condition, as opposed to a sudden release resulting from rupture of the can(s), was beyond the scope of this investigation. Several parameters from the original

  20. Disposition Options for Uranium-233

    International Nuclear Information System (INIS)

    Beahm, E.C.; Dole, L.R.; Forsberg, C.W.; Icenhour, A.S.; Storch, S.N.

    1999-01-01

    The U.S. Department of Energy (DOE) Fissile Materials Disposition Program (MD), in support of the U.S. arms-control and nonproliferation policies, has initiated a program to disposition surplus weapons-usable fissile material by making it inaccessible and unattractive for use in nuclear weapons. Weapons-usable fissile materials include plutonium, high-enriched uranium (HEU), and uranium-233 (sup 233)U. In support of this program, Oak Ridge National Laboratory led DOE's contractor efforts to identify and characterize options for the long-term storage and disposal of excess (sup 233)U. Five storage and 17 disposal options were identified and are described herein

  1. Void reactivity feedback analysis for U-based and Th-based LWR incineration cycles

    Energy Technology Data Exchange (ETDEWEB)

    Lindley, B.A.; Parks, G.T. [Cambridge University Engineering Department, Trumpington Street, Cambridge, CB2 1PZ (United Kingdom); Franceschini, F. [Westinghouse Electric Company LLC, Cranberry Township, PA (United States)

    2013-07-01

    In reduced-moderation LWRs, an external supply of transuranic (TRU) can be incinerated by mixing it with a fertile isotope ({sup 238}U or {sup 232}Th) and recycling all the actinides after each cycle. Performance is limited by coolant reactivity feedback - the moderator density coefficient (MDC) must be kept negative. The MDC is worse when more TRU is loaded, but TRU feed is also needed to maintain criticality. To assess the performance of this fuel cycle in different neutron spectra, three LWRs are considered: 'reference' PWRs and reduced-moderation PWRs and BWRs. The MDC of the equilibrium cycle is analysed by reactivity decomposition with perturbed coolant density by isotope and neutron energy. The results show that using {sup 232}Th as a fertile isotope yields superior performance to {sup 238}U. This is due essentially to the high resonance η of U bred from Th (U3), which increases the fissility of the U3-TRU isotope vector in the Th-fueled system relative to the U-fueled system, and also improves the MDC in a sufficiently hard spectrum. Spatial separation of TRU and U3 in the Th-fueled system renders further improvement by hardening the neutron spectrum in the TRU and softening it in the U3. This improves the TRU η and increases the negative MDC contribution from reduced thermal fission in U3. (authors)

  2. Intergranular tellurium cracking of nickel-based alloys in molten Li, Be, Th, U/F salt mixture

    Science.gov (United States)

    Ignatiev, Victor; Surenkov, Alexander; Gnidoy, Ivan; Kulakov, Alexander; Uglov, Vadim; Vasiliev, Alexander; Presniakov, Mikhail

    2013-09-01

    In Russia, R&D on Molten Salt Reactor (MSR) are concentrated now on fast/intermediate spectrum concepts which were recognized as long term alternative to solid fueled fast reactors due to their attractive features: strong negative feedback coefficients, easy in-service inspection, and simplified fuel cycle. For high-temperature MSR corrosion of the metallic container alloy in primary circuit is the primary concern. Key problem receiving current attention include surface fissures in Ni-based alloys probably arising from fission product tellurium attack. This paper summarizes results of corrosion tests conducted recently to study effect of oxidation state in selected fuel salt on tellurium attack and to develop means of controlling tellurium cracking in the special Ni-based alloys recently developed for molten salt actinide recycler and tranforming (MOSART) system. Tellurium corrosion of Ni-based alloys was tested at temperatures up to 750 °C in stressed and unloaded conditions in molten LiF-BeF2 salt mixture fueled by about 20 mol% of ThF4 and 2 mol% of UF4 at different [U(IV)]/[U(III)] ratios: 0.7, 4, 20, 100 and 500. Following Ni-based alloys (in mass%): HN80М-VI (Mo—12, Cr—7.6, Nb—1.5), HN80МТY (Mo—13, Cr—6.8, Al—1.1, Ti—0.9), HN80МТW (Mo—9.4, Cr—7.0, Ti—1.7, W—5.5) and ЕМ-721 (W—25.2, Cr—5.7, Ti—0.17) were used for the study in the corrosion facility. If the redox state the fuel salt is characterized by uranium ratio [U(IV)]/[U(III)] uranium intermetallic compounds and alloys with nickel and molybdenum. This leads to spontaneous behavior of alloy formation processes on the specimens' surface and further diffusion of uranium deep into the metallic phase. As consequence of this films of intermetallic compounds and alloys of nickel, molybdenum, tungsten with uranium are formed on the alloys specimens' surface, and intergranular corrosion does not take place. In the fuel salt with [U(IV)]/[U(III)] = 4-20 the potentials of uranium

  3. Nuclear performance optimization of the Be/Li/Th blanket for the fusion breeder

    International Nuclear Information System (INIS)

    Lee, J.D.; Bandini, B.R.

    1985-01-01

    More rigorous nuclear analysis, including treatment of resonance self-shielding effects coupled with an optimization procedure, has resulted in improved performance of the Be/Li/Th blanket. Net U-233 breeding ratio has increased 36% (to 0.84) while at an average U-233/Th ratio of 0.5 a/o average energy multiplication has increased only 12% (to 2.1) compared with earlier results

  4. Cross sections and neutron yields for U233, U235 and Pu239 at 2200 m/sec

    International Nuclear Information System (INIS)

    Sjoestrand, N.G.; Story, J.S.

    1960-04-01

    The experimental information on the 2200 m/sec values for σ abs , σ f , α, ν and η for 233 U , 235 U and 23 been collected and discussed. The values will later be used in an evaluation of a 'best' set of data. In appendix the isotopic abundances of the uranium isotopes are discussed and also the alpha activities of the uranium isotopes and Pu-239

  5. Plutonium and minor actinides utilization in Thorium molten salt reactor

    International Nuclear Information System (INIS)

    Waris, Abdul; Aji, Indarta K.; Novitrian,; Kurniadi, Rizal; Su'ud, Zaki

    2012-01-01

    FUJI-12 reactor is one of MSR systems that proposed by Japan. The original FUJI-12 design considers Th/ 233 U or Th/Pu as main fuel. In accordance with the currently suggestion to stay away from the separation of Pu and minor actinides (MA), in this study we evaluated the utilization of Pu and MA in FUJI-12. The reactor grade Pu was employed in the present study as a small effort of supporting THORIMS-NES scenario. The result shows that the reactor can achieve its criticality with the Pu and MA composition in the fuel of 5.96% or more.

  6. Mechanism study of freeze-valve for molten salt reactor (MSR)

    International Nuclear Information System (INIS)

    Qinhua, Zhang

    2014-01-01

    Molten salt reactor (MSR) is one of the fourth generation nuclear reactor, ordinary nuclear grade valve is unsuitable for MSR due to its special coolant and extraordinary working temperature. Freeze-valve is proposed as the most appropriate valve for MSR, but the technology issue about freeze-valve has not been report in recent decades. Its significance to test the comprehensive property of freeze-valve for the application in MSR. A high temperature molten salt test loop was built which the physics property of salt is similar to the coolant of MSR. The results indicate that freeze-valve has a good performance use in the molten salt circumstances of high temperature (max 700 deg. C) and strong corrosion (authors)

  7. Thermal-Neutron-Induced Fission of U235, U233 and Pu239

    International Nuclear Information System (INIS)

    Thomas, T.D.; Gibson, W.M.; Safford, G.J.

    1965-01-01

    We have used solid-state detectors to measure the kinetic energies of the coincident fission fragments in the thermal-neutron-induced fission of U 235 , U 233 and Pu 239 . Special care has been taken to eliminate spurious-events near symmetry to give an accurate measure of such quantities as the average total kinetic energy at symmetry. For each fissioning system over 10 6 events were recorded. As a result the statistics are good enough to see definite evidence for fine structure over a wide range of masses and energies. The data have been analysed to give mass yield curves, average kinetic energies as a function of mass, and other quantities of interest. For each fissioning system the average total kinetic energy goes through a maximum for a heavy fragment mass of about 132 and for the corresponding light fragment mass. There is a pronounced minimum at symmetry, although not as deep as that found in time-of-flight experiments. The difference between the maximum average kinetic energy and that at symmetry is about 32 MeV for U 235 , 18 MeV for U 233 and 20 MeV for Pu 239 . The dispersion of kinetic energies at symmetry is also smaller than that found in time-of-flight experiments. Fine structure is apparent in two different representations of the data. The energy spectrum of heavy fragments in coincidence with light fragment energies is greater than the most probable value. This structure becomes more pronounced as the light fragment energy increases. The mass yield curves for a given total kinetic energy show a structure suggesting a preference for fission fragments with masses ∼134, ∼140 and ∼145 (and their light fragment partners). Much of the structure observed can be understood by considering a semi-empirical mass surface and a simple model for the nuclear configuration at the saddle point. (author) [fr

  8. Ecdysone Induction of MsrA Protects Against Oxidative Stress in Drosophila

    Energy Technology Data Exchange (ETDEWEB)

    Roesijadi, Guri; Rezvankhah, Saeid; Binninger, David M.; Weissbach, Herbert

    2007-03-09

    The methionine sulfoxide reductases MsrA and MsrB reduce Met(O) to Met in epimer-specific fashion. In Drosophila, the major ecdysone induced protein is MsrA, which is regulated by the EcR-USP complex. We tested Kc cells for induction of MsrA, MsrB, EcR. and CAT by ecdysone and found that MsrA and the EcR were induced by ecdysone, but MsrB and CAT were not. When we tested for resistance to 20 mM H2O2 toxicity, viability of Kc cells was reduced threefold. After pretreatment with 0.2 μM ecdysone for 48 h, then exposed to H2O2, viability of Kc cells increased to 77% of controls. The EcR-deficient L57-3-11 knockout line was not responsive to ecdysone, and H2O2 resistance of both control and ecdysone-treated L57-3-11 cells was similar to that of the ecdysone-untreated Kc cells. These results show that hormonal regulation of MsrA is implicated in conferring protection against oxidative stress in the Drosophila model.

  9. U and Th background contents in the Crimea rocks, Th/U ratio interpretation

    International Nuclear Information System (INIS)

    Gerasimov, Yu.G.; Voronova, M.A.; AN Ukrainskoj SSR, Kiev. Inst. Geologicheskikh Nauk)

    1981-01-01

    Radiogeochemical sampling of rocks in the Crimea is carried out. Analyses for U are made by luminescence method, while for Th by X-ray one. About 1000 samples, characterizing the whole stratigraphical cross section except Pz of crystalline rocks, have been analyzed. Low background values for U and Th are established; U content of 1.3-2.3 g/kg and Th content of 7.0-9.0 g/kg are predominant, that is lower than the clark of the earth crust. The distribution of radioelements in stepper and submontane Crimea is gradual, while in mountain part - the mosaic one. Results of Th/U ratios interpretation are presented [ru

  10. Influence of form factors and multistep effects on the 232Th(d,t) 231Th and 230Th(d,p) 231Th reactions

    International Nuclear Information System (INIS)

    Wilcke, W.; Felix, W.; Elze, Th.W.; Huizenga, J.R.; Thompson, R.C.; Dreisler, R.M.

    1977-01-01

    Tables of spectroscopic information are given for the nuclei 233 , 235 , 237 Pa, and 229 , 231 Ac studied in the reactions 234 , 236 , 238 U, and 230 , 232 Th(t,α), including spin, parity, differential cross sections, and spectra. These quantities are discussed and plotted

  11. Virginia ADS consortium - thorium utilization

    International Nuclear Information System (INIS)

    Myneni, Ganapati

    2015-01-01

    A Virginia ADS consortium, consisting of Virginia Universities (UVa, VCU, VT), Industry (Casting Analysis Corporation, GEM*STAR, MuPlus Inc.), Jefferson Lab and not-for-profit ISOHIM, has been organizing International Accelerator-Driven Sub-Critical Systems (ADS) and Thorium Utilization (ThU) workshops. The third workshop of this series was hosted by VCU in Richmond, Virginia, USA Oct 2014 with CBMM and IAEA sponsorship and was endorsed by International Thorium Energy Committee (IThEC), Geneva and Virginia Nuclear Energy Consortium Authority. In this presentation a brief summary of the successful 3 rd International ADS and ThU workshop proceedings and review the worldwide ADS plans and/or programs is given. Additionally, a report on new start-ups on Molten Salt Reactor (MSR) systems is presented. Further, a discussion on potential simplistic fertile 232 Th to fissile 233 U conversion is made

  12. Data Analysis with the Morse-Smale Complex: The msr Package for R

    KAUST Repository

    Gerber, Samuel

    2012-01-01

    In many areas, scientists deal with increasingly high-dimensional data sets. An important aspect for these scientists is to gain a qualitative understanding of the process or system from which the data is gathered. Often, both input variables and an outcome are observed and the data can be characterized as a sample from a high-dimensional scalar function. This work presents the R package msr for exploratory data analysis of multivariate scalar functions based on the Morse-Smale complex. The Morse-Smale complex provides a topologically meaningful decomposition of the domain. The msr package implements a discrete approximation of the Morse-Smale complex for data sets. In previous work this approximation has been exploited for visualization and partition-based regression, which are both supported in the msr package. The visualization combines the Morse-Smale complex with dimension-reduction techniques for a visual summary representation that serves as a guide for interactive exploration of the high-dimensional function. In a similar fashion, the regression employs a combination of linear models based on the Morse-Smale decomposition of the domain. This regression approach yields topologically accurate estimates and facilitates interpretation of general trends and statistical comparisons between partitions. In this manner, the msr package supports high-dimensional data understanding and exploration through the Morse-Smale complex.

  13. System Requirements Document for the Molten Salt Reactor Experiment 233U conversion system

    International Nuclear Information System (INIS)

    Aigner, R.D.

    2000-01-01

    The purpose of the conversion process is to convert the 233 U fluoride compounds that are being extracted from the Molten Salt Reactor Experiment (MSRE) equipment to a stable oxide for long-term storage at Bldg. 3019

  14. Cross sections and neutron yields for U-233, U-235 and Pu-239 at 2200 m/sec

    Energy Technology Data Exchange (ETDEWEB)

    Sjoestrand, N G; Story, J S

    1960-04-15

    The experimental information on the 2200 m/sec values for {sigma}{sub abs}, {sigma}{sub f}, {alpha}, {nu} and {eta} for {sup 233}U , {sup 235}U and {sup 23} been collected and discussed. The values will later be used in an evaluation of a 'best' set of data. In appendix the isotopic abundances of the uranium isotopes are discussed and also the alpha activities of the uranium isotopes and Pu-239.

  15. Comparison of Hansen--Roach and ENDF/B-IV cross sections for 233U criticality calculations

    International Nuclear Information System (INIS)

    McNeany, S.R.; Jenkins, J.D.

    1976-01-01

    A comparison is made between criticality calculations performed using ENDF/B-IV cross sections and the 16-group Hansen-- Roach library at ORNL. The area investigated is homogeneous systems of highly enriched 233 U in simple geometries. Calculations are compared with experimental data for a wide range of H/ 233 U ratios. Results show that calculations of k/sub eff/ made with the Hansen--Roach cross sections agree within 1.5 percent for the experiments considered. Results using ENDF/B-IV cross sections were in good agreement for well-thermalized systems, but discrepancies up to 7 percent in k/sub eff/ were observed in fast and epithermal systems

  16. Feasibility to convert an advanced PWR from UO2 to a mixed (U,Th)O2 core

    International Nuclear Information System (INIS)

    Stefani, Giovanni Laranjo de; Maiorino, José Rubens; Moreira, João Manoel de Losada; Santos, Thiago Augusto dos; Rossi, Pedro Carlos Russo

    2017-01-01

    This work presents the neutronics and thermal hydraulics feasibility to convert the UO2 core of the Westinghouse AP1000 in a (U-Th)O 2 core, rather than the traditional uranium dioxide, for the purpose of reducing long-lived actinides, especially plutonium, and generates a stock pile of 233 U, which could in the future be used in advanced fuel cycles, in a more sustainable process and taking advantage of the large stock of thorium available on the planet and especially in Brazil. The reactor chosen as reference was the AP1000, which is considered to be one of the most reliable and modern reactor of the current Generation III, and its similarity to the reactors already consolidated and used in Brazil for electric power generation. The results show the feasibility and potentiality of the concept, without the necessity of changes in the core of the AP1000, and even with advantages over this. The neutron calculations were made by the SERPENT code. The results provided a maximum linear power density lower than the AP1000, favoring safety. In addition, the delayed neutron fraction and the reactivity coefficients proved to be adequate to ensure the safety of the concept. The results show that a production of about 260 Kg of 233 U per cycle is possible, with a minimum production of fissile plutonium that favors the use of the concept in U-Th cycles. (author)

  17. Dicty_cDB: SHH233 [Dicty_cDB

    Lifescience Database Archive (English)

    Full Text Available SH (Link to library) SHH233 (Link to dictyBase) - - - Contig-U11264-1 - (Link to Or...c08.g1 Strongyloides ratti whole genome shotgun library (SRAAGSS 004) Strongyloides ratti genomic...iginal site) - - SHH233Z 563 - - - - Show SHH233 Library SH (Link to library) Clone ID SHH233 (Link to dicty...631_5( AY458631 |pid:none) Uncultured marine bacterium 159 cl... 84 3e-15 CU92816...86F1 NIH_MGC_58 Homo sapiens cDNA clone IMAGE:4069772 5', mRNA sequence. 46 0.86 1 AP008210 |AP008210.1 Oryza sativa (japonica culti

  18. Use of sup(233)U for high flux reactors

    International Nuclear Information System (INIS)

    Sekimoto, Hiroshi; Liem, P.H.

    1991-01-01

    The feasibility design study on the graphite moderated gas cooled reactor as a high flux reactor has been performed. The core of the reactor is equipped with two graphite reflectors, i.e., the inner reflector and the outer reflector. The highest value of the thermal neutron flux and moderately high thermal neutron flux are expected to be achieved in the inner reflector region and in the outer reflector region respectively. This reactor has many merits comparing to the conventional high flux reactors. It has the inherent safety features associated with the modular high temperature reactors. Since the core is composed with pebble bed, the on-power refueling can be performed and the experiment time can be chosen as long as necessary. Since the thermal-to-fast flux ratio is large, the background neutron level is low and material damage induced by fast neutrons are small. The calculation was performed using a four groups diffusion approximation in a one-dimensional spherical geometry and a two-dimensional cylindrical geometry. By choosing the optimal values of the core-reflector geometrical parameters and moderator-to-fuel atomic density, high thermal neutron flux can be obtained. Because of the thermal neutron flux can be obtained. Because of the thermal design constraint, however, this design will produce a relatively large core volume (about 10 7 cc) and consequently a higher reactor power (100 MWth). Preliminary calculational results show that with an average power density of only 10 W/cc, maximum thermal neutron flux of 10 15 cm -2 s -1 can be achieved in the inner reflector. The eta value of 233 U is larger than 235 U. By introducing 233 U as the fissile material for this reactor, the thermal neutron flux level can be increased by about 15%. (author). 3 refs., 2 figs., 4 tabs

  19. A study of rates of (n, f), (n, γ), and (n, 2n) reactions in natU and 232Th produced by the neutron fluence in the graphite set-up (gamma-3) irradiated by 2.33 GeV deuteron beam

    International Nuclear Information System (INIS)

    Adam, J.; Chitra Bhatia; Katovskij, K.

    2011-01-01

    Spallation neutrons produced in a collision of 2.33 GeV deuteron beam with the large lead target are moderated by the thick graphite block surrounding the target and used to activate the radioactive samples of nat U and Th put at the three different positions, identified as holes 'a', 'b' and 'c' in the graphite block. Rates of the (n, f), (n, γ), and (n, 2n) reactions in the two samples are determined using the gamma spectrometry. Ratio of the experimental reaction rates, R(n, 2n)/R(n, f) for the 232 Th and nat U are estimated in order to understand the role of reactions of (n, xn) type in Accelerator Driven Subcritical Systems. For the Th-sample, the ratio is ∼ 54(10)% in case of hole 'a' and ∼ 95(57)% in case of hole 'b' compared to 1.73(20)% for the hole 'a' and 0.710(9)% for the hole 'b' in case of the nat U sample. Also the ratio of fission rates in uranium to thorium, nat U(n, f)/ 232 Th(n, f), is ∼ 11.2(17) in case of hole 'a' and 26.8(85) in hole 'b'. Similarly, ratio 238 U(n, 2n)/ 232 Th(n, 2n) is 0.36(4) for the hole 'a' and 0.20(10) for the hole 'b' showing that 232 Th is more prone to the (n, xn) reaction than 238 U. All the experimental reaction rates are compared with the simulated ones by generating neutron fluxes at the three holes from MCNPX 2.6c and making use of LA150 library of cross sections. The experimental and calculated rates of all the three reactions are in good agreement. The transmutation power of the set-up is estimated using the rates of (n, γ) and (n, 2n) reactions for both the samples in the three holes and compared with some of the results of the 'Energy plus Transmutation' set-up and TARC experiment

  20. Evaluation of fission cross sections and covariances for 233U, 235U, 238U, 239Pu, 240Pu, and 241Pu

    International Nuclear Information System (INIS)

    Kawano, Toshihiko; Matsunobu, Hiroyuki; Murata, Toru

    2000-02-01

    A simultaneous evaluation code SOK (Simultaneous evaluation on KALMAN) has been developed, which is a least-squares fitting program to absolute and relative measurements. The SOK code was employed to evaluate the fission cross sections of 233 U, 235 U, 238 U, 239 Pu, 240 Pu, and 241 Pu for the evaluated nuclear data library JENDL-3.3. Procedures of the simultaneous evaluation and the experimental database of the fission cross sections are described. The fission cross sections obtained were compared with evaluated values given in JENDL-3.2 and ENDF/B-VI. (author)

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

    International Nuclear Information System (INIS)

    Zhang, Jiahua

    1994-01-01

    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 232 Th or 238 U 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- 233 U fueled reactor will be discussed. 3 refs., 1 tab., 2 figs

  2. Core design options for high conversion BWRs operating in Th–233U fuel cycle

    International Nuclear Information System (INIS)

    Shaposhnik, Y.; Shwageraus, E.; Elias, E.

    2013-01-01

    Highlights: • BWR core operating in a closed self-sustainable Th– 233 U fuel cycle. • Seed blanket optimization that includes assembly size array and axial dimensions. • Fully coupled MC with fuel depletion and thermo-hydraulic feedback modules. • Thermal-hydraulic analysis includes MCPR observation. -- Abstract: Several options of fuel assembly design are investigated for a BWR core operating in a closed self-sustainable Th– 233 U fuel cycle. The designs rely on an axially heterogeneous fuel assembly structure consisting of a single axial fissile zone “sandwiched” between two fertile blanket zones, in order to improve fertile to fissile conversion ratio. The main objective of the study was to identify the most promising assembly design parameters, dimensions of fissile and fertile zones, for achieving net breeding of 233 U. The design challenge, in this respect, is that the fuel breeding potential is at odds with axial power peaking and the core minimum critical power ratio (CPR), hence limiting the maximum achievable core power rating. Calculations were performed with the BGCore system, which consists of the MCNP code coupled with fuel depletion and thermo-hydraulic feedback modules. A single 3-dimensional fuel assembly having reflective radial boundaries was modeled applying simplified restrictions on the maximum centerline fuel temperature and the CPR. It was found that axially heterogeneous fuel assembly design with a single fissile zone can potentially achieve net breeding, while matching conventional BWR core power rating under certain restrictions to the core loading pattern design

  3. Neutronic studies of a 233U breeder

    International Nuclear Information System (INIS)

    Hansen, L.F.; Maniscalco, J.A.

    1978-09-01

    Neutronic calculations have been carried out to design a laser fusion driven hybrid blanket which maximizes 233 U production per unit of thermal energy (>1 kg/MW/sub T/-year) with acceptable fusion energy multiplication (M/sub F/ approx. 4). Two hybrid blankets, a thorium and a uranium--thorium blanket, are discussed in detail and their performance is evaluated by incorporating them into an existing hybrid design (the LLL/Bechtel design). The performance of these two blankets is discussed in terms of their energy multiplication, tritium breeding and fissile fuel production. The neutronic calculations have been done for two neutron libraries, the ENDF/B-IV and the ENDL with differences no larger than 10% in the results. An estimate is given of the number of equivalent thermal power fission reactors (LWR, HWR, SSCR, and HTGR) that these fusion breeders can fuel

  4. Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

    Energy Technology Data Exchange (ETDEWEB)

    Mohammed, Abdul Aziz, E-mail: azizM@uniten.edu.my; Rahman, Shaik Mohmmed Haikhal Abdul [Universiti Tenaga Nasional. Jalan Ikram-UNITEN, 43000 Kajang, Selangor (Malaysia); Pauzi, Anas Muhamad, E-mail: anas@uniten.edu.my; Zin, Muhamad Rawi Muhammad; Jamro, Rafhayudi; Idris, Faridah Mohamad [Malaysian Nuclear Agency, Bangi, 43000 Kajang, Selangor (Malaysia)

    2016-01-22

    In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 ({sup 233}U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintaining the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.

  5. Photonuclear reactions of U-233 and Pu-239 near threshold induced by thermal neutron capture gamma rays

    International Nuclear Information System (INIS)

    Moraes, M.A.P. de.

    1990-01-01

    The photonuclear cross sections of U-293 and Pu-239 have been studied by using monochromatic and discrete photons, in the energy interval from 5.49 to 9.72 MeV, produced by thermal neutron capture. The gamma fluxes incident on the samples were measured using a ( 3 x 3 )'' NaI (TI) crystal. The photofission fragments were detected in Makrofol-Kg (SSNTD). A possible structure was observed in the U-233 cross sections, near 7.23 MeV. The relative fissionability of the nuclides was determined at each excitation energy and shown to be energy independent: ( 2.12 ± 0.25) for U-233 and ( 3.32 ± 0.41 ) for Pu-239. The angular distribution of photofission fragments of Pu-239 were measured at two mean excitation energies of 5.43 and 7.35 MeV. An anisotropic distribution of ( 12.2 ± 3.6 ) % was observed at 5.43 MeV. The total neutron cross sections were measured by using a long counter detector. The photoneutron cross sections were calculated by using energy dependent neutron multiplicities values, γ(E), obtained in the literature. The competition Γn/γf was also determined at each excitation energy, and shown to be energy independent: ( 0.54 ± 0.05 ) for U-233 and ( 0.44 ± 0.05 ) for Pu-239, and were correlated to the parameters Z sup(2)/A, ( Ef'-Bn'), A. According to the FUJIMOTO-YAMAGUCHI and CONSTANT NUCLEAR TEMPERATURE models, the nuclear temperatures were calculated. The total photoabsorption cross sections were also calculated as a sum of the photofission and photoneutron cross sections at each energy excitation. From these results the competition Γf/ΓA, called fission probability Pf, were obtained: ( 0.66 ± 0.02) for U-233 and ( 0.70 ± 0.02 ) for Pu-239. (author)

  6. A unified form of exact-MSR codes via product-matrix frameworks

    KAUST Repository

    Lin, Sian Jheng

    2015-02-01

    Regenerating codes represent a class of block codes applicable for distributed storage systems. The [n, k, d] regenerating code has data recovery capability while possessing arbitrary k out of n code fragments, and supports the capability for code fragment regeneration through the use of other arbitrary d fragments, for k ≤ d ≤ n - 1. Minimum storage regenerating (MSR) codes are a subset of regenerating codes containing the minimal size of each code fragment. The first explicit construction of MSR codes that can perform exact regeneration (named exact-MSR codes) for d ≥ 2k - 2 has been presented via a product-matrix framework. This paper addresses some of the practical issues on the construction of exact-MSR codes. The major contributions of this paper include as follows. A new product-matrix framework is proposed to directly include all feasible exact-MSR codes for d ≥ 2k - 2. The mechanism for a systematic version of exact-MSR code is proposed to minimize the computational complexities for the process of message-symbol remapping. Two practical forms of encoding matrices are presented to reduce the size of the finite field.

  7. A unified form of exact-MSR codes via product-matrix frameworks

    KAUST Repository

    Lin, Sian Jheng; Chung, Weiho; Han, Yunghsiangsam; Al-Naffouri, Tareq Y.

    2015-01-01

    Regenerating codes represent a class of block codes applicable for distributed storage systems. The [n, k, d] regenerating code has data recovery capability while possessing arbitrary k out of n code fragments, and supports the capability for code fragment regeneration through the use of other arbitrary d fragments, for k ≤ d ≤ n - 1. Minimum storage regenerating (MSR) codes are a subset of regenerating codes containing the minimal size of each code fragment. The first explicit construction of MSR codes that can perform exact regeneration (named exact-MSR codes) for d ≥ 2k - 2 has been presented via a product-matrix framework. This paper addresses some of the practical issues on the construction of exact-MSR codes. The major contributions of this paper include as follows. A new product-matrix framework is proposed to directly include all feasible exact-MSR codes for d ≥ 2k - 2. The mechanism for a systematic version of exact-MSR code is proposed to minimize the computational complexities for the process of message-symbol remapping. Two practical forms of encoding matrices are presented to reduce the size of the finite field.

  8. Possible viscosity effects in neutron-induced fission of 232Th and 238U

    International Nuclear Information System (INIS)

    Gindler, J.E.; Glendenin, L.E.; Wilkins, B.D.

    1979-01-01

    Fission yields induced in the 238 U(n,f) and 232 Th(n,f) reactions were determined as a function of incident neutron energy (E/sub n/). The ratio of 115 Cd-to- 140 Ba yields as a function of E/sub n/ is analyzed by means of the equation Y 1 /Y 2 = exp[2(a 1 (E/sub n/+E 1 )/sup 1/2/ -2(a 2 (E/sub n/+E 2 )/sup 1/2/] to give values of a/sub i/, the level density parameter, and E/sub i/, the excitation energy for E/sub n/=0. The energies E/sub i/ are interpreted on the basis of the liquid drop model with shell and pairing corrections. Values are deduced for the energy dissipated by viscosity effects in the descent from the saddle point to the point where masses are fixed in the fissioning nucleus. These values are 1.7 MeV for 232 Th(n,f) and 4.8 MeV for 238 U(n,f). These values are consistent with the experimental observation that anti ν/sub p/ is approx. 0.6 neutron greater for 239 U fission than for 233 Th fission and that strong odd--even (nucleon pairing) effects are found in the fragment total kinetic energy distribution for 230 Th fission but not for 234 U fission. The low dissipation energy values together with the low values of pre-scission kinetic energy deduced by Guet, et al., [Nucl. Phys. A134 (1971)1] indicate a shorter path from the saddle point of the fissioning nucleus to scission than is generally assumed in theoretical calculations. 31 references

  9. 230Th/238U and 226Ra/230Th fractionation in young basaltic glasses from the East Pacific Rise

    International Nuclear Information System (INIS)

    Reinitz, I.; Turekian, K.K.

    1989-01-01

    Mid-ocean ridge basalt (MORB) glasses treated to remove manganese and iron oxyhydroxide coatings containing U and Th decay chain nuclides sequestered from the ambient aqueous environment show little fractionation of ( 230 Th) from ( 238 U), but extensive enrichment of ( 226 Ra) over ( 230 Th). Chronometry based on ( 230 Th)-( 238 U) disequilibrium yields a mean age of 151,000 ± 55,000 (2σ) years ago for Th-U fractionation, from a source region with Th/U (atom ratio) = 1.58 ± 0.43 (2σ). ( 226 Ra) excesses over ( 230 Th) indicate an enrichment event more recent than the fractionation of Th from U. ( 226 Ra/ 230 Th) correlates well with K/U, providing a rough means of estimating the age since eruption of MORB glasses. (orig.)

  10. Feasibility to convert an advanced PWR from UO{sub 2} to a mixed (U,Th)O{sub 2} core

    Energy Technology Data Exchange (ETDEWEB)

    Stefani, Giovanni Laranjo de; Maiorino, José Rubens; Moreira, João Manoel de Losada; Santos, Thiago Augusto dos, E-mail: giovanni_laranjo@yahoo.com.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil); Rossi, Pedro Carlos Russo [Department of Energy, System, Territory, and Construction Engineering (DESTEC), Pisa (Italy)

    2017-07-01

    This work presents the neutronics and thermal hydraulics feasibility to convert the UO2 core of the Westinghouse AP1000 in a (U-Th)O{sub 2} core, rather than the traditional uranium dioxide, for the purpose of reducing long-lived actinides, especially plutonium, and generates a stock pile of {sup 233}U, which could in the future be used in advanced fuel cycles, in a more sustainable process and taking advantage of the large stock of thorium available on the planet and especially in Brazil. The reactor chosen as reference was the AP1000, which is considered to be one of the most reliable and modern reactor of the current Generation III, and its similarity to the reactors already consolidated and used in Brazil for electric power generation. The results show the feasibility and potentiality of the concept, without the necessity of changes in the core of the AP1000, and even with advantages over this. The neutron calculations were made by the SERPENT code. The results provided a maximum linear power density lower than the AP1000, favoring safety. In addition, the delayed neutron fraction and the reactivity coefficients proved to be adequate to ensure the safety of the concept. The results show that a production of about 260 Kg of {sup 233}U per cycle is possible, with a minimum production of fissile plutonium that favors the use of the concept in U-Th cycles. (author)

  11. Relics of continental lithosphere in the atlantic by data of (Th/U)th, (Th/U)pb, and K/Ti systematics

    International Nuclear Information System (INIS)

    Titayeva, N.A.; Mironov, Y.

    1996-01-01

    The problem of mantle heterogeneity of the Atlantic Ocean has been investigated. Th/U ratio is a sensitive geochemical tracer. This ratio can be calculated from 232Th/230Th or from 208Pb/206Pb since 230Th and 206Pb are decay product of 238U and 208Pb is a product of 232Th decay. The former way to get Th/U ratio is valid for quaternary volcanic rocks because of the relatively short 230Th half life (80000 years), the latter way gives an integral quantity characterizing the pre-oceanic history beginning from the onset of earth formation and ending about 150 million years ago. Clear distinctions between ''depleted oceanic'' and ''enriched continental'' reservoirs have been drawn from comparative analysis of rocks. (A.C.)

  12. The use in nuclear reactors of plutonium and U233 produced in accelerators

    International Nuclear Information System (INIS)

    Gambier, G.

    1983-01-01

    After a review of the presently known energy production systems and the estimated world's energy cumulative consumption during the next century, the author considers the production of fertile isotopes Pu239 and U233 in proton accelerators and finally their different uses in conventional PWR or FBR and the thorium cycle. (A.F.)

  13. Uranium-233 waste definition: Disposal options, safeguards, criticality control, and arms control

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Storch, S.N.; Lewis, L.C.

    1998-01-01

    The US investigated the use of 233 U for weapons, reactors, and other purposes from the 1950s into the 1970s. Based on the results of these investigations, it was decided not to use 233 U on a large scale. Most of the 233 U-containing materials were placed in long-term storage. At the end of the cold war, the US initiated, as part of its arms control policies, a disposition program for excess fissile materials. Other programs were accelerated for disposal of radioactive wastes placed in storage during the cold war. Last, potential safety issues were identified related to the storage of some 233 U-containing materials. Because of these changes, significant activities associated with 233 U-containing materials are expected. This report is one of a series of reports to provide the technical bases for future decisions on how to manage this material. A basis for defining when 233 U-containing materials can be managed as waste and when they must be managed as concentrated fissile materials has been developed. The requirements for storage, transport, and disposal of radioactive wastes are significantly different than those for fissile materials. Because of these differences, it is important to classify material in its appropriate category. The establishment of a definition of what is waste and what is fissile material will provide the guidance for appropriate management of these materials. Wastes are defined in this report as materials containing sufficiently small masses or low concentrations of fissile materials such that they can be managed as typical radioactive waste. Concentrated fissile materials are defined herein as materials containing sufficient fissile content such as to warrant special handling to address nuclear criticality, safeguards, and arms control concerns

  14. 47 CFR 90.233 - Base/mobile non-voice operations.

    Science.gov (United States)

    2010-10-01

    ... 47 Telecommunication 5 2010-10-01 2010-10-01 false Base/mobile non-voice operations. 90.233... SERVICES PRIVATE LAND MOBILE RADIO SERVICES Non-Voice and Other Specialized Operations § 90.233 Base/mobile non-voice operations. The use of A1D, A2D, F1D, F2D, G1D, or G2D emission may be authorized to base...

  15. Strategy for the future use and disposition of uranium-233: Technical information

    International Nuclear Information System (INIS)

    Bereolos, P.J.; Forsberg, C.W.; Kocher, D.C.; Krichinsky, A.M.

    1998-04-01

    This document provides a summary of technical information on the synthetic radioisotope 233 U. It is one of a series of four reports that map out a national strategy for the future use and disposition of 233 U. The technical information on 233 U in this document falls into two main areas. First, material characteristics are presented along with the contrasts of 233 U to the more well known strategic fissile materials, 235 U and plutonium (Pu). Second, information derived from the scientific information, such as safeguards, waste classifications, material form, and packaging, is presented. Throughout, the effects of isotopically diluting 233 U with nonfissile, depleted uranium (DU) are examined

  16. A 233U/236U/242Pu/244Pu spike for isotopic and isotope dilution analysis by mass spectrometry with internal calibration

    International Nuclear Information System (INIS)

    Stepanov, A.; Belyaev, B.; Buljanitsa, L.

    1989-11-01

    The Khlopin Radium Institute prepared on behalf of the IAEA a synthetic mixture of 233 U, 236 U, 242 Pu and 244 Pu isotopes. The isotopic composition and elemental concentration of uranium and plutonium were certified on the basis of analyses done by four laboratories of the IAEA Network, using mass spectrometry with internal standardization. The certified values for 233 U/ 236 U ratio and the 236 U chemical concentration have a coefficient of variation of 0.05%. The latter is fixed by the uncertainty in the 235 U/ 238 U ratio of NBS500 used as internal standard. The coefficients of variation of the 244 Pu/ 242 Pu ratio and the 242 Pu chemical concentration are respectively 0.10% and 0.16% and limited by the uncertainty in the 240 Pu/ 239 Pu ratio of NBS947. This four isotope mixture was used as an internal standard as well as a spike, to analyze 30 batches of LWR spent fuel solutions. The repeatability of the mass spectrometric measurements have a coefficient of variation of 0.025% for the uranium concentration, and of 0.039% for the plutonium concentration. The spiking and treatment errors had a coefficient of variation of 0.048%. (author). Refs, figs and tabs

  17. [download] (941Data Analysis with the Morse-Smale Complex: The msr Package for R

    Directory of Open Access Journals (Sweden)

    Samuel Gerber

    2012-07-01

    Full Text Available In many areas, scientists deal with increasingly high-dimensional data sets. An important aspect for these scientists is to gain a qualitative understanding of the process or system from which the data is gathered. Often, both input variables and an outcome are observed and the data can be characterized as a sample from a high-dimensional scalar function. This work presents the R package msr for exploratory data analysis of multivariate scalar functions based on the Morse-Smale complex. The Morse-Smale complex provides a topologically meaningful decomposition of the domain. The msr package implements a discrete approximation of the Morse-Smale complex for data sets. In previous work this approximation has been exploited for visualization and partition-based regression, which are both supported in the msr package. The visualization combines the Morse-Smale complex with dimension-reduction techniques for a visual summary representation that serves as a guide for interactive exploration of the high-dimensional function. In a similar fashion, the regression employs a combination of linear models based on the Morse-Smale decomposition of the domain. This regression approach yields topologically accurate estimates and facilitates interpretation of general trends and statistical comparisons between partitions. In this manner, the msr package supports high-dimensional data understanding and exploration through the Morse-Smale complex.

  18. Preliminary radiological safety assessment for decommissioning of thoria dissolver of the 233U pilot plant, Trombay

    International Nuclear Information System (INIS)

    Priya, S.; Srinivasan, P.; Gopalakrishnan, R. K.

    2012-01-01

    The thoria dissolver, used for separation of 233 U from reactor-irradiated thorium metal and thorium oxide rods, is no longer operational. It was decided to carry out assessment of the radiological status of the dissolver cell for planning of the future decommissioning/dismantling operations. The dissolver interiors are expected to be contaminated with the dissolution remains of irradiated thorium oxide rods in addition to some of the partially dissolved thoria pellets. Hence, 220 Rn, a daughter product of 228 Th is of major radiological concern. Airborne activity of thoron daughters 212 Pb (Th-B) and 212 Bi (Th-C) was estimated by air sampling followed by high-resolution gamma spectrometry of filter papers. By measuring the full-energy peaks counts in the energy windows of 212 Pb, 212 Bi and 208 Tl, concentrations of thoron progeny in the sampled air were estimated by applying the respective intrinsic peak efficiency factors and suitable correction factors for the equilibration effects of 212 Pb and 212 Bi in the filter paper during the delay between sampling and counting. Then the thoron working level (TWL) was evaluated using the International Commission on Radiological Protection (ICRP) methodology. Finally, the potential effective dose to the workers, due to inhalation of thoron and its progeny during dismantling operations was assessed by using dose conversion factors recommended by ICRP. Analysis of filter papers showed a maximum airborne thoron progeny concentration of 30 TWLs inside the dissolver. (authors)

  19. Safety analysis of thorium-based fuels in the General Electric Standard BWR

    International Nuclear Information System (INIS)

    Colby, M.J.; Townsend, D.B.; Kunz, C.L.

    1980-06-01

    A denatured (U-233/Th)O 2 fuel assembly has been designed which is energy equivalent to and hardware interchangeable with a modern boiling water reactor (BWR) reference reload assembly. Relative to the reference UO 2 fuel, the thorium fuel design shows better performance during normal and transient reactor operation for the BWR/6 product line and will meet or exceed current safety and licensing criteria. Power distributions are flattened and thermal operating margins are increased by reduced steam void reactivity coefficients caused by U-233. However, a (U-233/Th)O 2 -fueled BWR will likely have reduced operating flexibility. A (U-235/Th)O 2 -fueled BWR should perform similar to a UO 2 -fueled BWR under all operating conditions. A (Pu/Th)O 2 -fueled BWR may have reduced thermal margins and similar accident response and be less stable than a UO 2 -fueled BWR. The assessment is based on comparisions of point model and infinite lattice predictions of various nuclear reactivity parameters, including void reactivity coefficients, Doppler reactivity coefficients, and control blade worths

  20. Storage and disposition of weapons usable fissile materials (FMD) PEIS: Blending of U-233 to <12% or <5% enrichment at the Idaho National Engineering Laboratory. Data report, Draft: Version 1

    International Nuclear Information System (INIS)

    Shaber, E.L.

    1995-08-01

    Uranium-233 (U-233), a uranium isotope, is a fissionable material capable of fueling nuclear reactors or being utilized in the manufacturing of nuclear weapons. As such, it is controlled as a special nuclear material. The Idaho National Engineering Laboratory (INEL) and Oak Ridge National Laboratory (ORNL) currently store the Department of Energy's (DOE's) supply of unirradiated U-233 fuel materials. Irradiated U-233 is covered by the national spent nuclear fuel (SNF) program and is not in the scope of this report. The U-233 stored at ORNL is relatively pure uranium oxide in the form of powder or monolithic solids. This material is currently stored in stainless steel canisters of variable lengths measuring about 3 inches in diameter. The ORNL material enrichment varies with some material containing considerable amounts of U-235. The INEL material is fuel from the Light Water Breeder Reactor (LWBR) Program and consists of enriched uranium and thorium oxides in zircaloy cladding. The DOE inventory of U-233 contains trace quantities of U-232, and daughter products from the decay of U-232 and U-233, resulting in increased radioactivity over time. These increased levels of radioactivity generally result in the need for special handling considerations

  1. (232)Th(d,4n)(230)Pa cross-section measurements at ARRONAX facility for the production of (230)U.

    Science.gov (United States)

    Duchemin, C; Guertin, A; Haddad, F; Michel, N; Métivier, V

    2014-05-01

    (226)Th (T1/2=31 min) is a promising therapeutic radionuclide since results, published in 2009, showed that it induces leukemia cells death and activates apoptosis pathways with higher efficiencies than (213)Bi. (226)Th can be obtained via the (230)U α decay. This study focuses on the (230)U production using the (232)Th(d,4n)(230)Pa(β-)(230)U reaction. Experimental cross sections for deuteron-induced reactions on (232)Th were measured from 30 down to 19 MeV using the stacked-foil technique with beams provided by the ARRONAX cyclotron. After irradiation, all foils (targets as well as monitors) were measured using a high-purity germanium detector. Our new (230)Pa cross-section values, as well as those of (232)Pa and (233)Pa contaminants created during the irradiation, were compared with previous measurements and with results given by the TALYS code. Experimentally, same trends were observed with slight differences in orders of magnitude mainly due to the nuclear data change. Improvements are ongoing about the TALYS code to better reproduce the data for deuteron-induced reactions on (232)Th. Using our cross-section data points from the (232)Th(d,4n)(230)Pa reaction, we have calculated the thick-target yield of (230)U, in Bq/μA·h. This value allows now to a full comparison between the different production routes, showing that the proton routes must be preferred. Copyright © 2014 Elsevier Inc. All rights reserved.

  2. Migration of uranium process wastes from the uranium-233--thorium-232 cycle

    International Nuclear Information System (INIS)

    Fried, S.; Sabau, C.; Hines, J.; Friedman, A.

    1978-03-01

    With the advent of fuel loadings of 233 U in the Shippingport Reactor, it has become important to understand the migratory behavior of uranium. The purpose of this study is the determination of the parameters influencing the migration of U(VI), the most likely chemical form of uranium to be mobilized from a repository. Samples of rhyolite tuff were used to measure the absorption coefficients of solutions of U(VI) in ground waters. In addition, columns of tuff were used to measure the elution behavior of U(VI) at various conditions of pH, U(VI) concentration, and flow saturation. These results indicate that there are several elution peaks with values of K/sub d/ between 35 and 120. This behavior is not the same as that of Pu(VI) on tuff; and the experimental results to date have not revealed the reason for this difference. Values of K/sub d/ in this range imply that geological containment would be difficult in strata of this type. It may be possible to find more retentive strata than tuff. Rocks containing reducing components are the most likely candidates and further investigation is urgently needed if the 233 U-Th cycle is to be widely used

  3. The molten salt reactors (MSR) pyro chemistry and fuel cycle for innovative nuclear systems; Congres sur les reacteurs a sels fondus (RSF) pyrochimie et cycles des combustibles nucleaires du futur

    Energy Technology Data Exchange (ETDEWEB)

    Brossard, Ph. [GEDEON, Groupement de Recherche CEA CNRS EDF FRAMATOME (France); Garzenne, C.; Mouney, H. [and others

    2002-07-01

    In the frame of the studies on next generation nuclear systems, and especially for the molten salt reactors and for the integrated fuel cycle (as IFR), the fuel cycle constraints must be taken into account in the preliminary studies of the system to improve the cycle and reactor optimisation. Among the purposes for next generation nuclear systems, sustainability and waste (radio-toxicity and mass) management are important goals. These goals imply reprocessing and recycling strategies. The objectives of this workshop are to present and to share the different strategies and scenarios, the needs based on these scenarios, the experimental facilities available today or in the future and their capabilities, the needs for demonstration. It aims at: identifying the needs for fuel cycle based on solid fuel or liquid fuel, and especially, the on-line reprocessing or clean up for the molten salt reactors; assessing the state-of-the-art on the pyro-chemistry applied to solid fuel and to present the research activities; assessing the state-of-the-art on liquid fuels (or others), and to present the research activities; expressing the R and D programs for pyro-chemistry, molten salt, and also to propose innovative processes; and proposing some joint activities in the frame of GEDEON and PRACTIS programs. This document brings together the transparencies of 18 contributions dealing with: scenario studies with AMSTER concept (Scenarios, MSR, breeders (Th) and burners); fuel cycle for innovative systems; current reprocessing of spent nuclear fuel (SNF) in molten salts (review of pyro-chemistry processes (non nuclear and nuclear)); high temperature NMR spectroscopies in molten salts; reductive extraction of An from molten fluorides (salt - liquid metal extraction); electrochemistry characterisation; characterisation with physical methods - extraction coefficient and kinetics; electrolytic extraction; dissolution-precipitation of plutonium in the eutectic LiCl-KCl (dissolution and

  4. U-Th Burial Dates on Ostrich Eggshell

    Science.gov (United States)

    Sharp, W. D.; Fylstra, N. D.; Tryon, C. A.; Faith, J. T.; Peppe, D. J.

    2015-12-01

    Obtaining precise and accurate dates at archaeological sites beyond the range of radiocarbon dating is challenging but essential for understanding human origins. Eggshells of ratites (large flightless birds including ostrich, emu and others) are common in many archaeological sequences in Africa, Australia and elsewhere. Ancient eggshells are geochemically suitable for the U-Th technique (1), which has about ten times the range of radiocarbon dating (>500 rather than 50 ka), making eggshells attractive dating targets. Moreover, C and N isotopic studies of eggshell provide insights into paleovegetation and paleoprecipitation central to assessing past human-environment interactions (2,3). But until now, U-Th dates on ratite eggshell have not accounted for the secondary origin of essentially all of their U. We report a novel approach to U-Th dating of eggshell that explicitly accounts for secondary U uptake that begins with burial. Using ostrich eggshell (OES) from Pleistocene-Holocene east African sites, we have measured U and 232Th concentration profiles across OES by laser ablation ICP-MS. U commonly peaks at 10s to 100s of ppb and varies 10-fold or more across the ~2 mm thickness of OES, with gradients modulated by the layered structure of the eggshell. Common Th is high near the shell surfaces, but low in the middle "pallisade" layer of OES, making it optimal for U-Th dating. We determine U-Th ages along the U concentration gradient by solution ICP-MS analyses of two or more fractions of the pallisade layer. We then estimate OES burial dates using a simple model for diffusive uptake of uranium. Comparing such "U-Th burial dates" with radiocarbon dates for OES calcite from the same shells, we find good agreement in 7 out of 9 cases, consistent with rapid burial and confirming the accuracy of the approach. The remaining 2 eggshells have anomalous patterns of apparent ages that reveal they are unsuitable for U-Th dating, thereby providing reliability criteria innate

  5. Power level effects on thorium-based fuels in pressure-tube heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bromley, B.P.; Edwards, G.W.R., E-mail: blair.bromley@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Sambavalingam, P. [Univ. of Ontario Inst. of Technology, Oshawa, Ontario (Canada)

    2016-06-15

    Lattice and core physics modeling and calculations have been performed to quantify the impact of power/flux levels on the reactivity and achievable burnup for 35-element fuel bundles made with Pu/Th or U-233/Th. The fissile content in these bundles has been adjusted to produce on the order of 20 MWd/kg burnup in homogeneous cores in a 700 MWe-class pressure-tube heavy water reactor, operating on a once-through thorium cycle. Results demonstrate that the impact of the power/flux level is modest for Pu/Th fuels but significant for U-233/Th fuels. In particular, high power/flux reduces the breeding and burnup potential of U-233/Th fuels. Thus, there may be an incentive to operate reactors with U-233/Th fuels at a lower power density or to develop alternative refueling schemes that will lower the time-average specific power, thereby increasing burnup.(author)

  6. Power level effects on thorium-based fuels in pressure-tube heavy water reactors

    International Nuclear Information System (INIS)

    Bromley, B.P.; Edwards, G.W.R.; Sambavalingam, P.

    2016-01-01

    Lattice and core physics modeling and calculations have been performed to quantify the impact of power/flux levels on the reactivity and achievable burnup for 35-element fuel bundles made with Pu/Th or U-233/Th. The fissile content in these bundles has been adjusted to produce on the order of 20 MWd/kg burnup in homogeneous cores in a 700 MWe-class pressure-tube heavy water reactor, operating on a once-through thorium cycle. Results demonstrate that the impact of the power/flux level is modest for Pu/Th fuels but significant for U-233/Th fuels. In particular, high power/flux reduces the breeding and burnup potential of U-233/Th fuels. Thus, there may be an incentive to operate reactors with U-233/Th fuels at a lower power density or to develop alternative refueling schemes that will lower the time-average specific power, thereby increasing burnup.(author)

  7. Metasurface Reflector (MSR Loading for High Performance Small Microstrip Antenna Design.

    Directory of Open Access Journals (Sweden)

    Md Rezwanul Ahsan

    Full Text Available A meander stripline feed multiband microstrip antenna loaded with metasurface reflector (MSR structure has been designed, analyzed and constructed that offers the wireless communication services for UHF/microwave RFID and WLAN/WiMAX applications. The proposed MSR assimilated antenna comprises planar straight forward design of circular shaped radiator with horizontal slots on it and 2D metasurface formed by the periodic square metallic element that resembles the behavior of metamaterials. A custom made high dielectric bio-plastic substrate (εr = 15 is used for fabricating the prototype of the MSR embedded planar monopole antenna. The details of the design progress through numerical simulations and experimental results are presented and discussed accordingly. The measured impedance bandwidth, radiation patterns and gain of the proposed MSR integrated antenna are compared with the obtained results from numerical simulation, and a good compliance can be observed between them. The investigation shows that utilization of MSR structure has significantly broadened the -10 dB impedance bandwidth than the conventional patch antenna: from 540 to 632 MHz (17%, 467 to 606 MHz (29% and 758 MHz to 1062 MHz (40% for three distinct operating bands centered at 0.9, 3.5 and 5.5 GHz. Additionally, due to the assimilation of MSR, the overall realized gains have been upgraded to a higher value of 3.62 dBi, 6.09 dBi and 8.6 dBi for lower, middle and upper frequency band respectively. The measured radiation patterns, impedance bandwidths (S11<-10 dB and gains from the MSR loaded antenna prototype exhibit reasonable characteristics that can satisfy the requirements of UHF/microwave (5.8 GHz RFID, WiMAX (3.5/5.5 GHz and WLAN (5.2/5.8 GHz applications.

  8. Gamma Spectrometric Determination of U, Th, K and Some Geochemical Applications

    International Nuclear Information System (INIS)

    Dodona, A.; Tashko, A.

    2001-01-01

    The application of 'in situ' gamma-spectrometric method (''infinite'' environment), made possible the simultanious determination of U, Th and K. 4 channel gamma-spectrometric analyser with NaI(TI) scintilation counter crystal detector (103 cm 3 φ=50x50mm) was used to determin U, Th(more than 1-2 ppm) and K (more than 1%) in laboratory conditions. The detector was inserted into a lead camera and calibrated for measurement geometry with vessel of ''Marineli'' type of a 17o cm 3 volume. The study of main factors, which influence in the gamma spectrometric measurements, (the technical, physical, geometrical and time parameters) has been carried out. International standards of U, Th, K and internal monitoring standard samples are used for the calibration. External analytical control has been realized by other radiometric and chemical methods. The detection limits ( 1 ppm Th, 2ppm U and 1% K) and the relative errors (17-20% for 1-10 ppm U, Th and 10-15% for more than 10 ppm U, Th and more than 1% K) guarantee a quantitative analysis that may be used successfully in the geochemical studies. Some geochemical applications, based on the content of Th, U and Th/U ratio in rocks samples that we have we have analyzed with this method, are shown in this paper. U, Th and their ratio are used as trace elements to indicate the differences between the acidic magmatic rocks of Albania (Th/U ratio=2-6 and>10). The bimodal character of Th/U scattering in ignimbrides and monzonites (Korabi zone) shows that in addition to the ''normal'' rocks, there are also some ones enriched with Th, So, the differential analysis of Th, U, and K may be used as geochemical exploration criteria for the radioactive and non-radioactive mineralization, such as REE (Rare Earth Elements), phospghorites, bauxites, placers etc. (authors)

  9. Thermoionic emission characteristics of uranium with application to its determination by MSID technique using 233U tracer

    International Nuclear Information System (INIS)

    Shihomatsu, H.M.; Iyer, S.S.

    1988-01-01

    Experimental details of the uranium determination in geological samples (50-1500 ppm range) by mass spectrometric isotope dilution technique (MSID) employing 233 U tracer are presented. For this purpose the thermoionic emission characteristics of uranium in various filament arrangements like simple plane, filament boat, double, are studied and the most efficient one selected for the isotope dilution analysis. The various experimental procedures involved in the MSID like sample dissolution, chemical separation and mass spectrometric analysis are developed and optimised. The experimental results on the uranium determination by MSID with 233 U tracer yielded precision and accuracy of 0,5% and 1% respectively. The importance of the sampling in the precise and accuracy determination of uranium in geological samples, where it is heterogeneously distributed, is discussed. (author) [pt

  10. Reactor physical program in the frame of the MSR-SPHINX transmuter concept development

    International Nuclear Information System (INIS)

    Hron, M.; Mikisek, M.

    2008-01-01

    In the frame of the R and D program for the Molten Salt Reactor (MSR) - SPHINX (Spent Hot fuel Incinerator by Neutron flux) concept, which has been under development in the Czech Republic as an actinide burner in resonance neutron spectrum and a radionuclide transmuter in a well-thermalized neutron spectrum, and namely its reactor physical part, the relatively broad experimental activities have been involved in the program, recently, which will serve for a validation of computer codes and verification of design inputs for designing of a demonstration unit of the MSR-type. The experimental program, which has been focused, in its first stage, on a short-term irradiation of small size samples of molten-salt systems as well as structural materials proposed for the MSR blanket in the field of high neutron flux of research reactors, will be in the proposed next stage of the program focused on a large-scale experimental verification of design inputs by use of MSR-type inserted zones into the existing light water moderated experimental reactor LR-0, which may allow to modify it to experimental zero power salt reactor SR-0. There has been a preparatory stage of the project called EROS started in the year 2006 and new experiments with MSR-type zones irradiated by cyclotron based neutron source are planned at the end of 2007 and should go on in the year 2008. There will be a brief description of the so far prepared and performed experimental programs introduced in the paper. (authors)

  11. Management, Spirituality, and Religion (MSR) Ways and Means

    DEFF Research Database (Denmark)

    Tackney, Charles Thomas; Chappell, Stacie F.; Harris, Dan

    2017-01-01

    Despite 15 years of functioning as an interest group, our domain of inquiry is relatively young and there are limited theoretical boundaries to support, shape, and assist our efforts. This metaphorical “blank canvas” is both empowering, in that so many inquiries are open for exploration, and yet...... also limiting. In this document we highlight three critical elements to emphasize their importance in MSR research: (a) delineating and operationalizing the key terms of religion, spirituality, and workplace spirituality; (b) acknowledging the work to date in the MSR corpus around definitions...

  12. Metasurface Reflector (MSR) Loading for High Performance Small Microstrip Antenna Design.

    Science.gov (United States)

    Ahsan, Md Rezwanul; Islam, Mohammad Tariqul; Ullah, Mohammad Habib; Singh, Mandeep Jit; Ali, Mohd Tarmizi

    2015-01-01

    A meander stripline feed multiband microstrip antenna loaded with metasurface reflector (MSR) structure has been designed, analyzed and constructed that offers the wireless communication services for UHF/microwave RFID and WLAN/WiMAX applications. The proposed MSR assimilated antenna comprises planar straight forward design of circular shaped radiator with horizontal slots on it and 2D metasurface formed by the periodic square metallic element that resembles the behavior of metamaterials. A custom made high dielectric bio-plastic substrate (εr = 15) is used for fabricating the prototype of the MSR embedded planar monopole antenna. The details of the design progress through numerical simulations and experimental results are presented and discussed accordingly. The measured impedance bandwidth, radiation patterns and gain of the proposed MSR integrated antenna are compared with the obtained results from numerical simulation, and a good compliance can be observed between them. The investigation shows that utilization of MSR structure has significantly broadened the -10 dB impedance bandwidth than the conventional patch antenna: from 540 to 632 MHz (17%), 467 to 606 MHz (29%) and 758 MHz to 1062 MHz (40%) for three distinct operating bands centered at 0.9, 3.5 and 5.5 GHz. Additionally, due to the assimilation of MSR, the overall realized gains have been upgraded to a higher value of 3.62 dBi, 6.09 dBi and 8.6 dBi for lower, middle and upper frequency band respectively. The measured radiation patterns, impedance bandwidths (S11WLAN (5.2/5.8 GHz) applications.

  13. 230Th/U-dating of a late Holocene low uranium speleothem from Cuba

    International Nuclear Information System (INIS)

    Fensterer, Claudia; Mangini, Augusta; Scholz, Denis; Hoffmann, Derik; Pajon, Jesus M

    2010-01-01

    We present 22 U-series ages for a stalagmite from north-western Cuba based on multi-collector inductively coupled plasma mass spectrometry (MC-ICPMS) and thermal ionisation mass spectrometry (TIMS). Our results reveal that the stalagmite continuously grew within the last ∼1400a. Low uranium content of the sample and thus, extremely low 230 Th concentrations limit the precision and accuracy of 230 Th/U-dating by TIMS. Samples measured by MC-ICPMS show a high variability of 232 Th content along the growth axis with some sections significantly affected by initial 230 Th from a detrital phase. An a-priori bulk earth ratio for ( 238 U/ 232 Th) cannot be used to accurately account for this initial 230 Th. Using an age model based on the 230 Th/U ages determined on samples with low or negligible 232 Th concentration, we find that the ( 238 U/ 232 Th) activity ratio of the detrital phase is an order of magnitude larger than the bulk earth value, indicating the importance of an accurately determined correction factor.

  14. Uses for Uranium-233: What Should Be Kept for Future Needs?

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Lewis, L.C.

    1999-01-01

    Since the end of the cold war, the United States has been evaluating what fissile materials to keep for potential uses and what fissile materials to declare excess. There are three major fissile materials: high-enriched uranium (HEU), plutonium, and uranium-233 ( 233 U). Both HEU and plutonium were produced in large quantities for use in nuclear weapons and for reactor fuel. Uranium-233 was investigated for use in nuclear weapons and as a reactor fuel; however, it was never deployed in nuclear weapons or used commercially as a nuclear fuel. Uranium-233 has limited current uses, but it could have several future uses. Because of (1) the cost of storing 233 U and (2) arms control considerations, the U.S. government must decide how much of the existing 233 U inventory should be kept for future use and how much should be disposed of as waste. The objective of this report is to provide technical and economic input to make a use-or-dispose decision

  15. Storage and disposition of weapons usable fissile materials (FMD) PEIS: Blending of U-233 to {lt}12% or {lt}5% enrichment at the Idaho National Engineering Laboratory. Data report, Draft: Version 1

    Energy Technology Data Exchange (ETDEWEB)

    Shaber, E.L.

    1995-08-01

    Uranium-233 (U-233), a uranium isotope, is a fissionable material capable of fueling nuclear reactors or being utilized in the manufacturing of nuclear weapons. As such, it is controlled as a special nuclear material. The Idaho National Engineering Laboratory (INEL) and Oak Ridge National Laboratory (ORNL) currently store the Department of Energy`s (DOE`s) supply of unirradiated U-233 fuel materials. Irradiated U-233 is covered by the national spent nuclear fuel (SNF) program and is not in the scope of this report. The U-233 stored at ORNL is relatively pure uranium oxide in the form of powder or monolithic solids. This material is currently stored in stainless steel canisters of variable lengths measuring about 3 inches in diameter. The ORNL material enrichment varies with some material containing considerable amounts of U-235. The INEL material is fuel from the Light Water Breeder Reactor (LWBR) Program and consists of enriched uranium and thorium oxides in zircaloy cladding. The DOE inventory of U-233 contains trace quantities of U-232, and daughter products from the decay of U-232 and U-233, resulting in increased radioactivity over time. These increased levels of radioactivity generally result in the need for special handling considerations.

  16. Fabrication of zero power reactor fuel elements containing 233U3O8 powder

    International Nuclear Information System (INIS)

    Nicol, R.G.; Parrott, J.R.; Krichinsky, A.M.; Box, W.D.; Martin, C.W.; Whitson, W.R.

    1982-05-01

    Oak Ridge National Laboratory, under contract with Argonne National Laboratory, completed the fabrication of 1743 fuel elements for use in their Zero Power Reactor. The contract also included recovery of 20 kg of 233 U from rejected elements. This report describes the steps associated with conversion of purified uranyl nitrate (as solution) to U 3 O 8 powder (suitable for fuel) and subsequent charging, sealing, decontamination, and testing of the fuel elements (packets) preparatory to shipment. The nuclear safety, radiation exposures, and quality assurance aspects of the program are discussed

  17. Determination of 233U, 235U, 238U and 239Pu fission yields induced by fission and 14.7 MeV neutrons

    International Nuclear Information System (INIS)

    Laurec, Jean; Adam, Albert; Bruyne, Thierry de.

    1981-12-01

    The 233 U, 235 U, 238 U, 239 Pu fission yields have been determined by a radiochemical method. A target and a fission chamber made of same fissible material are irradied together. The total fission number is measured from the fission chamber. The fission product activities are directly measured on the target using calibrated Ge-Li detectors. The fissible material masses are determined by alpha and mass spectrometries. The irradiations were made on the critical assemblies PROSPERO and CALIBAN and on the 14 MeV neutron generator of C.E. VALDUC. 3 to 5% fission yield errors are got for the most measured nuclides: 95 Zr, 97 Zr, 99 Mo, 103 Ru, 131 I, 132 Te, 140 Ba, 141 Ce, 143 Ce, 144 Ce, 147 Nd [fr

  18. Water Ingress Testing of the Turbula Jar and U-233 Lead Pig Containers

    Energy Technology Data Exchange (ETDEWEB)

    Reeves, Kirk Patrick [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Karns, Tristan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Smith, Paul Herrick [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-11-02

    Understanding the water ingress behavior of containers used at the TA-55 Plutonium Facility has significant implications for criticality safety. The purpose of this report is to document the water ingress behavior of the Turbula Jar with Bakelite lid and Viton gaskets (Turbula Jar) used in oxide blending operations and the U-233 lead pig container used to store and transport U-233 material. The technical basis for water resistant containers at TA-55 is described in LA-UR-15-22781, “Water Resistant Container Technical Basis Document for the TA-55 Criticality Safety Program.” Testing of the water ingress behavior of various containers is described in LA-CP-13-00695, “Water Penetration Tests on the Filters of Hagan and SAVY Containers,” LA-UR-15-23121, “Water Ingress into Crimped Convenience Containers under Flooding Conditions,” and in LA-UR- 16-2411, “Water Ingress Testing for TA-55 Containers.” Water ingress criteria are defined in TA55-AP-522 “TA-55 Criticality Safety Program”, and in PA-RD-01009 “TA55 Criticality Safety Requirements.” The water ingress criteria for submersion is no more than 50 ml of water ingress at a 6” water column height for a period of 2 hours.

  19. The European Expression Of Interest For High Purity U-233 Materials

    Energy Technology Data Exchange (ETDEWEB)

    Giaquinto, Joseph M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Younkin, James R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-04-01

    The purpose of this letter report is to document the response for an Expression of Interest (EOI) sent to the European Safeguards and research and development (R&D) scientific communities for the distribution of small amounts of high purity 233U materials for use in safeguards, nonproliferation, and basic R&D in the nuclear disciplines. The intent for the EOI was to gauge the level of international interest for these materials from government and research institutions with programmatic missions in the nuclear security or nuclear R&D arena. The information contained herein is intended to provide information to assist key decision makers in DOE as to the ultimate disposition path for the high purity materials currently being recovered at Oak Ridge National Laboratory (ORNL) and only those items for which there is no United States (U.S.) sponsor identified.

  20. Extraction of Th and U from Swiss granites

    International Nuclear Information System (INIS)

    Bajo, C.

    1980-12-01

    The extraction, at the laboratory level, of U and Th from Swiss granites is discussed. The Mittagfluh, Bergell and Rotondo granites and the Giuv syenite offered a wide range of U and Th concentrations; 7.7 to 20.0 ppm U and 25.5 to 67.0 ppm Th. U and Th were determined in the leach solutions by the fission track method and by spectrophotometry, respectively. Samples containing less than 0.3 μg U and 4 μg Th, could be measured with an accuracy of 10% for U and 5% for Th. Leach tests were performed during which the following parameters were varied: granite-type, grain size, acid-type, acid concentration, temperature and time. There were very great leaching differences between the granites studied. Temperature was the most important parameter. Sharp differences in extraction occurred between 20 0 C, 50 0 C and 80 0 C. At 80 0 C, more than 85% U and Th were extracted. The extraction curve (percent extracted as a function of time) of aliquots sampled after 1, 2, 4, 8, 12 and 24 hours showed a plateau after 8 hours. The half life of the reaction was between one and two hours. As a general rule, Th was better extracted than U. (Auth.)

  1. Investigation of tritium and 233U breeding in a fission-fusion hybrid reactor fuelling with ThO2

    International Nuclear Information System (INIS)

    Yildiz, K.; Sahin, S.; Sahin, H. M.; Acir, A.; Yalcin, S.; Altinok, T.; Bayrak, M.; Alkan, M.; Durukan, O.

    2007-01-01

    In the world, thorium reserves are three times more than natural Uranium reserves. It is planned in the near future that nuclear reactors will use thorium as a fuel. Thorium is not a fissile isotope because it doesn't make fission with thermal neutrons so it could be converted to 2 33U isotope which has very high quality fission cross-section with thermal neutrons. 2 33U isotope can be used in present LWRs as an enrichment fuel. In the fusion reactors, tritium is the most important fossil fuel. Because tritium is not natural isotope, it has to be produced in the reactor. The purpose of this work is to investigate the tritium and 2 33U breeding in a fission-fusion hybrid reactor fuelling with ThO 2 for Δt=10 days during a reactor operation period in five years. The neutronic analysis is performed on an experimental hybrid blanket geometry. In the center of the hybrid blanket, there is a line neutron source in a cylindrical cavity, which simulates the fusion plasma chamber where high energy neutrons (14.1 MeV) are produced. The conventional fusion reaction delivers the external neutron source for blankets following, 2 D + 3 T →? 4 He (3.5 MeV) + n (14.1 MeV). (1) The fuel zone made up of natural-ThO 2 fuel and it is cooled with different coolants. In this work, five different moderator materials, which are Li 2 BeF 4 , LiF-NaF-BeF 2 , Li 2 0Sn 8 0, natural Lithium and Li 1 7Pb 8 3, are used as coolants. The radial reflector, called tritium breeding zones, is made of different Lithium compounds and graphite in sandwich structure. In the work, eight different Lithium compounds were used as tritium breeders in the tritium breeding zones, which are Li 3 N, Li 2 O, Li 2 O 2 , Li 2 TiO 3 , Li 4 SiO 3 , Li 2 ZrO 3 , LiBr and LiH. Neutron transport calculations are conducted in spherical geometry with the help of SCALE4.4A SYSTEM by solving the Boltzmann transport equation with code CSAS and XSDRNPM, under consideration of unresolved and resolved resonances, in S 8 -P 3

  2. Comparison of the U-233 dog data of Stevens et al. with uranium retention functions in ICRP Publication 30 and a 3-compartment mammillary model for uranium

    International Nuclear Information System (INIS)

    Bernard, S.R.

    1983-01-01

    Stevens measured the distribution, retention, and excretion of U-233 in seven beagles each given a single injection of U-233 citrate [2.8 μCi/kg U-233 (VI) (approx.3 mg/dog)]. These data, when plotted together with results obtained with the ICRP (Pub. 30) retention functions for purposes of comparison, are seen to differ only slightly from the ICRP-30 model. The number of transformations in the body, over a fifty-year period agree within a factor of 2. A three-compartment mammillary model has been parameterized from the data of Stevens by the method of Bernard. Retention in tissues of the body is represented by a linear combination of three compartments. The data plots for the dogs and ICRP-30 model will be presented and discussed together with the three compartment mammillary model for U-233 retention, distribution, and excretion. 3 figs., 2 tabs

  3. A study of sodium-cooled fast breeder reactor with thorium blanket for supply of U-233 to high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Yoshida, H.; Nishimura, H.; Osugi, T.

    1978-08-01

    Symbiotic energy system between fast breeder reactor and thermal reactor would have a potential merit for nuclear proliferation problem. And when using HTGR as the thermal reactor in the system, the energy system appears to be promising as an energy system self-sufficient in fuels, which can generate both electricity and high temperature process heat. In the system the fast breeder reactor has to supply sufficient amount of fissile plutonium to keep the reactor going, and also produce U-233 necessary to the associated U-233 fuelled process heat production HTGR. Three types of LMFBR concepts with thorium blanket, conventional homogeneous core LMFBR, and axial and radial parfait heterogeneous core LMFBRs, have been investigated to find out suitable configurations of LMFBR for supply of U-233 to the HTGR with relatively high conversion ratio of 0.85, in the symbiotic energy system between LMFBR and HTGR. The investigation on LMFBR has been made on fuel sufficiency of the system, inherent safety such as sodium-void and Doppler coefficients, and fuel cycle cost. The followings were revealed; (1) Conventional homogeneous core LMFBR with thorium radial blanket well satisfies the condition of fuel sufficiency, if adequate radial blanket thickness is chosen. However, the sodium-void coefficient and fuel cycle cost are inferior to the other concepts. (2) Axial parfait heterogeneous core LMFBR can be regarded as one of the best LMFBR concepts installed in the symbiotic energy system, from the viewpoints of fuel sufficiency, inherent safety and fuel cycle cost. However, further investigations should be needed on reliability and operationability of the concept. (3) Radial parfait heterogeneous core LMFBR seems inadequate as the LMFBR in the system, because the configurations based on this concept does not satisfy plutonium and U-233 breedings, simultaneously. This LMFBR concept, however, has excellent breeding performance in the internal radial blanket. So further

  4. 238U, 234U and 230Th in uranium miners' lungs

    International Nuclear Information System (INIS)

    Singh, M.P.; Wrenn, M.E.; Archer, V.E.; Saccomanno, G.

    1981-01-01

    Fourteen uranium miners' lungs from Colorado plateau were collected at autopsy and the concentrations of 238 U, 234 U and 230 Th were determined by radiochemical procedures utilizing solvent extraction - alpha spectrometric techniques. The uranium and thorium isotopes are in near equilibrium with average concentrations of 238 U, 234 U and 230 Th being 89.3, 95.2 and 91.1 pCi/kg respectively. The combined average radiation dose rate to lung from these three isotopes is about 24.2 mrad/year at death excluding the unmeasured contribution from the 226 Ra and daughters. The average concentration of 230 Th is about 65 times higher than the mean concentration of 230 Th in lungs of non-miners dying at comparable ages from the same region

  5. 238U, 234U and 232Th in seawater

    International Nuclear Information System (INIS)

    Chen, J.H.; Edwards, R.L.; Wasserburg, G.J.

    1986-01-01

    We have developed techniques to determine 238 U, 234 U and 232 Th concentrations in seawater by isotope dilution mass spectrometry. Using these techniques, we have measured 238 U, 234 U and 232 Th in vertical profiles of unfiltered, acidified seawater from the Atlantic and 238 U and 234 U in vertical profiles from the Pacific. Determinations of 234 U/ 238 U at depths ranging from 0 to 4900 m in the Atlantic (7 0 44'N, 40 0 43'W) and the Pacific (14 0 41'N, 160 0 01'W) Oceans are the same within experimental error (±5per mille, 2σ). The average of these 234 U/ 238 U measurements is 144±2per mille (2σ) higher than the equilibrium ratio of 5.472 x 10 -5 . U concentrations, normalized to 35per mille salinity, range from 3.162 to 3.281 ng/g, a range of 3.8%. The average concentration of the Pacific samples (31 0 4'N, 159 0 1'W) is ∝1% higher than that of the Atlantic (7 0 44'N, 40 0 43'W and 31 0 49'N, 64 0 6'W). 232 Th concentrations from an Atlantic profile range from 0.092 to 0.145 pg/g. The observed constancy of the 234 U/ 238 U ratio is consistent with the predicted range of 234 U/ 238 U using a simple two-box model and the residence time of deep water in the ocean determined from 14 C. The variation in salinity-normalized U concentrations suggests that U may be much more reactive in the marine environment than previously thought. (orig./WB)

  6. 39 CFR 233.12 - Civil penalties.

    Science.gov (United States)

    2010-07-01

    ... 39 Postal Service 1 2010-07-01 2010-07-01 false Civil penalties. 233.12 Section 233.12 Postal... Civil penalties. False representation and lottery orders— (a) Issuance. Pursuant to 39 U.S.C. 3005, the... be liable to the United States for a civil penalty in an amount not to exceed $11,000 for each day...

  7. Study of the variation with the energy of the fission cross-sections of {sup 233}U, {sup 235}U, {sup 239}Pu for the fast neutrons; Etude de la variation avec l'energie des sections efficaces de fission de {sup 233}U, {sup 235}U, {sup 239}Pu pour les neutrons rapides

    Energy Technology Data Exchange (ETDEWEB)

    Szteinsznaider, D; Naggiar, V; Netter, F [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1955-07-01

    This measurements have been done while taking the value of the fission cross-sections of {sup 238}U as reference. The neutrons are produced by the reaction {sup 7}Li(p,n) in the Van de Graaff generator of Saclay. The explored domain spreads from some tenths to 2000 keV. We find: for {sup 239}Pu: {sigma}{sub f} = 2,04 {+-} 0,12 barns, cross-section constant between 150 and 2000 keV, for {sup 235}U: {sigma}{sub f} = 1,15 {+-} 0,15 barns, cross-section constant between 700 and 1000 keV, for {sup 233}U: {sigma}{sub f} = 1,92 {+-} 0,25 barns, for neutrons of 850 keV. (authors) [French] Ces mesures ont ete effectuees en prenant la valeur de la section efficace de fission de {sup 238}U comme reference. Les neutrons sont produits par la reaction {sup 7}Li(p,n) au generateur Van de Graaff de Saclay. Le domaine explore s'etend de quelques dizaines de kev a 2000 kev. Nous trouvons: pour {sup 239}Pu: {sigma}{sub f} = 2,04 {+-} 0,12 barns, section efficace constante entre 150 et 2000 kev. pour {sup 235}U: {sigma}{sub f} = 1,15 {+-} 0,15 barns, section efficace constante entre 700 et 1000 kev. pour {sup 233}U: {sigma}{sub f} = 1,92 {+-} 0,25 barns, pour des neutrons de 850 kev. (auteurs)

  8. Fission cross section ratios for 233,234,236U relative to 235U from 0.5 to 400 MeV

    International Nuclear Information System (INIS)

    Lisowski, P.W.; Gavron, A.; Parker, W.E.; Balestrini, S.J.; Carlson, A.D.; Wasson, O.A.; Hill, N.W.

    1991-01-01

    Neutron-induced fission cross section ratios from 0.5 to 400 MeV for samples of 233, 234, 236 U relative to 235 U have been measured at the WNR neutron Source at Los Alamos. The fission reaction rate was determined using a fast parallel plate ionization chamber at a 20-m flight path. Cross sections over most the energy range were also extracted using the neutron fluence determined with three different proton telescope arrangements. Those data provided the shape of the 235 U(n,f) cross section relative to the hydrogen scattering cross section. That shape was then normalized to the very accurately known value for 235 U(n,f) at 14.1 MeV to allow us to obtain cross section section values from the ratio data and our values for 235 U(n,f). 6 refs., 1 fig

  9. Fission cross section ratios for 233,234,236U relative to 235U from 0.5 to 400 MeV

    International Nuclear Information System (INIS)

    Lisowski, P.W.; Gavron, A.; Parker, W.E.; Balestrini, S.J.; Carlson, A.D.; Wasson, O.A.; Hill, N.W.

    1992-01-01

    Neutron-induced fission cross section ratios from 0.5 to 400 MeV for samples of 233,234,236 U relative to 235 U have been measured at the WNR neutron Source at Los Alamos. The fission reaction rate was determined using a fast parallel plate ionization chamber at a 20-m flight path. Cross sections over most of the energy range were also extracted using the neutron fluence determined with three different proton telescope arrangements. Those data provided the shape of the 235 U(n, f) cross section relative to the hydrogen scattering cross section. That shape was then normalized to the very accurately known value for 235 U(n, f) at 14.1 MeV which will allow us to obtain cross section values from the ratio data and our values for 235 U(n, f). (orig.)

  10. 238U, 234U and 230Th in uranium miners' lungs

    International Nuclear Information System (INIS)

    Singh, N.P.; Wrenn, M.E.; Bennett, D.B.; Archer, V.; Saccomanno, G.

    1982-01-01

    Fourteen uranium miners' lungs from the Colorado Plateau were collected at autopsy and the concentrations of 238 U, 234 U and 230 Th were determined by radiochemical procedures utilizing solvent extraction and alpha spectrometric techniques. The uranium and thorium isotopes are in near equilibrium with average concentrations of 238 U, 234 U and 230 Th being 89.3, 95.2, and 91.1 pCi/kg respectively. The combined average radiation dose rate to lung from these three isotopes is about 24.1 mrad/year at death excluding the unmeasured contribution from the 226 Ra and daughters. The average concentration of 230 Th is about 65 times higher than the mean concentration of 230 Th in lungs of non-miners from the same region dying at comparable ages

  11. Reply to Comment on "Zircon U-Th-Pb dating using LA-ICP-MS: Simultaneous U-Pb and U-Th dating on the 0.1 Ma Toya Tephra, Japan"

    Science.gov (United States)

    Ito, Hisatoshi

    2015-04-01

    Guillong et al. (2015) mentioned that corrections for abundance sensitivity for 232Th and molecular zirconium sesquioxide ions (Zr2O3+) are critical for reliable determination of 230Th abundances in zircon for LA-ICP-MS analyses. There is no denying that more rigorous treatments are necessary to obtain more reliable ages than those in Ito (2014). However, as shown in Fig. 2 in Guillong et al. (2015), the uncorrected (230Th)/(238U) for reference zircons except for Mud Tank are only 5-20% higher than unity. Since U abundance of Toya Tephra zircons that have U-Pb ages Ito (2014) obtained U-Th ages of the Toya Tephra by comparison with Fish Canyon Tuff (FCT) data. Because both the FCT and the Toya Tephra have similar trends of overestimation of 230Th, the effect of overestimation of 230Th to cause overestimation of U-Th age should be cancelled out or negligible. Therefore the pivotal conclusion in Ito (2014) that simultaneous U-Pb and U-Th dating using LA-ICP-MS is possible and useful for Quaternary zircons holds true.

  12. Preliminary assessment of a symbiotic fusion--fission power system using the TH/U refresh fuel cycle

    International Nuclear Information System (INIS)

    Bender, D.J.; Lee, J.D.; Moir, R.W.

    1977-10-01

    Studies of the mirror hybrid reactor by LLL/GA have concluded that the most promising role for this reactor concept is that of a producer of fissile fuel for fission reactors. Studies to date have examined primarily the U/Pu fuel cycle with light-water reactors serving as the consumers of the hybrid-bred fissile fuel; the specific scenarios examined required reprocessing and refabrication of the bred fuel before introduction into the fission reactor. This combination of technologies was chosen to illustrate the manner in which the hybrid reactor concept could serve the needs of, and use the technology of, the fission reactor industry as it now exists (and as it was thought it would evolve). However, the current U.S. Administration has expressed strong concerns about proliferation of nuclear weapons capability and terrorist diversion of weapons-grade nuclear materials. These concerns are based on the projected technology for the light-water reactor/fast breeder reactor using the U/Pu fuel cycle and extensive reprocessing/refabrication. A symbiotic nuclear power generation concept (hybrid fissile producer plus fission burner reactors) is described which eliminates those aspects of the present nuclear fuel cycle that (may) represent significant proliferation/diversion risks. Specifically, the proposed concept incorporates the following features: (1)Th/U 233 fuel cycle, (2) no reprocessing or fabrication of fissile material, and (3) no fissile material in a weapons-grade state

  13. Community-based Approaches to Improving Accuracy, Precision, and Reproducibility in U-Pb and U-Th Geochronology

    Science.gov (United States)

    McLean, N. M.; Condon, D. J.; Bowring, S. A.; Schoene, B.; Dutton, A.; Rubin, K. H.

    2015-12-01

    The last two decades have seen a grassroots effort by the international geochronology community to "calibrate Earth history through teamwork and cooperation," both as part of the EARTHTIME initiative and though several daughter projects with similar goals. Its mission originally challenged laboratories "to produce temporal constraints with uncertainties approaching 0.1% of the radioisotopic ages," but EARTHTIME has since exceeded its charge in many ways. Both the U-Pb and Ar-Ar chronometers first considered for high-precision timescale calibration now regularly produce dates at the sub-per mil level thanks to instrumentation, laboratory, and software advances. At the same time new isotope systems, including U-Th dating of carbonates, have developed comparable precision. But the larger, inter-related scientific challenges envisioned at EARTHTIME's inception remain - for instance, precisely calibrating the global geologic timescale, estimating rates of change around major climatic perturbations, and understanding evolutionary rates through time - and increasingly require that data from multiple geochronometers be combined. To solve these problems, the next two decades of uranium-daughter geochronology will require further advances in accuracy, precision, and reproducibility. The U-Th system has much in common with U-Pb, in that both parent and daughter isotopes are solids that can easily be weighed and dissolved in acid, and have well-characterized reference materials certified for isotopic composition and/or purity. For U-Pb, improving lab-to-lab reproducibility has entailed dissolving precisely weighed U and Pb metals of known purity and isotopic composition together to make gravimetric solutions, then using these to calibrate widely distributed tracers composed of artificial U and Pb isotopes. To mimic laboratory measurements, naturally occurring U and Pb isotopes were also mixed in proportions to mimic samples of three different ages, to be run as internal

  14. {sup 230}Th/U-dating of a late Holocene low uranium speleothem from Cuba

    Energy Technology Data Exchange (ETDEWEB)

    Fensterer, Claudia; Mangini, Augusta [Forschungsstelle Radiometrie, Heidelberg Academy of Sciences, Im Neuenheimer Feld 229, 69120 Heidelberg (Germany); Scholz, Denis; Hoffmann, Derik [School of Geographical Sciences, University of Bristol, University Road, BS8 1SS, Bristol (United Kingdom); Pajon, Jesus M, E-mail: Claudia.Fensterer@iup.uni-heidelberg.d [Department of Archaeology, Cuban Institute of Anthropology, Amargura No. 203, e/n Habana y Aguiar, Ciudad de La Habana, CP: 10 100 (Cuba)

    2010-03-15

    We present 22 U-series ages for a stalagmite from north-western Cuba based on multi-collector inductively coupled plasma mass spectrometry (MC-ICPMS) and thermal ionisation mass spectrometry (TIMS). Our results reveal that the stalagmite continuously grew within the last {approx}1400a. Low uranium content of the sample and thus, extremely low {sup 230}Th concentrations limit the precision and accuracy of {sup 230}Th/U-dating by TIMS. Samples measured by MC-ICPMS show a high variability of {sup 232}Th content along the growth axis with some sections significantly affected by initial {sup 230}Th from a detrital phase. An a-priori bulk earth ratio for ({sup 238}U/{sup 232}Th) cannot be used to accurately account for this initial {sup 230}Th. Using an age model based on the {sup 230}Th/U ages determined on samples with low or negligible {sup 232}Th concentration, we find that the ({sup 238}U/{sup 232}Th) activity ratio of the detrital phase is an order of magnitude larger than the bulk earth value, indicating the importance of an accurately determined correction factor.

  15. Evaluation of fission cross sections and covariances for {sup 233}U, {sup 235}U, {sup 238}U, {sup 239}Pu, {sup 240}Pu, and {sup 241}Pu

    Energy Technology Data Exchange (ETDEWEB)

    Kawano, Toshihiko [Kyushu Univ., Fukuoka (Japan); Matsunobu, Hiroyuki [Data Engineering, Inc. (Japan); Murata, Toru [AITEL Corporation, Tokyo (JP)] [and others

    2000-02-01

    A simultaneous evaluation code SOK (Simultaneous evaluation on KALMAN) has been developed, which is a least-squares fitting program to absolute and relative measurements. The SOK code was employed to evaluate the fission cross sections of {sup 233}U, {sup 235}U, {sup 238}U, {sup 239}Pu, {sup 240}Pu, and {sup 241}Pu for the evaluated nuclear data library JENDL-3.3. Procedures of the simultaneous evaluation and the experimental database of the fission cross sections are described. The fission cross sections obtained were compared with evaluated values given in JENDL-3.2 and ENDF/B-VI. (author)

  16. Thermodynamic assessment of the HTGR fuel system Th-U-C-O

    International Nuclear Information System (INIS)

    Ugajin, M.; Shiba, K.

    1978-01-01

    Carbon monoxide pressures and uranium segregation at 2000 K have been calculated for the three-phase equilibria [(ThU)O 2 + (ThU)C 2 + C] in the Th-U-C-O system. This study is concerned with the thermochemical behavior of (Th, U)O 2 particle fuel for the high-temperature gas-cooled reactor (HTGR). The following two points are considered: (1) Reduction of the in-particle CO pressure of (Th, U)O 2 kernels by doping (Th, U)C 2 to make it an oxygen getter. (2) Prediction of U segregation between (Th, U)O 2 and (Th, U)C 2 , doped in the kernel. (Auth.)

  17. Strategy for the future use and disposition of Uranium-233: History, inventories, storage facilities, and potential future uses

    International Nuclear Information System (INIS)

    Bereolos, P.J.; Lewis, L.C.

    1998-06-01

    This document provides background information on the man-made radioisotope 233 U. It is one of a series of four reports that map out potential national strategies for the future use and disposition of 233 U pending action under the National Environmental Policy Act (NEPA). The scope of this report is separated 233 U, where separated refers to nonwaste 233 U or 233 U that has been separated from fission products. Information on other 233 U, such as that in spent nuclear fuel (SNF), is included only to recognize that it may be separated at a later date and then fall under the scope of this report. The background information in this document includes the historical production and current inventory of 233 U, the uses of 233 U, and a discussion of the available facilities for storing 233 U. The considerations for what fraction of the current inventory should be preserved for future use depend on several issues. First, 233 U always contains a small amount of the contaminant isotope 232 U. The decay products of 232 U are highly radioactive and require special handling. The current inventory has a variety of qualities with regard to 232 U content, ranging from 1 to about 200 ppm (on a total uranium basis). It is preferable to use 233 U with the minimum amount of 232 U in all applications. The second issue pertains to other isotopes of uranium mixed in with the 233 U, specifically 235 U and 238 U. A large portion of the inventory has a high quantity of 235 U associated with it. The presence of bulk amounts of 235 U complicates storage because of the added volume needing safeguards and criticality controls. Isotopic dilution using DU may remove safeguards and criticality concerns, but it increases the overall mass and may limit applications that depend on the fissile nature of 233 U. The third issue concerns the packaging of the material. There is no standard packaging (although one is being developed); consequently, the inventory exists in a variety of packages. For some

  18. 238U-234U-230Th-232Th systematics and the precise measurement of time over the past 500,000 years

    International Nuclear Information System (INIS)

    Edwards, R.L.; Chen, J.H.; Wasserburg, G.J.

    1987-01-01

    We have developed techniques to measure the 230 Th abundance in corals by isotope dilution mass spectrometry. This, coupled with our previous development of mass spectrometric techniques for 234 U and 232 Th measurement, has allowed us to reduce significantly the analytical errors in 238 U- 234 U- 230 Th dating and greatly reduce the sample size. We show that 6x10 8 atoms of 230 Th can be measured to ±30per mille (2 σ) and 2x10 10 atoms of 230 Th to ±2per mille. The time over which useful age data on corals can be obtained ranges from a few years to ≅ 500 ky. The uncertainty in age, based on analytical errors, is ±5 y(2 σ) for a 180 year old coral (3 g), ±44 y at 8294 years and ±1.1 ky at 123.1 ky (250 mg of coral). We also report 232 Th concentrations in corals (0.083-1.57 pmol/g) that are more than two orders of magnitude lower than previous values. Ages with high analytical precision were determined for several corals that grew during high sea level stands ≅ 120 ky ago. These ages lie specifically within or slightly postdate the Milankovitch insolation high at 128 ky and support the idea that the dominant cause of Pleistocene climate change is Milankovitch forcing. (orig.)

  19. U-Th isotopic systematics at 13deg N east Pacific Ridge segment

    International Nuclear Information System (INIS)

    Ben Othman, D.; Allegre, C.J.

    1990-01-01

    Fresh basaltic glasses have been analyzed for U-Th disequilibrium systematics as part of an extensive study on the East Pacific Rise (EPR) at 12deg 45'N. These samples are well described in terms of major and trace elements as well as in Nd, Pb and Sr isotopes. Our results show significant heterogeneities in the mantle source at a small scale, and show heterogeneities at larger scales also when compared to other EPR data. U and Th concentration and isotopic data rule out fractional crystallization as a main process and support a mixing model in agreement with the marble cake model developed by Allegre and Turcotte and constrained by trace elements and Nd, Sr and Pb isotopes on the same samples by Prinzhofer et al. Based on the high ( 230 Th/ 232 Th) isotopic ratios on recent tholeiites especially the Th/U values inferred for their sources clearly show that the upper mantle Th/U has decreased with time. (orig.)

  20. Response of gadolinium doped liquid scintillator to charged particles: measurement based on intrinsic U/Th contamination

    Science.gov (United States)

    Du, Q.; Lin, S. T.; He, H. T.; Liu, S. K.; Tang, C. J.; Wang, L.; Wong, H. T.; Xing, H. Y.; Yue, Q.; Zhu, J. J.

    2018-04-01

    A measurement is reported for the response to charged particles of a liquid scintillator named EJ-335 doped with 0.5% gadolinium by weight. This liquid scintillator was used as the detection medium in a neutron detector. The measurement is based on the in-situ α-particles from the intrinsic Uranium and Thorium contamination in the scintillator. The β–α and the α–α cascade decays from the U/Th decay chains were used to select α-particles. The contamination levels of U/Th were consequently measured to be (5.54±0.15)× 10‑11 g/g, (1.45±0.01)× 10‑10 g/g and (1.07±0.01)× 10‑11 g/g for 232Th, 238U and 235U, respectively, assuming secular equilibrium. The stopping power of α-particles in the liquid scintillator was simulated by the TRIM software. Then the Birks constant, kB, of the scintillator for α-particles was determined to be (7.28±0.23) mg/(cm2ṡMeV) by Birks' formulation. The response for protons is also presented assuming the kB constant is the same as for α-particles.

  1. Studies on use of reflector material and its position within FBR core for reducing U{sup 232} content of U produced in ThO{sub 2} radial blankets

    Energy Technology Data Exchange (ETDEWEB)

    Sen, Sujoy, E-mail: sujoy@igcar.gov.in [Core Design Group, IGCAR, Kalpakkam (India); Prasad, Rajeev Ranjan; Bagchi, Subhrojit [Core Design Group, IGCAR, Kalpakkam (India); Mohanakrishnan, P. [MCNS, Manipal University, Manipal (India); Arul, A. John; Puthiyavinayagam, P. [Core Design Group, IGCAR, Kalpakkam (India)

    2015-11-15

    Highlights: • Nuclear data processing for multigroup neutron transport calculation. • Discrete ordinate and Monte Carlo neutron transport. • Breeding of Thorium in Fast Reactor. • Minimization of U{sup 232} in U{sup 233}. • Fuel burn up using Neutron Diffusion. - Abstract: Presence of U{sup 232} in U{sup 233} bred in thorium blanket of fast reactor is a major concern in fuel reprocessing. The former's daughter products being hard gamma emitter and the isotope itself having substantial half life, its presence beyond 10 ppm makes fuel recycle complicated and expensive. In this study possibility of decreasing U{sup 232} production in a typical FBR blanket by means of spectrum modification is examined. SS, depleted B{sub 4}C, SiC, Mo and W regions were introduced between core and radial blanket and evolution of isotopes were studied to arrive at an optimal configuration that satisfies requirements of breeding U{sup 233} and lowering U{sup 232}concentration. SS, B{sub 4}C, SiC, Mo and W are known to be high temperature material with appropriate stability in harsh fast reactor environment. Study has shown that introducing two SS reflector rows can achieve the required low value of U{sup 232}concentration without greatly compromising the U{sup 233}production.

  2. Production of 231Pa and 232U by irradiation of 230Th/232Th mixtures

    International Nuclear Information System (INIS)

    Kluge, E.; Lieser, K.H.

    1981-01-01

    The production of 231 Pa and of 232 U by irradiating a 230 Th/ 232 Th mixture (containing 12 mol per cent 230 Th) in form of ThO 2 at a thermal neutron flux of 6.9 x 10 13 cm -2 s -1 for 4 months was investigated. Pa, U and Th were separated and the chemical yields were determined. 2.6% of the 230 Th were transformed into 231 Pa and 0.13% into 232 U. These values are higher than those calculated for a thermal flux, but lower than those calculated for a flux ratio epithermal to thermal = 0.03. 231 Pa and 232 U were isolated in form of a protactinium solution and of U 3 O 8 with 94.9 and 89.1% chemical yields, respectively. Foreign activities were not detected. Thorium was recuperated and isolated as ThO 2 , with a chemical yield of 93.6%. (orig.)

  3. Importance of coccolithophore-associated organic biopolymers for fractionating particle-reactive radionuclides (234Th, 233Pa, 210Pb, 210Po, and 7Be) in the ocean

    Science.gov (United States)

    Lin, Peng; Xu, Chen; Zhang, Saijin; Sun, Luni; Schwehr, Kathleen A.; Bretherton, Laura; Quigg, Antonietta; Santschi, Peter H.

    2017-08-01

    Laboratory incubation experiments using the coccolithophore Emiliania huxleyi were conducted in the presence of 234Th, 233Pa, 210Pb, 210Po, and 7Be to differentiate radionuclide uptake to the CaCO3 coccosphere from coccolithophore-associated biopolymers. The coccosphere (biogenic calcite exterior and its associated biopolymers), extracellular (nonattached and attached exopolymeric substances), and intracellular (sodium-dodecyl-sulfate extractable and Fe-Mn-associated metabolites) fractions were obtained by sequentially extraction after E. huxleyi reached its stationary growth phase. Radionuclide partitioning and the composition of different organic compound classes, including proteins, total carbohydrates (TCHO), and uronic acids (URA), were assessed. 210Po was closely associated with the more hydrophobic biopolymers (high protein/TCHO ratio, e.g., in attached exopolymeric substances), while 234Th and 233Pa showed similar partitioning behavior with most activity being distributed in URA-enriched, nonattached exopolymeric substances and intracellular biopolymers. 234Th and 233Pa were nearly undetectable in the coccosphere, with a minor abundance of organic components in the associated biopolymers. These findings provide solid evidence that biogenic calcite is not the actual main carrier phase for Th and Pa isotopes in the ocean. In contrast, both 210Pb and 7Be were found to be mostly concentrated in the CaCO3 coccosphere, likely substituting for Ca2+ during coccolith formation. Our results demonstrate that even small cells (E. huxleyi) can play an important role in the scavenging and fractionation of radionuclides. Furthermore, the distinct partitioning behavior of radionuclides in diatoms (previous studies) and coccolithophores (present study) explains the difference in the scavenging of radionuclides between diatom- and coccolithophore-dominated marine environments.

  4. Preparations of high density (Th,U)O2 pellets

    International Nuclear Information System (INIS)

    Akabori, Mitsuo; Ikawa, Katsuichi

    1986-07-01

    Preparations of high density and homogeneous (Th,U)O 2 pellets by a powder metallurgy method were examined. (Th,U)O 2 powders were prepared by calcining coprecipitates of ammonium uranate and thorium hydroxide derived from nitrates and mixed sols, and by calcining mixed oxalates precipitated from nitrates. (Th,U)O 2 pellets were characterized with respect to sinterability, lattice parameter, microstructure, homogeneity and stoichiometry. Sintering atmospheres had a significant effect upon all the properties of the derived pellets. The sinterability of (Th,U)O 2 was most favourable in oxidizing and reducing atmospheres for ThO 2 -rich and UO 2 -rich compositions, respectively, and can be enhanced by presence of water vapour in sintering atmospheres. In addition, highly homogeneous (Th,U)O 2 pellets with 99 % in theoretical density were derived from the sol powders. (author)

  5. On the Relative Signs of "ROT-Effects" in Ternary and Binary Fission of 233U and 235U Nuclei Induced by Polarized Cold Neutrons

    Science.gov (United States)

    Danilyan, G. V.

    2018-02-01

    Signs of the ROT-effects in ternary fission of 233U and 235U experimentally defined by PNPI group are the same, whereas in binary fission defined by ITEP group are opposite. This contradiction cannot be explained by the errors in the experiments of both groups, since such instrumental effects would be too large not to be noticed. Therefore, it is necessary to find the answer to this problem in the differences of the ternary and binary fission mechanisms.

  6. MSR implementation and progress

    Energy Technology Data Exchange (ETDEWEB)

    McKensey, B.; Ellicott, C.; Webb, N. [NSW Department of Mineral Resources, Sydney, NSW (Australia). Coal Mining Inspectorate and Engineering

    1998-12-31

    In November 1996 the NSW Minister for Mineral Resources, commissioned a review of safety within the New South Wales mining industry, as a response to a disturbing number of fatalities during the Minister`s term of office together with a number of alarming near misses. ACiL were given the following terms of reference for the Mine Safety Review (MSR): identify key issues which need to be addressed before a significant and measurable improvement in mine safety performance can be expected; explore options for addressing these key issues, through consultation; consider how the findings of the Warden`s Court Inquiry into the 1994 Moura Mine disaster should be applied in New South Wales; evaluate the role, activities, structure, employment conditions, and resourcing of the State`s Mines Inspectorates; evaluate existing legislative provisions in the light of the identified key issues; and provide Government with recommendations on how mine safety in New South Wales could be enhanced, with particular regard to the Inspectorates. The ACiL report titled `Review of Mine Safety in NSW` was tabled in Parliament by the Ministry in April 1997. The review considered industry safety performance and its measurement through considering available statistics, drawing on submissions and, predominantly seeking the views of a broad cross section of industry and Government personnel. This included mine managers, inspectors, union officials and the workforce. In its report, the MSR makes numerous observations concerning issues of safety and the mining industry. Arising from those observations the review made 44 recommendations for either consideration or action to address the issues identified. 2 refs., 1 fig.

  7. Powder preparation, compaction and sintering of ThO2 and (U,Th)O2 - a review

    International Nuclear Information System (INIS)

    Balakrishna, Palanki; Hemantha Rao, G.V.S.

    2012-01-01

    Sustainable development is the development that meets the needs of the present generation without compromising the needs of future generations. In spite of the disaster at Fukushima, nuclear power remains a clean viable alternative to fossil fuels that has wreaked havoc on the environment by way of global warming, climate change, melting ice caps and rising sea levels. Thorium fuels complement uranium fuels and ensure long term sustainability of nuclear power. Thorium is 3 to 4 times more abundant than uranium and widely distributed in nature as an easily exploitable resource. In recent times, the need for proliferation-resistance, longer fuel cycles, higher burn up, improved waste form characteristics, reduction of plutonium inventories and in situ use of bred-in fissile material has led to renewed interest in thorium-based fuels and fuel cycles in several countries. ThO 2 containing around 4 % 233 UO 2 along with ThO 2 - 3 to 4 % PuO 2 is the proposed fuel for the Indian Advanced Heavy Water Reactor

  8. Complexes of Cd sup 2 sup +, U(O sub 2 ) sup 2 sup + and Th sup 4 sup + at radiotracer levels with phosphoric acid

    International Nuclear Information System (INIS)

    Bouhlassa, S.; El-Yahyaoui, A.; Brillard, L.; Hussonnois, M.; Guillaumont, R.

    1994-01-01

    In this work, we have turn to account the radiochemical techniques in order to investigate the complexation of Cd sup 2 sup +, U(O sub 2) sup 2 sup + and Th sup 2 sup +, at strength Mu=0,2, in various phosphoric media characterized by C sub ( H sub 3 P O sub 4) <= 4 M and 0.7 <= pH <= 4. The method chosen for this purpose is the liquid-liquid extraction of radioisotopes at tracer scale, with di(2-ethyl hexyl) phosphoric acid dissolved in benzene. The radionuclides used are Cd-109, U-233, U-230 and Th-227. Their distribution between the two phases are established by alpha or gamma spectrometric analysis. The analysis of the distribution data allows to define, in addition of species extracted in organic phase, the nature of phosphoric complexes which take place in aqueous media. Stability constants of these complexes and associated thermodynamic data are determined. 2 tabs.; 2 refs. (author)

  9. 230Th-238U disequilibria in historical lavas from Iceland

    International Nuclear Information System (INIS)

    Condomines, M.; Morand, P.; Alleegre, C.J.; Sigvaldason, G.

    1981-01-01

    The 230 Th- 238 U disequilibrium studies on historical lavas from Iceland show a relative homogeneity for Th/U ratios and also a variation for ( 230 Th/ 232 Th) activity ratios at the scale of the island. The ( 230 Th/ 238 U) disequilibrium ratio is always greater than 1 which indicates that partial melting produces magmas with Th/U ratios greater than those of the mantle source. Furthermore, there seems to be a correlation between the variations of ( 230 Th/ 232 Th) (and delta 18 O) ratios and the geographical location of the samples along the active zones of Iceland. We develop and discuss several models in order to explain these variations. (orig.)

  10. Advantages and implications of U233 fueled thermionic space power energy conversion

    International Nuclear Information System (INIS)

    Terrell, C.W.

    1992-01-01

    In this paper two recent analyses are reported which demonstrate advantages of a U233 fueled thermionic fuel element (TFE) compared to 93 w/o U235, and that application (mission) has broad latitude in how space power reactor systems could or should be optimized. A reference thermionic reactor system was selected to provide the basis for the fuel comparisons. Both oxide and metal fuel forms were compared. Of special interest was to estimate the efficiencies of the four fuel forms to produce electrical power. A figure of merit (FOM) was defined which is directly proportional to the electrical average electrical power produced is proportional to the electrical power produced per unit uranium mass. In a TFE the average electrical power produced is proportional to the emitter surface area (Esa), hence the ratio Esa/Mu was selected as the FOM. Results indicate that the choice of fuel type and form leads to wide variations in critical and system masses FOM values, and system total power

  11. Study of the variation with the energy of the fission cross-sections of 233U, 235U, 239Pu for the fast neutrons

    International Nuclear Information System (INIS)

    Szteinsznaider, D.; Naggiar, V.; Netter, F.

    1955-01-01

    This measurements have been done while taking the value of the fission cross-sections of 238 U as reference. The neutrons are produced by the reaction 7 Li(p,n) in the Van de Graaff generator of Saclay. The explored domain spreads from some tenths to 2000 keV. We find: for 239 Pu: σ f = 2,04 ± 0,12 barns, cross-section constant between 150 and 2000 keV, for 235 U: σ f = 1,15 ± 0,15 barns, cross-section constant between 700 and 1000 keV, for 233 U: σ f = 1,92 ± 0,25 barns, for neutrons of 850 keV. (authors) [fr

  12. History of Uranium-233(233U)Processing at the Rocky Flats Plant. In support of the RFETS Acceptable Knowledge Program

    International Nuclear Information System (INIS)

    Moment, R.L.; Gibbs, F.E.; Freiboth, C.J.

    1999-01-01

    This report documents the processing of Uranium-233 at the Rocky Flats Plant (Rocky Flats Environmental Technology Site). The information may be used to meet Waste Isolation Pilot Plant (WIPP) Waste Acceptance Criteria (WAC)and for determining potential Uranium-233 content in applicable residue waste streams

  13. 234U/230Th ratio as an indicator of redox state, and U, Th and Ra behavior in briney aquifers

    International Nuclear Information System (INIS)

    Laul, J.C.; Smith, M.R.; Hubbard, N.

    1985-06-01

    The 234 U/ 230 Th ratio serves as an in-situ indicator of the redox state in groundwater aquifers. The higher this ratio, the more U there is in the +6 state and thus a lesser reducing environment. Radium is retarded in the shallow aquifer and its sorption is dependent on the CaSO 4 content and redox state. Relative to Ra, U and Th are highly sorbed. The total retardation factor for Th is approx.1400 and mean sorption time for 228 Th is approx.10 days in the shallow zone. The desorption rate of Ra is significantly slower in the shallow than in the deep aquifer. There is no effect of colloids in brines. 6 refs., 5 figs., 2 tabs

  14. U enrichment and Th/U fractionation in Archean boninites: Implications for paleo-ocean oxygenation and U cycling at juvenile subduction zones

    Science.gov (United States)

    Manikyamba, C.; Said, Nuru; Santosh, M.; Saha, Abhishek; Ganguly, Sohini; Subramanyam, K. S. V.

    2018-05-01

    Phanerozoic boninites record enrichments of U over Th, giving Th/U: 0.5-1.6, relative to intraoceanic island arc tholeiites (IAT) where Th/U averages 2.6. Uranium enrichment is attributed to incorporation of shallow, oxidized fluids, U-rich but Th-poor, from the slab into the melt column of boninites which form in near-trench to forearc settings of suprasubduction zone ophiolites. Well preserved Archean komatiite-tholeiite, plume-derived, oceanic volcanic sequences have primary magmatic Th/U ratios of 4.4-3.6, and Archean convergent margin IAT volcanic sequences, having REE and HFSE compositions similar to Phanerozoic IAT equivalents, preserve primary Th/U of 4-3.6. The best preserved Archean boninites of the 3.0 Ga Olondo and 2.7 Ga Gadwal greenstone belts, hosted in convergent margin ophiolite sequences, also show relative enrichments of U over Th, with low average Th/U ∼3 relative to coeval IAT, and Phanerozoic counterparts which are devoid of crustal contamination and therefore erupted in an intraoceanic setting, with minimal contemporaneous submarine hydrothermal alteration. Later enrichment of U is unlikely as Th-U-Nb-LREE patterns are coherent in these boninites whereas secondary effects induce dispersion of Th/U ratios. The variation in Th/U ratios from Archean to Phanerozoic boninites of greenstone belts to ophiolitic sequences reflect on genesis of boninitic lavas at different tectono-thermal regimes. Consequently, if the explanation for U enrichment in Phanerozoic boninites also applies to Archean examples, the implication is that U was soluble in oxygenated Archean marine water up to 600 Ma before the proposed great oxygenation event (GOE) at ∼2.4 Ga. This interpretation is consistent with large Ce anomalies in some hydrothermally altered Archean volcanic sequences aged 3.0-2.7 Ga.

  15. Evaluation of the root cause for MSR high level trip in Maanshan

    International Nuclear Information System (INIS)

    Liao, L.-Y.; Ferng, Y.-M.; Jange, S.J.; Ko, C.M.

    2004-01-01

    Reactor trip due to Moisture Separator Reheater (MSR) high water level has been a long time issue for Maanshan nuclear power plant. The operating experience shows that there are five reactor trips due to MSR high water level. Four out of the five reactor trips are generated when Combined Intermediate valve (CIV) no. 1 is closed during CIV closure test. The fifth reactor trip occurs when the reactor power is increasing from 99% to 100%. An extensive root cause analysis has been performed by Taipower Company. It is concluded that the water accumulated in the cross under leg between the exhaust of high pressure turbine and the inlet of MSR was the water source contributing to the MSR high level trip. Although, Maanshan does not have similar trip after the root cause analysis, it is interested to evaluate the proposed root cause from thermal hydraulic point of view. It is also hoped that some useful guidelines can be established. This paper includes a description of the scenario of reactor trips, a summary of the root cause analysis done by Taipower Company, an examination of possible mechanisms, an identification of key parameters and a presentation of major findings. In addition, the applicability of RELAP5/MOD3 under this condition is discussed. (author)

  16. Fission cross section ratios for sup 233,234,236 U relative to sup 235 U from 0. 5 to 400 MeV

    Energy Technology Data Exchange (ETDEWEB)

    Lisowski, P.W.; Gavron, A.; Parker, W.E.; Balestrini, S.J. (Los Alamos National Lab., NM (USA)); Carlson, A.D.; Wasson, O.A. (National Inst. of Standards and Technology, Gaithersburg, MD (USA)); Hill, N.W. (Oak Ridge National Lab., TN (USA))

    1991-01-01

    Neutron-induced fission cross section ratios from 0.5 to 400 MeV for samples of {sup 233, 234, 236}U relative to {sup 235}U have been measured at the WNR neutron Source at Los Alamos. The fission reaction rate was determined using a fast parallel plate ionization chamber at a 20-m flight path. Cross sections over most the energy range were also extracted using the neutron fluence determined with three different proton telescope arrangements. Those data provided the shape of the {sup 235}U(n,f) cross section relative to the hydrogen scattering cross section. That shape was then normalized to the very accurately known value for {sup 235}U(n,f) at 14.1 MeV to allow us to obtain cross section section values from the ratio data and our values for {sup 235}U(n,f). 6 refs., 1 fig.

  17. MsrA Overexpression Targeted to the Mitochondria, but Not Cytosol, Preserves Insulin Sensitivity in Diet-Induced Obese Mice.

    Directory of Open Access Journals (Sweden)

    JennaLynn Hunnicut

    Full Text Available There is growing evidence that oxidative stress plays an integral role in the processes by which obesity causes type 2 diabetes. We previously identified that mice lacking the protein oxidation repair enzyme methionine sulfoxide reductase A (MsrA are particularly prone to obesity-induced insulin resistance suggesting an unrecognized role for this protein in metabolic regulation. The goals of this study were to test whether increasing the expression of MsrA in mice can protect against obesity-induced metabolic dysfunction and to elucidate the potential underlying mechanisms. Mice with increased levels of MsrA in the mitochondria (TgMito MsrA or in the cytosol (TgCyto MsrA were fed a high fat/high sugar diet and parameters of glucose homeostasis were monitored. Mitochondrial content, markers of mitochondrial proteostasis and mitochondrial energy utilization were assessed. TgMito MsrA, but not TgCyto MsrA, mice remain insulin sensitive after high fat feeding, though these mice are not protected from obesity. This metabolically healthy obese phenotype of TgMito MsrA mice is not associated with changes in mitochondrial number or biogenesis or with a reduction of proteostatic stress in the mitochondria. However, our data suggest that increased mitochondrial MsrA can alter metabolic homeostasis under diet-induced obesity by activating AMPK signaling, thereby defining a potential mechanism by which this genetic alteration can prevent insulin resistance without affecting obesity. Our data suggest that identification of targets that maintain and regulate the integrity of the mitochondrial proteome, particular against oxidative damage, may play essential roles in the protection against metabolic disease.

  18. Characterization of the Metabolically Modified Heavy Metal-Resistant Cupriavidus metallidurans Strain MSR33 Generated for Mercury Bioremediation

    Science.gov (United States)

    Rojas, Luis A.; Yáñez, Carolina; González, Myriam; Lobos, Soledad; Smalla, Kornelia; Seeger, Michael

    2011-01-01

    Background Mercury-polluted environments are often contaminated with other heavy metals. Therefore, bacteria with resistance to several heavy metals may be useful for bioremediation. Cupriavidus metallidurans CH34 is a model heavy metal-resistant bacterium, but possesses a low resistance to mercury compounds. Methodology/Principal Findings To improve inorganic and organic mercury resistance of strain CH34, the IncP-1β plasmid pTP6 that provides novel merB, merG genes and additional other mer genes was introduced into the bacterium by biparental mating. The transconjugant Cupriavidus metallidurans strain MSR33 was genetically and biochemically characterized. Strain MSR33 maintained stably the plasmid pTP6 over 70 generations under non-selective conditions. The organomercurial lyase protein MerB and the mercuric reductase MerA of strain MSR33 were synthesized in presence of Hg2+. The minimum inhibitory concentrations (mM) for strain MSR33 were: Hg2+, 0.12 and CH3Hg+, 0.08. The addition of Hg2+ (0.04 mM) at exponential phase had not an effect on the growth rate of strain MSR33. In contrast, after Hg2+ addition at exponential phase the parental strain CH34 showed an immediate cessation of cell growth. During exposure to Hg2+ no effects in the morphology of MSR33 cells were observed, whereas CH34 cells exposed to Hg2+ showed a fuzzy outer membrane. Bioremediation with strain MSR33 of two mercury-contaminated aqueous solutions was evaluated. Hg2+ (0.10 and 0.15 mM) was completely volatilized by strain MSR33 from the polluted waters in presence of thioglycolate (5 mM) after 2 h. Conclusions/Significance A broad-spectrum mercury-resistant strain MSR33 was generated by incorporation of plasmid pTP6 that was directly isolated from the environment into C. metallidurans CH34. Strain MSR33 is capable to remove mercury from polluted waters. This is the first study to use an IncP-1β plasmid directly isolated from the environment, to generate a novel and stable bacterial strain

  19. Characterization of the metabolically modified heavy metal-resistant Cupriavidus metallidurans strain MSR33 generated for mercury bioremediation.

    Directory of Open Access Journals (Sweden)

    Luis A Rojas

    Full Text Available BACKGROUND: Mercury-polluted environments are often contaminated with other heavy metals. Therefore, bacteria with resistance to several heavy metals may be useful for bioremediation. Cupriavidus metallidurans CH34 is a model heavy metal-resistant bacterium, but possesses a low resistance to mercury compounds. METHODOLOGY/PRINCIPAL FINDINGS: To improve inorganic and organic mercury resistance of strain CH34, the IncP-1β plasmid pTP6 that provides novel merB, merG genes and additional other mer genes was introduced into the bacterium by biparental mating. The transconjugant Cupriavidus metallidurans strain MSR33 was genetically and biochemically characterized. Strain MSR33 maintained stably the plasmid pTP6 over 70 generations under non-selective conditions. The organomercurial lyase protein MerB and the mercuric reductase MerA of strain MSR33 were synthesized in presence of Hg(2+. The minimum inhibitory concentrations (mM for strain MSR33 were: Hg(2+, 0.12 and CH(3Hg(+, 0.08. The addition of Hg(2+ (0.04 mM at exponential phase had not an effect on the growth rate of strain MSR33. In contrast, after Hg(2+ addition at exponential phase the parental strain CH34 showed an immediate cessation of cell growth. During exposure to Hg(2+ no effects in the morphology of MSR33 cells were observed, whereas CH34 cells exposed to Hg(2+ showed a fuzzy outer membrane. Bioremediation with strain MSR33 of two mercury-contaminated aqueous solutions was evaluated. Hg(2+ (0.10 and 0.15 mM was completely volatilized by strain MSR33 from the polluted waters in presence of thioglycolate (5 mM after 2 h. CONCLUSIONS/SIGNIFICANCE: A broad-spectrum mercury-resistant strain MSR33 was generated by incorporation of plasmid pTP6 that was directly isolated from the environment into C. metallidurans CH34. Strain MSR33 is capable to remove mercury from polluted waters. This is the first study to use an IncP-1β plasmid directly isolated from the environment, to generate a novel

  20. Implantation of alpha spectrometry methodology for the determination of U and Th isotopes in igneous rocks: application to the study of radioactive desequilibrium in the Trindade Island, Brazil

    International Nuclear Information System (INIS)

    Santos, Rosana Nunes dos

    2001-01-01

    This work describes the implementation of experimental procedures for alpha spectrometry measurement of 238 U, 234 U and 230 Th activities in silicates. The best experimental conditions were defined using 233 U, 232 U and 229 Th radioactive tracers and simulating the usual conditions found in processing silicates. The chemical procedures consists of the following steps: radioactive tracer addition and sample dissolution by acid digestion, U and Th pre-concentration by co-precipitation, element separation and purification by ion exchange chromatography and electrodeposition in inox steel disks. In order to evaluate its effectiveness, the procedure was applied to the Brazilian geological standards BB-1 (basalt) and GB-1 (granite). The obtained chemical yields for uranium and thorium are of about 60% and 70%, respectively, for both matrices. The described methodology furnishes activity measurements with less than 4% relative precision and accuracies of about 1%, that are essential for petrogenetic applications. The 238 U and 232 Th series disequilibrium conditions were investigated by alpha spectrometry, together with neutron activation analysis and natural gamma-ray spectrometry. 234 U/ 238 U, 238 U/ 232 Th and 230 Th/ 232 Th activity ratios and the 234 Th, 214 Pb, 214 Bi, 235 U, 228 Ac, 212 Pb, 212 Bi and 208 Tl specific activities were obtained. These results were interpreted with the help of additional constraints given by the larger and smaller elements concentrations, measured by X-ray fluorescence. The 232 Th series is in secular radioactive equilibrium in all analysed samples. In the case of the 238 U series, the equilibrium condition was verified, as expected, in the oldest rocks from the Trindade Island (Trindade Complex and Desejado Sequence). On the other hand, the results show that, in the samples from the last three volcanic episodes in the island (Morro Vermelho Formation, Valado Formation and Vulcao do Paredao), the 230 Th and 238 U are not in

  1. Shippingport LWBR (Th/U Oxide) Fuel Characteristics for Disposal Criticality Analysis

    International Nuclear Information System (INIS)

    Taylor, L. L.; Loo, H. H.

    1999-01-01

    Department of Energy (DOE)-owned spent nuclear fuels encompass many fuel types. In an effort to facilitate criticality analysis for these various fuel types, they were categorized into eight characteristic fuel groups with emphasis on fuel matrix composition. Out of each fuel group, a representative fuel type was chosen for analysis as a bounding case within that fuel group. Generally, burnup data, fissile enrichments, and total fuel and fissile mass govern the selection of the representative or candidate fuel within that group. The Shippingport Light Water Breeder Reactor (LWBR) fuels incorporate more of the conventional materials (zirconium cladding/heavy metal oxides) and fabrication details (rods and spacers) that make them comparable to a typical commercial fuel assembly. The LWBR seed/blanket configuration tested a light-water breeder concept with Th-232/U-233 binary fuel matrix. Reactor design used several assembly configurations at different locations within the same core . The seed assemblies contain the greatest fissile mass per (displaced) unit volume, but the blanket assemblies actually contain more fissile mass in a larger volume; the atom-densities are comparable

  2. Simulation of Molten Salt Reactor dynamics

    International Nuclear Information System (INIS)

    Krepel, J.; Rohde, U.; Grundmann, U.

    2005-01-01

    Dynamics of the Molten Salt Reactor - one of the 'Generation IV' concepts - was studied in this paper. The graphite-moderated channel type MSR was selected for the numerical simulation of the reactor with liquid fuel. The MSR dynamics is very specific because of two physical peculiarities of the liquid fueled reactor: the delayed neutrons precursors are drifted by the fuel flow and the fission energy is immediately released directly into the coolant. Presently, there are not many accessible numerical codes appropriate for the MSR simulation, therefore the DYN3D-MSR code was developed based on the FZR in-house code DYN3D. It allows calculating of full 3D transient neutronics in combination with parallel channel type thermal-hydraulics. By means of DYN3D-MSR, several transients typical for the liquid fuel system were analyzed. Those transients were initiated by reactivity insertion, by overcooling of fuel at the core inlet, by the fuel pump start-up or coast-down, or by the blockage of selected fuel channels. In these considered transients, the response of the MSR is characterized by the immediate change of the fuel temperature with changing power and fast negative temperature feedback to the power. The response through the graphite temperature is slower. Furthermore, for big MSR cores fueled with U233 the graphite feedback coefficient can be positive. In this case the addition of erbium to the graphite can ensure the inherent safety features. The DYN3D-MSR code has been shown to be an effective tool for MSR dynamics studies. (author)

  3. U-Th-Ra disequilibria at the Masaya (Nicaragua); Desequilibres U-Th-Ra au Masaya (Nicaragua)

    Energy Technology Data Exchange (ETDEWEB)

    Sigmarsson, O; Condomines, M [Centre de Recherches Volcanologiques, CNRS URA-10, 63 - Clermont Ferrand (France)

    1997-12-31

    {sup 238}U-{sup 230}Th-{sup 226}Ra radioactive disequilibria were measured in several basalt samples of the post-caldera flows of the Masaya volcano (Nicaragua). {sup 230}Th/{sup 232}Th ratios are from the highest known in the world (about 2.53) with {sup 230}Th/{sup 238}U ratios close to 1. These exceptionally high isotopic thorium ratios from the Masaya and other neighboring volcanoes (Conception, Cerro Negro, Momotombo) are followed by very high {sup 10}Be/{sup 9}Be ratios (60 10{sup -11} for the 1722 flow). These geochemical characteristics with {delta}{sup 18}O of typical mantle origin (5.55) suggest an influence of the subducted sediments fluids in the magma source. The age of the metasomatism ranges from 10 to 0.3 Ma. Initial {sup 226}Ra/{sup 230}Th ratios measured in four historical flows vary from 1.3 to 1.4 and are anti-correlated with the Th content. These variations are probably linked to the fractionated crystallisation of plagioclase minerals. The initial {sup 226}Ra/Ba ratio remains constant and suggests the existence of a huge stationary magmatic reservoir. This hypothesis is also confirmed by the disproportion between the SO{sub 2} quantity emitted by the volcano and by the degassing of lavas on the ground. The {sup 226}Ra excess observed in the Masaya lavas can be the result of a second stage of metasomatism which occurred less than 8000 years B.P. during partial fusion. Abstract only. (J.S.).

  4. U-Th-Ra disequilibria at the Masaya (Nicaragua); Desequilibres U-Th-Ra au Masaya (Nicaragua)

    Energy Technology Data Exchange (ETDEWEB)

    Sigmarsson, O.; Condomines, M. [Centre de Recherches Volcanologiques, CNRS URA-10, 63 - Clermont Ferrand (France)

    1996-12-31

    {sup 238}U-{sup 230}Th-{sup 226}Ra radioactive disequilibria were measured in several basalt samples of the post-caldera flows of the Masaya volcano (Nicaragua). {sup 230}Th/{sup 232}Th ratios are from the highest known in the world (about 2.53) with {sup 230}Th/{sup 238}U ratios close to 1. These exceptionally high isotopic thorium ratios from the Masaya and other neighboring volcanoes (Conception, Cerro Negro, Momotombo) are followed by very high {sup 10}Be/{sup 9}Be ratios (60 10{sup -11} for the 1722 flow). These geochemical characteristics with {delta}{sup 18}O of typical mantle origin (5.55) suggest an influence of the subducted sediments fluids in the magma source. The age of the metasomatism ranges from 10 to 0.3 Ma. Initial {sup 226}Ra/{sup 230}Th ratios measured in four historical flows vary from 1.3 to 1.4 and are anti-correlated with the Th content. These variations are probably linked to the fractionated crystallisation of plagioclase minerals. The initial {sup 226}Ra/Ba ratio remains constant and suggests the existence of a huge stationary magmatic reservoir. This hypothesis is also confirmed by the disproportion between the SO{sub 2} quantity emitted by the volcano and by the degassing of lavas on the ground. The {sup 226}Ra excess observed in the Masaya lavas can be the result of a second stage of metasomatism which occurred less than 8000 years B.P. during partial fusion. Abstract only. (J.S.).

  5. Study of the mass, isotopic and kinetic energy distributions of the 233U(nth, f) and 241Pu(nth, f) fission products measured at the Lohengrin mass spectrometer (ILL)

    International Nuclear Information System (INIS)

    Martin, F.

    2013-01-01

    Fission product yields are significant nuclear data for neutronic simulations. The purpose of this work is to improve fission yield knowledge for two fissile nuclei: 241 Pu and 233 U. Those are respectively involved in the uranium and thorium nuclear fuel cycle. The measurements are performed at the Lohengrin mass spectrometer of the Institut Laue-Langevin (ILL) located in Grenoble. The spectrometer is combined with an ionization chamber to measure mass yields of 241 Pu and 233 U and with a gamma spectrometry set-up to determine isotopic yields of 233 U. A new analysis method of experimental data has been developed in order to control systematics and to reduce experimental biases. For the first time, the experimental variance-covariance matrix of our measured fission yields could be deduced. (author) [fr

  6. Age constraints for Palaeolithic cave art by U-Th dating of thin carbonate crusts

    Science.gov (United States)

    Hoffmann, Dirk; Pike, Alistair; Garcia-Diez, Marcos; Pettitt, Paul; Zilhão, João

    2015-04-01

    U-series dating is an important geochronological tool which is widely applied for instance in speleothem based palaeoclimate research. It has also great potential to provide age constraints for Archaeology, especially for sites or artefacts in cave environments. We present our methods to conduct precise U-Th dating of calcite crusts that formed on top of cave paintings. Recent developments in multi-collector (MC) inductively coupled plasma mass spectrometry (ICPMS) U-series dating greatly improved the precision of this method, and sample sizes needed to obtain reliable results were significantly reduced. Based on these developments the U-series technique can be applied for accurate dating of thin calcite crusts covering cave art at many sites, while taking care not to harm the art underneath. The method provides minimum ages for the covered art and, where possible, also maximum ages by dating the flowstone layer the art is painted on. The U-Th method has been used in a number of recent projects to date calcite precipitates above and occasionally below cave paintings in Spain. Initial results from Cantabria have shown that the earliest dated paintings are older than 41.4 ± 0.6 ka, dating at least to the Early Aurignacian period and present a far longer chronology than that based so far on radiocarbon dating. Here we outline our methodology and the steps we take to demonstrate the reliability of U-Th dates, and present some recent results of our ongoing U-Th dating programme.

  7. Dissolution behaviour of 238U, 234U and 230Th deposited on filters from personal dosemeters

    International Nuclear Information System (INIS)

    Beckova, V.; Malatova, I.

    2008-01-01

    Kinetics of dissolution of 238 U, 234 U and 230 Th dust deposited on filters from personal alpha dosemeters was studied by means of a 26-d in vitro dissolution test with a serum ultra-filtrate simulant. Dosemeters had been used by miners at the uranium mine 'Dolni Rozinka' at Rozna, Czech Republic. The sampling flow-rate as declared by the producer is 4 l h -1 and the sampling period is typically 1 month. Studied filters contained 125 ± 6 mBq 238 U in equilibrium with 234 U and 230 Th; no 232 Th series nuclides were found. Half-time of rapid dissolution of 1.4 d for 238 U and 234 U and slow dissolution half-times of 173 and 116 d were found for 238 U and 234 U, respectively. No detectable dissolution of 230 Th was found. (authors)

  8. Evaluation of neutron nuclear data for 233U in thermal and resonance regions

    International Nuclear Information System (INIS)

    Kikuchi, Yasuyuki

    1981-02-01

    The thermal and resonance cross sections of 233 U were evaluated for JENDL-2. The cross sections below 1 eV are given as point-wise data and were evaluated by the use of the measured fission and capture cross sections. The resolved resonance parameters are derived up to 100 eV. The parameters were obtained by using NDES so as to reproduce the measured total and fission cross sections. The cross sections from 100 eV to 30 keV are represented by the unresolved resonance parameters. The fission and capture resonance integrals calculated from these parameters are 771 and 138 barns, respectively, which agree with the measured data within the quoted errors. (author)

  9. Elevated uptake of Th and U by netted chain fern (Woodwardia areolata)

    International Nuclear Information System (INIS)

    Knox, A.S.; Kaplan, D.I.; Hinton, T.G.

    2008-01-01

    We assessed the ability of netted chain fern (Woodwardia areolata) to uptake U and Th from wetland soils on the U.S. Department of Energy's Savannah River Site in South Carolina. Netted chain fern had the highest Th and U concentrations of all plants collected from the wetland. Ferns grown in contaminated soil (329 mg x kg -1 Th, 44 mg x kg -1 U) in a greenhouse contained 6.4 mg x kg -1 Th and 5.3 mg x kg -1 U compared with 0.13 mg x kg -1 Th and 0.035 mg x kg -1 U in Bermuda grass (Cynodon dactylon). Netted chain fern has potential for the phytoremediation of soils contaminated with Th and U. (author)

  10. Dissolution behaviour of 238U, 234U and 230Th deposited on filters from personal dosemeters.

    Science.gov (United States)

    Becková, Vera; Malátová, Irena

    2008-01-01

    Kinetics of dissolution of (238)U, (234)U and (230)Th dust deposited on filters from personal alpha dosemeters was studied by means of a 26-d in vitro dissolution test with a serum ultrafiltrate simulant. Dosemeters had been used by miners at the uranium mine 'Dolní Rozínka' at Rozná, Czech Republic. The sampling flow-rate as declared by the producer is 4 l h(-1) and the sampling period is typically 1 month. Studied filters contained 125 +/- 6 mBq (238)U in equilibrium with (234)U and (230)Th; no (232)Th series nuclides were found. Half-time of rapid dissolution of 1.4 d for (238)U and (234)U and slow dissolution half-times of 173 and 116 d were found for (238)U and (234)U, respectively. No detectable dissolution of (230)Th was found.

  11. Laser Ablation in situ (U-Th-Sm)/He and U-Pb Double-Dating of Apatite and Zircon: Techniques and Applications

    Science.gov (United States)

    McInnes, B.; Danišík, M.; Evans, N.; McDonald, B.; Becker, T.; Vermeesch, P.

    2015-12-01

    We present a new laser-based technique for rapid, quantitative and automated in situ microanalysis of U, Th, Sm, Pb and He for applications in geochronology, thermochronometry and geochemistry (Evans et al., 2015). This novel capability permits a detailed interrogation of the time-temperature history of rocks containing apatite, zircon and other accessory phases by providing both (U-Th-Sm)/He and U-Pb ages (+trace element analysis) on single crystals. In situ laser microanalysis offers several advantages over conventional bulk crystal methods in terms of safety, cost, productivity and spatial resolution. We developed and integrated a suite of analytical instruments including a 193 nm ArF excimer laser system (RESOlution M-50A-LR), a quadrupole ICP-MS (Agilent 7700s), an Alphachron helium mass spectrometry system and swappable flow-through and ultra-high vacuum analytical chambers. The analytical protocols include the following steps: mounting/polishing in PFA Teflon using methods similar to those adopted for fission track etching; laser He extraction and analysis using a 2 s ablation at 5 Hz and 2-3 J/cm2fluence; He pit volume measurement using atomic force microscopy, and U-Th-Sm-Pb (plus optional trace element) analysis using traditional laser ablation methods. The major analytical challenges for apatite include the low U, Th and He contents relative to zircon and the elevated common Pb content. On the other hand, apatite typically has less extreme and less complex zoning of parent isotopes (primarily U and Th). A freeware application has been developed for determining (U-Th-Sm)/He ages from the raw analytical data and Iolite software was used for U-Pb age and trace element determination. In situ double-dating has successfully replicated conventional U-Pb and (U-Th)/He age variations in xenocrystic zircon from the diamondiferous Ellendale lamproite pipe, Western Australia and increased zircon analytical throughput by a factor of 50 over conventional methods

  12. 19 CFR 10.233 - Articles eligible for preferential tariff treatment.

    Science.gov (United States)

    2010-04-01

    ... control of the customs authority of the intermediate country; (ii) Did not enter into the commerce of the... 19 Customs Duties 1 2010-04-01 2010-04-01 false Articles eligible for preferential tariff treatment. 10.233 Section 10.233 Customs Duties U.S. CUSTOMS AND BORDER PROTECTION, DEPARTMENT OF HOMELAND...

  13. Study of electrodeposition technique to prepare alpha-counting plates of uranium 233

    International Nuclear Information System (INIS)

    Mertzig, W.

    1979-01-01

    The electrodeposition technique to prepare alpha-counting plates of 233 U for its determination is presented. To determine the optimum conditions for plating 233 U the effects of such parameters as current density, pH of eletrotype, salt concentration, time of electrolysis and distance electrodes were studied. A carrier method was developed to attain a quantitative electrodeposition of 233 U from aqueous solutions into alpha counting paltes. A single and incremental addition of natural uranium and thorium as carrier were studied. All samples were prepared using a electrodeposition cell manufactured at the IPEN, especially for use in electroplating tracer actinides. This cell is made of a metal-lucite to contain the electrolyte, which bottom is a polished brass disk coated with a Ni film serving as the cathode. A Pt wire anode is fixed on the top of the cell. The electroplated samples were alpha-counted using a surface barrier detector. A recovery of more than 99% was obtained in specific conditions. The plating procedure produced deposits which were firmly distributed over the plate area. The method was applied to determine tracer amounts of 233 U from oxalate and nitrate solutions coming from chemical processing irradiated thorium. (Author) [pt

  14. MSR redesign and reconstruction at Indiana Michigan Power Company's Donald C. Cook Nuclear Power Plant, Unit 1

    International Nuclear Information System (INIS)

    Yarden, A.L.; Tam, C.W.; Benes, J.D.; Arnold, W.E.

    1993-01-01

    When Indiana Michigan Power Company's (I and M) 1089- MWe, PWR, Donald C. Cook Nuclear Plant, Unit 1, (Cook 1) in Bridgeman, Michigan went into commercial operation in late 1975, its turbine generator included two Moisture Separator Reheater (MSR) vessels. Each of these original MSRs contained, in addition to the moisture separation section, a single stage 2-pass reheater consisting of 5/8 inch O.D., finned CuNi tubes with main heating steam as an energy source. The enormous size of the tube bank, with a vertical orientation of its tubes' U-bends, led the designer to choose two separate headers for the inlet side and outlet side of the tube bank. Over the years, these 2-pass reheaters had deteriorated mechanically such that maintenance costs had increased considerably. Also, the MSR performance in terms of MWe gain, had fallen off as a result of a gradual reduction of both superheat and moisture separation efficiency. In 1990, these MSRs were totally reconstructed with inherently different 4-pass reheaters and upgraded moisture separation systems. The performance and other direct parameters of these newly retrofitted and improved MSRs have exceeded original design specifications, and their operational stability has improved markedly. This MSR reconstruction at Cook 1 is the first of its kind to include a 4-pass reheater in association with a nuclear turbine generator of this design. This paper highlights the problems and solutions associated respectively with the original reheaters in the Cook 1 MSRs and their recent redesign, reconstruction, and performance

  15. Preparation of graphene oxide-manganese dioxide for highly efficient adsorption and separation of Th(IV)/U(VI).

    Science.gov (United States)

    Pan, Ning; Li, Long; Ding, Jie; Li, Shengke; Wang, Ruibing; Jin, Yongdong; Wang, Xiangke; Xia, Chuanqin

    2016-05-15

    Manganese dioxide decorated graphene oxide (GOM) was prepared via fixation of crystallographic MnO2 (α, γ) on the surface of graphene oxide (GO) and was explored as an adsorbent material for simultaneous removal of thorium/uranium ions from aqueous solutions. In single component systems (Th(IV) or U(VI)), the α-GOM2 (the weight ratio of GO/α-MnO2 of 2) exhibited higher maximum adsorption capacities toward both Th(IV) (497.5mg/g) and U(VI) (185.2 mg/g) than those of GO. In the binary component system (Th(IV)/U(VI)), the saturated adsorption capacity of Th(IV) (408.8 mg/g)/U(VI) (66.8 mg/g) on α-GOM2 was also higher than those on GO. Based on the analysis of various data, it was proposed that the adsorption process may involve four types of molecular interactions including coordination, electrostatic interaction, cation-pi interaction, and Lewis acid-base interaction between Th(IV)/U(VI) and α-GOM2. Finally, the Th(IV)/U(VI) ions on α-GOM2 can be separated by a two-stage desorption process with Na2CO3/EDTA. Those results displayed that the α-GOM2 may be utilized as an potential adsorbent for removing and separating Th(IV)/U(VI) ions from aqueous solutions. Copyright © 2016 Elsevier B.V. All rights reserved.

  16. Evolution of Th and U whole-rock contents in the Ilimaussaq intrusion

    International Nuclear Information System (INIS)

    Bailey, J.C.; Rose-Hansen, J.; Loevborg, L.; Soerensen, H.

    1981-01-01

    A great variety of investigations have been made on the distribution of Th and U in the Ilimaussaq alkaline, South Greenland. The major emphasis has been placed on economic assessment of the Kvanefjeld uranium deposit but attention has also been given to the Th and U contents of rocks and minerals outside the deposit. In the present study, we present Th and U values largely obtained by laboratory gamma-ray spectrometric (GRS) analysis of a large collection of representative samples taken from all rock types of the intrusion. The results are discussed in relation to current knowledge and ideas on the petrologic evolution of the Ilimaussaq intrusion. The behaviour of Th and U in igneous systems is moderately well known. During closed-system fractional crystallization, Th and U are generally excluded from the cumulus phases and attain higher levels in successive residual magmas. In most cumulate sequences, they are held in the trapped liquid (mesostasis). In both magmas and cumulates, the Th/U ratio remains virtually unchanged from the ratio of the parent magma. Only a few examples are known where significant amounts of Th-, U-rich cumulus phases (e.g. perovskite, eudialyte) crystallise and disturb the Th/U ratio. At many loctions, fractional crystallization occurred under open-system conditions and Th and U were redistributed by mobile fluids. These are frequently concentrated in roof zones or added to the surrounding country rocks. Elsewhere, post-magmatic Th-U metasomatism may be so intense that few of the primary, magmatic features are preserved. Previous invetigators of Th and U at Ilimaussaq have found evidence for closed- and open-system conditions at different stages of the evolution, and also for post-magmatic metasomatism. (author)

  17. Absolute M1 and E2 Transition Probabilities in 233U

    International Nuclear Information System (INIS)

    Malmskog, S.G.; Hoejeberg, M.

    1967-08-01

    Using the delayed coincidence technique, the following half lives have been determined for different excited states in 233 U: T 1/2 (311.9 keV level) = (1.20 ± 0.15) x 10 -10 sec, T 1/2 (340.5 keV level) = (5.2 ± 1.0) x 10 -11 sec, T 1/2 (398.6 keV level) = (5.5 ± 2.0) x 10 -11 sec and T 1/2 (415.8 keV level) -11 sec. From these half life determinations, together with earlier known electron intensities and conversion coefficients, 22 reduced B(Ml) and B(E2) transition probabilities (including 9 limits) have been deduced. The rotational transitions give information on the parameters δ and (g K - g R ) . The experimental M1 and E2 transition rates between members of different bands have been analysed in terms of the predictions of the Nilsson model, taking also pairing correlations and Coriolis coupling effects into account

  18. Preparation of UO{sub 2}, ThO{sub 2} and (Th,U)O{sub 2} pellets from photochemically-prepared nano-powders

    Energy Technology Data Exchange (ETDEWEB)

    Pavelková, Tereza [Czech Technical University in Prague, Faculty of Nuclear Sciences and Physical Engineering, Břehová 7, 115 19 Praha 1 (Czech Republic); Čuba, Václav, E-mail: vaclav.cuba@fjfi.cvut.cz [Czech Technical University in Prague, Faculty of Nuclear Sciences and Physical Engineering, Břehová 7, 115 19 Praha 1 (Czech Republic); Visser-Týnová, Eva de [Nuclear Research and Consultancy Group (NRG), Research & Innovation, Westerduinweg 3, 1755 LE Petten (Netherlands); Ekberg, Christian [Nuclear Chemistry/Industrial Materials Recycling, Chalmers University of Technology, SE-412 96 Göteborg (Sweden); Persson, Ingmar [Department of Chemistry and Biotechnology, Swedish University of Agricultural Sciences, P.O. Box 7015, SE-750 07 Uppsala (Sweden)

    2016-02-15

    Photochemically-induced preparation of nano-powders of crystalline uranium and/or thorium oxides and their subsequent pelletizing has been investigated. The preparative method was based on the photochemically induced formation of amorphous solid precursors in aqueous solution containing uranyl and/or thorium nitrate and ammonium formate. The EXAFS analyses of the precursors shown that photon irradiation of thorium containing solutions yields a compound with little long-range order but likely “ThO{sub 2} like” and the irradiation of uranium containing solutions yields the mixture of U(IV) and U(VI) compounds. The U-containing precursors were carbon free, thus allowing direct heat treatment in reducing atmosphere without pre-treatment in the air. Subsequent heat treatment of amorphous solid precursors at 300–550 °C yielded nano-crystalline UO{sub 2}, ThO{sub 2} or solid (Th,U)O{sub 2} solutions with high purity, well-developed crystals with linear crystallite size <15 nm. The prepared nano-powders of crystalline oxides were pelletized without any binder (pressure 500 MPa), the green pellets were subsequently sintered at 1300 °C under an Ar:H{sub 2} (20:1) mixture (UO{sub 2} and (Th,U)O{sub 2} pellets) or at 1600 °C in ambient air (ThO{sub 2} pellets). The theoretical density of the sintered pellets varied from 91 to 97%. - Highlights: • Photochemically prepared UO{sub 2}/ThO{sub 2} nano-powders were pelletized. • The nano-powders of crystalline oxides were pelletized without any binder. • Pellets were sintered at 1300 °C (UO{sub 2} and (Th,U)O{sub 2}) or 1600 °C (ThO{sub 2} pellets). • The theoretical density of the sintered pellets varies from 91 to 97%.

  19. High-resolution wave number spectrum using multi-point measurements in space – the Multi-point Signal Resonator (MSR technique

    Directory of Open Access Journals (Sweden)

    Y. Narita

    2011-02-01

    Full Text Available A new analysis method is presented that provides a high-resolution power spectrum in a broad wave number domain based on multi-point measurements. The analysis technique is referred to as the Multi-point Signal Resonator (MSR and it benefits from Capon's minimum variance method for obtaining the proper power spectral density of the signal as well as the MUSIC algorithm (Multiple Signal Classification for considerably reducing the noise part in the spectrum. The mathematical foundation of the analysis method is presented and it is applied to synthetic data as well as Cluster observations of the interplanetary magnetic field. Using the MSR technique for Cluster data we find a wave in the solar wind propagating parallel to the mean magnetic field with relatively small amplitude, which is not identified by the Capon spectrum. The Cluster data analysis shows the potential of the MSR technique for studying waves and turbulence using multi-point measurements.

  20. Aerosols generated by 239PU and 233U droplets burning in air

    International Nuclear Information System (INIS)

    Nelson, L.S.; Raabe, O.G.

    1978-01-01

    The inhalation hazards of radioactive aerosols produced by the explosive disruption and subsequent combustion of metallic plutonium in air are not adequately understood. Results of a study to determine whether uranium can be substituted for plutonium in such a situation in which experiments were performed under identical conditions with laser-ignited, single, freely falling droplets of 239 Pu and 233 U are reported. The total amounts of aerosol produced were studied quantitatively as a function of time during the combustion. Also, particle size distributions of selected aerosols were studied with aerodynamic particle separation techniques. Results showed that the ultimate quantity of aerosols, their final particle size distributions, and depositions as a function of time are not identical mainly because of the different vapor pressures of the metals, and the unlike degrees of violence of the explosions of the droplets

  1. U-TH-REE mobility and diffusion in granitic environments during alteration of accessory minerals and U-ores

    International Nuclear Information System (INIS)

    Cathelineau, M.; Vergneaud, M.

    1989-01-01

    U, Th and REE concentrations and distributions have been studied in granitic rocks, using a multidisciplinary approach involving micromapping of cracks in oriented samples, together with mineralogical and geochemical studies of the different U-Th-REE bearing phases. The behavior of U, Th and Nd, considered as chemical analogue elements of the radiotoxic nuclides, was investigated either in the vicinity of microsites (accessory mineral enviornment) or along plurimetric sections around U-ore bodies. The different granite minerals, especially the accessory minerals (uraninite, monazite, thorite, apatite, xeonotime), as well as U-ores, present different initial concentrations of U, Th and REE. Limitations to the analogy between these U-Th-REE concentrations and the radwastes is discussed as a function of their mineralogical features, chemical compostion, size and solubilities. These primary concentrations present different behavior when subjected to hydrothermal alteration, such as propylitization, phyllite type alteration, or clay alteration. Results show that in reduced media, in the temperature range 80-2000 0 C, the rate of mobilization of U, Th, REE is relatively moderate. However, fluids enriched in flourides, phosphates or carbonates may significantly solubilize and transport U and REE under specific conditions. In addition, the degree of opening of the microcracks and faults, as well as the oxidation-reduction processes, are critical parameters for the efficiency of the granitic geological barrier

  2. Feasibility to convert an advanced PWR from UO2 to a mixed U/ThO2 core – Part I: Parametric studies

    International Nuclear Information System (INIS)

    Maiorino, Jose R.; Stefani, Giovanni Laranjo; Moreira, João M.L.; Rossi, Pedro C.R.; Santos, Thiago A.

    2017-01-01

    Highlights: • Neutronics calculation using SERPENT code. • Conversion of an advanced PWR from a UO 2 to (U-Th)O 2 core. • AP 1000-advanced PWR. • Parametric studies to define a converted core. • Demonstration of the feasibility to convert the AP 1000 by using mixed uranium thorium oxide fuel with advantages. - Abstract: This work presents the neutronics and thermal hydraulics feasibility to convert the UO 2 core of the Westinghouse AP1000 in a (U-Th)O 2 core by performing a parametric study varying the type of geometry of the pins in fuel elements, using the heterogeneous seed blanket concept and the homogeneous concept. In the parametric study, all geometry and materials for the burnable poison were kept the same as the AP 1000, and the only variable was the fuel pin material, in which we use several mass proportion of uranium and thorium but keeping the enrichment in 235 U, as LEU (20 w/o). The neutronics calculations were made by SERPENT code, and to validate the thermal limits we used a homemade code. The optimization criteria were to maximize the 233 U, and conversion factor, and minimize the plutonium production. The results obtained showed that the homogeneous concept with three different mass proportion zones, the first containing (32% UO 2 -68%ThO 2 ); the second with (24% UO 2 -76% ThO 2 ), and the third with (20% UO 2 -80% ThO 2 ), using 235 U LEU (20 w/o), and corresponding with the 3 enrichment zones of the AP 1000 (4.45 w/o; 3.40 w/o; 2.35 w/o), satisfies the optimization criteria as well as attending all thermal constrain. The concept showed advantages compared with the original UO 2 core, such a lower power density, and keeping the same 18 months of cycle a reduction of B-10 concentration at the soluble poison as well as eliminating in the integral boron poison coated (IFBA).

  3. Correcting for long-alpha stopping distances in (U-Th)/He dating

    Science.gov (United States)

    Glotzbach, Christoph; Lang, Karl; Avdievitch, Nikita; Flowers, Rebecca; Metcalf, James; Ehlers, Todd

    2017-04-01

    Conventional (U-Th)/He dating requires a correction of measured He content for the effect of He loss by alpha particle ejection (e.g. Farley et al. 1996). Compared to typical mineral grain sizes ( 100 µm), the relatively long stopping distance of alpha particles ( 22 µm) results in a significant volume of lost He that systematically bias age calculations (e.g. Ketcham et al. 2011). For example, only 65% of radiogenic He ingrowth will remain within an apatite grain with a radius of 40 µm, assuming a spherical grain shape (Ft=0.65). With such a significant correction to (U-Th)/He age calculations, accurate characterization of grain shape and precise measurement of grain dimensions may often be the largest source of analytical uncertainty. Indeed, difficulty in calculating grain shape may explain at least part of commonly observed overdispersion in (U-Th)/He ages (e.g. Dobson et al. 2008; Horne et al. 2016). For example, the widely used Fish Canyon standard yields 11% dispersion in zircon (U-Th)/He ages(e.g. Dobson et al. 2008; Horne et al. 2016), although the analytical error in He and U-Th-Sm measurement is typically 2%. Most laboratories measure apatite and zircon grain dimensions with a stereo microscope under 200x magnification. Grains are often elongated and therefore measurements are often based on photomicrographs with the crystallographic c-axis parallel to the field of view. Grain dimensions measured this way cannot account for cross sectional variation perpendicular to the crystallographic c-axis, despite this assumption in commonly used analytical calculations of the Ft correction factor (e.g. Ketcham et al. 2011). Moreover, grains with morphologies not well described by frusta or pyramidal-terminated box, cylindrical or hexagonal shapes do not have simple analytical solutions for the Ft correction factor, and must be neglected from subsequent analysis. Here we introduce an advanced numerical approach to measure grain shape and calculate Ft correction

  4. Episodic growth and homogenization of plutonic roots in arc volcanoes from combined U-Th and (U-Th)/He zircon dating

    Science.gov (United States)

    Schmitt, Axel K.; Stockli, Daniel F.; Lindsay, Jan M.; Robertson, Richard; Lovera, Oscar M.; Kislitsyn, Roman

    2010-06-01

    Tracing the fate of unerupted magma is challenging because plutonic roots of young volcanoes are largely inaccessible. Here we develop the use of zircon age spectra to determine crystal provenance and source rocks for volcanic products, in analogy to detrital crystals in sediments. U-Th zircon crystallization ages for the Soufrière Volcanic Complex, Saint Lucia (Lesser Antilles) frequently predate their eruption as determined from combined U-Th and (U-Th)/He zircon dating. The oldest dated eruptions are 273 ± 15 ka and 264 ± 8 ka (1σ uncertainty) for Morne Bonin dacite and Bellevue pumice deposit, respectively. The most recent eruptions formed morphologically pristine domes in the center of the Qualibou depression (Belfond: 13.6 ± 0.4 ka; Terre Blanche: 15.3 ± 0.4 ka). U-Th (U-Pb) zircon crystallization ages determined for crystal rims and interiors range between near-eruption ages to ∼ 600 ka. Older xenocrysts are absent. Zircon crystallization age distributions are complex, yet systematic: crystal rim ages in the most recently erupted volcanic rocks match those of co-erupted plutonic inclusions, whereas crystal interiors are equivalent to the cumulative distribution of zircon ages from older eruptions. This is evidence that silicic lava domes and pyroclastic flows share a common source that is located underneath the Qualibou depression, where the intrusive roots of this long-lived arc volcanic system became homogenized through thermal and mechanical reprocessing of individual batches of unerupted magma from earlier volcanic episodes within timescales of < 100 ka.

  5. Development of sequential analytical method for the determination of U-238, U-234, Th-232, Th-230, Th-228, Ra-226 and Ra-228 and its application in mineral waters

    International Nuclear Information System (INIS)

    Costa Lauria, D. da.

    1986-01-01

    A sequential analytical method for the determination of U-238, U-234, Th-232, Th-230, Th-228, Ra-226 and Ra-228 in environmental samples and applied to the analysis of mineral waters is studied. Thorium isotopes are coprecipitated with lanthanium fluoride before counting in alpha spectrometer, the uranium isotopes are determined by alpha spectrometry following extraction with TOPO onto a polymenic membrane. Radium-226 is determined with the radom emanation technique. (M.J.C.) [pt

  6. Effects of transgenic methionine sulfoxide reductase A (MsrA expression on lifespan and age-dependent changes in metabolic function in mice

    Directory of Open Access Journals (Sweden)

    Adam B. Salmon

    2016-12-01

    Full Text Available Mechanisms that preserve and maintain the cellular proteome are associated with long life and healthy aging. Oxidative damage is a significant contributor to perturbation of proteostasis and is dealt with by the cell through regulation of antioxidants, protein degradation, and repair of oxidized amino acids. Methionine sulfoxide reductase A (MsrA repairs oxidation of free- and protein-bound methionine residues through enzymatic reduction and is found in both the cytosol and the mitochondria. Previous studies in Drosophila have shown that increasing expression of MsrA can extend longevity. Here we test the effects of increasing MsrA on longevity and healthy aging in two transgenic mouse models. We show that elevated expression of MsrA targeted specifically to the cytosol reduces the rate of age-related death in female mice when assessed by Gompertz analysis. However, neither cytosolic nor mitochondrial MsrA overexpression extends lifespan when measured by log-rank analysis. In mice with MsrA overexpression targeted to the mitochondria, we see evidence for improved insulin sensitivity in aged female mice. With these and our previous data, we conclude that the increasing MsrA expression in mice has differential effects on aging and healthy aging that are dependent on the target of its subcellular localization.

  7. 230Th-238U radioactive disequilibria in tholeiites from the FAMOUS zone (Mid-Atlantic Ridge, 36050'N): Th and Sr isotopic geochemistry

    International Nuclear Information System (INIS)

    Condomines, M.; Morand, P.; Allegre, C.J.

    1981-01-01

    We analyzed, U, Th and 230 Th/ 232 Th activity ratios for a few tholeiites from the Mid-Atlantic Ridge FAMOUS zone at 36 0 50'N. The results show a fairly wider scatter for both Th/U and ( 230 Th/ 232 Th) ratios. Seawater contamination appears to be responsible for this scatter and, for the uranium, produces an increase in content yielding a ( 234 U/ 238 U) ratio greater than 1 and, for the Th, an increase of the ( 230 Th/ 232 Th) ratio which is a very sensitive indicator for contamination. Also, the latter often is selective: U, Th and Sr are not affected in the same manner. When discarding all data for contaminated samples, the FAMOUS zone appears to be very homogeneous with a Th/U ratio value of 3.05 and a ( 230 Th/ 232 Th) ratio value of 1.24. Comparison with other active volcanic areas reveals a negative correlation between ( 230 Th/ 232 Th) and 87 Sr/ 86 Sr ratios for present lavas which is indicative of a consistency in Th-U and Rb-Sr fractionation in the source regions of these magmas. The Th isotopic geochemistry can thus provide useful information for the study of present volcanism, information as valuable as that from Sr, Pb or Nd isotopes. (orig.)

  8. Evolution of actinides in ThO2 blanket of prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Bachchan, Abhitab; Riyas, A.; Devan, K.; Puthiyavinayagam, P.

    2015-01-01

    The third stage of India's nuclear program focuses on fissile fuel production through Th- 233 U cycle in view of the better abundance and relative merits of thorium. For early introduction of Thorium into the nuclear energy system, several R and D program has started to find the best possible route of thorium utilization. Towards this, efforts were made to assess the feasibility of Th-U cycle in a fast spectrum reactor like Prototype Fast Breeder Reactor (PFBR). The effect on core neutronic parameters and actinide evolution with the replacement of depleted UO 2 in the PFBR blanket SA with thorium oxide has been studied using 3-D diffusion code FARCOB. Study shows that by the introduction of thorium blanket, core excess reactivity is coming down by ∼ 535 pcm and core breeding ratio is slightly lower than conventional oxide blanket. The distribution of region wise power production is slightly changed. Power from radial blanket is reduced from 3% to 2% while the core-1 power is increased from 49 % to 50 %. The estimated 233 U production is 7.6, 11.5 and 14.1 kg/t with 180, 360 and 540 days of irradiation respectively. (author)

  9. Peptide methionine sulfoxide reductase A (MsrA): direct electrochemical oxidation on carbon electrodes.

    Science.gov (United States)

    Enache, T A; Oliveira-Brett, A M

    2013-02-01

    The direct electrochemical behaviour of peptide methionine sulfoxide reductase A (MsrA) adsorbed on glassy carbon and boron doped diamond electrodes surface, was studied over a wide pH range by cyclic and differential pulse voltammetry. MsrA oxidation mechanism occurs in three consecutive, pH dependent steps, corresponding to the oxidation of tyrosine, tryptophan and histidine amino acid residues. At the glassy carbon electrode, the first step corresponds to the oxidation of tyrosine and tryptophan residues and occurs for the same potential. The advantage of boron doped diamond electrode was to enable the separation of tyrosine and tryptophan oxidation peaks. On the second step occurs the histidine oxidation, and on the third, at higher potentials, the second tryptophan oxidation. MsrA adsorbs on the hydrophobic carbon electrode surface preferentially through the three hydrophobic domains, C1, C2 and C3, which contain the tyrosine, tryptophan and histidine residues, and tryptophan exists only in these regions, and undergo electrochemical oxidation. Copyright © 2012 Elsevier B.V. All rights reserved.

  10. Determination of U, Th and K for optically stimulated luminescence dating by NAA

    International Nuclear Information System (INIS)

    Qin Yali; Chen Zhe; Wu Weiming

    2010-01-01

    Optically stimulated luminescence dating techniques have been widely used in northern Loess ancient soil series and recorded climate environment change, ancient earthquake, the ancients site and archaeology research. The determination of U, Th and K using neutron activation analysis (NAA) has been optimized for the samples related to OSL dating research. The procedure for determination of U, Th, K in loess have been fixed by using Miniature neutron source reactor. This procedure of NAA will provide a reliable data base for optically stimulated luminescence dating research. (authors)

  11. Asymmetrically deformed third minimum in the 231Th and 233Th fission barriers

    International Nuclear Information System (INIS)

    Blons, J.; Mazur, C.; Paya, D.; Ribrag, M.; Weigmann, H.

    1981-01-01

    Neutron induced fission cross-sections of 230 Th and 232 Th have been measured up to 5 MeV. The electron linear accelerator (GELINA) has been used as a neutron time of flight spectrometer with a nominal resolution of 84 psec/m for 230 Th(n,f) and 42 psec/m for 232 Th(n,f) reaction. The fission fragment detector was a 6 cell gas scintillator filled with xenon. The existence of fine structure peaks, a few keV wide, in both the 230 Th(n,f) and 232 Th(n,f) cross sections, is definitively confirmed. The analysis of the two vibrational resonances located respectively at 720 keV for 230 Th(the figure) and 1.6 MeV for 232 Th, shows clearly that these peaks can be interpreted, in terms of two rotational bands with opposite parities. This parity degeneracy is a consequence of the asymmetric, pear-like deformation of the excited nucleus [ru

  12. Development of fuel cycle technology for molten-salt reactor systems

    International Nuclear Information System (INIS)

    Uhlir, J.

    2006-01-01

    Full text: Full text: The Molten-Salt Reactor (MSR) represents one of promising advanced reactor type assigned to the GEN IV reactor systems. It can be operated either as thorium breeder within the Th -133U fuel cycle or as actinide transmuter incinerating transuranium fuel. Essentially the main advantage of MSR comes out from the prerequisite, that this reactor type should be directly connected with the 'on-line' reprocessing of circulating liquid (molten-salt) fuel. This principle should allow very effective extraction of freshly constituted fissile material (233U). Besides, the on-line fuel salt clean up is necessary within a long run to keep the reactor in operation. As a matter of principle, it permits to clear away typical reactor poisons like xenon, krypton, lanthanides etc. and possibly also other products of burned plutonium and transmuted minor actinides. The fuel salt clean up technology should be linked with the fresh MSR fuel processing to continuously refill the new fuel (thorium or transuranics) into the reactor system. On the other hand, the technologies of fresh transuranium molten-salt fuel processing from the current LWR spent fuel and of the on-line reprocessing of MSR fuel represent two killing points of the whole MSR technology, which have to be successfully solved before MSR deployment in the future. There are three main pyrochemical partitioning techniques proposed for processing and/or reprocessing of MSR fuel: Fluoride volatilization processes, Molten salt / liquid metal extraction processes and Electrochemical separation processes. Two of them - Fluoride Volatility Method and Electrochemical separation process from fluoride media are under development in the Nuclear Research Institute Rez pic. R and D in the field of Fluoride Volatility Method is concentrated to the development and verification of experimental semi-pilot technology for LWR spent fuel reprocessing, which may result in a product the form and composition of which might be

  13. The thermodynamics of extraction of U(VI) and Th(IV) from nitric acid by neutral phosphorus-based organic compounds

    International Nuclear Information System (INIS)

    Kalina, D.G.; Mason, G.W.; Horwitz, E.P.

    1981-01-01

    The extraction of Th(IV) and U(VI) from dilute nitric acid solution by several neutral phosphorus-based extractants has been studied as a function of temperature in the range of 0 to 50 0 C. From the variation of the distribution ratio (Ksub(d)) with temperature the thermodynamic quantities ΔG, ΔH and ΔS have been calculated for these extractions. The results of this study indicate that the steric bulk of the extractant plays a major role in determining how well Th(IV) is extracted. The size of the extractant appears to be of little or no importance in the extraction of U(VI). Similarly, the basicity of the extractant is of lesser importance in the extraction of uranyl ion relative to thorium ion. (author)

  14. Evolution of Th and U whole-rock contents in the Ilimaussaq intrusion

    International Nuclear Information System (INIS)

    Bailey, J.C.; Rose-Hansen, J.; Soerensen, H.

    1981-01-01

    Thorium and uranium values of a large collction of representative samples taken from all rock types of the Ilimaussaq alkaline intrusion, South Greenland, are presented. The values are largely obtained by laboratory gamma-ray spectrometric (GRS) analysis. The results are discussed in relation to current knowledge and ideas on the petrologic evolution of the Ilimaussaq intrusion. It is concluded that (1) Rocks from the Ilimaussaq alkaline intrusion evolve to extremely high Th and U contents; (2) The evolution is characterised by appearance of low-Th/U cumulates due to the appearance of low-Th/U eudialyte as a liquidus phase; (3) Fractionation of the observed cumulus assemblages fails to explain all features of the Th-U evolution; (4) Losses of mobile fluids, rich in Th/U, occur in the final stages. (BP)

  15. U4+ spectroscopic properties in Dsub(2d) with ThCl4, UCl4 and ThSiO4

    International Nuclear Information System (INIS)

    Khan Malek, C.

    1985-01-01

    This thesis is concerned with the study of the electronic structure of the tetravalent actinide ions in solid state. The technique used was high resolution optical spectroscopy. We deal with the U 4+ ion (sf 2 ) in the monocrystals ThCl 4 , UCl 4 , and ThSiO 4 where the U 4+ ion is substituded for the Th 4+ ion by doping. Visible and infrared optical spectra were recorded between 300 and 4.2K. With these three compounds, it is possible to compare the influence of different environments of Dsub(2d) symmetry: real symmetry for U 4+ in UCl 4 and ThSiO 4 ; approximate symmetry in ThCl 4 , whose structure is incommensurate and modulated at low temperature. The fitting of the data was carried out by diagonalizing the hamiltonian which describes the interactions of the U 4+ ion in a crystal field with its environment. This fitting procedure led to a coherent set of spectroscopic parameters. The fluorescence of U 4+ was observed in ThCl 4 and ThSiO 4 and the effect of the incommensurate structure of ThCl 4 on the optical spectra was studied. The symmetry of the U 4+ sites was identified by site selective excitation experiments and a relationship between the incommensurate structure and the lifetime of U 4+ energy levels was found. In conclusion, the U 4+ energy levels in a relatively low crystal field were determined for compounds that have a similar coordination polyhedron about the actinide ion. The values for these energy levels were then compared to those of lanthanide and 3d elements [fr

  16. In situ detrital zircon (U-Th)/He thermochronology

    Science.gov (United States)

    Tripathy, A.; Monteleone, B. D.; van Soest, M. C.; Hodges, K.; Hourigan, J. K.

    2010-12-01

    Detrital studies of both sand and rock are relevant to many problems, ranging from the climate and tectonics feedback debate to the long-term record of orogenic evolution. When applying the conventional (U-Th)/He technique to such studies, two important issues arise. Often, only euhedral grains are permissible for analysis in order to make simple geometric corrections for α-recoil. In detrital samples, this is problematic because euhedral grains can be scarce due to mechanical abrasion during transport, and potentially introduce bias in favour of more proximally sourced grains. Second, inherent to detrital studies is the need to date many grains (>100) per sample to ensure a representative sampling of the sediment source region, thus making robust conventional detrital studies both expensive and time-consuming. UV laser microprobes can improve this by permitting careful targeting of the grain interior away from the α-ejection zone, rendering the α-recoil correction unnecessary, thus eliminating bias toward euhedral grains. In the Noble Gas, Geochemistry, and Geochronology Laboratory at ASU, apatite and zircon have been successfully dated using in situ methods. For this study, the conventional and in situ techniques are compared by dating zircons from a modern river sand that drains a small catchment in the Mesozoic-Cenozoic Ladakh Batholith in NW India. This sample has a simple provenance, which allows us to demonstrate the robustness of the in situ method. Moreover, different microbeam techniques will be explored to establish the most efficient approach to obtain accurate and precise U-Th concentrations using synrock, which is our powdered, homogenized, and reconstituted zircon-rock standard. Without this, such in situ U-Th data would be difficult to obtain. 117 zircons were dated using the conventional (U-Th)/He method, revealing dates ranging from 9.70±0.35 to 106.6±3.5 Ma (2σ) with the major mode at 26 Ma. For comparison, 44 grains were dated using the in

  17. Application of spectrometer cropscan MSR 16R and Landsat imagery for identification the spectral characteristics of land cover

    Science.gov (United States)

    Tampubolon, Togi; Abdullah, Khiruddin bin; San, Lim Hwee

    2013-09-01

    The spectral characteristics of land cover are basic references in classifying satellite image for geophysics analysis. It can be obtained from the measurements using spectrometer and satellite image processing. The aims of this study to investigate the spectral characteristics of land cover based on the results of measurement using Spectrometer Cropscan MSR 16R and Landsat satellite imagery. The area of study in this research is in Medan, (Deli Serdang, North Sumatera) Indonesia. The scope of this study is the basic survey from the measurements of spectral land cover which is covered several type of land such as a cultivated and managed terrestrial areas, natural and semi-natural, cultivated aquatic or regularly flooded areas, natural and semi-natural aquatic, artificial surfaces and associated areas, bare areas, artificial waterbodies and natural waterbodies. The measurement and verification were conducted using a spectrometer provided their spectral characteristics and Landsat imagery, respectively. The results of the spectral characteristics of land cover shows that each type of land cover have a unique characteristic. The correlation of spectral land cover based on spectrometer Cropscan MSR 16R and Landsat satellite image are above 90 %. However, the land cover of artificial waterbodiese have a correlation under 40 %. That is because the measurement of spectrometer Cropscan MSR 16R and acquisition of Landsat satellite imagery has a time different.

  18. Influence of Contact Time on the Extraction of 233Uranyl Spike and Contaminant Uranium From Hanford Sediment

    International Nuclear Information System (INIS)

    Smith, Steven C.; Szecsody, James E.

    2011-01-01

    In this study 233Uranyl nitrate was added to uranium (U) contaminated Hanford 300 Area sediment and incubated under moist conditions for 1 year. It hypothesized that geochemical transformations and/or physical processes will result in decreased extractability of 233U as the incubation period increases, and eventually the extraction behavior of the 233U spike will be congruent to contaminant U that has been associated with sediment for decades. Following 1 week, 1 month, and 1 year incubation periods, sediment extractions were performed using either batch or dynamic (sediment column flow) chemical extraction techniques. Overall, extraction of U from sediment using batch extraction was less complicated to conduct compared to dynamic extraction, but dynamic extraction could distinguish the range of U forms associated with sediment which are eluted at different times.

  19. Isotopic exchange between 232Th and 234Th using ion exchange resins and its application for the radiochemical separation of thorium and europium

    International Nuclear Information System (INIS)

    Sepulveda Munita, C.J.A.; Atalla, L.T.

    1980-01-01

    The determination of thorium via the measurement of 233 Th activity (obtained by irradiating natural thorium with neutrons) may suffer the interference of various radioisotopes which may be also formed during irradiation, if their parent isotopes are present in the sample. Taking into account this possibility, another technique was chosen for the determination of thorium, based on isotopic exchange associated with ionic exchange. Conditions for the isotopic exchange between 234 Th in solution and 232 Th in the resin were optimized. It was verified that the behaviour of 233 Th and 234 Th is the same regarding isotopic exchange with 232 Th. 234 Th was chosen for the experiments since it has a longer half-life (24.1 days) than 233 Th (22.3 min), thus facilitating the performance of the work. As the major objective of this work is to separate thorium and europium isotopes, the behaviour of 152-154 Eu was studied in the same system used for thorium, envisaging a minimum retention of these radioisotopes in the resin. In order to establish the best conditions for separating 234-Th and 152/154-Eu, the following parameters were considered: the thorium concentration in the solution; the hydrochloric acid concentration in solution; the concentration of other elements in solution; the degree of cross-linking of the resin; the flow rate of the solution through the column. The other elements added to the elutant solution were: uranium, molybdenum, lanthanum, europium, ytterbium, bromine, cobalt, barium, manganese, indium, cesium and selenium. Europium was added so to dilute the 152/154-Eu tracer and avoid the retention of the latter in the resin. The other elements were added because they give rise to radioisotopes which interfere in the activation analysis of thorium when 233-Th activity is used and, the separation of these elements from thorium will also be subsequently studied by the method used in the present work. (C.L.B.) [pt

  20. Dissolution rates of unirradiated UO2, UO2 doped with 233U, and spent fuel under normal atmospheric conditions and under reducing conditions using an isotope dilution method

    International Nuclear Information System (INIS)

    Ollila, Kaija; Albinsson, Yngve; Oversby, Virginia; Cowper, Mark

    2003-10-01

    The experimental results given in this report allow us to draw the following conclusions. 1) Tests using unirradiated fuel pellet materials from two different manufacturers gave very different dissolution rates under air atmosphere testing. Tests for fragments of pellets from different pellets made by the same manufacturer gave good agreement. This indicates that details of the manufacturing process have a large effect on the behavior of unirradiated UO 2 in dissolution experiments. Care must be taken in interpreting differences in results obtained in different laboratories because the results may be affected by manufacturing effects. 2) Long-term tests under air atmosphere have begun to show the effects of precipitation. Further testing will be needed before the samples reach steady state. 3) Testing of unirradiated UO 2 in systems containing an iron strip to produce reducing conditions gave [U] less than detection limits ( 235 U added as spike was recovered, indicating that 90% of the spike had precipitated onto the solid sample or the iron strip. 9) Tests of UO 2 pellet materials containing 233 U to provide an alpha decay activity similar to that expected for spent fuel 3000 and 10,000 years after disposal showed that the pellet materials behaved as expected under air atmosphere conditions, showing that the manufacturing method was successful. 10) Early testing of the 233 U-doped materials under reducing conditions showed relatively rapid (30 minute) dissolution of small amounts of U at the start of the puff test procedure. Results of analyses of an acidified fraction of the same solutions after 1 or 2 weeks holding indicate that the solutions were inhomogeneous, indicating the presence of colloidal material or small grains of solid. 11) Samples from the 233 U-doped tests initially indicated dissolution of solid during the first week of testing, with some indication of more rapid dissolution of the material with the higher doping. 12) The second cycle of testing

  1. Distribution of U and Th in Growth Zones of Manganese Nodules

    DEFF Research Database (Denmark)

    Kunzendorf, Helmar; Friedrich, G. H. W.

    1976-01-01

    Growth zones and individual sublayers from one manganese nodule and three manganese crusts from an area south of Hawaii were analysed for U and Th by the delayed-neutron counting technique. The concentrations of uranium and thorium in the manganese nodule are highest in the outermost zone on top...... to the sediment which has low concentrations of Fe, relative to the zone last exposed to sea water, has also low U contents (2.7 ppm). Th concentrations are higher in the outermost zone on top of the nodule (40 to a maximum of 130 ppm) than in the zone last exposed to the sediment (about 20 ppm Th). Manganese...... crusts contain up to 9 ppm U in the outermost zones last exposed to the sea water. They also have higher concentrations of Th (up to 64 ppm) relative to the inner zones of the crust growing on altered andesitic rock, which contains about 8 ppm U and about 26 ppm Th as an average....

  2. Separation of Th(IV) and U(VI) by extraction chromatography

    International Nuclear Information System (INIS)

    Nadkarni, M.N.; Mayankutty, P.C.; Pillai, N.S.

    1984-01-01

    Application of extraction chromatography to the analytical separation of Th(IV) and U(VI) has been investigated. The stationary phase was a macroporous resin Amberlite XE-270 impregnated with undiluted tri-n-butylphosphate (TBP) and the mobile phase was either 5.0M HNO 3 or 6M HCl. Separation of traces of Th(IV) from large quantities of U(VI) was achieved on a laboratory column by elution of the absorbed Th(IV) with 6M HCl. (author)

  3. LA-ICP-MS and SIMS U-Pb and U-Th zircon geochronological data of Late Pleistocene lava domes of the Ciomadul Volcanic Dome Complex (Eastern Carpathians

    Directory of Open Access Journals (Sweden)

    Réka Lukács

    2018-06-01

    Full Text Available This article provides laser-ablation inductively coupled plasma mass spectrometry (LA-ICP-MS and secondary ionization mass spectrometry (SIMS U-Pb and U-Th zircon dates for crystals separated from Late Pleistocene dacitic lava dome rocks of the Ciomadul Volcanic Dome Complex (Eastern Carpathians, Romania. The analyses were performed on unpolished zircon prism faces (termed rim analyses and on crystal interiors exposed through mechanical grinding an polishing (interior analyses. 206Pb/238U ages are corrected for Th-disequilibrium based on published and calculated distribution coefficients for U and Th using average whole-rock and individually analyzed zircon compositions. The data presented in this article were used for the Th-disequilibrium correction of (U-Th/He zircon geochronology data in the research article entitled “The onset of the volcanism in the Ciomadul Volcanic Dome Complex (Eastern Carpathians: eruption chronology and magma type variation” (Molnár et al., 2018 [1].

  4. 238U-230Th radioactive disequilibria in the volcanic products from Izu arc volcanoes, Japan

    International Nuclear Information System (INIS)

    Kurihara, Yuichi; Takahashi, Masaomi; Sato, Jun

    2007-01-01

    The timescale of magmatic processes of Izu arc volcanoes, Japan, was estimated by the 238 U- 230 Th disequilibria in the volcanic products from the volcanoes. The majority of the 230 Th/ 238 U activity ratios of the products were less than unity, being enriched in 238 U relative to 230 Th. The ( 230 Th/ 232 Th)-( 238 U/ 232 Th)diagram for younger Fuji and Izu-Oshima volcanoes formed a whole rock isochrons, and the ages were 1x10 4 and 2x10 4 years, respectively. The ( 230 Th/ 232 Th) - ( 238 U/ 232 Th) data set for younger Fuji volcano formed a cluster on the diagram, while those of Izu-Oshima formed another cluster apparently apart from each other, suggesting that the concentration of U and Th may possibly be un-uniform in the mantle beneath Izu arc. (author)

  5. 238U-234U-230Th chronometry of Fe-Mn crusts: Growth processes and recovery of thorium isotopic ratios of seawater

    International Nuclear Information System (INIS)

    Chabaux, F.; Cohen, A.S.; O'Nions, R.K.; Hein, J.R.

    1995-01-01

    Comparison of ( 234 U) excess /( 238 U) and ( 230 Th)/( 232 Th) activity ratios in oceanic Fe-Mn deposits provides a method for assessing the closed-system behaviour of 238 U- 234 U- 230 Th, as well as variations in the initial uranium and thorium isotopic ratios of the precipitated metal oxides. This approach is illustrated using a Fe-Mn crust from Lotab seamount (Marshall Islands, west equatorial Pacific). Here we report uranium and thorium isotopic compositions in five subsamples from the surface of one large 5 cm diameter botyroid of this crust, and from two depth profiles of the outermost rim of the same botyroid. The decrease of ( 234 U) excess /( 238 U) and ( 230 Th/ 232 Th) activity ratio with depth in the two profiles gives mean growth rates, for the last 150 ka, of 7.8 ± 2 mm/Ma and 6.6 ± 1 mm/Ma, respectively. All data points (surface and core samples) but one, define a linear correlation in the Ln ( 230 Th/ 232 Th) - Ln [( 234 U) excess ( 238 U)] diagram. This correlation indicates that for all points the U-Th system remained closed after the Fe-Mn layer precipitated, and that the different samples possessed the same initial Uranium and thorium isotope ratios. Furthermore, these results show that the preserved surface of this Fe-Mn crust may not be the present-day growth surface, and that the thorium and uranium isotopic ratios of seawater in west equatorial Pacific have not changed during the past 150 ka. The initial thorium activity ratio is estimated from the correlation obtained between Ln( 230 Th/ 232 Th) and Ln [( 234 U) excess /( 238 U)

  6. Narrowband tunable laser for uranium-233 cleanup process

    International Nuclear Information System (INIS)

    Singh, Sunita; Sridhar, G.; Rawat, V.S.; Kawde, Nitin; Sinha, A.K.; Bhatt, S.; Gantayet, L.M.

    2009-01-01

    Design, development and technology demonstration of proto type Single Longitudinal Mode pulsed tunable laser is reported in this work. The tunable laser has a narrow bandwidth less than 400 MHz required for isotopic clean up of 233 U. (author)

  7. Fission cross sections of some thorium, uranium, neptunium and plutonium isotopes relative to /sup 235/U

    Energy Technology Data Exchange (ETDEWEB)

    Meadows, J W

    1983-10-01

    Earlier results from the measurements, at this Laboratory, of the fission cross sections of /sup 230/Th, /sup 232/Th, /sup 233/U, /sup 234/U, /sup 236/U, /sup 238/U, /sup 237/Np, /sup 239/Pu, /sup 240/Pu, and /sup 242/Pu relative to /sup 235/U are reviewed with revisions to include changes in data processing procedures, alpha half lives and thermal fission cross sections. Some new data have also been included. The current experimental methods and procedures and the sample assay methods are described in detail and the sources of error are presented in a systematic manner. 38 references.

  8. Potential use of thorium through fusion breeders in the Indian context

    International Nuclear Information System (INIS)

    Srinivasan, M.; Basu, T.K.; Subba Rao, K.

    1991-01-01

    The Indian Nuclear Programme is based on a three stage strategy: the first stage of about 10 GWe comprises of natural uranium fuelled Pressurised Heavy Water Reactors (PHWRs); the second stage would consist of Liquid Metal Cooled Fast Breeder Reactors (LMFBRs) to be fuelled with plutonium generated in the first stage PHWRs and the third stage is envisaged to be based on advanced converters/breeders operating on the Th/U-233 cycle. It has generally been assumed that the initial inventory of U-233 for the third stage reactors would be generated in the blankets of LMFBRs containing thorium. But the success of this strategy depends crucially on the attainment of LMFBR doubling times as short as 14 years. The progress registered in recent years in the magnetic confinement of fusion plasmas has opened up the prospects of developing Fusion Breeders for the direct conversion of fertile 232 Th into fissile 233 U using the 14 MeV neutron released in the (D-T) fusion reaction. A detailed study of the dependence of the 233 U production characteristics as well as energy cost of fissile fuel production of such systems on parameters such as plasma energy gain Q, blanket neutron multiplication has been carried out. The growth rate dynamics of the symbiotic combination of 233 U generating fusion breeders with PHWRs operating on the Th/U-233 cycle in the so called near-breeder regime has been examined. 95% of the energy generated by PHWRs operating with Th/ 233 U fuel would arise from thorium consumption rather than fission of the initially loaded 233 U. A few sub-engineering breakeven fusion breeders producing U-233 at an energy cost well under 200 MeV per atom are adequate to give the requisite nuclear capacity growth rates in conjunction with such near breeder PHWRs. This corresponds to only a 5% diversion of the grid electrical power for the operation of such fusion breeders. In summary the symbiotic combination of a few fusion breeders with a number of PHWRs gives fresh hopes

  9. Parametric systems analysis of the Modular Stellarator Reactor (MSR)

    International Nuclear Information System (INIS)

    Miller, R.L.; Krakowski, R.A.; Bathke, C.G.

    1982-05-01

    The close coupling in the stellarator/torsatron/heliotron (S/T/H) between coil design (peak field, current density, forces), magnetics topology (transform, shear, well depth), and plasma performance (equilibrium, stability, transport, beta) complicates the reactor assessment more so than for most magnetic confinement systems. In order to provide an additional degree of resolution of this problem for the Modular Stellarator Reactor (MSR), a parametric systems model has been developed and applied. This model reduces key issues associted ith plasma performance, first-wall/blanket/shield (FW/B/S), and coil design to a simple relationship between beta, system geometry, and a number of indicators of overall plant performance. The results of this analysis can then be used to guide more detailed, multidimensional plasma, magnetics, and coil design efforts towards technically and economically viable operating regimes. In general, it is shown that beta values > 0.08 may be needed if the MSR approach is to be substantially competitive with other approaches to magnetic fusion in terms of system power density, mass utilization, and cost for total power output around 4.0 GWt; lower powers will require even higher betas

  10. Coordination polymers: trapping of radionuclides and chemistry of tetravalent actinides (Th, U) carboxylates

    International Nuclear Information System (INIS)

    Falaise, Clement

    2014-01-01

    The use of nuclear energy obviously raises the question of the presence of radionuclides in the environment. Currently, their mitigation is a major issue associated with nuclear chemistry. This thesis focuses on both the trapping of radionuclides by porous solids called Metal-Organic Frameworks (MOF) and the crystal chemistry of the carboxylate of tetravalent actinides (AnIV). The academic knowledge of the reactivity of carboxylate of AnIV could help the understanding of actinides speciation in environment. We focused on the sequestration of iodine by aluminum based MOF. The functionalization (electron-donor group) of the MOF drastically enhances the iodine capture capacity. The removal of light actinides (Th and U) from aqueous solution was also investigated as well as the stability of (Al)-MOF under γ radiation. More than twenty coordination polymers based on tetravalent actinides have been synthesized and characterized by single crystal X-ray diffraction. The use of controlled hydrolysis promotes the formation of coordination polymers exhibiting polynuclear cluster ([U 4 ], [Th 6 ], [U 6 ] and [U 38 ]). In order to understand the formation of the largest cluster, the ex-situ study of the solvo-thermal synthesis of compound {U 38 } has also been investigated. (author)

  11. On the Use of 233U-236U Double-Spike for TIMS Measurements of Uranium Isotopes: A Simulation Study

    International Nuclear Information System (INIS)

    Williams, R W

    2004-01-01

    Synthetic ion beams with instantaneous and temporal characteristics appropriate to thermal ionization mass spectrometry (TIMS) were mathematically generated and analyzed to determine the effects of using a mixed 233 U- 236 U spike (double-spike) in the analysis of uranium isotopes. The instantaneous beam characteristics are the intensities (e.g., counts per second) modeled with a Poisson distribution plus a component of random noise that simulates the detection processes. Several beam intensity and mass fractionation vs. time functions were modeled to simulate a range of sample sizes and the commonly employed methods of data collection. These beam profiles were also generated with different noise levels, and signal-to-noise vs. analytical precision diagrams are presented. Modeling focused on natural uranium, where 238 U/ 235 U = 137.88, and on the ability of a given method to determine precisely and accurately small variations in this ratio. Practical limits on precision were determined to be 20-30 ppm, which is consistent with precision seen for other elements by state-of-the-art TIMS. The TIMS total evaporation method was compared directly with the double-spike method. While similar analytical precisions are obtained with either method, the double-spike method of correcting for analytical bias gives more accurate results. The results of a total evaporation analysis will deviate from true by more than the analytical precision if as little as 0.05% of the signal is not integrated, whereas the accuracy and precision of the double-spiked analyses are always linked

  12. Feasibility of Th-U separation through a pyrochemical route in molten LiCl-KCl eutectic

    International Nuclear Information System (INIS)

    Pakhui, Gurudas; Ghosh, Suddhasattwa; Prabhakara Reddy, B.; Nagarajan, K.

    2014-01-01

    Molten salt electrorefining is a high temperature electrometallurgical process developed for the reprocessing of spent UPu-Zr fuel employing LiCl-KCl as the electrolyte. A possible application of this high temperature electrochemical process could be in the separation of U from Th matrix for Th based fuels. It therefore becomes important to investigate the reduction behaviour of Th 4+ in the molten salt and compare with that of U 3+ . The present study is on the electrochemical behaviour of Th 4+ in LiCl-KCl. Electrochemical studies were also carried out on the LiCl-KCl-UCl 3 -ThCl 4 system using transient electrochemical techniques. The cyclic voltammograms for Th 4+ /Th redox couple at various scan rates at 723 K is shown. Reduction of Th 4+ to Th was found to be quasi-reversible at lower scan rates and irreversible at higher scan rates using cyclic voltammetry and convolution voltammetry. The number of electrons transferred for the reduction process was calculated to be ∼ 4 using various techniques like cyclic voltammetry, chronopotentiometry and convolution voltammetry

  13. Investigation of the fission yields of the fast neutron-induced fission of {sup 233}U; Mesure de la distribution en masse et en charge des produits de la fission rapide de l'{sup 233}U

    Energy Technology Data Exchange (ETDEWEB)

    Galy, J

    1999-09-01

    As a stars, a survey of the different methods of investigations of the fission product yields and the experimental data status have been studied, showing advantages and shortcomings for the different approaches. An overview of the existing models for the fission product distributions has been as well intended. The main part of this thesis was the measurement of the independent yields of the fast neutron-induced fission of{sup 233}U, never investigated before this work. The experiment has been carried out using the mass separator OSIRIS (Isotope Separator On-Line). Its integrated ion-source and its specific properties required an analysis of the delay-parameter and ionisation efficiency for each chemical species. On the other hand, this technique allows measurement of independent yields and cumulative yields for elements from Cu to Ba, covering most of the fission yield distribution. Thus, we measured about 180 independent yields from Zn (Z=30) to Sr (Z=38) in the mass range A=74-99 and from Pd (Z=46) to Ba (Z=56) in the mass range A=113-147, including many isomeric states. An additional experiment using direct {gamma}-spectroscopy of aggregates of fission products was used to determine more than 50 cumulative yields of element with half-life from 15 min to a several days. All experimental data have been compared to estimates from a semi-empirical model, to calculated values and to evaluated values from the European library JEF 2.2. Furthermore, a study of both thermal and fast neutron-induced fission of {sup 233}U measured at Studsvik, the comparison of the OSIRIS and LOHENGRIN facilities and the trends in new data for the Reactors Physics have been discussed. (author)

  14. Assessment of rigid multi-modality image registration consistency using the multiple sub-volume registration (MSR) method

    International Nuclear Information System (INIS)

    Ceylan, C; Heide, U A van der; Bol, G H; Lagendijk, J J W; Kotte, A N T J

    2005-01-01

    Registration of different imaging modalities such as CT, MRI, functional MRI (fMRI), positron (PET) and single photon (SPECT) emission tomography is used in many clinical applications. Determining the quality of any automatic registration procedure has been a challenging part because no gold standard is available to evaluate the registration. In this note we present a method, called the 'multiple sub-volume registration' (MSR) method, for assessing the consistency of a rigid registration. This is done by registering sub-images of one data set on the other data set, performing a crude non-rigid registration. By analysing the deviations (local deformations) of the sub-volume registrations from the full registration we get a measure of the consistency of the rigid registration. Registration of 15 data sets which include CT, MR and PET images for brain, head and neck, cervix, prostate and lung was performed utilizing a rigid body registration with normalized mutual information as the similarity measure. The resulting registrations were classified as good or bad by visual inspection. The resulting registrations were also classified using our MSR method. The results of our MSR method agree with the classification obtained from visual inspection for all cases (p < 0.02 based on ANOVA of the good and bad groups). The proposed method is independent of the registration algorithm and similarity measure. It can be used for multi-modality image data sets and different anatomic sites of the patient. (note)

  15. Structure and luminescence of α and β ThBr4: optical properties of U4+ in α ThBr4

    International Nuclear Information System (INIS)

    Simoni, E.

    1988-05-01

    The aim of this work is to understand the comparative structural and intrinsic luminescence properties of the pure matrices α and β - ThBr 4 , and to study the electronic structure by optical spectroscopy of the U 4+ ion in the α-ThBr 4 matrix. 1)Under U.V. excitation, βThBr 4 is intensively fluorescent in the blue-purple and α-ThBr 4 is fluorescent in the red. The main results concerning β-ThBr 4 are the following: -the optical absorption in the U.V. is under the form of a sudden absorption front and for a same temperature, its threshold energy has the same value as the threshold energy of the excitation function and of the photocurrent peak; -the intensity and the life time of the emission decrease when the temperature increases from 300 K until 400 K ( extinction temperature). All the obtained results have been explained either with the molecular orbitals levels of the ThBr 8 4- cluster or with the valence and conduction bands of the pure matrix. 2)The absorption and emission spectra of U 4+ in α-ThBr 4 (in which U 4+ has a point symmetry S 4 ) obtained between 300 K and 4.2 K have allowed to index 30 levels. The calculation of the spectroscopic parameters F k , ξ and B k q has been carried out in symmetry D 2d and S 4 . The comparison of these parameters with those calculated for U 4+ in β-ThBr 4 and β-ThCl 4 show that the global force of the crystalline field is practically the same in the three matrices, but that the structure transformation β→α occurs more on the values of these B k q than on the change of the ligands Br - →Cl - . On the other hand, it has been possible with the α-ThBr 4 matrix, or the β-ThBr 4 and the β-ThCl 4 , to observe the fluorescence spectra of the U 4+ ion (particularly weak phonons energies). (O.M.)

  16. Luminescent properties of UV excitable blue emitting phosphors MSr4(BO3)3:Ce3+ (M = Li and Na)

    International Nuclear Information System (INIS)

    Guo Chongfeng; Ding Xu; Seo, Hyo Jin; Ren Zhaoyu; Bai Jintao

    2011-01-01

    Research highlights: → Novel blue emitting phosphors borate MSr 4 (BO 3 ) 3 (M = Li or Na) were prepared first. → Luminescent properties of phosphors borate MSr 4 (BO 3 ) 3 (M = Li or Na) were investigated extensively as candidates of blue emitting phosphor used for UV excited LED. → The optimal concentrations of dopant Ce 3+ ions in compound MSr 4 (BO 3 ) 3 (M = Li or Na) were determined as 0.05 for Li and x = 0.09 for Na excited by UV light respectively. - Abstract: A series of Ce 3+ doped novel borate phosphors MSr 4 (BO 3 ) 3 (M = Li or Na) were successfully synthesized by traditional solid-state reaction. The crystal structures and the phase purities of samples were characterized by powder X-ray diffraction. The optimal concentrations of dopant Ce 3+ ions in compound MSr 4 (BO 3 ) 3 (M = Li or Na) were determined through the measurements of photoluminescence spectra of phosphors. Ce 3+ doped phosphors MSr 4 (BO 3 ) 3 (M = Li or Na) show strong broad band absorption in UV spectral region and bright blue emission under the excitation of 345 nm light. In addition, the temperature dependences of emission spectra of M 1+x Sr 4-2x Ce x (BO 3 ) 3 (M = Li or Na) phosphors with optimal composition x = 0.05 for Li and x = 0.09 for Na excited under 355 nm pulse laser were also investigated. The experimental results indicate that the M 1+x Sr 4-2x Ce x (BO 3 ) 3 (M = Li or Na) phosphors are promising blue emitting phosphors pumped by UV light.

  17. Overview of the recovery and processing of 233U from the Oak Ridge molten salt reactor experiment (MSRE) remediation activities

    International Nuclear Information System (INIS)

    Del Cul, G.D.; Icenhour, A.S.; Simmons, D.W.; Trowbridge, L.D.; Williams, D.F.; Toth, L.M.; Dai, S.

    2001-01-01

    The Molten Salt Reactor Experiment (MSRE) was operated at Oak Ridge National Laboratory (ORNL) from 1965 to 1969 to test the concept of a high-temperature, homogeneous, fluid-fueled reactor. The discovery that UF 6 and F 2 migrated from the storage tanks into distant pipes and a charcoal bed resulted in significant activities to remove and recover the 233 U and to decommission the reactor. The recovered fissile uranium will be converted into uranium oxide (U 3 O 8 ), which is a suitable form for long-term storage. This publication reports the research and several new developments that were needed to carry out these unique activities. (author)

  18. The oceanic chemistry of the U- and Th-series nuclides

    International Nuclear Information System (INIS)

    Cochran, J.K.

    1982-01-01

    The subject is discussed under the headings: input and removal of U- and Th-series nuclides in the oceans; uranium (input to the oceans; in the coastal ocean; in the open ocean; in sediment pore water; removal from the oceans; sources and sinks of 234 U in the oceans); thorium (scavenging in the deep sea; 230 Th and 231 Pa balance; removal from the coastal and surface ocean); Ra-226 and Ra-228; radon (in surface waters; near bottom 222 Rn as a tracer for vertical mixing); lead-210; polonium-210. (U.K.)

  19. U-Th series nuclides in the Gulf of Mexico

    International Nuclear Information System (INIS)

    Scott, M.R.

    1981-01-01

    A study of U and Th series nuclides is being conducted on sediments from the Gulf of Mexico. Uranium concentrations as a function of depth have been determined, as well as changes in the 234 U/ 238 U activity ratio. The geochemical behavior of uranium in shelf sediments is discussed

  20. U, Th, and Pb isotopes in hot springs on the Juan de Fuca Ridge

    International Nuclear Information System (INIS)

    Chen, J.H.

    1987-01-01

    The concentrations and isotopic compositions of U, Th, and Pb in three hydrothermal fluids from the Juan de Fuca Ridge were determined. The samples consisted of 10.2--57.6% of the pure hydrothermal end-members based on Mg contents. The Pb contents of the samples ranged from 34 to 87 ng/g, U from 1.3 to 3.0 ng/g, and Th from 0.2 to 7.7 pg/g. These samples showed large enrichments of Pb and Th relative to deep-sea water and some depletion of U. They did not show coherent relationships with Mg, however, indicating nonideal mixings between the hot hydrothermal fluids and cold ambient seawater. Particles filtered from these hydrothermal fluids contained significant amounts of Th and Pb which may effectively increase the concentration of these elements in the fluids when acidified. The /sup 234/U//sup 238/U values in all samples show a /sup 234/U enrichment relative to the equilibrium value and have a seawater signature. The Pb isotopic composition of the Juan de Fuca hydrothermal fluids resembles that of 21 0 N East Pacific Rise and has a uniform mid-ocean ridge basalt signature. The hydrothermal systems at oceanic spreading ridges have circulated through a large volume of basalts. Therefore Pb in these fluids may represent the best average value of the local oceanic crust. From the effects of U deposition from seawater to the crust and Pb extraction from rock to the ocean, the U/Pb ratio in the hydrothermally altered oceanic crust may be increased significantly. copyright American Geophysical Union 1987

  1. PWR core design, neutronics evaluation and fuel cycle analysis for thorium-uranium breeding recycle

    International Nuclear Information System (INIS)

    Bi, G.; Liu, C.; Si, S.

    2012-01-01

    This paper was focused on core design, neutronics evaluation and fuel cycle analysis for Thorium-Uranium Breeding Recycle in current PWRs, without any major change to the fuel lattice and the core internals, but substituting the UOX pellet with Thorium-based pellet. The fuel cycle analysis indicates that Thorium-Uranium Breeding Recycle is technically feasible in current PWRs. A 4-loop, 193-assembly PWR core utilizing 17 x 17 fuel assemblies (FAs) was taken as the model core. Two mixed cores were investigated respectively loaded with mixed reactor grade Plutonium-Thorium (PuThOX) FAs and mixed reactor grade 233 U-Thorium (U 3 ThOX) FAs on the basis of reference full Uranium oxide (UOX) equilibrium-cycle core. The UOX/PuThOX mixed core consists of 121 UOX FAs and 72 PuThOX FAs. The reactor grade 233 U extracted from burnt PuThOX fuel was used to fabrication of U 3 ThOX for starting Thorium-. Uranium breeding recycle. In UOX/U 3 ThOX mixed core, the well designed U 3 ThOX FAs with 1.94 w/o fissile uranium (mainly 233 U) were located on the periphery of core as a blanket region. U 3 ThOX FAs remained in-core for 6 cycles with the discharged burnup achieving 28 GWD/tHM. Compared with initially loading, the fissile material inventory in U 3 ThOX fuel has increased by 7% via 1-year cooling after discharge. 157 UOX fuel assemblies were located in the inner of UOX/U 3 ThOX mixed core refueling with 64 FAs at each cycle. The designed UOX/PuThOX and UOX/U 3 ThOX mixed core satisfied related nuclear design criteria. The full core performance analyses have shown that mixed core with PuThOX loading has similar impacts as MOX on several neutronic characteristic parameters, such as reduced differential boron worth, higher critical boron concentration, more negative moderator temperature coefficient, reduced control rod worth, reduced shutdown margin, etc.; while mixed core with U 3 ThOX loading on the periphery of core has no visible impacts on neutronic characteristics compared

  2. Compositional characterization of sintered (U,Th)O2 pellets by EDXRF using fused bead specimens

    International Nuclear Information System (INIS)

    Sanjay Kumar, S.; Dhara, Sangita; Misra, N.L.; Aggarwal, S.K.

    2015-01-01

    Fused bead specimens were used for analyzing sintered (U,Th)O 2 pellets by Energy Dispersive X-Ray Fluorescence (EDXRF) spectrometry. The bead specimens of calibration mixtures U 3 O 8 and ThO 2 were made by fusing them in Lithium Tetraborate/Metaborate fusion mixtures using a fusion bead machine. The EDXRF spectra of these beads were used for making calibration plot for U% determination in (U+Th) amounts. Using these calibration plots and EDXRF spectra of bead of sintered (U,Th)O 2 pellets, U% in these pellets was successfully determined. (author)

  3. Three dimensional neutronic/thermal-hydraulic coupled simulation of MSR in transient state condition

    International Nuclear Information System (INIS)

    Zhou, Jianjun; Zhang, Daling; Qiu, Suizheng; Su, Guanghui; Tian, Wenxi; Wu, Yingwei

    2015-01-01

    Highlights: • Developed a three dimensional neutronic/thermal-hydraulic coupled transient analysis code for MSR. • Investigated the neutron distribution and thermal-hydraulic characters of the core under transient condition. • Analyzed three different transient conditions of inlet temperature drop, reactivity jump and pump coastdown. - Abstract: MSR (molten salt reactor) use liquid molten salt as coolant and fuel solvent, which was the only one liquid reactor of six Generation IV reactor types. As a liquid reactor the physical property of reactor was significantly influenced by fuel salt flow and the conventional analysis methods applied in solid fuel reactors are not applicable for this type of reactors. The present work developed a three dimensional neutronic/thermal-hydraulic coupled code investigated the neutronics and thermo-hydraulics characteristics of the core in transient condition based on neutron diffusion theory and numerical heat transfer. The code consists of two group neutron diffusion equations for fast and thermal neutron fluxes and six group balance equations for delayed neutron precursors. The code was separately validated by neutron benchmark and flow and heat transfer benchmark. Three different transient conditions was analyzed with inlet temperature drop, reactivity jump and pump coastdown. The results provide some valuable information in design and research this kind of reactor

  4. Three dimensional neutronic/thermal-hydraulic coupled simulation of MSR in transient state condition

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, Jianjun [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China); College of Mechanical and Power Engineering, China Three Gorges University, No 8, Daxue road, Yichang, Hubei 443002 (China); Zhang, Daling, E-mail: dlzhang@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China); Qiu, Suizheng; Su, Guanghui; Tian, Wenxi; Wu, Yingwei [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China)

    2015-02-15

    Highlights: • Developed a three dimensional neutronic/thermal-hydraulic coupled transient analysis code for MSR. • Investigated the neutron distribution and thermal-hydraulic characters of the core under transient condition. • Analyzed three different transient conditions of inlet temperature drop, reactivity jump and pump coastdown. - Abstract: MSR (molten salt reactor) use liquid molten salt as coolant and fuel solvent, which was the only one liquid reactor of six Generation IV reactor types. As a liquid reactor the physical property of reactor was significantly influenced by fuel salt flow and the conventional analysis methods applied in solid fuel reactors are not applicable for this type of reactors. The present work developed a three dimensional neutronic/thermal-hydraulic coupled code investigated the neutronics and thermo-hydraulics characteristics of the core in transient condition based on neutron diffusion theory and numerical heat transfer. The code consists of two group neutron diffusion equations for fast and thermal neutron fluxes and six group balance equations for delayed neutron precursors. The code was separately validated by neutron benchmark and flow and heat transfer benchmark. Three different transient conditions was analyzed with inlet temperature drop, reactivity jump and pump coastdown. The results provide some valuable information in design and research this kind of reactor.

  5. Mass spectrometric 230Th-234U-238U dating of the Devils Hole calcite vein

    International Nuclear Information System (INIS)

    Ludwig, K.R.; Simmons, K.R.; Szabo, B.J.; Riggs, A.C.; Winograd, I.J.; Landwehr, J.M.; Hoffman, R.J.

    1992-01-01

    The Devils Hole calcite vein contains a long-term climatic record, but requires accurate chronologic control for its interpretation. Mass-spectrometric U-series ages for samples from core DH-11 yielding 230 Th ages with precisions ranging from less than 1,000 years (2σ) for samples younger than ∼140 ka (thousands of years ago) to less than 50,000 years for the oldest samples (∼566 ka). The 234 U/ 238 U ages could be determined to a precision of ∼20,000 years for all ages. Calcite accumulated continuously from 566 ka until ∼60 ka at an average rate of 0.7 millimeter per 10 3 years. The precise agreement between replicate analyses and the concordance of the 230 Th/ 238 U and 234 U/ 238 U ages for the oldest samples indicate that the DH-11 samples were closed systems and validate the dating technique in general

  6. New zircon (U-Th)/He and U/Pb eruption age for the Rockland tephra, western USA

    Science.gov (United States)

    Coble, Matthew A.; Burgess, Seth D.; Klemetti, Erik W.

    2017-09-01

    Eruption ages of a number of prominent Quaternary volcanic deposits remain inaccurately and/or imprecisely constrained, despite their importance as regional stratigraphic markers in paleo-environment reconstruction and as evidence of climate-altering eruptions. Accurately dating volcanic deposits presents challenging analytical considerations, including poor radiogenic yield, scarcity of datable minerals, and contamination of crystal populations by magma, eruption, and transport processes. One prominent example is the Rockland tephra, which erupted from the Lassen Volcanic Center in the southern Cascade arc. Despite a range in published eruption ages from 0.40 to 0.63 Ma, the Rockland tephra is extensively used as a marker bed across the western United States. To more accurately and precisely constrain the age of the Rockland tephra-producing eruption, we report U/Pb crystallization dates from the outermost ∼2 μm of zircon crystal faces (surfaces) using secondary ion mass spectrometry (SIMS). Our new weighted mean 238U/206Pb age for Rockland tephra zircon surfaces is 0.598 ± 0.013 Ma (2σ) and MSWD = 1.11 (mean square weighted deviation). As an independent test of the accuracy of this age, we obtained new (U-Th)/He dates from individual zircon grains from the Rockland tephra, which yielded a weighted mean age of 0.599 ± 0.012 Ma (2σ, MSWD = 5.13). We also obtained a (U-Th)/He age of 0.628 ± 0.014 Ma (MSWD = 1.19) for the Lava Creek Tuff member B, which was analyzed as a secondary standard to test the accuracy of the (U-Th)/He technique for Quaternary tephras, and to evaluate assumptions made in the model-age calculation. Concordance of new U/Pb and (U-Th)/He zircon ages reinforces the accuracy of our preferred Rockland tephra eruption age, and confirms that zircon surface dates sample zircon growth up to the time of eruption. We demonstrate the broad applicability of coupled U/Pb zircon-surface and single-grain zircon (U-Th)/He geochronology to accurate

  7. U and Th thin film neutron dosimetry for fission-track dating: application to the age standard Moldavite

    International Nuclear Information System (INIS)

    Iunes, P.J.; Bigazzi, G.; Hadler Neto, J.C.; Laurenzi, M.A.; Balestrieri, M.L.; Norelli, P.; Osorio Araya, A.M.; Guedes, S.; Tello S, C.A.; Paulo, S.R.; Moreira, P.A.F.P.; Palissari, R.; Curvo, E.A.C.

    2005-01-01

    Neutron dosimetry based on U and Th thin films was used for fission-track dating of the age standard Moldavite, the central European tektite, from the Middle Miocene deposit of Jankov (southern Bohemia, Czech Republic). Our fission-track age (13.98+/-0.58Ma) agrees with a recent 40 Ar/ 39 Ar age, 14.34+/-0.04Ma, based on several determinations on Moldavites from different sediments, including the Jankov deposit. This result indicates that the U and Th thin film neutron dosimetry represents a reliable alternative for an absolute approach in fission-track dating

  8. Levels in 223Th populated by α decay of 227U

    Science.gov (United States)

    Kalaninová, Z.; Antalic, S.; Heßberger, F. P.; Ackermann, D.; Andel, B.; Kindler, B.; Laatiaoui, M.; Lommel, B.; Maurer, J.

    2015-07-01

    Levels in 223Th populated by the α decay of 227U were investigated using α -γ decay spectroscopy. The 227U isotope was produced in the fusion-evaporation reaction 22Ne +208Pb at the velocity filter separator for heavy-ion reaction products at Gesellschaft für Schwerionenforschung in Darmstadt (Germany). Several new excited levels and γ transitions were identified in 223Th . An improved α -decay scheme of 227U was suggested. The experimental α -decay energy spectrum of 227U was compared with the Monte Carlo simulation performed using the toolkit geant4.

  9. Los Alamos National Laboratory Site Integrated Management plan, uranium 233 storage and disposition. Volume 1: Project scope and description

    International Nuclear Information System (INIS)

    Nielsen, J.B.; Erickson, R.

    1997-01-01

    This Site Integration Management plan provides the Los Alamos Response to the Defense Nuclear Facility Safety Board (DNFSB) Recommendation 97-1. This recommendation addresses the safe storage and management of the Departments uranium 233 ( 233 U) inventory. In the past, Los Alamos has used 233 U for a variety of different weapons related projects. The material was used at a variety of sites in varying quantities. Now, there is a limited need for this material and the emphasis has shifted from use to storage and disposition of the material. The Los Alamos program to address the DNFSB Recommendation 97-1 has two emphases. First, take corrective action to address near term deficiencies required to provide safe interim storage of 233 U. Second, provide a plan to address long term storage and disposition of excess inventory at Los Alamos

  10. Fission Cross-section Measurements of (233)U, (245)Cm and (241,243)Am at CERN n_TOF Facility

    CERN Document Server

    Calviani, M; Andriamonje, S; Chiaveri, E; Vlachoudis, V; Colonna, N; Meaze, M H; Marrone, S; Tagliente, G; Terlizzi, R; Belloni, F; Abbondanno, U; Fujii, K; Milazzo, P M; Moreau, C; Aerts, G; Berthoumieux, E; Dridi, W; Gunsing, F; Pancin, J; Perrot, L; Plukis, A; Alvarez, H; Duran, I; Paradela, C; Alvarez-Velarde, F; Cano-Ott, D; Gonzalez-Romero, E; Guerrero, C; Martinez, T; Villamarin, D; Vicente, M C; Andrzejewski, J; Marganiec, J; Assimakopoulos, P; Karadimos, D; Karamanis, D; Papachristodoulou, C; Patronis, N; Audouin, L; David, S; Ferrant, L; Isaev, S; Stephan, C; Tassan-Got, L; Badurek, G; Jericha, E; Leeb, H; Oberhummer, H; Pigni, M T; Baumann, P; Kerveno, M; Lukic, S; Rudolf, G; Becvar, F; Krticka, M; Calvino, F; Capote, R; Carrillo De Albornoz, A; Marques, L; Salgado, J; Tavora, L; Vaz, P; Cennini, P; Dahlfors, M; Ferrari, A; Gramegna, F; Herrera-Martinez, A; Kadi, Y; Mastinu, P; Praena, J; Sarchiapone, L; Wendler, H; Chepel, V; Ferreira-Marques, R; Goncalves, I; Lindote, A; Lopes, I; Neves, F; Cortes, G; Poch, A; Pretel, C; Couture, A; Cox, J; O'brien, S; Wiescher, M; Dillman, I; Heil, M; Kappeler, F; Mosconi, M; Plag, R; Voss, F; Walter, S; Wisshak, K; Dolfini, R; Rubbia, C; Domingo-Pardo, C; Tain, J L; Eleftheriadis, C; Savvidis, I; Frais-Koelbl, H; Griesmayer, E; Furman, W; Konovalov, V; Goverdovski, A; Ketlerov, V; Haas, B; Haight, R; Reifarth, R; Igashira, M; Koehler, P; Kossionides, E; Lampoudis, C; Lozano, M; Quesada, J; Massimi, C; Vannini, G; Mengoni, A; Oshima, M; Papadopoulos, C; Vlastou, R; Pavlik, A; Pavlopoulos, P; Plompen, A; Rullhusen, P; Rauscher, T; Rosetti, M; Ventura, A

    2011-01-01

    Neutron-induced fission cross-sections of minor actinides have been measured using the n_TOF white neutron source at CERN, Geneva, as part of a large experimental program aiming at collecting new data relevant for nuclear astrophysics and for the design of advanced reactor systems. The measurements at n_TOF take advantage of the innovative features of the n_TOF facility, namely the wide energy range, high instantaneous neutron flux and good energy resolution. Final results on the fission cross-section of 233U, 245Cm and 243Am from thermal to 20 MeV are here reported, together with preliminary results for 241Am. The measurement have been performed with a dedicated Fast Ionization Chamber (FIC), a fission fragment detector with a very high efficiency, relative to the very well known cross-section of 235U, measured simultaneously with the same detector.

  11. Measurements of 234U, 238U and 230Th in excreta of uranium-mill crushermen

    International Nuclear Information System (INIS)

    Fisher, D.R.; Jackson, P.O.; Brodacynski, G.G.; Scherpelz, R.I.

    1982-07-01

    Uranium and thorium levels in excreta of uranium mill crushermen who are routinely exposed to airborne uranium ore dust were measured. The purpose was to determine whether 230 Th was preferentially retained over either 234 U or 238 U in the body. Urine and fecal samples were obtained from fourteen active crushermen with long histories of exposure to uranium ore dust, plus four retired crushermen and three control individuals for comparison. Radiochemical procedures were used to separate out the uranium and thorium fractions, which were then electroplated on stainless steel discs and assayed by alpha spectrometry. Significantly greater activity levels of 234 U and 238 U were measured in both urine and fecal samples obtained from uranium mill crushermen, indicating that uranium in the inhaled ore dust was cleared from the body with a shorter biological half-time than the daughter product 230 Th. The measurements also indicated that uranium and thorium separate in vivo and have distinctly different metabolic pathways and transfer rates in the body. The appropriateness of current ICRP retention and clearance parameters for 230 Th in ore dust is questioned

  12. Comparative study of α and β-ThBr4: structure and luminescence. Spectroscopy of U4+ in α-ThBr4

    International Nuclear Information System (INIS)

    Simoni, E.

    1988-05-01

    UV absorption of β-ThBr 4 : presents a plain absorption front and for the same temperature the threshold energy has the same value than the threshold energy of excitation function and photocurrent peak. Emission intensity and lifetime decrease when temperature increases from 300 K to 400K (extinction temperature). Results are interpreted either by molecular orbital levels of the ThBr 8 4- cluster or either by conduction and valence bands of the matrix above. Absorption and emission spectra of U 4+ in α-ThBr 4 (where U 4+ has a S 4 symmetry) between 300 K and 4.2 K allow indexation of 30 levels. Spectroscopic parameters are calculated in D 2d and S 4 symmetry. Comparison of these parameters with those of U 4+ in β-ThBr 4 and β-ThCl 4 shows that crystal field force is practically the same in the three matrices but the structure transformation from β to α has more influence on B q k than ligand change from Br - to Cl - . Owing to very low phonon energy, fluorescence spectra of U 4+ is easy to observe in α-ThBr 4 as it is in β-ThBr 4 and ThCl 4 [fr

  13. Crystallographic Study of U-Th bearing minerals in Tranomaro, Anosy Region-Madagascar

    International Nuclear Information System (INIS)

    Sahoa, F.E.; Rabesiranana, N.; Raoelina Andriambololona; Geckeis, H.; Marquardt, C.; Finck, K.

    2011-01-01

    As an alternative to conventional fossil fuel, there is a renewed interest in the nuclear fuel to support increasing energy demand. New studies are then undertaken to characterize Madagascar U-Th bearing minerals. This is the case for the urano-thorianite bearing pyroxenites in the south East of Madagascar. In this region, several quarries were abandoned, after being mined by the French Atomic Energy Commission (C.E.A) in the fifties and sixties and are now explored by new mining companies. For this purpose, seven U-Th bearing mineral samples from old abandoned uranium quarries in Tranomaro, Amboasary Sud, Madagascar, have been collected. To determine the mineral microstructure, they were investigated for qualitative and quantitative identification of crystalline compounds using X-ray powder diffraction analytical method (XRD). Results showed that the U and Th compounds, as minor elements, are present in various crystalline structures. This is important to understand their environmental behaviours, in terms of crystallographic dispersion of U-Th minerals and their impacts on human health.

  14. The Anti-Oxidant Defense System of the Marine Polar Ciliate Euplotes nobilii: Characterization of the MsrB Gene Family

    Directory of Open Access Journals (Sweden)

    Francesca Ricci

    2017-01-01

    Full Text Available Organisms living in polar waters must cope with an extremely stressful environment dominated by freezing temperatures, high oxygen concentrations and UV radiation. To shed light on the genetic mechanisms on which the polar marine ciliate, Euplotes nobilii, relies to effectively cope with the oxidative stress, attention was focused on methionine sulfoxide reductases which repair proteins with oxidized methionines. A family of four structurally distinct MsrB genes, encoding enzymes specific for the reduction of the methionine-sulfoxide R-forms, were identified from a draft of the E. nobilii transcriptionally active (macronuclear genome. The En-MsrB genes are constitutively expressed to synthesize proteins markedly different in amino acid sequence, number of CXXC motifs for zinc-ion binding, and presence/absence of a cysteine residue specific for the mechanism of enzyme regeneration. The En-MsrB proteins take different localizations in the nucleus, mitochondria, cytosol and endoplasmic reticulum, ensuring a pervasive protection of all the major subcellular compartments from the oxidative damage. These observations have suggested to regard the En-MsrB gene activity as playing a central role in the genetic mechanism that enables E. nobilii and ciliates in general to live in the polar environment.

  15. Production of Thorium-229 at the ORNL High Flux Isotope Reactor

    International Nuclear Information System (INIS)

    Boll, Rose Ann; Garland, Marc A.; Mirzadeh, Saed

    2008-01-01

    The investigation of targeted cancer therapy using -emitters has developed considerably in recent years and clinical trials have generated promising results. In particular, the initial clinical trials for treatment of acute myeloid leukemia have demonstrated the effectiveness of the -emitter 213Bi in killing cancer cells. Pre-clinical studies have also shown the potential application of both 213Bi and its 225Ac parent radionuclide in a variety of cancer systems and targeted radiotherapy. Bismuth-213 is obtained from a radionuclide generator system from decay of the 10-d 225Ac parent, a member of the 7340-y 229Th chain. Currently, 233U is the only viable source for high purity 229Th; however, due to increasing difficulties associated with 233U safeguards, processing additional 233U is presently unfeasible. The recent decision to downblend and dispose of enriched 233U further diminished the prospects for extracting 229Th from 233U stock. Nevertheless, the anticipated growth in demand for 225Ac may soon exceed the levels of 229Th (∼40 g or ∼8 Ci; ∼80 times the current ORNL 229Th stock) present in the aged 233U stockpile. The alternative routes for the production of 229Th, 225Ra and 225Ac include both reactor and accelerator approaches. Here, we describe production of 229Th via neutron transmutation of 226Ra targets in the ORNL High Flux Isotope Reactor (HFIR).

  16. Radioisotope dilution analyses of geological samples using 236U and 229Th

    International Nuclear Information System (INIS)

    Rosholt, J.N.

    1984-01-01

    The use of 236 U and 229 Th in alpha spectrometric measurements has some advantages over the use of other tracers and measurement techniques in isotope dilution analyses of most geological samples. The advantages are: 1) these isotopes do not occur in terrestrial rocks, 2) they have negligible decay losses because of their long half lives, 3) they cause minimal recoil contamination to surface-barrier detectors, 4) they allow for simultaneous determination of the concentration and isotopic composition of uranium and thorium in a variety of sample types, and 5) they allow for simple and constant corrections for spectral interferences, 0.5% of the 238 U activity is subtracted for the contribution of 235 U in the 236 U peak and 1% of the 229 Th activity is subtracted from the 230 Th activity. Disadvantages in using 236 U and 229 Th are: 1) individual separates of uranium and thorium must be prepared as very thin sources for alpha spectrometry, 2) good resolution in the spectrometer system is required for thorium isotopic measurements where measurement times may extend to 300 h, and 3) separate calibrations of the 236 U and 229 Th spike solution with both uranium and thorium standards are required. The use of these tracers in applications of uranium-series disequilibrium studies has simplified the measurements required for the determination of the isotopic composition of uranium and thorium because of the minimal corrections needed for alpha spectral interferences. (orig.)

  17. Radioisotope dilution analyses of geological samples using 236U and 229Th

    Science.gov (United States)

    Rosholt, J.N.

    1984-01-01

    The use of 236U and 229Th in alpha spectrometric measurements has some advantages over the use of other tracers and measurement techniques in isotope dilution analyses of most geological samples. The advantages are: (1) these isotopes do not occur in terrestrial rocks, (2) they have negligible decay losses because of their long half lives, (3) they cause minimal recoil contamination to surface-barrier detectors, (4) they allow for simultaneous determination of the concentration and isotopic composition of uranium and thorium in a variety of sample types, and (5) they allow for simple and constant corrections for spectral inferences, 0.5% of the 238U activity is subtracted for the contribution of 235U in the 236U peak and 1% of the 229Th activity is subtracted from the 230Th activity. Disadvantages in using 236U and 229Th are: (1) individual separates of uranium and thorium must be prepared as very thin sources for alpha spectrometry, (2) good resolution in the spectrometer system is required for thorium isotopic measurements where measurement times may extend to 300 h, and (3) separate calibrations of the 236U and 229Th spike solution with both uranium and thorium standards are required. The use of these tracers in applications of uranium-series disequilibrium studies has simplified the measurements required for the determination of the isotopic composition of uranium and thorium because of the minimal corrections needed for alpha spectral interferences. ?? 1984.

  18. Engineering evaluation/cost analysis for the 233-S Plutonium Concentration Facility

    International Nuclear Information System (INIS)

    Rugg, J.E.

    1996-08-01

    The 100, 200, 300 and 1100 Areas of the Hanford Site were placed on the U. S. Environmental Protection Agency's National Priorities List in November 1989 under the Comprehensive Environmental Response, Compensation, and Liability Act of 1980 (CERCLA). Located in the 200 Area is the deactivated 233-S Plutonium Concentration Facility (used in the REDOX process). The facility has undergone severe degradation due to exposure to extreme weather conditions. An expedited response is proposed to ensure protection of human health and the environment. The Department of Energy, Richland Operations Office (RL) in cooperation with the Washington State Department of Ecology, has prepared this Engineering Evaluation/Cost Analysis pursuant to CERCLA. Based on the evaluation, RL has determined that hazardous substances in the 233-S Facility may present a potential threat to human health or the environment, and that an expedited removal action is warranted for decommissioning of the facility

  19. Rates of carbonate soil evolution from carbon, U- and Th-series isotope studies: Example of the Astian sands (SE France)

    Science.gov (United States)

    Barbecot, Florent; Ghaleb, Bassam; Hillaire-Marcel, Claude

    2015-04-01

    In carbonate rich soils, C-isotopes (14C, 13C) and carbonate mass budget may inform on centennial to millennial time scale dissolution/precipitation processes and weathering rates, whereas disequilibria between in the U- and Th-decay series provide tools to document high- (228Ra-228Th-210Pb) to low- (234U, 230Th, 231Pa, 226Ra) geochemical processes rate, covering annual to ~ 1Ma time scales, governing both carbonate and silicate soil fractions. Because lithology constitutes a boundary condition, we intend to illustrate the behavior of such isotopes in soils developed over Astian sands formation (up to ~ 30% carbonate) from the Béziers area (SE France). A >20 m thick unsaturated zone was sampled firstly along a naturally exposed section, then in a cored sequence. Geochemical and mineralogical analyses, including stable isotopes and 14C-measurements, were complemented with 228U, 234U, 230Th, 226Ra, 210Pb and 228Th, 232Th measurements. Whereas the upper 7 m depict geochemical and isotopic features forced by dissolution/precipitation processes leading to variable radioactive disequilibria, but overall deficits in more soluble elements of the decay series, the lower part of the sequence shows strong excesses in 234U and 230Th over parent isotopes (i.e., 238U and 234U, respectively). These features might have been interpreted as the result of successive phases of U-loss and gains. However, 226Ra and 230Th are in near-equilibrium, thus leading to conclude at a more likely slow enrichment process in both 234Th(234U) and 230Th, which we link to dissolved U-decay during groundwater recharge events. In addition, 210Pb deficits (vs parent 226Ra) are observed down to 12 m along the natural outcropping section and below the top-soil 210Pb-excess in the cored sequence, due to gaseous 222Rn-diffusion over the cliff outcrop. Based on C-isotope and chemical analysis, reaction rates at 14C-time scale are distinct from those estimates at the short- or long-lived U-series isotopes

  20. U-Th-Pb systematics in zircon and titanite

    International Nuclear Information System (INIS)

    Zartman, R.E.; Kwak, L.M.

    1990-01-01

    U-Th-Pb isotopic analyses of zircon and titanite were made for two core samples of granite from borehole ATK-1 drilled into the Eye-Dashwa Lakes pluton. One of the samples from near the bottom of the hole (990.97 to 996.78 m) yielded zircon and titanite that were slightly to severely disturbed isotopically. Eight fractions of zircon give an upper concordia intercept age of 2625 ± 16 Ma (MSWD = 34), which, based on an evaluation of the more concordant data points and on other geochronological results, is interpreted as being slightly too young. The time of crystallization is probably better approximated by the 207 Pb/ 206 Pb age of 2665 Ma determined on a slightly (∼8 percent) discordant titanite. The other sample from near the surface (3.85 to 9.61 m) generally revealed even more severely disturbed isotopic systematics for both zircon and titanite. The complex nature of the disturbances probably resulted from the penetration of meteoritic water into rock already modified by post-crystallization hydrothermal alteration. Nuclide migration occurred in both minerals -- during the Middle or Late Proterozoic for the zircon and during the modern weathering cycle for the titanite. Material balance calculations are used to demonstrate a recent relative gain of radiogenic Pb and/or loss of Th and U from the freshest-looking, least-altered titanite by exchange with altered, leucoxenite-bearing titanite

  1. Evaluation of Codisposal Viability for TH/U Carbide (Fort Saint Vrain HTGR) DOE-Owned Fuel

    International Nuclear Information System (INIS)

    Radulescu, H.

    2001-01-01

    There are more than 250 forms of US Department of Energy (DOE)-owned spent nuclear fuel (SNF). Due to the variety of the spent nuclear fuel, the National Spent Nuclear Fuel Program has designated nine representative fuel groups for disposal criticality analyses based on fuel matrix, primary fissile isotope, and enrichment. The Fort Saint Vrain reactor (FSVR) SNF has been designated as the representative fuel for the Th/U carbide fuel group. The FSVR SNF consists of small particles (spheres of the order of 0.5-mm diameter) of thorium carbide or thorium and high-enriched uranium carbide mixture, coated with multiple, thin layers of pyrolytic carbon and silicon carbide, which serve as miniature pressure vessels to contain fission products and the U/Th carbide matrix. The coated particles are bound in a carbonized matrix, which forms fuel rods or ''compacts'' that are loaded into large hexagonal graphite prisms. The graphite prisms (or blocks) are the physical forms that are handled in reactor loading and unloading operations, and which will be loaded into the DOE standardized SNF canisters. The results of the analyses performed will be used to develop waste acceptance criteria. The items that are important to criticality control are identified based on the analysis needs and result sensitivities. Prior to acceptance to fuel from the Th/U carbide fuel group for disposal, the important items for the fuel types that are being considered for disposal under the Th/U carbide fuel group must be demonstrated to satisfy the conditions determined in this report

  2. Evaluation of Codisposal Viability for TH/U Carbide (Fort Saint Vrain HTGR) DOE-Owned Fuel

    Energy Technology Data Exchange (ETDEWEB)

    H. radulescu

    2001-09-28

    There are more than 250 forms of US Department of Energy (DOE)-owned spent nuclear fuel (SNF). Due to the variety of the spent nuclear fuel, the National Spent Nuclear Fuel Program has designated nine representative fuel groups for disposal criticality analyses based on fuel matrix, primary fissile isotope, and enrichment. The Fort Saint Vrain reactor (FSVR) SNF has been designated as the representative fuel for the Th/U carbide fuel group. The FSVR SNF consists of small particles (spheres of the order of 0.5-mm diameter) of thorium carbide or thorium and high-enriched uranium carbide mixture, coated with multiple, thin layers of pyrolytic carbon and silicon carbide, which serve as miniature pressure vessels to contain fission products and the U/Th carbide matrix. The coated particles are bound in a carbonized matrix, which forms fuel rods or ''compacts'' that are loaded into large hexagonal graphite prisms. The graphite prisms (or blocks) are the physical forms that are handled in reactor loading and unloading operations, and which will be loaded into the DOE standardized SNF canisters. The results of the analyses performed will be used to develop waste acceptance criteria. The items that are important to criticality control are identified based on the analysis needs and result sensitivities. Prior to acceptance to fuel from the Th/U carbide fuel group for disposal, the important items for the fuel types that are being considered for disposal under the Th/U carbide fuel group must be demonstrated to satisfy the conditions determined in this report.

  3. Thorex reprocessing characterization

    International Nuclear Information System (INIS)

    1978-11-01

    The purpose of this report is to bring together, in highly condensed form, information which would need to be considered in planning a commercial reprocessing plant for recovering 233 U-Th reactor fuel. This report does not include a discussion of process modifications which would be required for thorium-base fuels that contain plutonium (such as would be required for thorium fuels containing 235 U or 233 U denatured with 238 U). It is the intent of this paper to address only the basic Thorex process for treating 233 U-Th fuels. As will be pointed out, the degree of development of the various proposed operations varies widely, from preliminary laboratory experiments for the dissolution of Zircaloy-clad thoria to engineering scale demonstration of the recovery of moderately irradiated thorium by a solvent extraction process (Thorex)

  4. Transport of Th(IV) and U(VI) through barium silico-phosphate composite membrane using electric field

    International Nuclear Information System (INIS)

    Zaki, E.E.

    2002-01-01

    The present paper describes the preparation of a novel barium silico-phosphate filter paper supported membrane. It is based on precipitation reaction of barium silico-phosphate on the outer surface and in the interstices of a filter paper by means of electrodialysis. The main physical and electrical properties of the membrane are given and its electrodialysis behaviour is assessed for Th(IV) and U(VI). The transport of Th(IV) in presence of U(VI) was studied. The cationic fluxes of Th(IV) and U(VI) were found to be 1.2 x 10 -8 and 6.5 x 10 -9 g eq cm -2 s -1 , respectively. Transport of Th(IV) and U(VI) in presence of EDTA was investigated. The cationic flux of U(VI) is found to be 9.8 x 10 -9 g eq cm -2 s -1 at a current density of 25 mA/cm 2 . A comparative study on the electro osmotic effect was carried out using the developed membrane and commercially available Nafion membranes. In this context, different parameters like current density, electrolyte concentration, etc. were investigated. The electro-osmotic permeability coefficient, D e , of Th(IV) through barium silico-phosphate and Nafion membranes were 6.9 x 10 -2 and 1.0 x 10 -2 cm 3 /As, respectively. It can be concluded that inorganic membranes have very marked electro-osmotic properties unlike their organic counterparts. (orig.)

  5. Chain and independent fission product yields adjusted to conform with physical conservation laws. Part 2

    International Nuclear Information System (INIS)

    Crouch, E.A.C.

    1976-01-01

    Previously reported adjustments to the chain yields and independent yields for the thermal neutron induced fission of 233 U, 235 U, 239 Pu and 241 Pu, the fast neutron induced fission of 232 Th, 233 U, 235 U, 238 U, 239 Pu, 240 Pu and 241 Pu, and the 14 MeV neutron induced fission of 232 Th, 233 U, 235 U and 238 U, have been recalculated using the principle of least squares. The adjustments to the chain yields so found are much smaller than those previously reported. (author)

  6. Complexation of Eu(III), Th(IV) and U(VI) by humic substances

    International Nuclear Information System (INIS)

    Moulin, V.; Reiller, P.; Dautel, C.; Plancque, G.; Laszak, I.; Moulin, C.

    1999-01-01

    Complexation of actinides by humic substances has been studied by different techniques depending on the actinide and its oxidation state. For trivalent actinide (using a rare earth element, Eu as an analogue of trivalent actinide), Time-Resolved Laser-Induced Fluorescence (TRLIF) has been retained as a method for direction speciation at low level. By varying pH and physicochemical conditions (absence of carbonate ions) and at fixed ionic strength, it is possible together to identify spectrally and temporally, all the hydroxo and carbonato complexes. This approach has also been retained for U(VI) as a model of hexavalent actinide, for which hydroxo complexes have been characterized by TRLIF (the simple carbonato complexes are not fluorescent). In the case of U(VI), titrations hy humic acids of U(VI) solutions at various pH have allowed to characterize organic complexes formed with U(VI): single complexes (UO 2 HA) and mixed complexes (UO 2 (OH) 3 HA). The impact on U(VI) speciation has then been identified. In the case of Th(IV) as a model of tetravalent actinides, a competitive method has been used to obtain data on the Th-HA system by studying the ternary system silica colloid/HA/Th at constant pH (Schubert method). Apparent interaction constants have been calculated depending on Th hydrolysis constants used. A study of the system Th/HA/silica has a function of pH and for different HA concentrations has shown the strong complexing character of humic acids towards Th in the pH range 4-9. (orig.)

  7. Thorium utilization in a small long-life HTR. Part I: Th/U MOX fuel blocks

    Energy Technology Data Exchange (ETDEWEB)

    Ding, Ming, E-mail: dingm2005@gmail.com [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB, Delft (Netherlands); Harbin Engineering University, Nantong Street 145, 150001 Harbin (China); Kloosterman, Jan Leen, E-mail: j.l.kloosterman@tudelft.nl [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB, Delft (Netherlands)

    2014-02-15

    Highlights: • We propose thorium MOX (TMOX) fuel blocks for a small block-type HTR. • The TMOX fuel blocks with low-enriched uranium are recommended. • More thorium decreases the reactivity swing of the TMOX fuel blocks. • Thorium reduces the negative temperature coefficient of the TMOX fuel blocks. • Thorium increases the conversion ratio of the TMOX fuel blocks. - Abstract: The U-Battery is a small, long-life and transportable high temperature gas-cooled reactor (HTR). The neutronic features of a typical fuel block with uranium and thorium have been investigated for a application of the U-Battery, by parametrically analyzing the composition and geometric parameters. The type of fuel block is defined as Th/U MOX fuel block because uranium and thorium are assumed to be mixed in each fuel kernel as a form of (Th,U)O{sub 2}. If the initially loaded mass of U-235 is mostly consumed in the early period of the lifetime of Th/U MOX fuel block, low-enriched uranium (LEU) as ignited fuel will not largely reduce the neutronic performance of the Th/U MOX fuel block, compared with high-enriched uranium. The radii of fuel kernels and fuel compacts and packing fraction of TRISO particles determine the atomic ratio of the carbon to heavy metal. When the ratio is smaller than 400, the difference among them due to double heterogeneous effects can be neglected for the Th/U MOX fuel block. In the range between 200 and 400, the reactivity swing of the Th/U MOX fuel block during 10 years is sufficiently small. The magnitude of the negative reactivity temperature coefficients of the Th/U MOX fuel block decreases by 20–45%, which is positive to reduce temperature defect of the Th/U MOX fuel block. The conversion ratio (CR) of the fuel increases from 0.48 (typical CR of the LEU-fueled U-Battery) to 0.78. The larger conversion ratio of the Th/U MOX fuel block reduces the reactivity swing during 10 years for the U-Battery.

  8. U, TH and lanthanides in street soils of Sao Paulo city, Brazil

    International Nuclear Information System (INIS)

    Ticianelli, R.B.; Ribeiro, A.P.; Figueiredo, A.M.G.; Zanh, G.S.

    2013-01-01

    The study of lanthanide distribution in urban environments has become of interest over the last years, due to the increased industrial use of these elements. Sao Paulo is the 6th largest metropolitan region of the world, with about 20 million inhabitants in its metropolitan area, more than 9 million motor vehicles and intense industrial activity. There is little information on U, Th, and lanthanide contents in urban soils, and there are as of yet no reference values for these elements in soils of Sao Paulo city. The present study aimed to determine U, Th and lanthanide concentrations in soils adjacent to avenues of highly dense traffic downtown in Sao Paulo city, to assess their possible sources and potential environmental impacts. The analytical technique employed was Instrumental Neutron Activation Analysis (INAA). Th and U levels ranged from 4.0 to 37.0 mg kg -1 and from 1.6 to 18.7 mg kg -1 , respectively. These values are higher than literature values for U and Th in Brazilian superficial soils. The results obtained for the lanthanides indicate enrichment in La and Ce. However, a possible anthropogenic source should be investigated since high background values of these elements may be associated to the natural geological composition of the soils. (author)

  9. Microstructural evolution and thermophysical property evaluation of Th-U alloys

    International Nuclear Information System (INIS)

    Das, Santanu; Kaity, Santu; Bannerjee, Joydipto; Kumar, Raj; Roy, S.B.; Chaudhari, G.P.; Daniel, B.S.S.

    2015-01-01

    Thorium-uranium alloy fuel has not received much research attention mainly because of easy availability of uranium and military incentive offered by U-Pu cycle. Moreover, (i) lack of a consistent systematic effort to develop the alloys and define the limitations of these fuels, (ii) dearth of initiatives to define its microstructures that can result from composition and fabrication variables are prime reasons for this system not having witnessed much developmental research endeavour. Hence, it seems prudent to explore few compositions selected from thorium-uranium phase diagram keeping two primary objectives in view viz. (i) establishing its microstructural features and to study the variations in those, if any, brought about by processing variables etc. and (ii) to assess few thermal properties relevant to fuel applications. This experimental work aims at addressing gap in research on thorium-uranium alloys. Selected compositions of thorium-uranium alloy have been taken for microstructural study and evaluation of thermophysical properties. Based on the microstructural features and thermophysical property evaluation it is seen that high thorium Th-U alloys have appreciable thermal conductivity and low thermal expansion coefficient. It can reasonably be concluded that high thorium Th-U alloy can be used for possible nuclear fuel application in reactors provided other factors (e.g. reactor physics, post irradiation examinations etc.) are also seen to be favourable. (author)

  10. Measurements of the prompt neutron spectra in 233U, 235U, 239Pu thermal neutron fission in the energy range of 0.01-5 MeV and in 252Cf spontaneous fission in the energy range of 0.01-10 MeV

    International Nuclear Information System (INIS)

    Starostov, B.I.; Semenov, A.F.; Nefedov, V.N.

    1978-01-01

    The measurement results on the prompt neutron spectra in 233 U, 235 U, 239 Pu thermal neutron fission in the energy range of 0.01-5 MeV and in 252 Cf spontaneous fission in the energy range of 0.01-10 MeV are presented. The time-of-flight method was used. The exceeding of the spectra over the Maxwell distributions is observed at E 252 Cf neutron fission spectra. The spectra analysis was performed after normalization of the spectra and corresponding Maxwell distributions for one and the same area. In the range of 0.05-0.22 MeV the yield of 235 U + nsub(t) fission neutrons is approximately 8 and approximately 15 % greater than the yield of 252 Cf and 239 Pu + nsub(t) fission neutrons, respectively. In the range of 0.3-1.2 MeV the yield of 235 U + nsub(t) fission neutrons is 8 % greater than the fission neutron yield in case of 239 Pu + nsub(t) fission. The 235 U + nsub(t) and 233 U + nsub(t) fission neutron spectra do not differ from one another in the 0.05-0.6 MeV range

  11. U/Th-isotopes as natural analogues for the mobility of actinides in granitic rocks

    International Nuclear Information System (INIS)

    Mengel, K.; Gerdes, A.

    2001-01-01

    The short-lived decay products of 238 U ( 234 U and 230 Th) can be used as natural analogues for actinides in a hard rock repository. Their mobility in the past may serve as a key for understanding actinide migration in the future. For generally old calcites of the HRL Aespoethe age of disturbance of 238 U/ 234 U and 234 U/ 230 Th activity ratios ranges from 30 000 to 436 000 years at degrees of disturbance ranging from 0.5 to 6.7. The results obtained imply that during the past 440 000 years U was mobile throughout the tunnel sections of the HRL Aespoeinvestigated here. For the FL Grimsel, the disequilibrium states of the 234 U/ 238 U and 230 Th/ 234 U activity ratios in fracture minerals (calcites silicates) also imply that the reactions causing isotopic disturbances have occurred within the past 500 000 years. The U/Th-isotope data of both the samples from the HRL Aespoeand the FL Grimsel have in common the mobilization of U in secondary fracture minerals by migrating solutions within the past 500 000 years. As for the question of a final disposal of radioactive waste in granite host rocks, the transport of U - and thus of similarly behaving actinides - in migrating underground solutions can therefore not be ruled out, if suitable hydraulic systems are considered. (orig.)

  12. Hydrogeochemical utilization of natural isotopes from U(4n+2) and Th(4n) series at Morro do Ferro, Pocos de Caldas (MG)

    International Nuclear Information System (INIS)

    Bonotto, D.M.

    1991-01-01

    Uranium and thorium isotopic analysis were performed on well spoils of the main ore body at Morro do Ferro, Pocos de Caldas (MG), using groundwater from several boreholes in the area and surface water from a steam that originates at the base of the hill. For extraction of uranium and thorium a long chemical process was applied to samples; activities of 228Th and 232 Th isotopes (4n series) and also of 238U, 234U and 230Th isotopes (4n+2 series)were determined by the alpha spectrometry method. The ratios 234U?238U determined for well spoils did not show marked disequilibria between these isotopes. However, the ratios 228Th/232Th and 230Th/234U obtained in some samples showed a high disequilibrium between these isotopes, associated with the presence of possible zones of removal of uranium and precipitation of radium.(author)

  13. 230Th/234U activity ratio in the products from Izu-Bonin island-arc volcanoes

    International Nuclear Information System (INIS)

    Kurihara, Y.; Takahashi, M.; Sato, J.

    2006-01-01

    Magma genesis at subduction zone is generally inferred to be induced by the partial melting of the mantle wedge by the addition of fluid derived from the subducting slab. Uranium-series disequilibria in the volcanic products are a useful tracer to understand various magma processes. Young island-arc volcanic rocks showed the characteristic feature of excess 234 U over 230 Th. Observation was carried out on the radioactive disequilibrium between 234 U and 230 Th in the volcanic products from Asama volcano and Izu-Mariana island-arc volcanoes. Thorium and Uranium in the volcanic rock samples were separated by anion-exchange resin and purified by TEVA·Spec. and UTEVA·Spec. resins, respectively. Purified Th and U were electrodeposited onto a stainless steel planchet for α-ray counting. U-234 and 230 Th in volcanic rock samples were determined by isotope dilution method coupled with α-ray spectrometry. 230 Th/ 234 U activity ratio in the volcanic products from Asama volcano and Izu-Bonin island-arc volcanoes were in radioactive disequilibrium, enriched in 234 U relative to 230 Th, which is often observed for volcanic products from young island-arc volcanic products. (author)

  14. Uranium-233 analysis of biological samples

    International Nuclear Information System (INIS)

    Gies, R.A.; Ballou, J.E.; Case, A.C.

    1979-01-01

    Two liquid scintillation techniques were compared for 233 U analysis: a two-phase extraction system (D2EHPA) developed by Keough and Powers, 1970, for Pu analysis; and a single-phase emulsion system (TT21) that holds the total sample in suspension with the scintillator. The first system (D2EHPA) was superior in reducing background (two- to threefold) and in accommodating a larger sample volume (fivefold). Samples containing > 50 mg/ml of slats were not extracted quantitatively by D2EHPA

  15. Radiological impact due to natural radionuclides (U and Th-isotopes) in soils from Salamanca, Mexico

    Energy Technology Data Exchange (ETDEWEB)

    Mandujano G, C. D.; Sosa, M. [Universidad de Guanajuato, Division de Ciencias e Ingenierias, Loma del Bosque 103, Col. Lomas del Campestre, 37150 Leon, Guanajuato (Mexico); Mantero, J.; Manjon, G.; Garcia T, R. [Universidad de Sevilla, Grupo en Fisica Nuclear Aplicada, Av. Reina Mercedes No. 2, 41012 Sevilla (Spain); Costilla, R., E-mail: cmandujano@fisica.ugto.mx [Universidad de Guanajuato, Division de Ciencias de la Vida, Departamento de Ciencias Ambientales, Ex-Hacienda El Copal Km 9 Irapuato-Silao, 36500 Irapuato, Guanajuato (Mexico)

    2015-10-15

    Full text: Activity concentrations of U ({sup 238}U, {sup 234}U) and Th ({sup 232}Th, {sup 230}Th) radionuclides in samples of superficial urban soils surrounding an industrial complex in Salamanca, Mexico have been determined. Levels of naturally occurring radionuclides (Norm) in the environment may be affected due to the presence of different industrial activities in this zone, representing a potential radiological risk for the population which should be evaluated. Alpha-particle Spectrometry with Pips detectors has been used for the radiometric characterization. A well established radiochemical procedure was used for the isolation of the radionuclides of interest. Alkali fusion for sample digestion, liquid-liquid extraction with Tbp (tri-butyl-phosphate) for U and Th isolation and electrodeposition in stainless steel dishes for measurement conditioning has been used. The results cover the ranges of 10-42, 12-60, 12-52 and 11-51 Bq·kg{sup -1} for {sup 238}U, {sup 234}U, {sup 230}Th, and {sup 232}Th respectively, being not observed any clear anthropogenic increments in relation with the values normally found in unaffected soils. Although there is disequilibrium between U isotopes and {sup 230}Th in some soil samples, it can be attributed to natural processes. The radiological impact of the industrial activities in the surrounding soils can be then evaluated as very low. Hence, from the Radiological Protection point of view, the soils studied do not represent a radiological risk for the health of the population. (Author)

  16. U,Th-21Ne dating and its applications

    International Nuclear Information System (INIS)

    Basu, Sudeshna; Murty, S.V.S.; Anil Kumar

    2003-01-01

    The potential of radiogenic and fissiogenic noble gas isotopes as dating tools has been well exploited. U, Th- 4 He , K- 40 Ar and U- fission Xe pairs as well as their variants like 39 Ar- 40 Ar and induced fission Xe- spontaneous fission Xe pairs have been extensively used as geochronological tools. A new dating method that utilizes the nucleogenic isotope 21 Ne and demonstrate its application for an apatite separate from a carbonatite is proposed

  17. Study on domestic material purchasing in MSR manufacture of conventional island

    International Nuclear Information System (INIS)

    Xie Zhengmao

    2010-01-01

    Combining the real case of Dongfang Electric (Guangzhou) Heavy Machinery Co., Ltd. trying to purchase the domestic sealing gasket as needed in the MSR of the conventional island, this paper describes the trends and relevant experience about nuclear power equipment manufacturers purchasing materials in the domestic market, and provides a reference to broadening the procurement channels of the purchasing departments of nuclear equipment manufacturers. (author)

  18. HPLC method for determination of Th, U and Pu in irradiated (Th,Pu)O{sub 2} using mandelic acid as an eluent

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Pranaw; Paul, Sumana; Jaison, Perumpillil George; Telmore, Vijay Madhukar; Alamelu, Devanathan; Aggarwal, Suresh Kumar [Bhabha Atomic Research Centre, Mumbai (India). Fuel Chemistry Div.

    2014-07-01

    Studies for chromatographic separation of Th, U(VI) and Pu(IV) were carried out using mandelic acid as an eluent. The different chromatographic conditions like concentration of mandelic acid, pH of the mobile phase, presence of MeOH and effect of ion interaction reagent (IIR) were studied. The method was optimized for the separation of Th, U(VI) and Pu(IV). At pH<3.5 of mobile phase, Pu(IV) was more retained compared to U(VI) whereas at pH>3.5, reverse trend was observed. The optimized parameters were employed for the separation and determination of Th, U(VI) and Pu(IV) in a dissolved solution of irradiated (Th,Pu)O{sub 2}. Sample treatment was optimized to minimize loss of Pu and Th during chromatographic determination. Studies were carried out using two IIRs to understand the anomalous chromatographic behavior of Pu(IV). Retention behavior of different oxidation states of Pu viz. Pu(III), Pu(IV) and Pu(VI) was also studied in mandelic acid.

  19. Superconductivity in the U(Th)-Y-Ba-Cu-O systems

    International Nuclear Information System (INIS)

    Qin Qizong; He Adi; Jia Weijie; Ma Lidun; Cheng Huansheng; Hua Zhongyi

    1989-01-01

    High T c superconductivity has been observed both resistively and magnetically in the new U(Th)-Y-Ba-Cu-O systems. The zero resistance temperature of the three samples with nominal composition of U 0.1 Y 1.1 Ba 0.8 Cu O 4-z , U 0.15 Y 1.05 Ba 0.8 Cu 4-z and Th 0.3 Y 0.8 Ba 0.8 Cu 4-z is 87K, 79K and 74K, respectively. The result of ac magnetic susceptibility measurement implies that the superconducting state is realized in the U-doped samples below 90K. The stoichimetry of the U-Y-Ba-Cu O superconductors has the aid of Rutherford backscattering and nuclear reaction 16 O(d, p) 17 O. The analytical results show that the 'real composition' of somples may be different from that of the nomial composition. The X ray diffraction analysis of the U-Y-Ba-Cu-O system shows that none of the peaks can be fitted to the uranium oxides and the other raw materiale structure, and its pattern may be attributed to new multiple phases with structure different from the known singlephase superconducting oxides

  20. Alecto - results obtained with homogeneous critical experiments on plutonium 239, uranium 235 and uranium 233; Alecto - resultats des experiences critiques homogenes realisees sur le plutonium 239, l'uranium 235 et l'uranium 233

    Energy Technology Data Exchange (ETDEWEB)

    Bruna, J G; Brunet, J P; Caizegues, R; Clouet d' Orval, Ch; Kremser, J; Tellier, H; Verriere, Ph [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1965-07-01

    In this report are given the results of the homogeneous critical experiments ALECTO, made on plutonium 239, uranium 235 and uranium 233. After a brief description of the equipment, the critical masses for cylinders of diameters varying from 25 to 42 cm, are given and compared with other values (foreign results, criticality guide). With respect to the specific conditions of neutron reflection in the ALECTO experiments the minimal values of critical masses are: Pu239 M{sub c} = 910 {+-} 10 g, U235 M{sub c} = 1180 {+-} 12 g and U233 M{sub c} = 960 {+-} 10 g. Experiments relating to cross sections and constants to be used on these materials are presented. Lastly, kinetic experiments allow to compare pulsed neutron methods to fluctuation methods. [French] On presente dans ce rapport les resultats des experiences critiques homogenes ALECTO, effectuees sur le plutonium 239, l'uranium 235 et l'uranium 233. Apres avoir rappele la description des installations, on donne les masses critiques pour des cylindres de diametres variant entre 25 et 42 cm, qui sont comparees avec d'autres chiffres (resultats etrangers, guide de criticite). Dans les gammes des diametres etudies pour des cuves a fond plat reflechies lateralement, la valeur minimale des masses critiques est la suivante: Pu239 M{sub c} = 910 {+-} 10 g, U235 M{sub c} = 1180 {+-} 12 g et U233 M{sub c} 960 {+-} 10 g. Des experiences portant sur les sections efficaces et les constantes a utiliser sur ces milieux sont ensuite presentees. Enfin des experiences de cinetique permettent une comparaison entre la methode des neutrons pulses et la methode des fluctuations. (auteur)

  1. Preliminary study of the {alpha} ratio measurement, ratio of the neutron capture cross section to the fission one for {sup 233}U, on the PEREN platform. Development and study of the experimental setup; Etude preliminaire de la mesure du rapport {alpha}, rapport de la section efficace moyenne de capture sur celle de fission de l'{sup 233}U, sur la plateforme PEREN. Developpement et etude du dispositif experimental

    Energy Technology Data Exchange (ETDEWEB)

    Cognet, M A

    2007-12-15

    Producing nuclear energy in order to reduce anthropic CO{sub 2} emission and to meet high energy demand, implies three conditions to the nuclear plants of the IV. generation: safety improvements, radioactive waste minimization, and fuel breeding for a sustainable use of the resources. The Thorium fuel cycle used in Molten Salt Reactors seems promising. Many numerical studies based on probabilistic codes are carried out in order to analyse the behaviour of such reactors. Nevertheless, one of the most important parameters is badly known: the alpha ratio of {sup 233}U, ratio of the neutron capture cross section to fission one for {sup 233}U. This key-parameter is necessary to calculate the breeding ratio and thus, the deployment capacities of those reactors. This Ph-D thesis was intended to prepare a precise measurement of the alpha ratio of {sup 233}U between 1 eV and 10 keV. Preliminary measurements have been performed on the experimental platform PEREN. This experimental environment is composed of a lead slowing-down time spectrometer associated with an intense pulsed neutron generator. Capture and fission rates are measured thanks to eight scintillators with their photomultipliers, surrounding a fission chamber. A software analysis sets the coincidence rate between the scintillators. In order to understand perfectly the experimental setup, preliminary tests using a {sup 235}U fission chamber have been done. This experiment resulted in a very low signal to background ratio (1 %). The background coming from the scintillators themselves seriously handicapped the measurement. Another series of experiment has been done with scintillators 5 times thinner. Nevertheless, the signal to background ratio should still be increased to measure the capture of {sup 235}U. To make sure that the experimental setup has totally been understood, we made many comparisons between experimental results and simulations. Two simulation codes were mainly used: MCNP and GEANT4. We paid

  2. Preliminary study of the {alpha} ratio measurement, ratio of the neutron capture cross section to the fission one for {sup 233}U, on the PEREN platform. Development and study of the experimental setup; Etude preliminaire de la mesure du rapport {alpha}, rapport de la section efficace moyenne de capture sur celle de fission de l'{sup 233}U, sur la plateforme PEREN. Developpement et etude du dispositif experimental

    Energy Technology Data Exchange (ETDEWEB)

    Cognet, M.A

    2007-12-15

    Producing nuclear energy in order to reduce anthropic CO{sub 2} emission and to meet high energy demand, implies three conditions to the nuclear plants of the IV. generation: safety improvements, radioactive waste minimization, and fuel breeding for a sustainable use of the resources. The Thorium fuel cycle used in Molten Salt Reactors seems promising. Many numerical studies based on probabilistic codes are carried out in order to analyse the behaviour of such reactors. Nevertheless, one of the most important parameters is badly known: the alpha ratio of {sup 233}U, ratio of the neutron capture cross section to fission one for {sup 233}U. This key-parameter is necessary to calculate the breeding ratio and thus, the deployment capacities of those reactors. This Ph-D thesis was intended to prepare a precise measurement of the alpha ratio of {sup 233}U between 1 eV and 10 keV. Preliminary measurements have been performed on the experimental platform PEREN. This experimental environment is composed of a lead slowing-down time spectrometer associated with an intense pulsed neutron generator. Capture and fission rates are measured thanks to eight scintillators with their photomultipliers, surrounding a fission chamber. A software analysis sets the coincidence rate between the scintillators. In order to understand perfectly the experimental setup, preliminary tests using a {sup 235}U fission chamber have been done. This experiment resulted in a very low signal to background ratio (1 %). The background coming from the scintillators themselves seriously handicapped the measurement. Another series of experiment has been done with scintillators 5 times thinner. Nevertheless, the signal to background ratio should still be increased to measure the capture of {sup 235}U. To make sure that the experimental setup has totally been understood, we made many comparisons between experimental results and simulations. Two simulation codes were mainly used: MCNP and GEANT4. We paid

  3. Retention and translocation of inhaled uranyl nitrate (233U and 232U) in rats

    International Nuclear Information System (INIS)

    Ballou, J.E.; Gies, R.A.; Wogman, N.A.

    1978-01-01

    The uranium-thorium breeder reactors proposed for nuclear power production, and other thorium fuel systems in conventional reactors, utilize fuels and fuel recycle process solutions that have not been evaluated for biological hazard. This project emphasizes studies of the metabolism of the oxide fuels and the nitrate solutions of the major radionuclides, following inhalation, ingestion, or cutaneous application in rodents. Preliminary data are reported for the clearance of inhaled 233 UO 2 (NO 3 ) 2 and 232 UO 2 (NO 3 ) 2 from the lung and their translocation to skeleton

  4. Aquifer Transport of Th, U, Ra and Rn in solution and on colloids

    International Nuclear Information System (INIS)

    Wsserburg, G.J.; Baskaran, Mahalingam

    1999-01-01

    We have completed a theoretical study of the U-Th and radioactive decay series in an aquifer (Tricca et al, in press). Using this model as a guide, we have reassessed our results on the aquifer associated with Brookhaven National Laboratory. Based on our study and analyses of the U-Th series nuclei in this aquifer and the theoretical results, it was considered mandatory that a new set of samples be acquired. In addition to data from the groundwater samples, information concerning the addition of nuclides into the ground water system from the vadose zone was needed. A field study was then carried out and the samples returned to the laboratory. Sampling was done in consultation with hydrologists from the Brookhaven National Laboratory staff. The analyses are now almost completed and a draft of the final report is in preparation

  5. Preliminary study of the α ratio measurement, ratio of the neutron capture cross section to the fission one for 233U, on the PEREN platform. Development and study of the experimental setup

    International Nuclear Information System (INIS)

    Cognet, M.A.

    2007-12-01

    Producing nuclear energy in order to reduce anthropic CO 2 emission and to meet high energy demand, implies three conditions to the nuclear plants of the IV. generation: safety improvements, radioactive waste minimization, and fuel breeding for a sustainable use of the resources. The Thorium fuel cycle used in Molten Salt Reactors seems promising. Many numerical studies based on probabilistic codes are carried out in order to analyse the behaviour of such reactors. Nevertheless, one of the most important parameters is badly known: the alpha ratio of 233 U, ratio of the neutron capture cross section to fission one for 233 U. This key-parameter is necessary to calculate the breeding ratio and thus, the deployment capacities of those reactors. This Ph-D thesis was intended to prepare a precise measurement of the alpha ratio of 233 U between 1 eV and 10 keV. Preliminary measurements have been performed on the experimental platform PEREN. This experimental environment is composed of a lead slowing-down time spectrometer associated with an intense pulsed neutron generator. Capture and fission rates are measured thanks to eight scintillators with their photomultipliers, surrounding a fission chamber. A software analysis sets the coincidence rate between the scintillators. In order to understand perfectly the experimental setup, preliminary tests using a 235 U fission chamber have been done. This experiment resulted in a very low signal to background ratio (1 %). The background coming from the scintillators themselves seriously handicapped the measurement. Another series of experiment has been done with scintillators 5 times thinner. Nevertheless, the signal to background ratio should still be increased to measure the capture of 235 U. To make sure that the experimental setup has totally been understood, we made many comparisons between experimental results and simulations. Two simulation codes were mainly used: MCNP and GEANT4. We paid special attention to quantify the

  6. Fast Thorium Molten Salt Reactors Started with Plutonium

    International Nuclear Information System (INIS)

    Merle-Lucotte, E.; Heuer, D.; Le Brun, C.; Brissot, R.; Liatard, E.; Meplan, O.; Nuttin, A.; Mathieu, L.

    2006-01-01

    One of the pending questions concerning Molten Salt Reactors based on the 232 Th/ 233 U fuel cycle is the supply of the fissile matter, and as a consequence the deployment possibilities of a fleet of Molten Salt Reactors, since 233 U does not exist on earth and is not yet produced in the current operating reactors. A solution may consist in producing 233 U in special devices containing Thorium, in Pressurized Water or Fast Neutrons Reactors. Two alternatives to produce 233 U are examined here: directly in standard Molten Salt Reactors started with Plutonium as fissile matter and then operated in the Th/ 233 U cycle; or in dedicated Molten Salt Reactors started and fed with Plutonium as fissile matter and Thorium as fertile matter. The idea is to design a critical reactor able to burn the Plutonium and the minor actinides presently produced in PWRs, and consequently to convert this Plutonium into 233 U. A particular reactor configuration is used, called 'unique channel' configuration in which there is no moderator in the core, leading to a quasi fast neutron spectrum, allowing Plutonium to be used as fissile matter. The conversion capacities of such Molten Salt Reactors are excellent. For Molten Salt Reactors only started with Plutonium, the assets of the Thorium fuel cycle turn out to be quickly recovered and the reactor's characteristics turn out to be equivalent to Molten Salt Reactors operated with 233 U only. Using a combination of Molten Salt Reactors started or operated with Plutonium and of Molten Salt Reactors started with 233 U, the deployment capabilities of these reactors fully satisfy the condition of sustainability. (authors)

  7. Determination of Sintered (Th,U)O2 Pellet at the Grain Growth Step

    International Nuclear Information System (INIS)

    Indrati-Y, Tundjung; Pristi-Hartati, Murdani; Ari-Handayani; Ginting, Aslina Br

    2000-01-01

    The determination of sintered (Th,U)O 2 pellet at the grain growth stephave been done by dilatometer and Scanning Electron Microscope (SEM). Thecalculation method based on the densification curve and quantitativemetallurgy. The green pellet be produced by single action compaction. Itspellet was heated on the dilatometer with heating rate 11 o C/minute and inthe argon atmosphere, 2 liters/hour. The activation energy at thedensification step can be calculated by densification curve only, but theactivation energy at the grain growth step can be calculated by densificationcurve or quantitative metallurgy. The capability of the dilatometer can beoperated until 1200 o C, so the densification curve based on the experiencecan be used to calculate activation energy at the densification step, 4.492kcal/mole. The activation energy at the grain growth step, which is 25.277kcal/mole, can be predicted by trial and error on n value. That activationenergy is almost the same with activation energy that based on thequantitative metallurgy method 25.042 kcal/mole. All of the activation energyfor the (Th,U)O 2 pellet sintering process is 29.769 kcal/mole. (author)

  8. 15 CFR 23.3 - Plan.

    Science.gov (United States)

    2010-01-01

    ... 15 Commerce and Foreign Trade 1 2010-01-01 2010-01-01 false Plan. 23.3 Section 23.3 Commerce and... MISSING CHILDREN § 23.3 Plan. (a) The Department of Commerce will supplement and expand the national... biannual meetings of departmental representatives to discuss the current plan and recommendations for...

  9. Study of characteristics of Th-U cycle in CANDU SCWR

    International Nuclear Information System (INIS)

    Shi, J.; Shi, G.

    2010-01-01

    The flexibility of CANDU technology allows the use of different fuel cycles including various uranium-driven thorium cycles. Direct self-recycle method and heterogeneous cycle modes with supercritical water as coolant were studied for (U,Th)O 2 CANFLEX fuel bundle. Lattice pitch and enrichment of driver fuel were treated as independent variables, taking account of coolant void reactivity, fuel burnup, and linear power uneven factor. In the end, appropriate cycle mode and parameters of bundle were chosen for (U,Th)O 2 cycle in CANDU SCWR. Calculations were processed by the two-dimensional multigroup neutron transport code WIMS-AECL release 3.1.2.1. (author)

  10. Th and U in the Paleozoic and Mesozoic systems of Kitakami range (preliminary report)

    International Nuclear Information System (INIS)

    Katada, Masato; Kanaya, Hiroshi; Sato, Choji.

    1984-01-01

    The research of Th and U in Kitakami range was commenced during the period of late 1950s and early 1960s. Following the exploration, the studies on Th and U in sedimentary rocks in Kitakami have been continued systematically. The data of Th and U covered whole Kitakami range by the addition of the newly obtained analytical data of northern part. The behaviors of Th and U during deposition and their contents in source rocks were studied by the analytical data. 75 samples of mudstone, sandstone, the matrix of conglomerate and limestone from south Kitakami, and 180 samples of mudstone, cherty clay stone, limestone, chert and green rock from northern part of Kitakami were analyzed. U/K 2 O ratio was constant regardless of the stratigraphy in the samples of southern Kitakami. This suggests that the major portion of U was initially dissolved in seawater, adsorbed by sericite, which is the only K 2 O -bearing mineral of sediment, and deposited. The values of Th and U in the sedimentary rocks in southern Kitakami were nearly the same as those of common sedimentary rocks in the world. It is supposed that the formation of K 2 O-bearing mineral was small, and the contents in source rocks affected. On the contrary, the Th values of sedimentary rocks in northern Kitakami, were higher than those of south, and it is supposed that this is attributable to the felsitic nature of source rocks. The mudstone of Matsumae, Hokkaido, differed from that of Kitakami, which means that they were not in same sedimentary basin. (Ishimitsu, A.)

  11. Differential geochemical behaviour of natural isotopes of U and Th in an aquifer in humid tropical terrain

    International Nuclear Information System (INIS)

    Bonotto, D.M.

    1989-01-01

    Uranium and thorium isotopic analyses were performed on spoil samples from the saturated zone of a borehole drilled in the main ore body of a high grade thorium/rare earth ore, and on groundwaters from a borehole drilled in the zone. The deposit is located at Morro do Ferro, a hill near the centre of the Pocos de Caldas Plateau (MG), where an aquifer system developed in the weathered mantle due to in situ intense alteration. For extraction of uranium and thorium a long chemical process was applied to the samples; activities of Th-228 and Th-232 isotopes (4n series) and also of U-238, U-234 and Th-230 isotopes (4n+2 series) were determined by the alpha spectrometry method. U-234/U-238 activity ratios in groundwaters were between 1 and 2 but Th-228/Th-232 activity ratios showed marked isotopic fractionation between these nuclides. The mechanism of mobilization of uranium by complexation with humic substances is considered. U-234/U-238, Th-228/Th-232 and Th-230/U-234 activity ratios in soil samples allowed consider action of other possible mechanisms related to the mobilization of uranium, such as, ion-exchange reaction and adsorption by Fe and Mn oxides. (author) [pt

  12. Study of bone microarchitecture abnormalities in mice through microct due to U and Th chemical contamination

    International Nuclear Information System (INIS)

    Taam, Pedro; Lopes, Ricardo Tadeu; Lima, Inaya

    2009-01-01

    The di calcium phosphate, widely used to manufacture fertilizers and animal ration, is extracted from rock minerals. Some of these rocks, as fluoroapatite and collophanite, had together with the calcium phosphate, traces of elements as Fe, F, Mg, Mn, Th and U. Most of these elements are considered to be proper additives for fertilizers and animal ration. In the same time, the presence of U and Th is inappropriate and potentially harmful. The risks posed are more than radioactive exposure, it is rather chemical contamination and its biological effects, since U and Th have strong chemical affinity with many substances present in live organisms, specially phosphorus. The effects of U and Th in bone microarchiteture are still unknown. The aim of this work was to study bone microarchiteture changes in mice fed with animal ration enriched with uranyl phosphate and thorium nitrate, both compounds present in the nuclear fuel cycle. At regular intervals(24, 72, 120 and 168 hours after beginning of the enriched feeding) subjects were sacrificed, blood and bone samples were collected and U and Th levels measured through wavelength dispersive X ray fluorescence (WDXRF). We present the data of U and Th blood level and microarchiteture evaluation through micro computed tomography (microCT) for each mice studied. The results showed that the intake of U and Th does indeed affect bone porosity. (author)

  13. Study of bone microarchitecture abnormalities in mice through microct due to U and Th chemical contamination

    Energy Technology Data Exchange (ETDEWEB)

    Taam, Pedro; Lopes, Ricardo Tadeu, E-mail: taam@lin.ufrj.b, E-mail: ricardo@lin.ufrj.b [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Inst. Alberto Luiz Coimbra de Pos Graduacao e Pesquisa de Engenharia. Lab. de Instrumentacao Nuclear; Lima, Inaya, E-mail: inaya@lin.ufrj.b [Universidade do Estado do Rio de Janeiro (UERJ), Nova Friburgo, RJ (Brazil). Instituto Politecnico

    2009-07-01

    The di calcium phosphate, widely used to manufacture fertilizers and animal ration, is extracted from rock minerals. Some of these rocks, as fluoroapatite and collophanite, had together with the calcium phosphate, traces of elements as Fe, F, Mg, Mn, Th and U. Most of these elements are considered to be proper additives for fertilizers and animal ration. In the same time, the presence of U and Th is inappropriate and potentially harmful. The risks posed are more than radioactive exposure, it is rather chemical contamination and its biological effects, since U and Th have strong chemical affinity with many substances present in live organisms, specially phosphorus. The effects of U and Th in bone microarchiteture are still unknown. The aim of this work was to study bone microarchiteture changes in mice fed with animal ration enriched with uranyl phosphate and thorium nitrate, both compounds present in the nuclear fuel cycle. At regular intervals(24, 72, 120 and 168 hours after beginning of the enriched feeding) subjects were sacrificed, blood and bone samples were collected and U and Th levels measured through wavelength dispersive X ray fluorescence (WDXRF). We present the data of U and Th blood level and microarchiteture evaluation through micro computed tomography (microCT) for each mice studied. The results showed that the intake of U and Th does indeed affect bone porosity. (author)

  14. 49 CFR 233.9 - Reports.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 4 2010-10-01 2010-10-01 false Reports. 233.9 Section 233.9 Transportation Other... TRANSPORTATION SIGNAL SYSTEMS REPORTING REQUIREMENTS § 233.9 Reports. Not later than April 1, 1997 and every 5 years thereafter, each carrier shall file with FRA a signal system status report “Signal System Five...

  15. Fission fragments mass distributions of nuclei populated by the multinucleon transfer channels of the 18O+232Th reaction

    Directory of Open Access Journals (Sweden)

    R. Léguillon

    2016-10-01

    Full Text Available It is shown that the multinucleon transfer reactions is a powerful tool to study fission of exotic neutron-rich actinide nuclei, which cannot be accessed by particle-capture or heavy-ion fusion reactions. In this work, multinucleon transfer channels of the 18O+232Th reaction are used to study fission of fourteen nuclei 231,232,233,234Th, 232,233,234,235,236Pa, and 234,235,236,237,238U. Identification of fissioning nuclei and of their excitation energy is performed on an event-by-event basis, through the measurement of outgoing ejectile particle in coincidence with fission fragments. Fission fragment mass distributions are measured for each transfer channel, in selected bins of excitation energy. In particular, the mass distributions of 231,234Th and 234,235,236Pa are measured for the first time. Predominantly asymmetric fission is observed at low excitation energies for all studied cases, with a gradual increase of the symmetric mode towards higher excitation energy. The experimental distributions are found to be in general agreement with predictions of the fluctuation–dissipation model.

  16. The 230Th correction is the 1st priority for accurate dates of young zircons: U/Th partitioning experiments and measurements

    Science.gov (United States)

    Krawczynski, M.; McLean, N.

    2017-12-01

    One of the most accurate and useful ways of determining the age of rocks that formed more than about 500,000 years ago is uranium-lead (U-Pb) geochronology. Earth scientists use U-Pb geochronology to put together the geologic history of entire regions and of specific events, like the mass extinction of all non-avian dinosaurs about 66 million years ago or the catastrophic eruptions of supervolcanoes like the one currently centered at Yellowstone. The mineral zircon is often utilized because it is abundant, durable, and readily incorporates uranium into its crystal structure. But it excludes thorium, whose isotope 230Th is part of the naturally occurring isotopic decay chain from 238U to 206Pb. Calculating a date from the relative abundances of 206Pb and 238U therefore requires a correction for the missing 230Th. Existing experimental and observational constraints on the way U and Th behave when zircon crystallizes from a melt are not known precisely enough, and thus currently the uncertainty in dates introduced by they `Th correction' is one of the largest sources of systematic error in determining dates. Here we present preliminary results on our study of actinide partitioning between zircon and melt. Experiments have been conducted to grow zircon from melts doped with U and Th that mimic natural magmas at a range of temperatures, and compositions. Synthetic zircons are separated from their coexisting glass and using high precision and high-spatial-resolution techniques, the abundance and distribution of U and Th in each phase is determined. These preliminary experiments are the beginning of a study that will result in precise determination of the zircon/melt uranium and thorium partition coefficients under a wide variety of naturally occurring conditions. This data will be fit to a multidimensional surface using maximum likelihood regression techniques, so that the ratio of partition coefficients can be calculated for any set of known parameters. The results of

  17. Protactinium-231 found in natural thorium irradiated in JMTR

    International Nuclear Information System (INIS)

    Suzuki, Susumu; Mitsugashira, Toshiaki; Hara, Mitsuo; Satoh, Isamu; Shiokawa, Yoshinobu; Satoh, Michiko

    1987-01-01

    Natural thorium dioxides, which differed in the content of 230 Th, were irradiated in JMTR(Japan Material Testing Reactor). 232 U, 233 U, 231 Pa, 233 Pa, and remaining Th were measured radiometrically. High production of 231 Pa and high consumption of 230 Th were observed and it was necessary to assume large resonance capture of 230 Th in order to explain the production of 231 Pa and the consumption of 230 Th. (author)

  18. Crystal structures of Th(OH)PO4, U(OH)PO4 and Th2O(PO4)2. Condensation mechanism of M(IV)(OH)PO4 (M= Th, U) into M2O(PO4)2

    International Nuclear Information System (INIS)

    Dacheux, N.; Clavier, N.; Wallez, G.; Quarton, M.

    2007-01-01

    Three new crystal structures, isotypic with β-Zr 2 O(PO 4 ) 2 , have been resolved by the Rietveld method. All crystallize with an orthorhombic cell (S.G.: Cmca) with a = 7.1393(2) Angstroms, b = 9.2641(2) Angstroms, c 12.5262(4) Angstroms, V = 828.46(4) (Angstroms) 3 and Z = 8 for Th(OH)PO 4 ; a = 7.0100(2) Angstroms, b = 9.1200(2) Angstroms, c = 12.3665(3) Angstroms, V 790.60(4) (Angstroms) 3 and Z = 8 for U(OH)PO 4 ; a 7.1691(3) Angstroms, b 9.2388(4) Angstroms, c = 12.8204(7) Angstroms, V 849.15(7) (Angstroms) 3 and Z = 4 for Th 2 O(PO 4 ) 2 . By heating, the M(OH)PO 4 (M Th, U) compounds condense topotactically into M 2 O(PO 4 ) 2 , with a change of the environment of the tetravalent cation that lowers from 8 to 7 oxygen atoms. The lower stability of Th 2 O(PO 4 ) 2 compared to that of U 2 O(PO 4 ) 2 seems to result from this unusual environment for tetravalent thorium. (authors)

  19. Research on solvent extraction process for reprocessing of Th-U fuel from HTGR

    International Nuclear Information System (INIS)

    Bao Borong; Wang Gaodong; Qian Jun

    1992-05-01

    The unique properties of spent fuel from HTGR (high temperature gas cooled reactor) have been analysed. The single solvent extraction process using 30% TBP for separation and purification of Th-U fuel has been studied. In addition, the solvent extraction process for second uranium purification is also investigated to meet different needs of reprocessing and reproduction of Th-U spent fuel from HTGR

  20. Combined analysis of KamLAND and Borexino neutrino signals from Th and U decays in the Earth's interior

    CERN Document Server

    Fogli, G L; Palazzo, A; Rotunno, A M

    2010-01-01

    The KamLAND and Borexino experiments have detected electron antineutrinos produced in the decay chains of natural thorium and uranium (Th and U geoneutrinos). We analyze the energy spectra of current geoneutrino data in combination with solar and long-baseline reactor neutrino data, with marginalized three-neutrino oscillation parameters. We consider the case with unconstrained Th and U event rates in KamLAND and Borexino, as well as cases with fewer degrees of freedom, as obtained by successively assuming for both experiments a common Th/U ratio, a common scaling of Th+U event rates, and a chondritic Th/U value. In combination, KamLAND and Borexino can reject the null hypothesis (no geoneutrino signal) at 5 sigma. Interesting bounds or indications emerge on the Th+U geoneutrino rates and on the Th/U ratio, in broad agreement with typical Earth model expectations. Conversely, the results disfavor the hypothesis of a georeactor in the Earth's core, if its power exceeds a few TW. The interplay of KamLAND and Bo...

  1. Production and release of the fission gas in (Th U)O2 fuel rods

    International Nuclear Information System (INIS)

    Dias, Marcio S.

    1982-06-01

    The volume, composition and release of the fission gas products were caculated for (Th, U)O 2 fuel rods. The theorectical calculations were compared with experimental results available on the literature. In ThO 2 + 5% UO 2 fuel rods it will be produced approximated 5% more fission gas as compared to UO 2 fuel rods. The fission gas composition or Xe to Kr ratio has showed a decreasing fuel brunup dependence, in opposition to that of UO 2 . Under the same fuel rod operational conditions, the (Th, U)O 2 fission gas release will be smaller as compared to UO 2 . This behaviour of (Th, U)O 2 fuel comes from smallest gas atom difusivity and higher activation energies of the processes that increase the fission gas release. (Author) [pt

  2. Genetic modification of potato against microbial diseases: in vitro and in planta activity of a dermaseptin B1 derivative, MsrA2.

    Science.gov (United States)

    Osusky, Milan; Osuska, Lubica; Kay, William; Misra, Santosh

    2005-08-01

    Dermaseptin B1 is a potent cationic antimicrobial peptide found in skin secretions of the arboreal frog Phyllomedusa bicolor. A synthetic derivative of dermaseptin B1, MsrA2 (N-Met-dermaseptin B1), elicited strong antimicrobial activities against various phytopathogenic fungi and bacteria in vitro. To assess its potential for plant protection, MsrA2 was expressed at low levels (1-5 microg/g of fresh tissue) in the transgenic potato (Solanum tuberosum L.) cv. Desiree. Stringent challenges of these transgenic potato plants with a variety of highly virulent fungal phytopathogens--Alternaria, Cercospora, Fusarium, Phytophthora, Pythium, Rhizoctonia and Verticillium species--and with the bacterial pathogen Erwinia carotovora demonstrated that the plants had an unusually broad-spectrum and powerful resistance to infection. MsrA2 profoundly protected both plants and tubers from diseases such as late blight, dry rot and pink rot and markedly extended the storage life of tubers. Due to these properties in planta, MsrA2 is proposed as an ideal antimicrobial peptide candidate to significantly increase resistance to phytopathogens and improve quality in a variety of crops worldwide with the potential to obviate fungicides and facilitate storage under difficult conditions.

  3. @u234@@Th scavenging and particle export fluxes from the upper 100 m of the Arabian Sea

    Digital Repository Service at National Institute of Oceanography (India)

    Sarin, M.M.; Rengarajan, R.; Ramaswamy, V.

    for the upper 100 m yields a mean scavenging residence time of ~k30 days and a removal rate of ~k 3400 dpm m@u-2@@ d@u-1@@ for @u234@@Th, from dissolved to particulate phases. The deficiency of total @u234@@Th (dissolved + particulate) relative to @u238@@U...

  4. Thermophysical and anion diffusion properties of (U x ,Th1-x )O2.

    Science.gov (United States)

    Cooper, Michael W D; Murphy, Samuel T; Fossati, Paul C M; Rushton, Michael J D; Grimes, Robin W

    2014-11-08

    Using molecular dynamics, the thermophysical properties of the (U x ,Th 1- x )O 2 system have been investigated between 300 and 3600 K. The thermal dependence of lattice parameter, linear thermal expansion coefficient, enthalpy and specific heat at constant pressure is explained in terms of defect formation and diffusivity on the oxygen sublattice. Vegard's law is approximately observed for solid solution thermal expansion below 2000 K. Different deviations from Vegard's law above this temperature occur owing to the different temperatures at which the solid solutions undergo the superionic transition (2500-3300 K). Similarly, a spike in the specific heat, associated with the superionic transition, occurs at lower temperatures in solid solutions that have a high U content. Correspondingly, oxygen diffusivity is higher in pure UO 2 than in pure ThO 2 . Furthermore, at temperatures below the superionic transition, oxygen mobility is notably higher in solid solutions than in the end members. Enhanced diffusivity is promoted by lower oxygen-defect enthalpies in (U x ,Th 1- x )O 2 solid solutions. Unlike in UO 2 and ThO 2 , there is considerable variety of oxygen vacancy and oxygen interstitial sites in solid solutions generating a wide range of property values. Trends in the defect enthalpies are discussed in terms of composition and the lattice parameter of (U x ,Th 1- x )O 2 .

  5. MSR Founding Narratives and Content Analysis of Best/Dexter Award Nominee Papers (2001-2015)

    DEFF Research Database (Denmark)

    Tackney, Charles T.; Chappell, Stacie; Sato, Toyoko

    data and founder interviews. The individuals interviewed were identified through preliminary inquiry and from archival data. As a complement of and extension to this inclusive founding narrative, we conducted a content and discourse analysis of the 15 years of MSR Best Papers and Carolyn Dexter Best...

  6. 39 CFR 233.5 - Requesting financial records from a financial institution.

    Science.gov (United States)

    2010-07-01

    ... INSPECTION SERVICE AUTHORITY § 233.5 Requesting financial records from a financial institution. (a... Department of the U.S. Postal Service to request financial records from a financial institution pursuant to... authorized to request financial records of any customer from a financial institution pursuant to a formal...

  7. A road map for the realization of global-scale thorium breeding fuel cycle by single molten-fluoride flow

    International Nuclear Information System (INIS)

    Furukawa, K.; Arakawa, K.; Erbay, L. B.

    2007-01-01

    For global survival in this century, we urgently need to launch a completely new global nuclear fission industry. To get worldwide public acceptance of nuclear energy, improvements are essential not only on safety, radio-waste management and economy but also especially nuclear proliferation resistance and safeguards. However, such global fission industry cannot replace the present fossil fuel industry in the next 50 years, unless the doubling-time of nuclear energy is less than 10 years, preferably 5-7 years. Such a doubling-time cannot be established by any kind of classical 'Fission Breeding Power Station' concept. We need a symbiotic system which couples fission power reactors with a system which can convert fertile thorium to fissile U-233, such as a spallation or D/T fusion (if and when it becomes available). For such a purpose, THORIMS-NES [Thorium Molten-Salt Nuclear Energy Synergetic System] has been proposed, which is composed of simple thermal fission power stations (FUJI) and fissile producing Accelerator Molten-Salt Breeder (AMSB). Its system functions are very ambitious, delicate and complex, but can be realized in the form of simple hardware applying the multifunctional 'single-phase molten-fluoride' circulation system. This system has no difficulties relating with 'radiation-damage', 'heat-removal' and 'chemical processing' owing to the simple 'idealistic ionic liquid' character. FUJI is size-flexible (economical even in smaller sizes), fuel self-sustaining without any continuous chemical processing and without core-graphite replacement, and AMSB is based on a single-fluid molten-salt target/blanket concept, which solves most engineering difficulties such as radiation-damage, heat-removal etc., except high-current proton accelerator development. Several AMSBs are accommodated in the regional centers (several ten sites in the world) with batch chemical processing plants including radio-waste management. The integrated thorium breeding fuel cycle is

  8. U-Th-Geochemistry of Permian and Triassic sediments of the Drauzug, Carinthia, Austria

    International Nuclear Information System (INIS)

    Kurat, G.; Korkisch, J.; Niedermayr, G.; Seemann, R.

    1976-05-01

    Chemical analysis of samples of Triassic and Permian rocks from the Drauzug, Carinthia and Austria was carried out. U concentration was measured by flurimetry, Th and Cu by spectrophotometry using Thoronol method or Arsenazo III method for Th, Fe by titrimetry and V, Ba, Sr by atomic absorption spectrophotometry. The average U concentration ranged from 0.8 to 4.6 ppm and the Th concentration from 3.2 to 15.6 ppm depending upon the mineral material. The quartzporphyries contained the highest concentration of both. It was concluded that the Permian-Triassic series are very similar to the equivalent deposits in Northern Italy. Assuming a lateral displacement, the former represent the Northern marginal part of the latter and therefore are inferior in thickness and thus not favourable for larger U mineralization

  9. Determination of soil weathering rates with U-Th series disequilibria: approach on bulk soil and selected mineral phases

    International Nuclear Information System (INIS)

    Gontier, Adrien

    2014-01-01

    The aim of the present study was to evaluate weathering and soil formation rates using U-Th disequilibria in bulk soil or separated minerals. The specific objectives of this work were to evaluate the use of U-Th chronometric tools 1) regarding the impact of a land cover change and the bedrock characteristics 2) in selected secondary mineral phases and 3) in primary minerals. On the Breuil-Chenue (Morvan) site, no vegetation effect neither a grain size effect was observed on the U-Th series in the deepest soil layers (≤ 40 cm). The low soil production rate (1-2 mm/ka) is therefore more affected by regional geomorphology than by the underlying bedrock texture. In the second part of this work, based on a thorough evaluation of different techniques, a procedure was retained to extract Fe-oxides without chemical fractionation. Finally, the analysis of biotites hand-picked from one of the studied soil profile showed that U-series disequilibria allow to independently determinate the field-weathering-rate of minerals. (author)

  10. 230Th-238U disequilibrium and the melting processes beneath ridge axes

    International Nuclear Information System (INIS)

    McKenzie, D.

    1985-01-01

    The activity ratio ( 230 Th/ 238 U) is calculated for a simple model of melting, for which the melt fraction in chemical and radioactive equilibrium with the solid residium remains constant as melting proceeds. The activity ratio in the melt is only significantly different from unity if the melting is slow compared with the half-life of 230 Th and if the melt fraction present at any time does not exceed a few percent. The observation that ( 230 Th/ 238 U) is about 1.25 for many ocean ridge basalts is therefore most easily explained if the melt fraction in the source region is less than 2% and if the melting occurs in a broad region more than 100 km wide beneath the ridge axis. These results are compatible with other geophysical observations. Measurements of ( 226 Ra/ 238 U) might provide useful constraints on the time required to reach chemical equilibrium between the melt and the matrix. (orig.)

  11. 40 CFR 233.31 - Coordination requirements.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 24 2010-07-01 2010-07-01 false Coordination requirements. 233.31 Section 233.31 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) OCEAN DUMPING 404 STATE PROGRAM REGULATIONS Program Operation § 233.31 Coordination requirements. (a) If a proposed...

  12. Thermal diffusivity and thermal conductivity of (Th,U)O2 fuels

    International Nuclear Information System (INIS)

    Sengupta, A.K.; Jarvis, T.; Nair, M.R.; Ramachandran, R.; Mujumdar, S.; Purushotham, D.S.C.

    2000-05-01

    India has vast reserves of thorium (> 460,000 tons) and sustained work on all aspects of thorium utilization has been initiated. In this context work on fabrication of sintered thoria and mixed (Th,U)O 2 pellets and evaluation of their thermophysical properties have been taken up in Radiometallurgy Division. Thermal conductivity, being the most important thermal properties, has been calculated using the experimentally measured thermal diffusivity, density and literature values of specific heats for ThO 2 and thoria containing 2,4,6,10 and 20% UO 2 . Thermal diffusivity was measured experimentally by the laser flash method from 600 to 1600 deg C in vacuum. It was observed that thermal conductivity of ThO 2 and mixed (Th,U)O 2 decrease with increase in temperature. It was also observed that the conductivity decreases with increase in UO 2 content, the decrease being more at lower temperature than that at higher temperatures. Empirical relations correlating thermal conductivity to temperatures have been generated by the least square fit method and reported. (author)

  13. The effect of dissolved hydrogen on the dissolution of 233U doped UO2(s) high burn-up spent fuel and MOX fuel

    International Nuclear Information System (INIS)

    Carbol, P.; Spahiu, K.

    2005-03-01

    In this report the results of the experimental work carried out in a large EU-research project (SFS, 2001-2004) on spent fuel stability in the presence of various amounts of near field hydrogen are presented. Studies of the dissolution of 233 U doped UO 2 (s) simulating 'old' spent fuel were carried out as static leaching tests, autoclave tests with various hydrogen concentrations and electrochemical tests. The results of the leaching behaviour of a high burn-up spent fuel pellet in 5 M NaCl solutions in the presence of 3.2 bar H 2 pressure and of MOX fuel in dilute synthetic groundwater under 53 bar H 2 pressure are also presented. In all the experimental studies carried out in this project, a considerable effect of hydrogen in the dissolution rates of radioactive materials was observed. The experimental results obtained in this project with a-doped UO 2 , high burn-up spent fuel and MOX fuel together with literature data give a reliable background to use fractional alteration/dissolution rates for spent fuel of the order of 10 -6 /yr - 10 -8 /yr with a recommended value of 4x10 -7 /yr for dissolved hydrogen concentrations above 10 -3 M and Fe(II) concentrations typical for European repository concepts. Finally, based on a review of the experimental data and available literature data, potential mechanisms of the hydrogen effect are also discussed. The work reported in this document was performed as part of the Project SFS of the European Commission 5th Framework Programme under contract no FIKW-CT-2001-20192 SFS. It represents the deliverable D10 of the experimental work package 'Key experiments using a-doped UO 2 and real spent fuel', coordinated by SKB with the participation of ITU, FZK-INE, ENRESA, CIEMAT, ARMINES-SUBATECH and SKB

  14. Mass determination of U-233 and Pu-239 by gamma spectrometry technique

    International Nuclear Information System (INIS)

    Moraes, M.A.P.V. de; Pugliesi, R.

    1988-09-01

    The gamma spectrometry technique has been used for masses determinations of uranium-233 and plutonium-239, granted by AERE-HARWELL. A high purity Ge semicondutor detector was used and the total efficiency curve was obtained for the counting system in the energy range 13 KeV to 135 KeV. The calculated values for the masses compared with that obtained by means of gravimetry technique. (author) [pt

  15. U and Th in some brown coals of Serbia and Montenegro and their environmental impact.

    Science.gov (United States)

    Zivotić, Dragana; Grzetić, Ivan; Lorenz, Hans; Simić, Vladimir

    2008-03-01

    The objective of this paper is to determine and compare the concentrations of U and Th in soft to hard brown (lignite to sub-bituminous) coals of Serbia and Montenegro. It also presents comparison of the obtained data on U and Th concentrations with the published data on coals located in some other countries of the world. Almost the whole coal production of Serbia and Montenegro is used as feed coals for combustion in thermal power plants. Channel samples from open pit and underground mines and core samples were collected for hard and soft brown coals. For the analysis the samples were decomposed using microwave technique. Obtained solutions containing U and Th were analyzed by inductively coupled plasma mass spectroscopy (ICP-MS) using NIST standards. Concentration of U from the investigated basins and the corresponding mine fields ranges within 0.60-70.10 mg/kg, 0.65-3.20 mg/kg, 0.95-6.59 mg/kg, 1.20-6.05 mg/kg, 0.80-6.66 mg/kg, 0.18-89.90 mg/kg, 0.19-4.14 mg/kg, and 0.28-3.52 mg/kg for the Kostolac, Kolubara, Krepoljin, Sjenica, Soko Banja, Bogovina East field, Senje-Resavica and Pljevlja basins, respectively. Concentration of Th ranges within 0.20-2.60 mg/kg, 0.84-6.57 mg/kg, 1.48-6.48 mg/kg, 0.12-2.71 mg/kg, 0.13-4.95 mg/kg, 0.14-3.48 mg/kg, 0.29-3.56 mg/kg, and 0.17-1.89 mg/kg for the Kostolac, Kolubara, Krepoljin, Sjenica, Soko Banja, Bogovina East field, Senje-Resavica and Pljevlja basins, respectively. Brown coal from Senje-Resavica, Kolubara, Kostolac and Pljevlja is characterized by low U concentration. Coals form the Krepoljin, Soko Banja and Sjenica basins have slightly higher U concentrations than the mentioned group. The highest concentration of U is characteristic for the coal from the Bogovina East field. Concentration of Th in coals from Serbia and Montenegro has proved to be low. Out of all investigated coal basins, only the coal from the Krepoljin and Kolubara basins has high concentration of Th. The hydrothermally altered rocks of the Timok

  16. Optical spectroscopy of an atomic nucleus: Progress toward direct observation of the {sup 229}Th isomer transition

    Energy Technology Data Exchange (ETDEWEB)

    Hehlen, Markus P., E-mail: hehlen@lanl.gov [Los Alamos National Laboratory, Mailstop E549, Los Alamos, NM 87545 (United States); Greco, Richard R. [Los Alamos National Laboratory, Mailstop E549, Los Alamos, NM 87545 (United States); Rellergert, Wade G.; Sullivan, Scott T. [Department of Physics and Astronomy, University of California, Los Angeles, CA 90095 (United States); DeMille, David [Department of Physics, Yale University, New Haven, CT 06511 (United States); Jackson, Robert A. [School of Physical and Geographical Sciences, Keele University, Keele, Staffordshire ST5 5BG (United Kingdom); Hudson, Eric R. [Department of Physics and Astronomy, University of California, Los Angeles, CA 90095 (United States); Torgerson, Justin R. [Los Alamos National Laboratory, Mailstop E549, Los Alamos, NM 87545 (United States)

    2013-01-15

    The nucleus of the thorium-229 isotope possesses a first excited nuclear state ({sup 229m}Th) at an exceptionally low energy of 7.8{+-}0.5 eV above the nuclear ground state ({sup 229g}Th), as determined by earlier indirect measurements. This is the only nuclear excited state known that is within the range of optical spectroscopy. This paper reports progress toward detecting the {sup 229m}Th state directly by luminescence spectroscopy in the vacuum ultraviolet spectral region. The estimated natural linewidth of the {sup 229g}Th{r_reversible}{sup 229m}Th isomer transition of 2{pi} Multiplication-Sign 0.1 to 2{pi} Multiplication-Sign 10 mHz is expected to broaden to {approx}10 kHz for {sup 229}Th{sup 4+} doped into a suitable crystal. The factors governing the choice of crystal system and the substantial challenges in acquiring a sufficiently large quantity of {sup 229}Th are discussed. We show that the {sup 229g}Th{r_reversible}{sup 229m}Th transition energy can be identified to within 0.1 nm by luminescence excitation and luminescence spectroscopy using the Advanced Light Source (ALS) at Lawrence Berkeley National Laboratory. This would open the door for subsequent laser-based measurements of the isomer transition and future applications of {sup 229}Th in nuclear clocks. We also show that {sup 233}U-doped materials should produce an intrinsic, continuous, and sufficiently high rate of {sup 229m}Th{yields}{sup 229g}Th luminescence and could be a useful aid in the initial direct search of the isomer transition. - Highlights: Black-Right-Pointing-Pointer Thorium-229 has a long-lived nuclear excited state. Black-Right-Pointing-Pointer It is the only known nuclear isomer within the reach of optical spectroscopy. Black-Right-Pointing-Pointer Thorium-229 doped fluoride crystals may offer sufficiently high luminescence rates. Black-Right-Pointing-Pointer Luminescence excitation spectroscopy in the vacuum ultraviolet spectral region may enable the first direct observation of

  17. The Effect of fO2 on Partition Coefficients of U and Th between Garnet and Silicate Melt

    Science.gov (United States)

    Huang, F.; He, Z.; Schmidt, M. W.; Li, Q.

    2014-12-01

    Garnet is one of the most important minerals controlling partitioning of U and Th in the upper mantle. U is redox sensitive, while Th is tetra-valent at redox conditions of the silicate Earth. U-series disequilibria have provided a unique tool to constrain the time-scales and processes of magmatism at convergent margins. Variation of garnet/meltDU/Th with fO2 is critical to understand U-series disequilibria in arc lavas. However, there is still no systematic experimental study about the effect of fO2 on partitioning of U and Th between garnet and melt. Here we present experiments on partitioning of U, Th, Zr, Hf, Nb, Ta, and REE between garnet and silicate melts at various fO2. The starting material was hydrous haplo-basalt. The piston cylinder experiments were performed with Pt double capsules with C-CO, MnO-Mn3O4 (MM), and hematite-magnetite (HM) buffers at 3 GPa and 1185-1230 oC. The experiments produced garnets with diameters > 50μm and quenched melt. Major elements were measured by EMPA at ETH Zurich. Trace elements were determined using LA-ICP-MS at Northwestern University (Xi'an, China) and SIMS (Cameca1280 at the Institute of Geology and Geophysics, Beijing, China), producing consistent partition coefficient data for U and Th. With fO2 increasing from CCO to MM and HM, garnet/meltDU decreases from 0.041 to 0.005, while garnet/meltDTh ranges from 0.003 to 0.007 without correlation with fO2. Notably, garnet/meltDTh/U increases from 0.136 at CCO to 0.41 at HM. Our results indicate that U is still more compatible than Th in garnet even at the highest fO2 considered for the subarc mantle wedge (~NNO). Therefore, we predict that if garnet is the dominant phase controlling U-Th partitioning during melting of the mantle wedge, melts would still have 230Th excess over 238U. This explains why most young continental arc lavas have 230Th excess. If clinopyroxene is the dominant residual phase during mantle melting, U could be more incompatible than Th at high fO2

  18. Neutronic simulation of a research reactor core of (232Th, 235U)O2 fuel using MCNPX2.6 code

    International Nuclear Information System (INIS)

    Feghhi, Seyed Amir Hossein; Rezazadeh, Marzieh; Kadi, Yacine; ); Tenreiro, Claudio; Aref, Morteza; Gholamzadeh, Zohreh

    2013-01-01

    The small reactor design for the remote and less developed areas of the user countries should have simple features in view of the lack of infra-structure and resources. Many researchers consider long core life with no on-site refuelling activity as a primary feature for the small reactor design. Long core life can be achieved by enhancing internal conversion rate of fertile to fissile materials. For that purpose, thorium cycle can he adopted because a high fissile production rate of 233 U converted from 232 Th can be expected in the thermal energy region. A simple nuclear reactor core arranged 19 assemblies in hexagonal structure, using thorium-based fuel and heavy water as coolant and moderator was simulated using MCNPX2.6 code, aiming an optimized critical assembly. Optimized reflector thickness and gap between assemblies were determined to achieve minimum neutron leakage and void reactivity. The result was a more compact core, where assemblies were designed having 19-fuel pins in 1.25 pitch-to-diameter ratio. Optimum reflector thickness of 15 cm resulted in minimal neutron leakage in view of economic limitations. A 0.5 cm gap between assembles achieved more safety and 2.2 % enrichment requirements. The present feasibility study suggests a thermal core of acceptable neutronic parameters to achieve a simple and safe core. (author)

  19. Application of the 226Ra-230Th-234U and 227Ac-231Pa-235U radiochronometers to uranium certified reference materials

    International Nuclear Information System (INIS)

    Rolison, J.M.; Treinen, K.C.; McHugh, K.C.; Gaffney, A.M.; Williams, R.W.

    2017-01-01

    Uranium certified reference materials (CRM) issued by New Brunswick Laboratory were subjected to dating using four independent uranium-series radiochronometers. In all cases, there was acceptable agreement between the model ages calculated using the 231 Pa- 235 U, 230 Th- 234 U, 227 Ac- 235 U or 226 Ra- 234 U radiochronometers and either the certified 230 Th- 234 U model date (CRM 125-A and CRM U630), or the known purification date (CRM U050 and CRM U100). The agreement between the four independent radiochronometers establishes these uranium certified reference materials as ideal informal standards for validating dating techniques utilized in nuclear forensic investigations in the absence of standards with certified model ages for multiple radiochronometers. (author)

  20. 7 CFR 58.233 - Skim milk.

    Science.gov (United States)

    2010-01-01

    ... 7 Agriculture 3 2010-01-01 2010-01-01 false Skim milk. 58.233 Section 58.233 Agriculture Regulations of the Department of Agriculture (Continued) AGRICULTURAL MARKETING SERVICE (Standards... Materials § 58.233 Skim milk. The skim milk shall be separated from whole milk meeting the requirements as...

  1. 14 CFR 23.3 - Airplane categories.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Airplane categories. 23.3 Section 23.3... STANDARDS: NORMAL, UTILITY, ACROBATIC, AND COMMUTER CATEGORY AIRPLANES General § 23.3 Airplane categories. (a) The normal category is limited to airplanes that have a seating configuration, excluding pilot...

  2. Evaluation of daily intake of 238U and 232Th in a Korean mixed diet sample using RNAA

    International Nuclear Information System (INIS)

    Chung, Yong Sam; Moon, Jong Hwa; Kim, Sun Ha; Park, Kwang Won; Kang, Sang Hoon; Cho, Seung Yeon

    2000-01-01

    To estimate the degree of intake of 238 U and 232 Th through daily diet, a Korean mixed diet sample was prepared after the investigation of the amount of consumption of the daily diet which corresponds to the age of 20 to 60 years. For the analysis of U and Th, the RNAA method was applied. Two standard reference materials were used for quality control and assurance and the analytical results were compared with a certified value. The determination of U and Th in the Korean mixed diet sample was carried out under the same analytical conditions and procedures with SRM. It is found that the concentration of U and Th in a Korean mixed diet was about 35.4 ppb and 3.4 ppb. From these results, the daily intake of 238 U and 232 Th by diet is evaluated to be 6.98 and 0.67 μg per day, respectively. Radioactivities related to the intake of 238 U and 232 Th were estimated to be about 86 mBq per person per day and the annual dose equivalents from 238 U and 232 Th revealed as 3.18 μSv and 0.29 μSv per person, respectively

  3. Study on the effect of UO2 composition on dissolution of sintered (Th-U)O2 MOX by microwave heating

    International Nuclear Information System (INIS)

    Singh, G.; Malav, R.K.; Fulzele, A.K.; Prakash, A.; Afzal, Md.; Panakkal, J.P.

    2010-01-01

    Full text: Complete dissolution of sample is a prerequisite for any chemical analysis in liquid form. Dissolution of ThO 2 based mixed oxide sample like (Th-U)O 2 , (Th-Pu)O 2 is a challenging job due to single oxidation state of thorium (IV). The present paper describes a study carried out on effect of UO 2 composition on dissolution of sintered (Th-U)O 2 mixed oxide pellets, in 0.05M HF prepared in 16 M HNO 3 . The experiments were performed in PTFE pressure vessels which could stand up to ∼ 250 deg C and safely operated up to 120 psi in an indigenous 700 watts microwave digestion system. ThO 2 , ThO 2 -3.75%UO 2 and ThO 2 -5%UO 2 pellets (∼ 6 g each) were dissolved in 60 mL of 16M HNO 3 /HF mixtures (0.05M HF in 16 M HNO 3 ) in PTFE (teflon) made pressure vessels (each experiment triplicate) at a pressure of ∼ 120 psi. Samples (two at an instant) were withdrawn after each hour and Th in the solution was determined by EDTA complexometric titration where end point was detected visually. Table 1 shows the results of percent dissolution of Th (mean of three experiments) for the sintered pellet after each interval of time until 100% dissolution. The plot for percent dissolution of Th (mean Th %) against time taken for sintered pellets is shown. Application of microwave heating has been applied for the dissolution of uncrushed sintered ThO 2 and (Th-U)O 2 pellets. It is quite evident from Th% dissolved versus time curves that the dissolution is faster as percentage of UO 2 in (Th-U)O 2 MOX solid solution increases. This is attributed to UO 2 as it can easily absorb microwave energy, leading to high temperature

  4. 230Th-238U disequilibrium systematics in oceanic tholeiites from 210N on the East Pacific Rise

    International Nuclear Information System (INIS)

    Newman, S.; Finkel, R.C.; MacDougall, J.D.

    1983-01-01

    Significant disequilibrium occurs between 230 Th and its parent, 238 U, in a suite of fresh basalt glasses from the RISE Project study area at 21 0 N on the East Pacific Rise. The ( 230 Th/ 232 Th) activity ratios observed for eight of nine samples from the crest of the axis at this site are constant within analytical uncertainty, with a value of 1.22. This observed homogeneity of ( 230 Th/ 232 Th) has two possible interpretations. First, the measured ( 230 Th/ 232 Th) can be considered to indicate a mantle-source for the RISE basalts with Th/U of 2.5. This interpretation, however, conflicts with the proposed correlation between ( 230 Th/ 232 Th) and 87 Sr/ 86 Sr which predicts that ( 230 Th/ 232 Th) should equal 1.33 at the RISE site. A second possible interpretation is that radioactive decay of 230 Th, in the basalts themselves or in a magma chamber, has decreased ( 230 Th/ 232 Th) from 1.33 to the observed values. The required time span is 11,000 to > 100,000 years. However, petrologic arguments rule against long residence time in a magma chamber, and the spreading rate of this section of the East Pacific Rise (6 cm/yr) predicts that the maximum age for axis basalts is 27,000 years. Both interpretations of the measured ( 230 Th/ 232 Th) imply a low Th/U ratio for the RISE basalt source and suggest that the MORB source at this location is depleted in Th with respect to U relative to primitive mantle or bulk earth. (orig./WL)

  5. Monazite processing of tin mining waste : rare earth separation from U and Th

    International Nuclear Information System (INIS)

    Hafni, L.N.; Faizal, R.; Sugeng, W.; Budi, S.; Arif, S.; Susilaningtyas

    2000-01-01

    Separation of Rare Earths from U and Th of Bangka monazite digestion solution, by using NaOH reagent and precipitation system has been carried out. The aim of the experiment is to find a condition of RE(OH) sub.3 precipitation to produce maximal RE recovery and high purity of RE that are free from radioelements U and Th. Parameters studied were pH, NaOH normality and precipitation time. The optimal conditions obtained were pH 9.8, 1N NaOH and 3 hours precipitation time. At this condition recovery of the RE(OH) sub.3 is 99.79 % and Th 4.52 %. However uranium and phosphate were not detected. Purity of the products are RE(OH) sub.3 98.868 %, Th(OH) sub.4 0.009 % and the others 1.123 %. (author)

  6. Origin of the {sup 238}U-{sup 230}Th disequilibrium in magmas from subduction zones: the Arenal example; Origine du desequilibre {sup 238}U-{sup 230}TH dans les magmas des zones de subduction: exemple de l`Arenal

    Energy Technology Data Exchange (ETDEWEB)

    Villemant, B [Paris-6 Univ., 75 (France)

    1997-12-31

    The existence in some volcanic products of strong excess of {sup 238}U with respect to {sup 230}Th is one of the characteristics of volcanic arc magmas. These excesses are generally attributed to fluid additions inside mantellic sources before magma segregation, differentiation and eruption. These fluids should be linked to the dehydration of the subducted rocks. These hypotheses are essentially based on correlations between {sup 10}Be, {sup 87}Sr anomalies, Ba/La ratios and on the distribution of volcanic centers with respect to the subduction zone. Recent studies suggest an evolution of the composition of volcanic sources in Central America from a depleted mantle type (MORB) in the North (Nicaragua) to a less transformed enriched type (OIB) in the South (Costa Rica). The Arenal volcano belongs to a transition zone between these two types. The preliminary study of trace elements and {sup 238}U-{sup 230}Th disequilibria in recent volcanic products (1968-1993) indicates a more complex situation. At least two different mantle sources were successively involved characterized by different Th/La and La/Yb ratios and very different to the OIB type. Also most lavas are in equilibrium with {sup 238}U/{sup 232}Th ratios of about 1.2 to 1.3. However, in eruptive cycle, some lavas are characterized by a strong {sup 238}U excess with respect to {sup 230}Th with cannot be linked to the sources, even when modified by fluids in depth. These results are interpreted in terms of heterogeneities of mantle sources and low depths late interactions with hydrothermal fluids during eruptions. Abstract only. (J.S.). 2 refs.

  7. Origin of the {sup 238}U-{sup 230}Th disequilibrium in magmas from subduction zones: the Arenal example; Origine du desequilibre {sup 238}U-{sup 230}TH dans les magmas des zones de subduction: exemple de l`Arenal

    Energy Technology Data Exchange (ETDEWEB)

    Villemant, B. [Paris-6 Univ., 75 (France)

    1996-12-31

    The existence in some volcanic products of strong excess of {sup 238}U with respect to {sup 230}Th is one of the characteristics of volcanic arc magmas. These excesses are generally attributed to fluid additions inside mantellic sources before magma segregation, differentiation and eruption. These fluids should be linked to the dehydration of the subducted rocks. These hypotheses are essentially based on correlations between {sup 10}Be, {sup 87}Sr anomalies, Ba/La ratios and on the distribution of volcanic centers with respect to the subduction zone. Recent studies suggest an evolution of the composition of volcanic sources in Central America from a depleted mantle type (MORB) in the North (Nicaragua) to a less transformed enriched type (OIB) in the South (Costa Rica). The Arenal volcano belongs to a transition zone between these two types. The preliminary study of trace elements and {sup 238}U-{sup 230}Th disequilibria in recent volcanic products (1968-1993) indicates a more complex situation. At least two different mantle sources were successively involved characterized by different Th/La and La/Yb ratios and very different to the OIB type. Also most lavas are in equilibrium with {sup 238}U/{sup 232}Th ratios of about 1.2 to 1.3. However, in eruptive cycle, some lavas are characterized by a strong {sup 238}U excess with respect to {sup 230}Th with cannot be linked to the sources, even when modified by fluids in depth. These results are interpreted in terms of heterogeneities of mantle sources and low depths late interactions with hydrothermal fluids during eruptions. Abstract only. (J.S.). 2 refs.

  8. 49 CFR 233.11 - Civil penalties.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 4 2010-10-01 2010-10-01 false Civil penalties. 233.11 Section 233.11..., DEPARTMENT OF TRANSPORTATION SIGNAL SYSTEMS REPORTING REQUIREMENTS § 233.11 Civil penalties. Any person (an... subject to a civil penalty of at least $650 and not more than $25,000 per violation, except that...

  9. Determination of the concentration of 238U, 234U, 232Th, 228Th, 228Ra, 226Ra and 210Pb in the feces of workers from a mining company of niobium and their families

    International Nuclear Information System (INIS)

    Oliveira, Roges de; Lopes, Ricardo T.; Melo, Dunstana R.; Juliao, Ligia M.Q.C.

    2005-01-01

    The object of this study consists of an open mine from which Niobium ore (pyrochlore) is extracted and a metallurgy company, where Fe-Nb alloys are produced for export. For geological reasons, the main ore is associated to natural radionuclides U and Th, and its decay products. The concentration of 234 U, 238 U, 232 Th, 226 Ra and 228 Ra, 228 Th, including 210 Pb in fecal excretion of 12:0 am, 29 workers and 13 family members were determined. The technique employed for the determination of the elements was the sequential method of radiochemical separation, followed by alpha spectrometry and counting α and β in proportional detector. Statistically significant difference was observed in the concentration of 234 U and 238 U, in feces samples, among the group of mining workers and family members; as well as for 232 Th in the feces of workers of crushing and metallurgy groups when compared with the Family Group. No statistically significant difference was detected at a concentration of 226 Ra, 228 Ra and 210 Pb, in feces of any group of workers of the installation in relation to the family group

  10. Studies on 232Th and 238U levels in marine algae collected from the coast of Niigata Prefecture

    International Nuclear Information System (INIS)

    Kato, Kenji; Tonouchi, Shigemasa; Maruta, Fumiyuki; Ebata, Hidekazu

    2001-01-01

    To evaluate the properties of algae to concentrate radioactive elements, 14 species of algae like Sargassum were collected in the Prefecture and analyzed for their 232 Th and 238 U levels with Yokogawa HP4500 ICP-MS apparatus. The places of collection included those near the water discharge of an atomic power station. Mean 232 Th and 238 U levels were found to be 120 and 260 ng/g dry wt, respectively, and Phaeophyta showed more than several times higher 238 U level than Chlorophyta and Rhodophyta. There was no clear difference in 232 Th levels. No difference between places of collection was observed in Sargassum 232 Th or 238 U level. Adsorption of 232 Th particle to and incorporation of soluble 238 U into algae body were suggested. Mean 232 Th and 238 U radioactivities were found 73 and 510 μBq/g wet wt, respectively, and the respective annual committed effective doses, 0.2 and 0.3 μSv, calculated from those values were confirmed to be enough lower than the annual public dose limit, 1 mSv. (K.H.)

  11. Thermal expansion data of (Th,U)O2 fuels

    International Nuclear Information System (INIS)

    Sengupta, A.K.; Banerjee, J.; Bhagat, R.K.; Ramachandran, R.; Majumdar, S.; Purushotham, D.S.C.

    2000-04-01

    Thermal expansion data for sintered ThO 2 and ThO 2 containing 2, 4, 6, 10 and 20% UO 2 pellets were measured using a high temperature dilatometer in the temperature range from ambient to 1773 K. The dilatometer was first calibrated using a standard graphite sample as reference material. The reproducibility of the dilatometer was tested by measuring the coefficient of expansion of tungsten (NBS SRM 737) and comparing the data with that recommended by National Bureau of Standard. It was observed that there is close agreement between the experimental and reported data. The coefficient of expansion data of (Th,U)O 2 fuel indicate that out of all the six compositions, ThO 2 +2%UO 2 showed the maximum expansion of around 1.75% at 1773 K. However, the expansion data for all the compositions were very close to each other. Empirical equation correlating thermal expansion and temperature for all six compositions have been generated and reported. (author)

  12. Actinide-pnictide (An-Pn) bonds spanning non-metal, metalloid, and metal combinations (An=U, Th; Pn=P, As, Sb, Bi)

    Energy Technology Data Exchange (ETDEWEB)

    Rookes, Thomas M.; Wildman, Elizabeth P.; Gardner, Benedict M.; Wooles, Ashley J.; Gregson, Matthew; Tuna, Floriana; Liddle, Stephen T. [School of Chemistry, The University of Manchester (United Kingdom); Balazs, Gabor; Scheer, Manfred [Institute of Inorganic Chemistry, University of Regensburg (Germany)

    2018-01-26

    The synthesis and characterisation is presented of the compounds [An(Tren{sup DMBS}){Pn(SiMe_3)_2}] and [An(Tren{sup TIPS}){Pn(SiMe_3)_2}] [Tren{sup DMBS}=N(CH{sub 2}CH{sub 2}NSiMe{sub 2}Bu{sup t}){sub 3}, An=U, Pn=P, As, Sb, Bi; An=Th, Pn=P, As; Tren{sup TIPS}=N(CH{sub 2}CH{sub 2}NSiPr{sup i}{sub 3}){sub 3}, An=U, Pn=P, As, Sb; An=Th, Pn=P, As, Sb]. The U-Sb and Th-Sb moieties are unprecedented examples of any kind of An-Sb molecular bond, and the U-Bi bond is the first two-centre-two-electron (2c-2e) one. The Th-Bi combination was too unstable to isolate, underscoring the fragility of these linkages. However, the U-Bi complex is the heaviest 2c-2e pairing of two elements involving an actinide on a macroscopic scale under ambient conditions, and this is exceeded only by An-An pairings prepared under cryogenic matrix isolation conditions. Thermolysis and photolysis experiments suggest that the U-Pn bonds degrade by homolytic bond cleavage, whereas the more redox-robust thorium compounds engage in an acid-base/dehydrocoupling route. (copyright 2018 The Authors. Published by Wiley-VCH Verlag GmbH and Co. KGaA.)

  13. Two types of adakites revealed by 238U-230Th disequilibrium from Daisen volcano, southwestern Japan

    International Nuclear Information System (INIS)

    Tokunaga, Saimi; Nakai, Shun'ichi; Orihashi, Yuji

    2010-01-01

    Daisen volcano is located on the Quaternary volcanic front in southwestern Japan. The volcano is composed mainly of andesite and dacite, which chemically resemble adakites, with high Al 2 O 3 and Sr/Y, steep REE patterns, and no negative Eu anomaly. ( 238 U/ 230 Th) disequilibrium (herein, a ratio in parentheses denotes the activity ratio) and trace element analyses of adakites from two volcanic domes, Karasugasen and Misen, indicate two adakite types. Adakite from Karasugasen is characterized by excess ( 230 Th) over ( 238 U), typical of most adakites, whereas adakite from Misen is characterized by excess ( 238 U) over ( 230 Th). The latter is consistent with enrichment in fluid-mobile elements relative to fluid immobile elements compared to rocks from Karasugasen. The values of ( 230 Th/ 232 Th) of adakites from Karasugasen and Misen are, respectively, around 0.75 and 0.81. These low ( 230 Th/ 232 Th) ratios result from the incorporation of subducted sedimentary material. The ratios, nevertheless, are higher than that for the estimate of lower crustal material suggesting significant incorporation of lower crust is unlikely. As adakites from Misen have ( 238 U) excess over ( 230 Th), adakite magma must have interacted with wedge mantle metasomatized by a slab-derived fluid, confirming the presence of a fluid-metasomatized mantle beneath Daisen volcano. (author)

  14. Dissolution of unirradiated UO{sub 2} and UO{sub 2} doped with {sup 233}U under reducing conditions

    Energy Technology Data Exchange (ETDEWEB)

    Ollila, K. [VTT Processes (Finland); Oversby, V.M. [VMO Konsult (Sweden)

    2005-01-01

    Experiments have been conducted to determine an upper limit to the dissolution rate of UO{sub 2} under reducing conditions appropriate to those in a geologic repository for spent fuel disposal in Finland and Sweden. Test duration ranged from 52 to 140 days. The total amount of U recovered in each test was converted into a dissolution rate per year for the sample. The dissolution rate was then used to calculate an expected lifetime for the samples under the test conditions. The dissolution rate did not depend on the length of the testing period. Rather, the dissolution rate appeared to decrease as the samples were exposed to sequential testing periods. This indicates that the results are still influenced by transient effects such as high-energy surface sites, which implies that the dissolution rates measured are upper limits. The sample lifetimes calculated from the last two testing periods, which had a total of 269 days, ranged from 7 to 10 million years. There was no indication of an effect of alpha radiolysis on the dissolution rate results for samples with doping levels of 0, 5, and 10% {sup 233}U.

  15. Integrated Laser Ablation U/Pb and (U-Th)/He Dating of Detrital Accessory Minerals from the Naryani River, Central Nepal

    Science.gov (United States)

    Horne, A.; Hodges, K. V.; Van Soest, M. C.

    2015-12-01

    The newly developed 'laser ablation double dating' (LADD) technique, an integrated laser microprobe U/Pb and (U-Th)/He dating method, could be an exceptionally valuable tool in detrital thermochronology for identifying sedimentary provenance and evaluating the exhumation history of a source region. A recent proof-of-concept study has used LADD to successfully date both zircon and titanite crystals from the well-characterized Fish Canyon tuff, but we also believe that another accessory mineral, rutile, could be amenable to dating via the LADD technique. To continue the development of the method, we present an application of LADD to detrital zircon, titanite, and rutile from a sample collected on the lower Naryani River of central Nepal. Preliminary analyses of the sample have yielded zircon U/Pb dates ranging from 31.4 to 2405 Ma; zircon (U-Th)/He from 1.8 to 15.4 Ma; titanite U/Pb between 18 and 110 Ma; titanite (U-Th)/He between 1 and 16 Ma; rutile U/Pb from 6 to 45 Ma; and rutile (U-Th)/He from 2 to 25 Ma. In addition to the initial data, we can use Ti-in-zircon, Zr-in-titanite, and Zr-in-rutile thermometers to determine the range of possible long-term cooling rates from grains with U/Pb ages younger than collision. Thus far our results from zircon analyses imply a cooling rate of approximately 15°C/Myr; titanite analyses imply between 10 and 67°C/Myr; and rutile between 9 and 267°C/Myr. This spread in potential cooling rates, especially in the order of magnitude differences of cooling rates calculated from the rutile grains, suggests that the hinterland source regions of the Naryani river experienced dramatically different exhumation histories during Himalayan orogenisis. Ongoing analyses will expand the dataset such that we can more adequately characterize the range of possibilities represented in the sample.

  16. Local structure of Th1-xMO2 solid solutions (M = U, Pu)

    International Nuclear Information System (INIS)

    Hubert, S.; Heisbourg, G.; Moisy, Ph.; Dacheux, N.; Purans, J.E.

    2004-01-01

    X-ray absorption spectroscopy of Th 1-x U x O 2 and Th 1-x Pu x O 2 solid solutions was carried out on the Th, U L 3 -edges, and Pu L 3 edge to study the local structure environment of actinide mixed oxides. Various compositions of Th 1-x M x O 2 solid solutions have been prepared through the coprecipitation of the mixed oxalates from chloride or nitrate solutions: x = 0.11, 0.24, 0.37, 0.53, 0.67, 0.81, 0.91 and 1 for Th 1-x U x O 2 , and x = 0.13, 0.32, 0.66 and 1 for Th 1-x Pu x O 2 . They were characterized using X- ray diffraction. XRD analysis allowed to confirm that the variation of the lattice parameters varies linearly with the composition between the end members, suggesting that the atomic volume was conserved regardless of the details of the local distortions of the lattice, following the Vegard's law. Extending X-ray absorption fine structure (EXAFS) provides a direct characterization of the local distortions present in solid solutions. We found that opposite to the lattice parameter obtained by XRD, the interatomic distances given by EXAFS do not follow completely to neither the Vegard's law nor the virtual crystal approximation (VCA). However, the average lattice parameter obtained from EXAFS data for the first and the second shells agrees well with the one calculated from XRD data. (authors)

  17. Spectra of Th/Ar and U/Ne hollow cathode lamps for spectrograph calibration

    Science.gov (United States)

    Nave, Gillian; Shlosberg, Ariel; Kerber, Florian; Den Hartog, Elizabeth; Neureiter, Bianca

    2018-01-01

    Low-current Th/Ar hollow cathode lamps have long been used for calibration of astronomical spectrographs on ground-based telescopes. Thorium is an attractive element for calibration as it has a single isotope, has narrow spectral lines, and has a dense spectrum covering the whole of the visible region. However, the high density of the spectrum that makes it attractive for calibrating high-resolution spectrographs is a detriment for lower resolution spectrographs and this is not obvious by examination of existing linelists. In addition, recent changes in regulations regarding the handling of thorium have led to a degradation in the quality of Th/Ar calibration lamps, with contamination by molecular ThO lines that are strong enough to obscure the calibration lines of interest.We are pursuing two approaches to these problems. First, we have expanded and improved the NIST Standard Reference Database 161, "Spectrum of Th-Ar Hollow Cathode Lamps" to cover the region 272 nm to 5500 nm. Spectra of hollow cathode lamps at up to 3 different currents can now be displayed simultaneously. Interactive zooming and the ability to convolve any of the spectra with a Gaussian or uploaded instrument profile enable the user to see immediately what the spectrum would look like at the particular resolution of their spectrograph. Second, we have measured the spectrum of a recent, contaminated Th/Ar hollow cathode lamp using a high-resolution Echelle spectrograph (Madison Wisconsin) at a resolving power (R~ 250,000). This significantly exceeds the resolving power of most astronomical spectrographs and resolves many of the molecular lines of ThO. With these spectra we are measuring and calibrating the positions of these molecular lines in order to make them suitable for spectrograph calibration.In the near infrared region, U/Ne hollow cathode lamps give a higher density of calibration lines than Th/Ar lamps and will be implemented on the upgraded CRIRES+ spectrograph on ESO’s Very Large

  18. Study of dissolution factors of U, Th and Ta

    International Nuclear Information System (INIS)

    Santos, Maristela; Medeiros, Geiza; Zouain, Felipe; Cunha, Kenya Dias da; Pitassi, Gabriel; Lima, Cintia; Leite, Carlos Vieira Barros; Nascimento, Jose Eduardo; Dalia, Kely Cristina

    2009-01-01

    Air pollution can be a problem in industrial processes, but monitoring and controlling the aerosols in the work place is not enough to estimate the occupational risk due to dust particle inhalation. The solubility in lung fluid is considered to estimate this risk. The aim of this study is to determine in vitro specific dissolution parameters for thorium (Th), uranium (U) and tantalum (Ta) associated to crystal lattice of a niobium mineral (pyrochlore). Th, U and Ta dissolution factors in vitro were obtained using the Gamble solution (Simulant Lung Fluid, SLF), PIXE (Particle Induced X ray Emission) and alpha spectrometry as analytical techniques. Ta, Th and U are present in the pyrochlore crystal lattice as oxide; however they have shown different dissolution parameters. The rapid dissolution fraction (fr), rapid dissolution rate (λr); slow dissolution rate (fs) and slow dissolution fraction ((λs) measured for tantalum oxide were equal to 0.1, 0.45 d -1 and 0.00007 d -1 , respectively; for uranium oxide fr was equal to 0.05, (λr equal to 1.1 d -1 ; (λs equal to 0.000068 d -1 ; for thorium oxide fr was 0.025, (λr was 1.5 d -1 and (λs: 0.000065 d -1 . These results show that chemical behavior of these 3 compounds in the SLF could not be represented by the same parameter. The ratio of uranium concentration in urine and feces samples from workers exposed to pyrochlore dust particle was determined. These values agree with the theoretical values of estimated uranium concentration using specific parameters for uranium oxide present in pyrochlore. (author)

  19. Transgenic Brassica juncea plants expressing MsrA1, a synthetic cationic antimicrobial peptide, exhibit resistance to fungal phytopathogens.

    Science.gov (United States)

    Rustagi, Anjana; Kumar, Deepak; Shekhar, Shashi; Yusuf, Mohd Aslam; Misra, Santosh; Sarin, Neera Bhalla

    2014-06-01

    Cationic antimicrobial peptides (CAPs) have shown potential against broad spectrum of phytopathogens. Synthetic versions with desirable properties have been modeled on these natural peptides. MsrA1 is a synthetic chimera of cecropin A and melittin CAPs with antimicrobial properties. We generated transgenic Brassica juncea plants expressing the msrA1 gene aimed at conferring fungal resistance. Five independent transgenic lines were evaluated for resistance to Alternaria brassicae and Sclerotinia sclerotiorum, two of the most devastating pathogens of B. juncea crops. In vitro assays showed inhibition by MsrA1 of Alternaria hyphae growth by 44-62 %. As assessed by the number and size of lesions and time taken for complete leaf necrosis, the Alternaria infection was delayed and restricted in the transgenic plants with the protection varying from 69 to 85 % in different transgenic lines. In case of S. sclerotiorum infection, the lesions were more severe and spread profusely in untransformed control compared with transgenic plants. The sclerotia formed in the stem of untransformed control plants were significantly more in number and larger in size than those present in the transgenic plants where disease protection of 56-71.5 % was obtained. We discuss the potential of engineering broad spectrum biotic stress tolerance by transgenic expression of CAPs in crop plants.

  20. Delayed neutron kinetic functions for /sup 232/Th and /sup 238/U mixtures

    Energy Technology Data Exchange (ETDEWEB)

    Ganich, P P; Goshovskij, M V; Lendel, A I; Lomonosov, V I; Sikora, D I; Sychev, S I

    1984-11-01

    In order to investigate the applicability of the method based on using kinetic functions, describing the emission of delayed neutrons by samples for determination of the content of fissionable nuclides in binary mixtures, the /sup 232/Th+/sup 238/U mixtures have been analyzed with the M-30 microtron. Fresh samples containing ThO/sub 2/, U/sub 3/O/sub 8/ and their mixtures are irradiated by bremstrahlung at the 15.5 MeV energy of accelerated electrons and 9 ..mu..A average current. The mass of samples is about 6 g. To determine the kinetic functions, temporal distributions of delayed neutron pulses are used, their maximum number for different samples being (1.7-3.0) x 10/sup 4/. In processing the data obtained two methods of normalization of the delayed neutron number in the kinetic functions are used: to the total yield of delayed neutrons and to the yield of /sup 133/I ..gamma..-quanta. The conclusion is drawn that the method investigated permits to determine relative /sup 238/U concentrations in the mixtures considered with 0.06-0.2 errors. Error reduction is achieved during the normalization of the number of delayed neutrons to the yield of /sup 130/I ..gamma..-quanta.

  1. Alecto - results obtained with homogeneous critical experiments on plutonium 239, uranium 235 and uranium 233

    International Nuclear Information System (INIS)

    Bruna, J.G.; Brunet, J.P.; Caizegues, R.; Clouet d'Orval, Ch.; Kremser, J.; Tellier, H.; Verriere, Ph.

    1965-01-01

    In this report are given the results of the homogeneous critical experiments ALECTO, made on plutonium 239, uranium 235 and uranium 233. After a brief description of the equipment, the critical masses for cylinders of diameters varying from 25 to 42 cm, are given and compared with other values (foreign results, criticality guide). With respect to the specific conditions of neutron reflection in the ALECTO experiments the minimal values of critical masses are: Pu239 M c = 910 ± 10 g, U235 M c = 1180 ± 12 g and U233 M c = 960 ± 10 g. Experiments relating to cross sections and constants to be used on these materials are presented. Lastly, kinetic experiments allow to compare pulsed neutron methods to fluctuation methods [fr

  2. Metal-Silicate-Sulfide Partitioning of U, Th, and K: Implications for the Budget of Volatile Elements in Mercury

    Science.gov (United States)

    Habermann, M.; Boujibar, A.; Righter, K.; Danielson, L.; Rapp, J.; Righter, M.; Pando, K.; Ross, D. K.; Andreasen, R.

    2016-01-01

    During formation of the solar system, the Sun produced strong solar winds, which stripped away a portion of the volatile elements from the forming planets. Hence, it was expected that planets closest to the sun, such as Mercury, are more depleted in volatile elements in comparison to other terrestrial planets. However, the MESSENGER mission detected higher than expected K/U and K/Th ratios on Mercury's surface, indicating a volatile content between that of Mars and Earth. Our experiments aim to resolve this discrepancy by experimentally determining the partition coefficients (D(sup met/sil)) of K, U, and Th between metal and silicate at varying pressure (1 to 5 GPa), temperature (1500 to 1900 C), oxygen fugacity (IW-2.5 to IW-6.5) and sulfur-content in the metal (0 to 33 wt%). Our data show that U, Th, and K become more siderophile with decreasing fO2 and increasing sulfur-content, with a stronger effect for U and Th in comparison to K. Using these results, the concentrations of U, Th, and K in the bulk planet were calculated for different scenarios, where the planet equilibrated at a fO2 between IW-4 and IW-7, assuming the existence of a FeS layer, between the core and mantle, with variable thickness. These models show that significant amounts of U and Th are partitioned into Mercury's core. The elevated superficial K/U and K/Th values are therefore only a consequence of the sequestration of U and Th into the core, not evidence of the overall volatile content of Mercury.

  3. Study of Crud Formation Using One Stage Mixer Settler for U-Th Extraction

    International Nuclear Information System (INIS)

    Busron-Masduki; Mashudi; Didiek-Herhady, R; Endang-Susiantini

    2000-01-01

    It was carried out solvent extraction of used fuel simulation solution ofU-Th using one stage of mixer settler. The ratio of U/Th was 1/9. Thesolution of U- Th and extractant of 30% TBP diluted in the diluent ofn-dodecane filled in the mixer chamber with the ratio of 1/1 then stirred.The first experiment determined equilibrium time and optimum rpm and thensearched the influenced parameter of crud formation of thorium, zirconium(fission product), phosphate acid, butanol, bentonite powder (represent offines solid), ferrum, silicium according to the TBP degradation of DBP.Zirconium and thorium are significant parameter of crud formation. Theequilibrium time was 1.5 hour, optimum rpm was 1800. The weightest crud wasobtained related to the cumulative parameter which result of 250 gram crud.According to this result and for radiation dose of 1 watt, the extractantmust be regenerated before exceed 48 days to hold the crud formation whichdisturbance the extraction process. (author)

  4. Hypothesis of ion migration and ion metallogeny. Taking U, Th, K as an example

    International Nuclear Information System (INIS)

    Peng Xuncai

    2006-01-01

    In the crust rocks of the earth, there occurs characteristic U, Th, K distribution belts, and it is very difficult to explain these belts and their features with traditional theories, The ion migration hypothesis is proposed in this paper based on the b=fact that there are free ions and three types of electronic voltages in the rock and that the voltage rises with the increase of mechanical power, and some arguments and reasons have been expound. The author makes the hypothesis on the belt characteristics extracted from larger amount data. Ion metallogeny is based on the metallogenetic theory, which is believed that the metallogenetic mass is migrated by forms of ion. They are precipitated and concentrated by chemical reaction, then form a deposit in favorable areas. The author provides five laws on ion metallogeny hypothesis. The practical results of two studied areas in southern China initially prove that ion metallogeny and its deduction are correct. the hypothesis explains not only the characteristic belts of U, Th, K distribution, but also the reason why the geochemical exploration method obtained is better effects in exploration. (authors)

  5. Processing Th C{sub 2} - UC{sub 2} fuel extracted from high temperature reactors HTGCR; Etude du traitement des combustibles Th C{sub 2} - UC{sub 2} issus de reacteurs a haute temperature

    Energy Technology Data Exchange (ETDEWEB)

    Derrien, C; Lessart, P; Pianezza, E; Verry, C; Villain, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1969-07-01

    The object of this investigation is solubilisation head-end (from crushing and grinding phase to non included first purification phase) of pulverulent ({sup 233}U/{sup 232}Th)C{sub 2} (200 - 500 microns diameter) contained in a graphite matrix extracted from a 4.10{sup 13} n.cm{sup -2}.s{sup -1} thermalized neutrons average flux with an irradiation of 80000 MWjT{sup -1} HTGCR reactor. After having succinctly described different bibliographic processes we have chosen the burn - leach of reactor fuel and graphite matrix containing it. The technology of burner is original in nuclear field and still more by utilizing ultra-sounds to intensify burning reaction and to minimize the weight of unburnables. The mixture of ThO{sub 2}, U{sub 3}O{sub 8}, and fission products oxides is solubilized by boiling HNO{sub 3} 13 M + HF 0.05 M. This process is profit-learning in a thorium recuperation and reprocessing point of view. In the contrary-case it would be interesting to consider a dry-process which would permit to separate solid ThF{sub 4} from gaseous UF{sub 6}. (authors) [French] Cette etude a pour objet le traitement initial de mise en solution ou 'head-end' (allant de la phase broyag-concassage a la phase de premiere purification exclue) d'un combustible ({sup 233}U/{sup 232}Th)C{sub 2} pulverulent (de 200 a 500 {mu} de diametre) contenu dans une matrice de graphite issu d'un reacteur HTGCR surgenerateur a neutrons thermiques de flux moyen 4. l0{sup 13} n.cm{sup -2}.s{sup -1} et taux d'irradiation 80000 MWjT{sup -1}. Apres exposition succincte des differents procedes bibliographiques decrits, nous avons finalement choisi le traitement par combustion-attaque ('Burn-Leach') du combustible et de la matrice etanche graphite qui le contient. La technologie du bruleur est originale dans le domaine nucleaire d'autant qu'elle utilise les ultra-sons pour ameliorer le rendement de la reaction de combustion et reduire au minimum le poids des imbrules. Le melange ThO{sub 2}, U{sub 3}O

  6. Analysis of a sustainable gas cooled fast breeder reactor concept

    International Nuclear Information System (INIS)

    Kumar, Akansha; Chirayath, Sunil S.; Tsvetkov, Pavel V.

    2014-01-01

    Highlights: • A Thorium-GFBR breeder for actinide recycling ability, and thorium fuel feasibility. • A mixture of 232 Th and 233 U is used as fuel and LWR used fuel is used. • Detailed neutronics, fuel cycle, and thermal-hydraulics analysis has been presented. • Run this TGFBR for 20 years with breeding of 239 Pu and 233 U. • Neutronics analysis using MCNP and Brayton cycle for energy conversion are used. - Abstract: Analysis of a thorium fuelled gas cooled fast breeder reactor (TGFBR) concept has been done to demonstrate the self-sustainability, breeding capability, actinide recycling ability, and thorium fuel feasibility. Simultaneous use of 232 Th and used fuel from light water reactor in the core has been considered. Results obtained confirm the core neutron spectrum dominates in an intermediate energy range (peak at 100 keV) similar to that seen in a fast breeder reactor. The conceptual design achieves a breeding ratio of 1.034 and an average fuel burnup of 74.5 (GWd)/(MTHM) . TGFBR concept is to address the eventual shortage of 235 U and nuclear waste management issues. A mixture of thorium and uranium ( 232 Th + 233 U) is used as fuel and light water reactor used fuel is utilized as blanket, for the breeding of 239 Pu. Initial feed of 233 U has to be obtained from thorium based reactors; even though there are no thorium breeders to breed 233 U a theoretical evaluation has been used to derive the data for the source of 233 U. Reactor calculations have been performed with Monte Carlo radiation transport code, MCNP/MCNPX. It is determined that this reactor has to be fuelled once every 5 years assuming the design thermal power output as 445 MW. Detailed analysis of control rod worth has been performed and different reactivity coefficients have been evaluated as part of the safety analysis. The TGFBR concept demonstrates the sustainability of thorium, viability of 233 U as an alternate to 235 U and an alternate use for light water reactor used fuel as a

  7. A study on the determination of phosphorus in Th-U matrix by a high resolution spectroanalyser (ICP-AES)

    International Nuclear Information System (INIS)

    Save, Neeta; Kumar, Neeraj; Jaiswal, Rajesh; Ghosh, Seema; Malav, R.K.; Das, D.K.; Prakash, Amrit; Behere, P.G.; Afzal, Mohd

    2013-01-01

    Phosphorus is present as an impurity in Thoria and constitutes an important chemical specification of the Thoria based fuel. The present paper depicts the importance of determination of phosphorus, which forms an important quality control step during the fabrication of mixed oxide (Th-U)O 2 pellet. A high resolution spectroanalyser using ICP-AES has been used for the determination of Phosphorus in (ThO 2 -3.25%UO 2 ). (author)

  8. Design of a thorium fuelled Advanced Heavy Water Reactor

    International Nuclear Information System (INIS)

    Krishnani, P.D.

    2009-01-01

    Full text: The main objective for development of Advanced Heavy Water Reactor (AHWR) is to demonstrate thorium fuel cycle technologies, along with several other advanced technologies required for next generation reactors, so that these are readily available in time for launching the third stage. The AHWR under design is a 300 MWe vertical pressure tube type thorium-based reactor cooled by boiling light water and moderated by heavy water. The fuel consists of (Th-Pu)O 2 and ( 233 ThU)O 2 pins. The fuel cluster is designed to generate maximum energy out of 233 U, which is bred in-situ from thorium and has a slightly negative void coefficient of reactivity, negative fuel temperature coefficient and negative power coefficient. For the AHWR, the well -proven pressure tube technology and online fuelling have been adopted. Core heat removal is by natural circulation of coolant during normal operation and shutdown conditions. Thus, it combines the advantages of light water reactors and PHWRs and removes the disadvantages of PHWRs. It has several passive safety systems for reactor normal operation, decay heat removal, emergency core cooling, confinement of radioactivity etc. The fuel cycle is based on the in-situ conversion of naturally available thorium into fissile 233 U in self sustaining mode. The uranium in the spent fuel will be reprocessed and recycled back into the reactor. The plutonium inventory will be kept a minimum and will come from fuel irradiated in Indian PHWRs. The 233 U required initially can come from the fast reactor programme or it can be produced by specially designing the initial core of AHWR using (Th,Pu)MOX fuel. There will be gradual transition from the initial core which will not contain any 233 U to an equilibrium core, which will have ( 233 U, Th) MOX fuel pins also in a composite cluster. The self sustenance is being achieved by a differential fuel loading of low and a relatively higher Pu in the composite clusters. The AHWR burns the

  9. 238U-230Th dating of chevkinite in high-silica rhyolites from La Primavera and Yellowstone calderas

    Science.gov (United States)

    Vazquez, Jorge A.; Velasco, Noel O.; Schmitt, Axel K.; Bleick, Heather A.; Stelten, Mark E.

    2014-01-01

    Application of 238U-230Th disequilibrium dating of accessory minerals with contrasting stabilities and compositions can provide a unique perspective on magmatic evolution by placing the thermochemical evolution of magma within the framework of absolute time. Chevkinite, a Th-rich accessory mineral that occurs in peralkaline and metaluminous rhyolites, may be particularly useful as a chronometer of crystallization and differentiation because its composition may reflect the chemical changes of its host melt. Ion microprobe 128U-230Th dating of single chevkinite microphenocrysts from pre- and post-caldera La Primavera, Mexico, rhyolites yields model crystallization ages that are within 10's of k.y. of their corresponding K-Ar ages of ca. 125 ka to 85 ka, while chevkinite microphenocrysts from a post-caldera Yellowstone, USA, rhyolite yield a range of ages from ca. 110 ka to 250 ka, which is indistinguishable from the age distribution of coexisting zircon. Internal chevkinite-zircon isochrons from La Primavera yield Pleistocene ages with ~5% precision due to the nearly two order difference in Th/U between both minerals. Coupling chevkinite 238U-230Th ages and compositional analyses reveals a secular trend of Th/U and rare earth elements recorded in Yellowstone rhyolite, likely reflecting progressive compositional evolution of host magma. The relatively short timescale between chevkinite-zircon crystallization and eruption suggests that crystal-poor rhyolites at La Primavera were erupted shortly after differentiation and/or reheating. These results indicate that 238U-230Th dating of chevkinite via ion microprobe analysis may be used to date crystallization and chemical evolution of silicic magmas.

  10. Kinetics and thermodynamics of the dissolution of Th1-xMxO2 solid solutions (M = U, Pu)

    International Nuclear Information System (INIS)

    Hubert, S.; Heisbourg, G.; Dacheux, N.; Moisy, Ph.; Purans, J.

    2004-01-01

    Kinetics of the dissolution of Th 1-x M x O 2 (M = U, Pu) solid solutions was investigated as a function of several chemical parameters such as pH, substitution ratio, temperature, ionic strength, and electrolyte. Several compositions of Th 1-x U x O 2 and Th 1-x Pu x O 2 were synthesized and characterized before and after leaching by using several methods such as XRD, EXAFS, BET, PIXE, SEM, and XPS. Leaching tests were performed in nitric, hydrochloric or sulfuric media and groundwater. The normalized dissolution rates were evaluated for Th 1-x U x O 2 , and Th 0.88 Pu 0.12 O 2 leading to the determination of the partial order related to the proton concentration, n, and to the corresponding normalized dissolution rate constant at pH = 0, k'T. While for Th enriched solids, the solid solutions Th 1-x U x O 2 have the same dissolution behaviour than ThO 2 with a partial order n ∼ 0.3, in the case of uranium enriched solids, Th 1-x U x O 2 has the same dissolution behaviour than UO 2 with a partial order of n = 1, indicating that uranium oxidation rate becomes the limiting step of the dissolution process. The stoichiometry of the release of both actinides (U or Pu, Th) was verified until the precipitation of thorium occurred in the leachate for pH > 2, while uranium was released in the solution as an uranyl form. For uranium enriched solid solutions, thermodynamic equilibrium was reached after 100 days, and solubility constant of secondary phase was determined. In the case of Th 1-x Pu x O 2 , the dissolution behaviour is similar to that of ThO 2 , but only kinetic aspect of the dissolution can be studied. From the analysis of XPS and EXAFS data on leached and un-leached Th 1-x U x O 2 samples, the dissolution mechanism of solid solutions was explained and will be discussed. The role of the electrolytes on the dissolution of the solid solutions is discussed. Kinetics parameters of dissolution are also given in groundwater and in neutral media

  11. 230Th-234U Model-Ages of Some Uranium Standard Reference Materials

    International Nuclear Information System (INIS)

    Williams, R.W.; Gaffney, A.M.; Kristo, M.J.; Hutcheon, I.D.

    2009-01-01

    The 'age' of a sample of uranium is an important aspect of a nuclear forensic investigation and of the attribution of the material to its source. To the extent that the sample obeys the standard rules of radiochronometry, then the production ages of even very recent material can be determined using the 230 Th- 234 U chronometer. These standard rules may be summarized as (a) the daughter/parent ratio at time=zero must be known, and (b) there has been no daughter/parent fractionation since production. For most samples of uranium, the 'ages' determined using this chronometer are semantically 'model-ages' because (a) some assumption of the initial 230 Th content in the sample is required and (b) closed-system behavior is assumed. The uranium standard reference materials originally prepared and distributed by the former US National Bureau of Standards and now distributed by New Brunswick Laboratory as certified reference materials (NBS SRM = NBL CRM) are good candidates for samples where both rules are met. The U isotopic standards have known purification and production dates, and closed-system behavior in the solid form (U 3 O 8 ) may be assumed with confidence. We present here 230 Th- 234 U model-ages for several of these standards, determined by isotope dilution mass spectrometry using a multicollector ICP-MS, and compare these ages with their known production history

  12. U-Th-He dating of apatite: A potential thermochronometer

    International Nuclear Information System (INIS)

    Zeitler, P.K.; Herczeg, A.L.; McDougall, I.; Honda, M.

    1987-01-01

    The authors found a gem quality crystal of Durango fluorapatite to have a 4 He content consistent with complete retention of radiogenic helium since its formation at about 31 Ma. Isothermal heating and step-heating analysis reveal 4 He loss to occur systematically by volume diffusion at low temperatures. The linear, low-temperature portion of the diffusion data yields an activation energy of 38.5 ± 8.1 kcal/mol and a frequency factor of 1n (D 0 /a 2 ) = 16.4 ± 2.8 sec -1 , corresponding to a closure temperature of 105 degree C ± 30 degree C (cooling rate 10 degree C/m.y.). It appears that U-Th-He dating of apatite might represent a useful new thermochronometer with a range similar to that of fission-track dating of apatite. From these results, they infer that a number of the too-young U-Th-He dates reported in the literature on minerals such as zircon and magnetite may in fact represent valuable records of the low-temperature thermal history of their host rocks

  13. Radioelement (U,Th,Rn) concentrations in Norwegian bedrock groundwaters

    International Nuclear Information System (INIS)

    Banks, D.; Roeyset, O.; Strand, T.; Skarphagen, H.

    1993-12-01

    Samples of groundwater from bedrock boreholes in three Norwegian geological provinces have been analysed for content of 222 Rn, U and Th. Median values of 290 Bq/l, 7.6 μg/l and 0.02 μg/l were obtained for Rn, U and Th, respectively, while maximum values were 8500 Bq/l, 170 μg/l and 2.2 μg/l. Commonly suggested drinking water limits range from 8 to 1000 Bq/l for radon and 14 to 160 μg/l for uranium. Radioelement content was closely related to lithology, the lowest concentrations being derived from the largely Caledonian rocks of the Troendelag area, and the highest from the Precambrian Iddefjord Granite of South East Norway where median values of 2500 Bq/l, 15 μg/l and 0.38 μg/l, respectively, were obtained. The Iddefjord Granite is not believed to be unique in Norway yielding high dissolved radionuclide contents in groundwaters, and several other granitic aquifers warrant further investigation in this respect. 63 refs., 13 figs., 8 tabs

  14. Study of the desintegration of short-life fission products. Application to the mass distribution in the fission of 238U and 233U induced by 14MeV neutrons

    International Nuclear Information System (INIS)

    Cavallini, Pierre.

    1975-01-01

    Nuclear spectrometry of short-life fission products was investigated, together with direct applications to the study of mass and charge distribution in fission reactions. It is shown that, by choosing judiciously the target in which the fission product is created and owing to the differences in stabilities and evaporation temperatures of the compounds obtained, it is possible to separate some elements. For example, niobium was separated by heating after irradiation of a mixture of UC and RuCl 3 , and sublimation in a tube with temperature gradient. It was thus possible to study the 99 Nb isotope. Other classical chemical separation processes were used for yttrium and strontium. The half-lifes beta and gamma spectra, decay schemes of 93 Sr, 94 Y and 95 Y were studied. It was shown how to obtain mass distribution in fission using a nondestructive gamma analysis method. As an application, results obtained in the fission of 233 U and 238 U at 14 MeV are given [fr

  15. K-U-Th systematics of terrestrial igneous rocks for planetological comparisons: volcanic rocks of the Earth oceanic island arc and Venus surface material

    International Nuclear Information System (INIS)

    Nikolaeva, O.V.

    1997-01-01

    Principles of the formation o data base for 339 samples of oceanic island arc (OIA) igneous rocks of the Earth available in literature are described as well as of the formation of fresh rock sample, characteristics of this sample, and K-U-Th-systematics of the fresh igneous rocks of Earth OIA. Results of comparison of the Venus measured rocks and Earth OIA rocks by K, U, Th

  16. Study of the uranium availability through the research method Th/U

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez, Zahily Herrero; Santos Junior, Jose Araujo dos; Amaral, Romilton dos Santos; Santos, Josineide Marques do Nascimento; Damascena, Kennedy Francys Rodrigues; Medeiros, Nilson Vicente da Silva; Maciel Neto, Jose de Almeida, E-mail: zahily1985@gmail.com, E-mail: jaraujo@ufpe.br, E-mail: romilton@ufpe.br, E-mail: neideden@hotmail.com, E-mail: kennedy.eng.ambiental@gmail.com, E-mail: nvsmedeiros@gmail.com, E-mail: profjosemaciel@gmail.com [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Centro de Tecnologia e Geociencias. Departamento de Energia Nuclear; Alvarez, Juan Reinaldo Estevez, E-mail: jestevez@ceaden.cu [Centro de Aplicaciones Tecnologicas y Desarrollo Nuclear (CEADEN), Havana (Cuba); Silva, Alberto Antonio da, E-mail: alberto.silva@barreiros.ifpe.edu.br [Instituto Federal de Educacao, Ciencia e Tecnologia de Pernambuco (IFPE), Barreiros, PE (Brazil)

    2015-07-01

    The uranium and thorium, precursors of the main natural radioactive series, have different concentrations in the Earth's crust. The ratio Th/U has been used as an indicator of oxidizing and reducing conditions, whose factors suggest availability of uranium to displacement in the environment and incorporations in different matrices. This parameter become essential to determine possible conditions of availability by the chemical form and incorporation in the food chain. The state of Paraiba, in northeastern Brazil, has a uranium deposits located in Sao Jose de Espinharas, where there are agricultural practices in areas surrounding the deposit of natural uranium. The Environmental Monitoring Program and Radioecological, making it an area that offers all the features for research mobility of uranium, chemical form and availability of incorporation, in addition to understanding the kinetics and transport of this natural radionuclide in the environment. Soil samples were collected from agricultural areas, close to the uraniferous occurrences where the samples were analyzed in the Laboratorio de Radioecologia e Controle Ambiental (LARCA) of the Departamento de Energia Nuclear at the Universidade Federal de Pernambuco (UFPE) by High Resolution Gamma Spectrometry, obtaining the experimental activities of {sup 238}U and {sup 232}Th using indirect gamma measures. The obtained findings show that the ratio Th/U varied from 0.11 to 1.33, with an average of 0.69. (author)

  17. Study of the uranium availability through the research method Th/U

    International Nuclear Information System (INIS)

    Fernandez, Zahily Herrero; Santos Junior, Jose Araujo dos; Amaral, Romilton dos Santos; Santos, Josineide Marques do Nascimento; Damascena, Kennedy Francys Rodrigues; Medeiros, Nilson Vicente da Silva; Maciel Neto, Jose de Almeida; Silva, Alberto Antonio da

    2015-01-01

    The uranium and thorium, precursors of the main natural radioactive series, have different concentrations in the Earth's crust. The ratio Th/U has been used as an indicator of oxidizing and reducing conditions, whose factors suggest availability of uranium to displacement in the environment and incorporations in different matrices. This parameter become essential to determine possible conditions of availability by the chemical form and incorporation in the food chain. The state of Paraiba, in northeastern Brazil, has a uranium deposits located in Sao Jose de Espinharas, where there are agricultural practices in areas surrounding the deposit of natural uranium. The Environmental Monitoring Program and Radioecological, making it an area that offers all the features for research mobility of uranium, chemical form and availability of incorporation, in addition to understanding the kinetics and transport of this natural radionuclide in the environment. Soil samples were collected from agricultural areas, close to the uraniferous occurrences where the samples were analyzed in the Laboratorio de Radioecologia e Controle Ambiental (LARCA) of the Departamento de Energia Nuclear at the Universidade Federal de Pernambuco (UFPE) by High Resolution Gamma Spectrometry, obtaining the experimental activities of 238 U and 232 Th using indirect gamma measures. The obtained findings show that the ratio Th/U varied from 0.11 to 1.33, with an average of 0.69. (author)

  18. Synthesis, characterization, kinetic and thermodynamic studies of the dissolution of ThO2 and of solid solutions Th1-xMxO2 (M = U, Pu)

    International Nuclear Information System (INIS)

    Heisbourg, G.

    2003-12-01

    The aim of this work was to understand the mechanisms of dissolution of ThO 2 and of thorium mixed oxides such as Th 1-x U x O 2 and Th 1-x Pu x O 2 in aqueous, oxygenated or inert media. Several solids have been synthesized by precipitation in oxalic medium: Th 1-x U x O 2 (x= 0.11; 0.24; 0.37; 0.53; 0.67; 0.81 and 0.91) and Th 1-x Pu x O 2 (x= 0.13; 0.32 and 0.66). They have been characterized by XRD, SEM, TEM, XPS, XAS, PIXE and EPMA. The sintering conditions of these materials have been studied and optimized in order to obtain sintered samples with a measured density very near the theoretical densities. A kinetic study of the dissolution of ThO 2 and of solid solutions Th 1-x U x O 2 has been carried out in several aqueous media (HNO 3 , HCl, H 2 SO 4 ) in terms of several parameters: protons concentration, temperature, pH, ionic strength, nature of the electrolyte solution and uranium molar ratio for the solid solutions Th 1-x U x O 2 in order to determine the kinetic laws of dissolution of the solid solutions having different compositions comparatively to ThO 2 . The leaching tests carried out in natural waters of compositions near those of the deep geologic sites considered for the storage of nuclear wastes have shown that the dissolution of the solids was bound to the complexing effect of the constitutional ions of the water considered. The leaching tests carried out on sintered samples of the same composition have led to the same normalized dissolution velocities. The thermodynamic aspect of the dissolution of the solid solutions Th 1-x U x O 2 in nitric medium has been studied at last. (O.M.)

  19. Burn-up calculation of different thorium-based fuel matrixes in a thermal research reactor using MCNPX 2.6 code

    Directory of Open Access Journals (Sweden)

    Gholamzadeh Zohreh

    2014-12-01

    Full Text Available Decrease of the economically accessible uranium resources and the inherent proliferation resistance of thorium fuel motivate its application in nuclear power systems. Estimation of the nuclear reactor’s neutronic parameters during different operational situations is of key importance for the safe operation of nuclear reactors. In the present research, thorium oxide fuel burn-up calculations for a demonstrative model of a heavy water- -cooled reactor have been performed using MCNPX 2.6 code. Neutronic parameters for three different thorium fuel matrices loaded separately in the modelled thermal core have been investigated. 233U, 235U and 239Pu isotopes have been used as fissile element in the thorium oxide fuel, separately. Burn-up of three different fuels has been calculated at 1 MW constant power. 135X and 149Sm concentration variations have been studied in the modelled core during 165 days burn-up. Burn-up of thorium oxide enriched with 233U resulted in the least 149Sm and 135Xe productions and net fissile production of 233U after 165 days. The negative fuel, coolant and void reactivity of the used fuel assures safe operation of the modelled thermal core containing (233U-Th O2 matrix. Furthermore, utilisation of thorium breeder fuel demonstrates several advantages, such as good neutronic economy, 233U production and less production of long-lived α emitter high radiotoxic wastes in biological internal exposure point of view

  20. Evidence for contamination of recent Hawaiian lavas from 230Th-238U data

    International Nuclear Information System (INIS)

    Condomines, M.; Bernat, M.; Allegre, C.J.

    1976-01-01

    230 Th- 238 U radioactive disequilibrium was studied in the historical lava flows of the Mauna Loa and Kilauea, Hawaii. Large variations of the ( 230 Th/ 232 Th) ratio among lavas of the same volcano that were erupted at a few years' interval are interpreted as due to contamination. The contamination probably occurs by assimilation of zeolitic minerals formed by seawater interaction while the magma resides in a superficial chamber. (Auth.)

  1. 234Th/238U disequilibrium in near-shore sediment: particle reworking and diagenetic time scales

    International Nuclear Information System (INIS)

    Aller, R.C.; Cochran, J.K.

    1976-01-01

    The distribution of 234 Th (tsub(1.2)=24.1 days) in excess of its parent 238 U in the upper layers of near-shore sediment makes possible the evaluation of short-term sediment reworking and diagenetic rates. 234 Th has a maximum residence time in Long Island Sound water of 1.4 days. Seasonal measurement of 234 Th/ 238 U disequilibrium in sediment at a single station in central Long Island Sound demonstrates rapid particle reworking and high 234 Thsub(XS)(>1 dpm/g) in the upper 4 cm of sediment with slower, irregular reworking and low 234 Thsub(XS) to at least 12 cm. The rate of rapid particle reworking varies seasonally and is highest in the fall. The rapidly mixed zone is characterized by steep gradients in sediment chemistry implying fast reactions spanned by 234 Th decay time scales. 238 U is depleted in the upper mixed zone and shows addition in reducing sediment at depth. (Auth.)

  2. 40 CFR 233.4 - Conflict of interest.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 24 2010-07-01 2010-07-01 false Conflict of interest. 233.4 Section 233.4 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) OCEAN DUMPING 404 STATE PROGRAM REGULATIONS General § 233.4 Conflict of interest. Any public officer or employee who has a direct...

  3. Physical and chemical feasibility of fueling molten salt reactors with TRU's trifluorides

    International Nuclear Information System (INIS)

    Ignatiev, V.; Feinberg, O.; Konakov, S.; Subbotine, S.; Surenkov, A.; Zakirov, R.

    2001-01-01

    The molten salt reactor (MSR) concept is very important for consideration as an element of future nuclear energy systems. These reactor systems are unique in many ways. Particularly, the MSRs appear to have substantial promise not only as advanced TRU free system operating in U-Th cycle, but also as transmuter of TRU. Physical and chemical feasibility of fueling MSR with TRU trifluorides is examined. Solvent compositions with and without U-Th as fissile / fertile addition are considered. The principle reactor and fuel cycle variables available for optimizing the performance of MSR as TRU transmuting system are discussed. These efforts led to the definition in minimal TRU mass flow rate, reduced total losses to waste and maximum possible burn up rate for the molten salt transmuter. The current status of technology and prospects for revisited interest are summarized. Significant chemical problems are remain to be resolved at the end of prior MSRs programs, notably, graphite life durability, tritium control, fate of noble metal fission products. Questions arising from plutonium and minor actinide fueling include: corrosion and container chemistry, new redox buffer for systems without uranium, analytical chemistry instrumentation, adequate constituent solubilities, suitable fuel processing and waste form development. However these problems appear to be soluble. (author)

  4. Concept of a demonstrational hybrid reactor—a tokamak with molten-salt blanket for {sup 233}U fuel production: 1. Concept of a stationary Tokamak as a neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Azizov, E. A.; Gladush, G. G., E-mail: gladush@triniti.ru; Dokuka, V. N.; Khayrutdinov, R. R. [State Research Center of the Russian Federation, Troitsk Institute for Innovation and Fusion Research (Russian Federation)

    2015-12-15

    On the basis of current understanding of physical processes in tokamaks and taking into account engineering constraints, it is shown that a low-cost facility of a moderate size can be designed within the adopted concept. This facility makes it possible to achieve the power density of neutron flux which is of interest, in particular, for solving the problem of {sup 233}U fuel production from thorium. By using a molten-salt blanket, the important task of ensuring the safe operation of such a reactor in the case of possible coolant loss is accomplished. Moreover, in a hybrid reactor with the blanket based on liquid salts, the problem of periodic refueling that is difficult to perform in solid blankets can be solved.

  5. Determination of U and Th α-emitters in NORM samples through extraction chromatography by using new and recycled UTEVA resins

    International Nuclear Information System (INIS)

    Casacuberta, N.; Lehritani, M.; Mantero, J.; Masqué, P.; Garcia-Orellana, J.; Garcia-Tenorio, R.

    2012-01-01

    This manuscript describes a protocol for the determination of U and Th isotopes via alpha spectrometry in NORM samples containing high concentrations of these radionuclides, up to kBq kg −1 . This technique is based on extraction chromatography with UTEVA (Triskem Int.) resins and it has been tested using both NORM samples from a phosphate industry and reference materials. The results proved that this method is highly optimized in terms of accuracy and precision when dealing with NORM samples. Recycling of UTEVA columns was also checked using NORM samples and successful results were obtained for both U and Th isotopes, thus proving the feasibility of re-using these type of columns. - Highlights: ► U and Th isotopes in NORM samples are determined via alpha spectrometry. ► The results show a highly optimized data in terms of accuracy and precision. ► Recycling of UTEVA columns was also checked and successful results were obtained.

  6. Study of the U and Th series in Crassostrea mangle shell

    Energy Technology Data Exchange (ETDEWEB)

    Farias, Wellington M.; Damatto, Sandra R.; Silva, Paulo S.C., E-mail: wellington.m@usp.br, E-mail: damatto@ipen.br, E-mail: pscsilva@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Simone, Luiz R.L.; Amaral, Vanessa S., E-mail: lrsimone@usp.br, E-mail: vanessamolusco@gmail.com [Universidade de Sao Paulo (MZ/USP), Sao Paulo, SP (Brazil). Museu de Zoologia

    2015-07-01

    Foraminifera, corals and mollusks shells have been used as proxies for environmental, paleoenvironmental and climatic change studies in marine system by using elemental and isotopic ratios as recorder of such events. Nevertheless, there is little information available on the U and Th radionuclides decay series applied on those fields. In this sense, the objective of this paper was to evaluate the activity concentrations of the U and Th nuclide decay series in Crassostrea mangle shell samples as a function of the geographic location. Samples from Sao Paulo, Parana, Alagoas, Rio Grande do Norte and Pernambuco states were analyzed by Neutron Activation Analysis and Gross Alpha and Beta Counting. Statistical analysis applied to the obtained results allowed differencing samples coming from Sao Paulo from that coming from Parana. (author)

  7. Study of the U and Th series in Crassostrea mangle shell

    International Nuclear Information System (INIS)

    Farias, Wellington M.; Damatto, Sandra R.; Silva, Paulo S.C.; Simone, Luiz R.L.; Amaral, Vanessa S.

    2015-01-01

    Foraminifera, corals and mollusks shells have been used as proxies for environmental, paleoenvironmental and climatic change studies in marine system by using elemental and isotopic ratios as recorder of such events. Nevertheless, there is little information available on the U and Th radionuclides decay series applied on those fields. In this sense, the objective of this paper was to evaluate the activity concentrations of the U and Th nuclide decay series in Crassostrea mangle shell samples as a function of the geographic location. Samples from Sao Paulo, Parana, Alagoas, Rio Grande do Norte and Pernambuco states were analyzed by Neutron Activation Analysis and Gross Alpha and Beta Counting. Statistical analysis applied to the obtained results allowed differencing samples coming from Sao Paulo from that coming from Parana. (author)

  8. Zircon, titanite, and apatite (U-Th)/He ages and age-eU correlations from the Fennoscandian Shield, southern Sweden

    Science.gov (United States)

    Guenthner, William R.; Reiners, Peter W.; Drake, Henrik; Tillberg, Mikael

    2017-07-01

    Craton cores far from plate boundaries have traditionally been viewed as stable features that experience minimal vertical motion over 100-1000 Ma time scales. Here we show that the Fennoscandian Shield in southeastern Sweden experienced several episodes of burial and exhumation from 1800 Ma to the present. Apatite, titanite, and zircon (U-Th)/He ages from surface samples and drill cores constrain the long-term, low-temperature history of the Laxemar region. Single grain titanite and zircon (U-Th)/He ages are negatively correlated (104-838 Ma for zircon and 160-945 Ma for titanite) with effective uranium (eU = U + 0.235 × Th), a measurement proportional to radiation damage. Apatite ages are 102-258 Ma and are positively correlated with eU. These correlations are interpreted with damage-diffusivity models, and the modeled zircon He age-eU correlations constrain multiple episodes of heating and cooling from 1800 Ma to the present, which we interpret in the context of foreland basin systems related to the Neoproterozoic Sveconorwegian and Paleozoic Caledonian orogens. Inverse time-temperature models constrain an average burial temperature of 217°C during the Sveconorwegian, achieved between 944 Ma and 851 Ma, and 154°C during the Caledonian, achieved between 366 Ma and 224 Ma. Subsequent cooling to near-surface temperatures in both cases could be related to long-term exhumation caused by either postorogenic collapse or mantle dynamics related to the final assembly of Rodinia and Pangaea. Our titanite He age-eU correlations cannot currently be interpreted in the same fashion; however, this study represents one of the first examples of a damage-diffusivity relationship in this system, which deserves further research attention.

  9. 238U-230Th-226Ra radioactive disequilibria in the products from 1707 eruption of Fuji volcano, Japan

    International Nuclear Information System (INIS)

    Kurihara, Yuichi; Takahashi, Masaomi; Sato, Jun

    2008-01-01

    Time scale of magmatic processes in the 1707 eruptive activity of Fuji volcano, Japan, was estimated by the 238 U- 230 Th- 226 Ra disequilibria observed in the 1707 volcanic products. The activity ratios of 226 Ra/ 230 Th in the products were larger than unity, being enriched in 226 Ra relative to 230 Th. The decay-corrected 226 Ra/ 230 Th activity ratio to the time of the eruption versus 238 U/ 230 Th activity ratio diagram for the 1707 volcanic products showed a positive correlation, suggesting that the 238 U/ 230 Th- 226 Ra disequilibria occurred during the magma genesis of Fuji volcano. The 230 Th- 226 Ra disequilibria in the 1707 volcanic products suggested that the time scale from the magma genesis to the eruption, including the melting of the mantle wedge, magma storage and magmatic differentiation from basalt to andesite, was less than 8000 years. (author)

  10. Calculation of U, Ra, Th and K contents in uranium ore by multiple linear regression method

    International Nuclear Information System (INIS)

    Lin Chao; Chen Yingqiang; Zhang Qingwen; Tan Fuwen; Peng Guanghui

    1991-01-01

    A multiple linear regression method was used to compute γ spectra of uranium ore samples and to calculate contents of U, Ra, Th, and K. In comparison with the inverse matrix method, its advantage is that no standard samples of pure U, Ra, Th and K are needed for obtaining response coefficients

  11. Preparations for an optical access to the lowest nuclear excitation in {sup 229}Th

    Energy Technology Data Exchange (ETDEWEB)

    Wense, Lars v.d.; Seiferle, Benedict; Thirolf, Peter [Ludwig-Maximilians-Universitaet Muenchen (Germany); Laatiaoui, Mustapha [GSI Helmholtzzentrum fuer Schwerionenforschung GmbH (Germany)

    2014-07-01

    The isomeric lowest excited nuclear level of {sup 229}Th has been indirectly measured to be 7.6±0.5 eV (163±11 nm). In order to improve the accuracy as prerequisite of an all-optical control, {sup 229m}Th is populated via a 2% decay branch in the α decay of {sup 233}U. The Thorium ions are extracted and cooled with the help of a buffer gas stopping cell and an RFQ-cooler. In order to suppress accompanying α decay chain products other than {sup 229}Th, a quadrupole mass spectrometer (QMS) is used, performance and extraction efficiency measurements were performed. Following the QMS, the Thorium isomers will be collected on a 50 μm micro electrode. The decay of these isomers can then be detected using deep UV optics, presently in the phase of preparation and adjustment. Newest results are presented.

  12. 10 CFR 600.233 - Supplies.

    Science.gov (United States)

    2010-01-01

    ... supplies exceeding $5,000 in total aggregate fair market value upon termination or completion of the award... 10 Energy 4 2010-01-01 2010-01-01 false Supplies. 600.233 Section 600.233 Energy DEPARTMENT OF... Supplies. (a) Title. Title to supplies acquired under a grant or subgrant will vest, upon acquisition, in...

  13. He, U, and Th Depth Profiling of Apatite and Zircon Using Laser Ablation Noble Gas Mass Spectrometry and SIMS

    Science.gov (United States)

    Monteleone, B. D.; van Soest, M. C.; Hodges, K. V.; Hervig, R.; Boyce, J. W.

    2008-12-01

    Conventional (U-Th)/He thermochronology utilizes single or multiple grain analyses of U- and Th-bearing minerals such as apatite and zircon and does not allow for assessment of spatial variation in concentration of He, U, or Th within individual crystals. As such, age calculation and interpretation require assumptions regarding 4He loss through alpha ejection, diffusive redistribution of 4He, and U and Th distribution as an initial condition for these processes. Although models have been developed to predict 4He diffusion parameters, correct for the effect of alpha ejection on calculated cooling ages, and account for the effect of U and Th zonation within apatite and zircon, measurements of 4He, U, and Th distribution have not been combined within a single crystal. We apply ArF excimer laser ablation, combined with noble gas mass spectrometry, to obtain depth profiles within apatite and zircon crystals in order to assess variations in 4He concentration with depth. Our initial results from pre-cut, pre-heated slabs of Durango apatite, each subjected to different T-t schedules, suggest a general agreement of 4He profiles with those predicted by theoretical diffusion models (Farley, 2000). Depth profiles through unpolished grains give reproducible alpha ejection profiles in Durango apatite that deviate from alpha ejection profiles predicted for ideal, homogenous crystals. SIMS depth profiling utilizes an O2 primary beam capable of sputtering tens of microns and measuring sub-micron resolution variation in [U], [Th], and [Sm]. Preliminary results suggest that sufficient [U] and [Th] zonation is present in Durango apatite to influence the form of the 4He alpha ejection profile. Future work will assess the influence of measured [U] and [Th] zonation on previously measured 4He depth profiles. Farley, K.A., 2000. Helium diffusion from apatite; general behavior as illustrated by Durango fluorapatite. J. Geophys. Res., B Solid Earth Planets 105 (2), 2903-2914.

  14. A dispersive optical model potential for nucleon induced reactions on 238U and 232Th nuclei with full coupling

    Directory of Open Access Journals (Sweden)

    Chiba Satoshi

    2013-03-01

    Full Text Available A dispersive coupled-channel optical model potential (DCCOMP that couples the ground-state rotational and low-lying vibrational bands of 238U and 232Th nuclei is studied. The derived DCCOMP couples almost all excited levels below 1 MeV of excitation energy of the corresponding even-even actinides. The ground state, octupole, beta, gamma, and non-axial bands are coupled. The first two isobar analogue states (IAS populated in the quasi-elastic (p,n reaction are also coupled in the proton induced calculation, making the potential approximately Lane consistent. The coupled-channel potential is based on a soft-rotor description of the target nucleus structure, where dynamic vibrations are considered as perturbations of the rigid rotor underlying structure. Matrix elements required to use the proposed structure model in Tamura coupled-channel scheme are derived. Calculated ratio R(U238/Th232 of the total cross-section difference to the averaged σT for 238U and 232Th nuclei is shown to be in excellent agreement with measured data.

  15. 230Th-234U Model-Ages of Some Uranium Standard Reference Materials

    Energy Technology Data Exchange (ETDEWEB)

    Williams, R W; Gaffney, A M; Kristo, M J; Hutcheon, I D

    2009-05-28

    The 'age' of a sample of uranium is an important aspect of a nuclear forensic investigation and of the attribution of the material to its source. To the extent that the sample obeys the standard rules of radiochronometry, then the production ages of even very recent material can be determined using the {sup 230}Th-{sup 234}U chronometer. These standard rules may be summarized as (a) the daughter/parent ratio at time=zero must be known, and (b) there has been no daughter/parent fractionation since production. For most samples of uranium, the 'ages' determined using this chronometer are semantically 'model-ages' because (a) some assumption of the initial {sup 230}Th content in the sample is required and (b) closed-system behavior is assumed. The uranium standard reference materials originally prepared and distributed by the former US National Bureau of Standards and now distributed by New Brunswick Laboratory as certified reference materials (NBS SRM = NBL CRM) are good candidates for samples where both rules are met. The U isotopic standards have known purification and production dates, and closed-system behavior in the solid form (U{sub 3}O{sub 8}) may be assumed with confidence. We present here {sup 230}Th-{sup 234}U model-ages for several of these standards, determined by isotope dilution mass spectrometry using a multicollector ICP-MS, and compare these ages with their known production history.

  16. Site specific health and safety plan, 233-S decontamination and decommissioning

    Energy Technology Data Exchange (ETDEWEB)

    J. E. Fasso

    1997-12-31

    The deactivated 233-S Plutonium Concentration Facility, located in the 200 Area at the Hanford Site, is the subject of this Health and Safety Plan.The 233-S Facility operated from January 1952 until July 1967 at which time the building entered the U.S. Department of Energy`s Surplus Facility Management Program as a retired facility. The facility has since undergone severe degradation due to exposure to extreme weather conditions. Additionally, the weather caused existing cracks in concrete structures of the building to lengthen, thereby increasing the potential for failed confinement of the radioactive material in the building. Differential settlement has also occurred causing portions of the facility to separate from the main building structure, increasing the potential for release of radioactive material to the environment. An expedited response is proposed to remove this threat and ensure protection of human health and the environment. On this premise it is intended that the 233-S Facility removal action be performed as a Comprehensive Environmental Response, Compensation, and Liability Act of 1980 Time-Critical Project being conducted under the Pilot Hanford Environmental Restoration (ER) Initiative

  17. Quantifying K, U and Th contents of marine sediments using shipboard natural gamma radiation spectra measured on DV JOIDES Resolution

    Science.gov (United States)

    De Vleeschouwer, David; Dunlea, Ann G.; Auer, Gerald; Anderson, Chloe H.; Brumsack, Hans; de Loach, Aaron; Gurnis, Michael C.; Huh, Youngsook; Ishiwa, Takeshige; Jang, Kwangchul; Kominz, Michelle A.; März, Christian; Schnetger, Bernhard; Murray, Richard W.; Pälike, Heiko; Expedition 356 shipboard scientists, IODP

    2017-04-01

    During International Ocean Discovery Program (IODP) expeditions, shipboard-generated data provide the first insights into the cored sequences. The natural gamma radiation (NGR) of the recovered material, for example, is routinely measured on the ocean drilling research vessel DV JOIDES Resolution. At present, only total NGR counts are readily available as shipboard data, although full NGR spectra (counts as a function of gamma-ray energy level) are produced and archived. These spectra contain unexploited information, as one can estimate the sedimentary contents of potassium (K), thorium (Th), and uranium (U) from the characteristic gamma-ray energies of isotopes in the 40K, 232Th, and 238U radioactive decay series. Dunlea et al. [2013] quantified K, Th and U contents in sediment from the South Pacific Gyre by integrating counts over specific energy levels of the NGR spectrum. However, the algorithm used in their study is unavailable to the wider scientific community due to commercial proprietary reasons. Here, we present a new MATLAB algorithm for the quantification of NGR spectra that is transparent and accessible to future NGR users. We demonstrate the algorithm's performance by comparing its results to shore-based inductively coupled plasma-mass spectrometry (ICP-MS), inductively coupled plasma-emission spectrometry (ICP-ES), and quantitative wavelength-dispersive X-ray fluorescence (XRF) analyses. Samples for these comparisons come from eleven sites (U1341, U1343, U1366-U1369, U1414, U1428-U1430, U1463) cored in two oceans during five expeditions. In short, our algorithm rapidly produces detailed high-quality information on sediment properties during IODP expeditions at no extra cost. Dunlea, A. G., R. W. Murray, R. N. Harris, M. A. Vasiliev, H. Evans, A. J. Spivack, and S. D'Hondt (2013), Assessment and use of NGR instrumentation on the JOIDES Resolution to quantify U, Th, and K concentrations in marine sediment, Scientific Drilling, 15, 57-63.

  18. Combined U-Th/He and 40Ar/39Ar geochronology of post-shield lavas from the Mauna Kea and Kohala volcanoes, Hawaii

    Energy Technology Data Exchange (ETDEWEB)

    Aciego, S.M.; Jourdan, F.; DePaolo, D.J.; Kennedy, B.M.; Renne, P.R.; Sims, K.W.W.

    2009-10-01

    Late Quaternary, post-shield lavas from the Mauna Kea and Kohala volcanoes on the Big Island of Hawaii have been dated using the {sup 40}Ar/{sup 39}Ar and U-Th/He methods. The objective of the study is to compare the recently demonstrated U-Th/He age method, which uses basaltic olivine phenocrysts, with {sup 40}Ar/{sup 39}Ar ages measured on groundmass from the same samples. As a corollary, the age data also increase the precision of the chronology of volcanism on the Big Island. For the U-Th/He ages, U, Th and He concentrations and isotopes were measured to account for U-series disequilibrium and initial He. Single analyses U-Th/He ages for Hamakua lavas from Mauna Kea are 87 {+-} 40 ka to 119 {+-} 23 ka (2{sigma} uncertainties), which are in general equal to or younger than {sup 40}Ar/{sup 39}Ar ages. Basalt from the Polulu sequence on Kohala gives a U-Th/He age of 354 {+-} 54 ka and a {sup 40}Ar/{sup 39}Ar age of 450 {+-} 40 ka. All of the U-Th/He ages, and all but one spurious {sup 40}Ar/{sup 39}Ar ages conform to the previously proposed stratigraphy and published {sup 14}C and K-Ar ages. The ages also compare favorably to U-Th whole rock-olivine ages calculated from {sup 238}U - {sup 230}Th disequilibria. The U-Th/He and {sup 40}Ar/{sup 39}Ar results agree best where there is a relatively large amount of radiogenic {sup 40}Ar (>10%), and where the {sup 40}Ar/{sup 36}Ar intercept calculated from the Ar isochron diagram is close to the atmospheric value. In two cases, it is not clear why U-Th/He and {sup 40}Ar/{sup 39}Ar ages do not agree within uncertainty. U-Th/He and {sup 40}Ar/{sup 39}Ar results diverge the most on a low-K transitional tholeiitic basalt with abundant olivine. For the most alkalic basalts with negligible olivine phenocrysts, U-Th/He ages were unattainable while {sup 40}Ar/{sup 39}Ar results provide good precision even on ages as low as 19 {+-} 4 ka. Hence, the strengths and weaknesses of the U-Th/He and {sup 40}Ar/{sup 39}Ar methods are

  19. Study of the production of uranium-233 in natural thorium subjected to a neutron flux of 14 MeV

    International Nuclear Information System (INIS)

    Abel, G.; Martel, J.G.; St Germain, J.P.

    1979-01-01

    This is a study of neutron flux and reactivity in several simplified models of a fusion reactor blanket composed of thorium, carbon and a stainless steel structure. The objective is the comparative determination, theoretical and experimental, of values for the conversion of nuclei of Th-232 into fissile U-233 nuclei in such a blanket. Theoretical calculations are carried out using the ANISN (transport equation) and MORSE (Monte Carlo) codes. Experimental values are measured in a stainless steel rack structure allowing simplified blanket models to be mounted around a 14 MeV neutron source. In the first year of work theoretical flux values have been obtained, the necesssary detectors constructed, the rack structure set up, programs worked out for the reactivity calculations, and finally verification methods worked out in conditions similar to those that will prevail. (LL) [fr

  20. 45 CFR 233.52 - Overpayment to aliens.

    Science.gov (United States)

    2010-10-01

    ... 45 Public Welfare 2 2010-10-01 2010-10-01 false Overpayment to aliens. 233.52 Section 233.52... ELIGIBILITY IN FINANCIAL ASSISTANCE PROGRAMS § 233.52 Overpayment to aliens. A State Plan under title IV-A of the Social Security Act, shall provide that: (a) Any sponsor of an alien and the alien shall be...

  1. D-MSR: A Distributed Network Management Scheme for Real-Time Monitoring and Process Control Applications in Wireless Industrial Automation

    Science.gov (United States)

    Zand, Pouria; Dilo, Arta; Havinga, Paul

    2013-01-01

    Current wireless technologies for industrial applications, such as WirelessHART and ISA100.11a, use a centralized management approach where a central network manager handles the requirements of the static network. However, such a centralized approach has several drawbacks. For example, it cannot cope with dynamicity/disturbance in large-scale networks in a real-time manner and it incurs a high communication overhead and latency for exchanging management traffic. In this paper, we therefore propose a distributed network management scheme, D-MSR. It enables the network devices to join the network, schedule their communications, establish end-to-end connections by reserving the communication resources for addressing real-time requirements, and cope with network dynamicity (e.g., node/edge failures) in a distributed manner. According to our knowledge, this is the first distributed management scheme based on IEEE 802.15.4e standard, which guides the nodes in different phases from joining until publishing their sensor data in the network. We demonstrate via simulation that D-MSR can address real-time and reliable communication as well as the high throughput requirements of industrial automation wireless networks, while also achieving higher efficiency in network management than WirelessHART, in terms of delay and overhead. PMID:23807687

  2. D-MSR: a distributed network management scheme for real-time monitoring and process control applications in wireless industrial automation.

    Science.gov (United States)

    Zand, Pouria; Dilo, Arta; Havinga, Paul

    2013-06-27

    Current wireless technologies for industrial applications, such as WirelessHART and ISA100.11a, use a centralized management approach where a central network manager handles the requirements of the static network. However, such a centralized approach has several drawbacks. For example, it cannot cope with dynamicity/disturbance in large-scale networks in a real-time manner and it incurs a high communication overhead and latency for exchanging management traffic. In this paper, we therefore propose a distributed network management scheme, D-MSR. It enables the network devices to join the network, schedule their communications, establish end-to-end connections by reserving the communication resources for addressing real-time requirements, and cope with network dynamicity (e.g., node/edge failures) in a distributed manner. According to our knowledge, this is the first distributed management scheme based on IEEE 802.15.4e standard, which guides the nodes in different phases from joining until publishing their sensor data in the network. We demonstrate via simulation that D-MSR can address real-time and reliable communication as well as the high throughput requirements of industrial automation wireless networks, while also achieving higher efficiency in network management than WirelessHART, in terms of delay and overhead.

  3. Distribution of naturally occurring radionuclides (U, Th) in Timahdit's black shale (Morocco)

    International Nuclear Information System (INIS)

    Galindo, C.; Mougin, L.; Nourreddine, A.; Fakhi, S.

    2006-01-01

    Attention has been recently focused on the use of Moroccan's black shale as the raw material for production of a new type of adsorbents. The purpose of the present work was to characterize a black shale specimen, collected in the region of Timahdit, in terms of the total uranium and thorium contents, measurements of some geochemically important elements (Al, Fe, Si, K, Mn, P, Ca), and XRD/SEM analysis. Selective leaching procedure, followed by radiochemical purification and alpha-counting, was also performed to assess the distribution of 238 U, 234 U, 235 U, 232 Th, 228 Th, 230 Th in the main structures. It was found that calcite, dolomite, quartz, clays constitute the main bulk composition of inorganic matrix. Organic matter counts for at least 15 wt. % of the sample. As in most other organic rich rocks, uranium is highly enriched in the black shale. It was interpreted to have been concentrated over a long period of time under anaerobic environment. This actinide is associated predominantly with humic acids, the precursor of kerogen. An integrated isotopic approach points out its mobilization from these humic acids to carbonates and apatite phases. The radionuclide that is the less mobile in this environment is 232 Th, as was expected from its chemical properties, and in agreement with the most common view in the literature. It is partitioned between silicate minerals (49%), pyrite and kerogen (51%). Speciation, chemical behaviour of uranium and thorium and alpha decay related processes are widely responsible for disequilibria in the uranium decay series. (author)

  4. 233S Decommissioning Project Environmental Control Plan

    International Nuclear Information System (INIS)

    Zoric, J.P.

    2000-01-01

    This Environmental Control Plan is for the 233S Decommissioning activities conducted under the removal action report for the 233S Decontamination and Demolition Project. The purpose of this ECP is to identify environmental requirements for the 233S project. The ECP is a compilation of existing environmental permit conditions, regulatory requirements, and environmental requirements applicable to the specific project or functional activity

  5. Transcriptional attenuation controls macrolide inducible efflux and resistance in Streptococcus pneumoniae and in other Gram-positive bacteria containing mef/mel(msr(D)) elements.

    Science.gov (United States)

    Chancey, Scott T; Bai, Xianhe; Kumar, Nikhil; Drabek, Elliott F; Daugherty, Sean C; Colon, Thomas; Ott, Sandra; Sengamalay, Naomi; Sadzewicz, Lisa; Tallon, Luke J; Fraser, Claire M; Tettelin, Hervé; Stephens, David S

    2015-01-01

    Macrolide resistance, emerging in Streptococcus pneumoniae and other Gram-positive bacteria, is increasingly due to efflux pumps encoded by mef/mel(msr) operons found on discrete mobile genetic elements. The regulation of mef/mel(msr) in these elements is not well understood. We identified the mef(E)/mel transcriptional start, localized the mef(E)/mel promoter, and demonstrated attenuation of transcription as a mechanism of regulation of macrolide-inducible mef-mediated macrolide resistance in S. pneumoniae. The mef(E)/mel transcriptional start site was a guanine 327 bp upstream of mef(E). Consensus pneumococcal promoter -10 (5'-TATACT-3') and -35 (5'-TTGAAC-3') boxes separated by 17 bp were identified 7 bp upstream of the start site. Analysis of the predicted secondary structure of the 327 5' region identified four pairs of inverted repeats R1-R8 predicted to fold into stem-loops, a small leader peptide [MTASMRLR, (Mef(E)L)] required for macrolide induction and a Rho-independent transcription terminator. RNA-seq analyses provided confirmation of transcriptional attenuation. In addition, expression of mef(E)L was also influenced by mef(E)L-dependent mRNA stability. The regulatory region 5' of mef(E) was highly conserved in other mef/mel(msr)-containing elements including Tn1207.1 and the 5612IQ complex in pneumococci and Tn1207.3 in Group A streptococci, indicating a regulatory mechanism common to a wide variety of Gram-positive bacteria containing mef/mel(msr) elements.

  6. Display of rotational levels near the fission threshold in 232Th(n,f) reaction

    International Nuclear Information System (INIS)

    Blons, J.; Mazur, C.; Paya, D.

    1975-01-01

    The 232 Th(n,f) cross section has been measured relative to that of 235 U up to 5MeV, with a neutron energy resolution of 3keV at 1.6MeV. The angular anisotropy of fission fragments has also been measured in the same energy range with an energy resolution of 6keV at 1,6MeV. The broad vibrational levels located above 1MeV are resolved into sharp structures which are interpreted as rotational states. The rotational constants h 2 /2J of highly deformed 233 Th are found to be 2.45 and 2.65keV at 1.5 and 1.6MeV respectively. These results are interpreted by the possibility of a third minimum in the fission barrier [fr

  7. Tracing changes in mantle and crustal influences in individual cone-building stages at Mt. Shasta using U-Th and Sr isotopes

    Science.gov (United States)

    Wende, Allison M.; Johnson, Clark M.; Beard, Brian L.

    2015-10-01

    230Th-excess is rare in most arc lavas, but common in the Cascades, yet the origin of such excesses remains unclear. At Mt. Shasta, age-corrected (230Th/232Th) and (238U/232Th) activity ratios range from 1.108 to 1.290 and from 0.987 to 1.309 (27.3% 230Th-excess to 6.1% 238U-excess), respectively. Although small degrees of zircon crystallization (ancestral cone (Sand Flat) was followed by four cone-building stages, three of which lie in the age range of U-series geochronology. Lavas within individual eruptive stages have relatively constant (230Th/232Th)0 ratios that are interpreted to reflect specific mixtures of mantle (m) and lower crustal (lc) melts that are characteristic of a specific stage (Mm:lc). High (230Th/232Th)0 ratios identify higher proportions of lower crust in the Misery Hill stage (Mm:lc = ∼ 85 : 15), whereas low (230Th/232Th)0 ratios reflect the more mantle-like composition of the Shastina lavas (Mm:lc = ∼ 95 : 5); in the case of Shastina lavas, very low 87Sr/86Sr ratios, down to 0.7029, support a substantial mantle contribution. Changes in (230Th/232Th)0 ratios correlate with eruptive volume, where the most voluminous stage (Misery Hill) is inferred to have the largest proportion of crustal melt and highest (230Th/232Th)0 ratios. Variable (230Th/238U)0 ratios within, and between, eruptive groups likely reflect a combination of residence time in the lower crust and differential assimilation of bulk, non-garnet-bearing crust that had (230Th/238U) = 1. The volume-(230Th/232Th)0 relations are accompanied by correlations with 87Sr/86Sr ratios, where the most radiogenic Sr is associated with the largest eruptive volumes, indicating that the largest magmatic episodes produced the largest amount of lower crustal interaction. The new U-Th and Sr isotope measurements of this study, along with U-series data for other Cascade centers suggest that interaction with the lower crust exerts greater control on Cascade magma chemistry than previously

  8. Neutron-induced reactions on U and Th - A new approach via AMS

    International Nuclear Information System (INIS)

    Wallner, A.; Capote, R.; Christl, M.; Fifield, L.K.; Srncik, M.; Tims, S.; Hotchkis, M.; Krasa, A.; Lachner, J.; Lippold, J.; Plompen, A.; Semkova, V.; Steier, P.; Winkler, S.

    2014-01-01

    Recent studies exhibit discrepancies at keV and MeV energies between major nuclear data libraries for 238 U(n,γ), 232 Th(n,γ) and also for (n,xn) reactions. We have extended our initial (n,γ) measurements on 235,238 U to higher neutron energies and to additional reaction channels. Neutron-induced reactions on 232 Th and 238 U were measured by a combination of the activation technique and atom counting of the reaction products using accelerator mass spectrometry (AMS). Natural thorium and uranium samples were activated with quasi-monoenergetic neutrons at IRMM. Neutron capture data were produced for neutron energies between 0.5 and 5 MeV. Fast neutron-induced reactions were studied in the energy range from 17 to 22 MeV. Preliminary data indicate a fair agreement with data libraries; however at the lower band of existing data. This approach represents a complementary method to on-line particle detection techniques and also to conventional decay counting. (authors)

  9. Preparations for an optical access to the lowest nuclear excitation in {sup 229}Th

    Energy Technology Data Exchange (ETDEWEB)

    Wense, Lars v.d.; Seiferle, Benedict; Thirolf, Peter G. [Ludwig-Maximilians-Universitaet Muenchen (Germany); Laatiaoui, Mustapha [GSI Helmholtzzentrum fuer Schwerionenforschung, Darmstadt (Germany)

    2015-07-01

    The isomeric lowest excited nuclear level of {sup 229}Th has been indirectly measured to be 7.6±0.5 eV (163±11 nm). This low transition energy, compared to energies typically involved in nuclear processes, would allow for the application of laser-spectroscopic methods. Also considering the isomeric lifetime of the excited state (estimated to be 10{sup 3} to 10{sup 4} s), which leads to an extremely sharp linewidth of Δω/ω ∝ 10{sup -20}, the isomer becomes a strong candidate for a nuclear-based frequency standard. In order to directly detect the isomeric ground-state decay and improve the accuracy of its energy as a prerequisite for an all-optical control, {sup 229m}Th is populated via a 2% decay branch in the α decay of {sup 233}U. The Thorium ions are extracted and cooled with the help of a buffer-gas stopping cell and an RFQ-cooler. In order to suppress accompanying α decay chain products other than {sup 229}Th, a quadrupole mass spectrometer (QMS) is used. Following the QMS, the Thorium isomeric decay is expected to be detectable. Internal conversion as well as photonic decay is probed via different detection techniques. Latest results are presented.

  10. Feasibility of Thorium Fuel Cycles in a Very High Temperature Pebble-Bed Hybrid System

    Directory of Open Access Journals (Sweden)

    L.P. Rodriguez

    2015-08-01

    Full Text Available Nuclear energy presents key challenges to be successful as a sustainable energy source. Currently, the viability of the use thorium-based fuel cycles in an innovative nuclear energy generation system is being investigated in order to solve these key challenges. In this work, the feasibility of three thorium-based fuel cycles (232Th-233U, 232Th-239Pu, and 232Th-U in a hybrid system formed by a Very High Temperature Pebble-Bed Reactor (VHTR and two Pebble-Bed Accelerator Driven Systems (ADSs was evaluated using parameters related to the neutronic behavior such as nuclear fuel breeding, minor actinide stockpile, the energetic contribution of each fissile isotope, and the radiotoxicity of the long lived wastes. These parameters were used to compare the fuel cycles using the well-known MCNPX ver. 2.6e computational code. The results obtained confirm that the 232Th-233U fuel cycle is the best cycle for minimizing the production of plutonium isotopes and minor actinides. Moreover, the inclusion of the second stage in the ADSs demonstrated the possibility of extending the burnup cycle duration and reducing the radiotoxicity of the discharged fuel from the VHTR.

  11. U-Th/ESR combined dating of faunal remains from the Mousterian open site of Beauvais (France)

    International Nuclear Information System (INIS)

    Michel, V.; Masaoudi, H.; Falgueres, Ch.; Yokoyama, Y.; Locht, J.L.; Antoine, P.

    1999-01-01

    Faunal remains from the Beauvais open site 'La Justice' (Oise) are the subject of an U-Th dating (disequilibrium of the uranium chain) and an ESR dating (electronic spin resonance). This study was performed in order to identify the chronological situation of the Middle Paleolithic levels of this site located in the north of France, supposedly aged stage 4 after stratigraphic correlations. U-Th ages of bones and dentine are between 20 and 200 ka and are scattered; however, the combined ESR/U-Th ages of rhinoceros dental enamels are homogeneous and indicate that the archaeological levels were deposited between 60 to 40 ka. This period corresponds to the end of the oxygen isotopic stage 4 to the beginning of stage 3. (authors) stage 3. (authors)

  12. Synthesis, characterization, kinetic and thermodynamic studies of the dissolution of ThO{sub 2} and of solid solutions Th{sub 1-x}M{sub x}O{sub 2} (M = U, Pu); Synthese, caracterisation et etudes cinetique et thermodynamique de la dissolution de ThO{sub 2} et des solutions solides Th{sub 1-x}M{sub x}O{sub 2} (M = U, Pu)

    Energy Technology Data Exchange (ETDEWEB)

    Heisbourg, G

    2003-12-01

    The aim of this work was to understand the mechanisms of dissolution of ThO{sub 2} and of thorium mixed oxides such as Th{sub 1-x}U{sub x}O{sub 2} and Th{sub 1-x}Pu{sub x}O{sub 2} in aqueous, oxygenated or inert media. Several solids have been synthesized by precipitation in oxalic medium: Th{sub 1-x}U{sub x}O{sub 2} (x= 0.11; 0.24; 0.37; 0.53; 0.67; 0.81 and 0.91) and Th{sub 1-x}Pu{sub x}O{sub 2} (x= 0.13; 0.32 and 0.66). They have been characterized by XRD, SEM, TEM, XPS, XAS, PIXE and EPMA. The sintering conditions of these materials have been studied and optimized in order to obtain sintered samples with a measured density very near the theoretical densities. A kinetic study of the dissolution of ThO{sub 2} and of solid solutions Th{sub 1-x}U{sub x}O{sub 2} has been carried out in several aqueous media (HNO{sub 3}, HCl, H{sub 2}SO{sub 4}) in terms of several parameters: protons concentration, temperature, pH, ionic strength, nature of the electrolyte solution and uranium molar ratio for the solid solutions Th{sub 1-x}U{sub x}O{sub 2} in order to determine the kinetic laws of dissolution of the solid solutions having different compositions comparatively to ThO{sub 2}. The leaching tests carried out in natural waters of compositions near those of the deep geologic sites considered for the storage of nuclear wastes have shown that the dissolution of the solids was bound to the complexing effect of the constitutional ions of the water considered. The leaching tests carried out on sintered samples of the same composition have led to the same normalized dissolution velocities. The thermodynamic aspect of the dissolution of the solid solutions Th{sub 1-x}U{sub x}O{sub 2} in nitric medium has been studied at last. (O.M.)

  13. Studies on {sup 232}Th and {sup 238}U levels in marine algae collected from the coast of Niigata Prefecture

    Energy Technology Data Exchange (ETDEWEB)

    Kato, Kenji; Tonouchi, Shigemasa; Maruta, Fumiyuki; Ebata, Hidekazu [Niigata Prefectural Inst. of Public Health and Environmental Sciences (Japan)

    2001-12-01

    To evaluate the properties of algae to concentrate radioactive elements, 14 species of algae like Sargassum were collected in the Prefecture and analyzed for their {sup 232}Th and {sup 238}U levels with Yokogawa HP4500 ICP-MS apparatus. The places of collection included those near the water discharge of an atomic power station. Mean {sup 232}Th and {sup 238}U levels were found to be 120 and 260 ng/g dry wt, respectively, and Phaeophyta showed more than several times higher {sup 238}U level than Chlorophyta and Rhodophyta. There was no clear difference in {sup 232}Th levels. No difference between places of collection was observed in Sargassum {sup 232}Th or {sup 238}U level. Adsorption of {sup 232}Th particle to and incorporation of soluble {sup 238}U into algae body were suggested. Mean {sup 232}Th and {sup 238}U radioactivities were found 73 and 510 {mu}Bq/g wet wt, respectively, and the respective annual committed effective doses, 0.2 and 0.3 {mu}Sv, calculated from those values were confirmed to be enough lower than the annual public dose limit, 1 mSv. (K.H.)

  14. New adsorbents from oil shales. Preparation, characterization and U, Th isotope adsorption tests

    International Nuclear Information System (INIS)

    Khouya, E.; Andres, Y.; Naslain, R.; Pailler, R.; Nourredine, A.

    2004-01-01

    New activated adsorbents for radionuclides have been produced from Moroccan oil shales by pyrolysis of the natural material at 550 deg C flowed by a KMnO 4 activation. The texture and composition of the native rock and the adsorbents were studied before their use in tests for adsorption of radionuclides from standard solutions prepared from uranylnitrate and thorium nitrate in equilibrium with their daughters. The distribution coefficients between solutions containing U, Th and Ra and the adsorbents were evaluated by means of specific activities, measured by γ-ray spectrometry. The adsorbents were observed to eliminate U, Th, Ra, Ac and Tl from aqueous solutions. (author)

  15. Effect of anthropogenic organic complexants on the solubility of Ni, Th, U(IV) and U(VI)

    Energy Technology Data Exchange (ETDEWEB)

    Felipe-Sotelo, M., E-mail: m.felipe-sotelo@lboro.ac.uk [Department of Chemistry, Loughborough University, LE11 3TU Loughborough, Leicestershire (United Kingdom); Edgar, M. [Department of Chemistry, Loughborough University, LE11 3TU Loughborough, Leicestershire (United Kingdom); Beattie, T. [MCM Consulting. Täfernstrasse 11, CH 5405 Baden-Dättwil (Switzerland); Warwick, P. [Enviras Ltd., LE11 3TU Loughborough, Leicestershire (United Kingdom); Evans, N.D.M.; Read, D. [Department of Chemistry, Loughborough University, LE11 3TU Loughborough, Leicestershire (United Kingdom)

    2015-12-30

    Highlights: • Citrate increases the solubility of Ni, Th and U between 3 and 4 orders of magnitude. • Theophrastite is the solubility controlling phase of Ni in 95%-saturated Ca(OH){sub 2}. • U(VI) and Ni may form Metal-citrate-OH complexes stabilised by the presence of Ca{sup 2+}. - Abstract: The influence of anthropogenic organic complexants (citrate, EDTA and DTPA from 0.005 to 0.1 M) on the solubility of nickel(II), thorium(IV) and uranium (U(IV) and U(VI)) has been studied. Experiments were carried out in 95%-saturated Ca(OH){sub 2} solutions, representing the high pH conditions anticipated in the near field of a cementitious intermediate level radioactive waste repository. Results showed that Ni(II) solubility increased by 2–4 orders of magnitude in the presence of EDTA and DTPA and from 3 to 4 orders of magnitude in the case of citrate. Citrate had the greatest effect on the solubility of Th(IV) and U(IV)/(VI). XRD and SEM analyses indicate that the precipitates are largely amorphous; only in the case of Ni(II), is there some evidence of incipient crystallinity, in the form of Ni(OH){sub 2} (theophrastite). A study of the effect of calcium suggests that U(VI) and Ni(II) may form metal-citrate-OH complexes stabilised by Ca{sup 2+}. Thermodynamic modelling underestimates the concentrations in solution in the presence of the ligands for all the elements considered here. Further investigation of the behaviour of organic ligands under hyperalkaline conditions is important because of the use of the thermodynamic constants in preparing the safety case for the geological disposal of radioactive wastes.

  16. 230Th/234U age of a Mousterian site in France

    International Nuclear Information System (INIS)

    Schwarcz, H.P.; Blackwell, B.

    1983-01-01

    The 230 Th/ 234 U dating method has been used to determine the age of a travertine layer in the Pech 1 cave of Sarlat, South-West France. The reported age of 123 +- 15 kyr is consistent with deposition of this travertine during isotope state 5e, the warmest substage of the last inter-glacial. It is shown that this result is consistent with geological and paleontological estimates of the age of sediments filling the cave. (U.K.)

  17. Thorium cycles and proliferation

    International Nuclear Information System (INIS)

    Lovins, A.B.

    1979-01-01

    This paper analyzes several prevalent misconceptions about nuclear fuel cycles that breed fissile uranium-233 from thorium. Its main conclusions are: U-233, despite the gamma radioactivity of associated isotopes, is a rather attractive material for making fission bombs, and is a credible material for subnational as well as national groups to use for this purpose; (2) pure thorium cycles, which in effect merely substitute U-233 for Pu, would take many decades and much U to establish, and offer no significant safeguards advantage over Pu, cycles; (3) denatured Th-U cycles, which dilute the U-233 with inert U-238 to a level not directly usable in bombs, are not an effective safeguard even against subnational bomb-making; (4) several other features of mixed Th-U cycles are rather unattractive from a safeguards point of view; (5) thus, Th cycles of any kind are not a technical fix for proliferation (national or subnational) and, though probably more safeguardable than Pu cycles, are less so than once-through U cycles that entail no reprocessing; (6) while thorium cycles have some potential technical advantages, including flexibility, they cannot provide major savings in nuclear fuel resources compared to simpler ways of saving neutrons and U; and (7) while advocates of nuclear power may find Th cycles worth exploring, such cycles do not differ fundamentally from U cycles in any of the respects--including safeguards and fuel resources--that are relevant to the broader nuclear debate, and should not be euphorically embraced as if they did

  18. Partitioning of U, Th and K Between Metal, Sulfide and Silicate, Insights into the Volatile-Content of Mercury

    Science.gov (United States)

    Habermann, M.; Boujibar, A.; Righter, K.; Danielson, L.; Rapp, J.; Righter, M.; Pando, K.; Ross, D. K.; Andreasen, R.; Chidester, B.

    2016-01-01

    During the early stages of the Solar System formation, especially during the T-Tauri phase, the Sun emitted strong solar winds, which are thought to have expelled a portion of the volatile elements from the inner solar system. It is therefore usually believed that the volatile depletion of a planet is correlated with its proximity to the Sun. This trend was supported by the K/Th and K/U ratios of Venus, the Earth, and Mars. Prior to the MESSENGER mission, it was expected that Mercury is the most volatile-depleted planet. However, the Gamma Ray Spectrometer of MESSENGER spacecraft revealed elevated K/U and K/Th ratios for the surface of Mercury, much higher than previous expectations. It is possible that the K/Th and K/U ratios on the surface are not a reliable gauge of the bulk volatile content of Mercury. Mercury is enriched in sulfur and is the most reduced of the terrestrial planets, with oxygen fugacity (fO2) between IW-6.3 and IW-2.6 log units. At these particular compositions, U, Th and K behave differently and can become more siderophile or chalcophile. If significant amounts of U and Th are sequestered in the core, the apparent K/U and K/Th ratios measured on the surface may not represent the volatile budget of the whole planet. An accurate determination of the partitioning of these elements between silicate, metal, and sulfide phases under Mercurian conditions is therefore essential to better constrain Mercury's volatile content and assess planetary formation models.

  19. Analysis of the running-in phase of a Passively Safe Thorium Breeder Pebble Bed Reactor

    International Nuclear Information System (INIS)

    Wols, F.J.; Kloosterman, J.L.; Lathouwers, D.; Hagen, T.H.J.J. van der

    2015-01-01

    Highlights: • This work analyzes important trends of the running-in phase of a thorium breeder PBR. • Depletion equations are solved for important actinides and a fission product pair. • Breeding U-233 is achieved in 7 years by cleverly adjusting the feed fuel enrichment. • A safety analysis shows the thorium PBR is passively safe during the running-in phase. - Abstract: The present work investigates the running-in phase of a 100 MW th Passively Safe Thorium Breeder Pebble Bed Reactor (PBR), a conceptual design introduced in previous equilibrium core design studies by the authors. Since U-233 is not available in nature, an alternative fuel, e.g. U-235/U-238, is required to start such a reactor. This work investigates how long it takes to converge to the equilibrium core composition and to achieve a net production of U-233, and how this can be accelerated. For this purpose, a fast and flexible calculation scheme was developed to analyze these aspects of the running-in phase. Depletion equations with an axial fuel movement term are solved in MATLAB for the most relevant actinides (Th-232, Pa-233, U-233, U-234, U-235, U-236 and U-238) and the fission products are lumped into a fission product pair. A finite difference discretization is used for the axial coordinate in combination with an implicit Euler time discretization scheme. Results show that a time dependent adjustment scheme for the enrichment (in case of U-235/U-238 start-up fuel) or U-233 weight fraction of the feed driver fuel helps to restrict excess reactivity, to improve the fuel economy and to achieve a net production of U-233 faster. After using U-235/U-238 startup fuel for 1300 days, the system starts to work as a breeder, i.e. the U-233 (and Pa-233) extraction rate exceeds the U-233 feed rate, within 7 years after start of reactor operation. The final part of the work presents a basic safety analysis, which shows that the thorium PBR fulfills the same passive safety requirements as the

  20. Distribution of K, Na, Th and U in sandstones and shales from western Shikoku, Japan

    International Nuclear Information System (INIS)

    Ishihara, Shunso; Sakamaki, Yukio; Mochizuki, Tsunekazu; Terashima, Shigeru; Endo, Yuji

    1981-01-01

    The regional variation of K, Na, Th and U distributions was studied on 58 sandstones, 81 shales and 3 green schists from the sedimentary terrains across western Shikoku. The geological structure of the studied district is explained. The regional characteristics of the sedimentary rocks are best demonstrated in the composition of the sandstones. The sandstones, in the source areas of which granitic and rhyolitic rocks exist and which have been deposited rapidly, were rich in K, whereas those derived mainly from mafic volcanic areas showed high Na content. The sandstones of the Shimanto Supergroup had the intermediate values, and K and K + Na contents became low in the south where the younger Upper Shimanto Group is exposed. Th and U in both sandstones and shales were highest in the Izumi Group, and generally low in the Shimanto Supergroup. The black shales of the Shimanto Supergroup did not show U-anomaly. In each group, highly matured rocks gave slightly higher Th/U ratio. Highly matured polycyclic sediments contained the least amount of radioactive elements. The radioactive anomaly due to the anomalous K contained in sericite, and that due to U in black shale were found in Chichibu and Sambosan belts. Similar anomaly was discovered in the foot wall of Mn deposits in the same zone. The possibility of anomalous U may be the least in the Shimanto Supergroup. (Kako, I.)

  1. Dissolution rates of unirradiated UO{sub 2}, UO{sub 2} doped with {sup 233}U, and spent fuel under normal atmospheric conditions and under reducing conditions using an isotope dilution method

    Energy Technology Data Exchange (ETDEWEB)

    Ollila, Kaija [VTT Processes, Helsinki (Finland); Albinsson, Yngve [Chalmers Univ. of Technology, Goeteborg (Sweden); Oversby, Virginia [VMO Konsult, Stockholm (Sweden); Cowper, Mark [AEA Technology, Harwell (United Kingdom)

    2003-10-01

    additional meaningful data. 8) A test procedure that used several short exposures of the sample to solution - the puff test procedure - gave results that showed very little recovery of the spike solution at the end of the tests. Only 10% of the {sup 235}U added as spike was recovered, indicating that 90% of the spike had precipitated onto the solid sample or the iron strip. 9) Tests of UO{sub 2} pellet materials containing {sup 233}U to provide an alpha decay activity similar to that expected for spent fuel 3000 and 10,000 years after disposal showed that the pellet materials behaved as expected under air atmosphere conditions, showing that the manufacturing method was successful. 10) Early testing of the {sup 233}U-doped materials under reducing conditions showed relatively rapid (30 minute) dissolution of small amounts of U at the start of the puff test procedure. Results of analyses of an acidified fraction of the same solutions after 1 or 2 weeks holding indicate that the solutions were inhomogeneous, indicating the presence of colloidal material or small grains of solid. 11) Samples from the {sup 233}U-doped tests initially indicated dissolution of solid during the first week of testing, with some indication of more rapid dissolution of the material with the higher doping. 12) The second cycle of testing of the {sup 233}U-doped materials also showed dissolution occurring during the dilution stages of the puff test. The subsequent week of testing also showed small amounts of further dissolution, with hints that the doped samples were dissolving faster than the undoped samples. 13) At the end of 2 weeks of cycle 2 the remaining solution and solid was transferred to a new reaction vessel, the solution was made up to original volume, and a new dose of spike was added. The results of analyses of [U] and isotopic composition show that the measured U is that expected from dilution of the original solution plus adding the spike. 14) Samples taken during 2 weeks of testing of

  2. Preliminary Report on U-Th-Pb Isotope Systematics of the Olivine-Phyric Shergottite Tissint

    Science.gov (United States)

    Moriwaki, R.; Usui, T.; Yokoyama, T.; Simon, J. I.; Jones, J. H.

    2014-01-01

    Geochemical studies of shergottites suggest that their parental magmas reflect mixtures between at least two distinct geochemical source reservoirs, producing correlations between radiogenic isotope compositions, and trace element abundances.. These correlations have been interpreted as indicating the presence of a reduced, incompatible-element- depleted reservoir and an oxidized, incompatible-element-rich reservoir. The former is clearly a depleted mantle source, but there has been a long debate regarding the origin of the enriched reservoir. Two contrasting models have been proposed regarding the location and mixing process of the two geochemical source reservoirs: (1) assimilation of oxidized crust by mantle derived, reduced magmas, or (2) mixing of two distinct mantle reservoirs during melting. The former clearly requires the ancient martian crust to be the enriched source (crustal assimilation), whereas the latter requires a long-lived enriched mantle domain that probably originated from residual melts formed during solidification of a magma ocean (heterogeneous mantle model). This study conducts Pb isotope and U-Th-Pb concentration analyses of the olivine-phyric shergottite Tissint because U-Th-Pb isotope systematics have been intensively used as a powerful radiogenic tracer to characterize old crust/sediment components in mantle- derived, terrestrial oceanic island basalts. The U-Th-Pb analyses are applied to sequential acid leaching fractions obtained from Tissint whole-rock powder in order to search for Pb isotopic source components in Tissint magma. Here we report preliminary results of the U-Th-Pb analyses of acid leachates and a residue, and propose the possibility that Tissint would have experienced minor assimilation of old martian crust.

  3. Critical Experiments With Aqueous Solutions of 233UO2(NO3)2

    International Nuclear Information System (INIS)

    Thomas, J.T.

    2001-01-01

    This report provides the critical experimenter's interpretations and descriptions of informal critical experiment logbook notes and associated information (e.g., experimental equipment designs/sketches, chemical and isotopic analyses, etc.) for the purpose of formally documenting the results of critical experiments performed in the late 1960s at the Oak Ridge Critical Experiments Facility. The experiments were conducted with aqueous solutions of 97.6 wt % 233 U uranyl nitrate having uranium densities varying between about 346 g U/l and 45 g U/l. Criticality was achieved with single simple units (e.g., cylinders and spheres) and with spaced subcritical simple cylindrical units arranged in unreflected, water-reflected, and polyethylene reflected critical arrays

  4. Preparation of some complexes of Th(IV) and U(IV) with tetradentate Schiff bases. The crystal structure of bis(N,N'-ethylenebis(3-methoxysalicylaldiminato)) thorium(IV) monopyridine

    Energy Technology Data Exchange (ETDEWEB)

    Hill, R J; Rickard, C E.F.; White, H E [Auckland Univ. (New Zealand). Dept. of Chemistry

    1981-01-01

    Complexes of Th(IV) and U(IV) with tetradentate Schiff bases derived from substituted salicylaldehydes have been prepared and characterised. The structure of bis(N,N'-ethylenebis(3-methoxysalicylaldiminato)) Th(IV) monopyridine has been determined by X-ray crystallographic methods. The crystals are triclinic, a = 13.468, b = 9.932, c = 16.552 A, ..cap alpha.. = 91.74, ..beta.. = 94.69, ..gamma.. = 93.03/sup 0/, space group P/sub 1//sup -/. The molecules are eight coordinate with a slightly distorted dodecahedral geometry with the imine nitrogen atoms in the dodecahedral A sites. The pyridine molecule is uncoordinated and functions in a space-filling role.

  5. Preparation of high density (Th, U)O2 pellets by sol-gel microsphere pelletization and 1300 C air sintering

    International Nuclear Information System (INIS)

    Yamagishi, Shigeru; Takahashi, Yoshihisa

    1994-01-01

    The fabrication of high density (Th, U)O 2 pellets by the sol-gel microsphere pelletization (SGMP) process was studied. To prepare source ThO 2 -UO 3 microspheres, isopropyl alcohol was substituted for the water in gel and thereafter removed by evacuating and subsequently by heating at 200 C in air. After humidifying the microspheres up to the moisture content ranging 10-21%, they were compacted into a pellet under 150-500 MPa and sintered in air at 1300 C. Even at the relatively low temperature, the maximum density reached 98% TD or higher for the U/(Th+U) ratios of 5-20 mol%. Such high density products survived as firm pellets with a similarly high density of 99% TD during the reduction into (Th, U)O 2 in Ar-4% H 2 at 1300 C. ((orig.))

  6. Rapid screening of natually occurring radioactive nuclides({sup 2}'3{sup 8}U, {sup 232}Th) in raw materials and by-products samples using XRF

    Energy Technology Data Exchange (ETDEWEB)

    Park, Ji Young; Lim, Chung Sup [Radiation Biotechnology and Applied Radioiostope Science, University of Science and Technology, Daejeon (Korea, Republic of); Lim, Jong Myoung; Ji, Young Yong; Chung, Kun Ho; Lee, Wan No; Kang, Mun Ja [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Jang, Byung Uck [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2016-12-15

    As new legislation has come into force implementing radiation safety management for the use of naturally occurring radioactive materials (NORM), it is necessary to establish a rapid and accurate measurement technique. Measurement of {sup 238}U and {sup 232}Th using conventional methods encounter the most significant difficulties for pretreatment (e.g., purification, speciation, and dilution/enrichment) or require time-consuming processes. Therefore, in this study, the applicability of ED-XRF as a non-destructive and rapid screening method was validated for raw materials and by-product samples. A series of experiments was conducted to test the applicability for rapid screening of XRF measurement to determine activity of {sup 238}U and {sup 23{sup 2}}Th based on certified reference materials (e.g., soil, rock, phosphorus rock, bauxite, zircon, and coal ash) and NORM samples commercially used in Korea. Statistical methods were used to compare the analytical results of ED-XRF to those of certified values of certified reference materials (CRM) and inductively coupled plasma mass spectrometry (ICP-MS). Results of the XRF measurement for {sup 238}U and {sup 232}Th showed under 20% relative error and standard deviation. The results of the U-test were statistically significant except for the case of U in coal fly ash samples. In addition, analytical results of {sup 238}U and {sup 232}Th in the raw material and by-product samples using XRF and the analytical results of those using ICP-MS (R{sup 2}≥0.95) were consistent with each other. Thus, the analytical results rapidly derived using ED-XRF were fairly reliable. Based on the validation results, it can be concluded that the ED-XRF analysis may be applied to rapid screening of radioactivities ({sup 238}U and {sup 232}Th) in NORM samples.

  7. Th, Pa and U isotopes in an echinoderm, Encope grandis

    International Nuclear Information System (INIS)

    Omura, Akio; Ku, Teh-Lung.

    1979-01-01

    The application of 230 Th and 231 Pa growth methods to the hard tissues of living things, which are effective for the radiometric age measurement for latter Quaternary period, has been limited to certain corals, therefore it has been scarcely utilized in other areas than coral reefs. Reef coral fossils (Porites) were obtained from terrace deposits of Magdalena Island in Southern Baja California, and the methods were applied to them. At the time, the isotope compositions of Th, Pa and U in the shells of echinoderm Encope Grandis and of the living samples were examined. The estimated ages were in agreement with those of coral. It suggests that the reliable 230 Th and 231 Pa ages of sea-urchin fossils were presented for the first time and that the method is applicable to such fossils only if the conditions can be met. The results are highly significant, since the method may be used in other areas than coral reefs. (J.P.N.)

  8. Release of U, Th, and REE from granitic rock: A mineralogical approach

    International Nuclear Information System (INIS)

    Markovaara-Koivisto, M.

    2006-01-01

    Finland plans to dispose of its spent nuclear fuel deep in the bedrock, and comprehensive assessment of the potential risks is required. One risk is glaciations induced by climate change, which might eventually cause malfunction of the engineered barrier system and breakdown of the copper-iron canisters containing the spent fuel. The fuel might then come into contact with groundwater. This groundwater might be acidic rain water, or oxygenated glacial melt water, which intrudes into the bedrock with hydrostatic pressure under the ice sheet. In this study, behaviour of uranium and rare earth elements was investigated in the Palmottu uranium deposit. Studies in the Palmottu deposit provide an indication of how uranium and other harmful elements could migrate from the repository to the surrounding bedrock in the event the canisters were breached. The spent fuel contains uranium and other actinides. The possible release of these elements and their behaviour after release in bedrock and groundwater were studied by means of chemical analogues occurring in nature, namely uranium (U), thorium (Th) and rare earth elements (REE). The study was focused on the mode of occurrence of these elements in granitic rocks. The chemistry of the mineral phases was explored by scanning electron microscopy and wavelength dispersive spectrometry, while the release of the elements was investigated with leaching experiments. In the first phase the samples were leached with artificial groundwater. In the second phase a HNO 3 solution of pH 5 was used, and in the final step a solution of pH 3. The U, Th and REE phases after each leaching were studied by fieldemission scanning electron microscopy and energy dispersive XRay microanalysis (EDAX), and the leachates were analysed by mass spectrometry (ICPMS and ICPAES). The aim of this study was to clarify how U, Th and REEs behave in the leaching processes associated with solutions simulating possible natural water conditions in the bedrock and to

  9. Concentration of radioactive elements (U, Th and K derived from phosphatic fertilizers in cultivated soils

    Directory of Open Access Journals (Sweden)

    Valter Antonio Becegato

    2008-12-01

    Full Text Available Gamma spectrometric measurements were obtained for the agricultural soils aiming at characterizing the spatial distribution of radionuclide concentrations (K, eU and eTh, as well for the samples of phosphatic fertilizers and agricultural gypsum. In the study areas, three types of soils occured: Eutrophic Red Nitosol (Alfisoil, Eutroferric Red Latosol of clayey texture (Oxisoil and Dystrophic Red Latosol of medium texture (Oxisoil. The results showed that the radionuclide concentrations in more clayey soils were higher than in more sandy soils, mainly as a function of a higher adsorption capacity of the former. For the area where human activity predominated, the average contents of K, eU and eTh were respectively 54.75; 10.22 and 7.27 Bq/Kg, significantly higher than those for the area where no fertilizers were used (34.15 Bq/Kg K; 1.69 Bq/Kg eU, and 5.36 Bq/Kg eTh. Variations in the radionuclide concentrations were also observed in various fertilizer formula used in soybean and wheat crops.Medições gamaespectrométricas foram obtidas em solos agrícolas objetivando caracterizar a distribuição espacial das concentrações de radionuclídeos (K, eU e eTh, bem como em amostras de fertilizantes fosfatados e gesso agrícola. Na área ocorrem três tipos de solos: Nitossolo Vermelho Eutrófico, Latossolo Vermelho Eutroférrico textura argilosa e Latossolo Vermelho Distrófico textura média. Constatou-se que as concentrações de radionuclídeos nos solos mais argilosos foram maiores do que nos solos mais arenosos, em função, principalmente, da maior adsorção pelos primeiros. Os teores médios em Bq/Kg de K, eU e eTh na área com atividade antrópica foram respectivamente de 54,75; 10,22 e 7,27, significativamente maiores do que em áreas virgens sem aplicação de fertilizantes (34,15 de K; 1,69 de eU e 5,36 de eTh. Foram também observadas variações nas concentrações de radionuclídeos em diferentes formulações de adubos utilizados nas

  10. The mobility of U and Th in subduction zone fluids: an indicator of oxygen fugacity and fluid salinity

    Science.gov (United States)

    Bali, Enikő; Audétat, Andreas; Keppler, Hans

    2011-04-01

    The solubility of U and Th in aqueous solutions at P-T-conditions relevant for subduction zones was studied by trapping uraninite or thorite saturated fluids as synthetic fluid inclusions in quartz and analyzing their composition by Laser Ablation-ICPMS. Uranium is virtually insoluble in aqueous fluids at Fe-FeO buffer conditions, whereas its solubility increases both with oxygen fugacity and with salinity to 960 ppm at 26.1 kbar, Re-ReO2 buffer conditions and 14.1 wt% NaCl in the fluid. At 26.1 kbar and 800°C, uranium solubility can be reproduced by the equation: log {{U}} = 2.681 + 0.1433log f{{O}}2 + 0.594{{Cl,}} where fO2 is the oxygen fugacity, and Cl is the chlorine content of the fluid in molality. In contrast, Th solubility is generally low (uranium increases strongly both with oxygen fugacity and with salinity. We show that reducing or NaCl-free fluids cannot produce primitive arc magmas with U/Th ratio higher than MORB. However, the dissolution of several wt% of oxidized, saline fluids in arc melts can produce U/Th ratios several times higher than in MORB. We suggest that observed U/Th ratios in arc magmas provide tight constraints on both the salinity and the oxidation state of subduction zone fluids.

  11. Contamination level of natural 238U and 232Th radionuclides in offshore of coal power plant (assessment at offshore of Panjang Island and Lada Bay, Banten)

    International Nuclear Information System (INIS)

    Sabam Parsaoran Situmorang; Harpasis Selamet Sanusi; June Mellawati

    2011-01-01

    This study had been carried out by collecting sample of the surficial sediments, sea water, seaweeds, anchovies (Stolephorus and Anchoa) and mussels (Codakia) from 4 locations in waters of Pulau Panjang and coastal of Lada Bay (as control/comparison site), Banten in June - July 2010. Natural radionuclides (Th) concentration in samples was measured using neutron activation analysis (NAA) method. The results showed that the total radionuclides concentration in sediment ( 238 U: 18.6160 - 35.0013 Bq/kg; 232 Th: 11.2020 - 35.6685 Bq/kg), seawater ( 238 U: undetected; 232 Th: 0.0790 - 0.1299 Bq/l), cultivation seaweeds ( 238 U: undetected; 232 Th: 3.6735 - 4.8345 Bq/kg), natural seaweeds ( 238 U: 3.6851 - 48.0430 Bq/kg; 232 Th: 3.9941 - 9.0788 Bq/kg), Stolephorus ( 238 U: undetected; 232 Th: 3.3078 Bq/kg) and Codakia ( 238 U: 6.8903 Bq/kg; 232 Th: 3.6023 Bq/kg) in Pulau Panjang, Banten around Suralaya coal power plant higher than control site that were around the Labuan coal power plant, namely in sediments ( 238 U: 10.4253 Bq/kg; 232 Th: 16.5952 Bq/kg), seawater( 238 U: undetected; 232 Th: 0.0671 Bq/l), cultivation seaweeds ( 238 U: undetected; 232 Th: 2.3005 Bq/kg), natural seaweeds ( 238 U: 19.5367 Bq/kg; 232 Th: 2.6729 Bq/kg) and Anchoa ( 238 U: undetected; 232 Th: 2.0603 Bq/kg). (author)

  12. MSR - SPHINX concept program Eros (Experimental zero power Salt reactor SR-0) - The proposed experimental program as a basis for validation of reactor physics methods

    Energy Technology Data Exchange (ETDEWEB)

    Hron, M.; Juricek, V.; Kyncl, J.; Mikisek, M.; Rypar, V. [Nuclear Research Institute Rez plc, Rez (Czech Republic)

    2007-07-01

    The Molten Salt Reactor (MSR) - SPHINX (SPent Hot fuel Incinerator by Neutron fluX) concept solves this principal problem of spent fuel treatment by means of so-called nuclear incineration. It means the burning of fissionable part of its inventory and transmutation of other problematic radionuclides by use of nuclear reactions with neutrons in a MSR-SPHINX system. This reactor system is an actinide burner (most in resonance neutron spectrum) and a radionuclide transmuter in a well-thermalized neutron spectrum. In the frame of the physical part, there are computational analyses and experimental activities. The experimental program has been focused, in its first stage, on a short-term irradiation of small size samples of molten-salt systems as well as structural materials proposed for the MSR blanket in the field of high neutron flux of research reactors. The proposed next stage of the program will focus on a large-scale experimental verification of design inputs by use of MSR-type inserting zones into the existing light water moderated experimental reactor LR-0, which may allow us to modify it into the experimental zero power salt reactor SR-0. There will be a detail description of the proposed program given in the paper together with the so far performed experiments and their first results. These realized experiments help us also to verify computational codes used, and to recognize some anomalies related to molten fluorides utilization. (authors)

  13. Calibration of the 14C timescale over the past 30,000 years using mass spectrometric U-Th ages from Barbados corals

    International Nuclear Information System (INIS)

    Bard, E.; Hamelin, B.; Fairbanks, R.G.; Zindler, A.

    1990-01-01

    Uranium-thorium ages obtained by mass spectrometry from corals raised off the island of Barbados confirm the high precision of this technique over at least the past 30,000 years. Comparison of the U-Th ages with 14 C ages obtained on the Holocene samples shows that the U-Th ages are accurate, because they accord with the dendrochronological calibration. Before 9,000 yr BP the 14 C ages are systematically younger than the U-Th ages, with a maximum difference of ∼3,500 yr at ∼20,000 yr BP. The U-Th technique thus provides a way of calibrating the radiocarbon timescale beyond the range of dendrochronological calibration. (author)

  14. Is the beagle dog an appropriate experimental animal for extrapolating data to humans on organ distribution patterns of U, Th, and Pu

    International Nuclear Information System (INIS)

    Singh, N.P.; Wrenn, M.E.

    1989-01-01

    Concentrations and organ distribution patterns of alpha-emitting isotopes of U (238U and 234U), Th (232Th, 230Th, and 228Th), and Pu (239,240Pu) were determined for beagle dogs of our colony. The dogs were exposed to environmental levels of U and Th isotopes through ingestion (food and water) and inhalation to stimulate environmental exposures of the general human population. The organ distribution patterns of these radionuclides in beagles are compared to patterns in humans to determine if it is appropriate to extrapolate organ content data from beagles to humans. The results indicated that approximately 80% of the U and Th accumulated in bone in both species. The organ content percentages of these radionuclides in soft tissues such as liver, kidney, etc. of both species were comparable. The human lung contained higher percentages of U and Th than the beagle lung, perhaps because the longer life span of humans resulted in a longer exposure time. If the U and Th content of dog lung is normalized to an exposure time of 58 y and 63 y, median ages of the U and Th study populations, respectively, the lung content for both species is comparable. The organ content of 239,240Pu in humans and beagles differed slightly. In the beagle, the liver contained more than 60%, and the skeleton contained less than 40% of the Pu body content. In humans, the liver contained approximately 37%, and the skeleton contained approximately 58% of the body content. This difference may have been due to differences in the mode of intake of Pu in each species or to differences in the chemical form of Pu. In general, the results suggest that the beagle may be an appropriate experimental animal from which to extrapolate data to humans with reference to the percentage of U, Th, and Pu found in the organs

  15. Trace analysis of U, Th and other heavy metals in high purity aluminium with isotope dilution mass spectrometry

    International Nuclear Information System (INIS)

    Beer, B.; Heumann, K.G.

    1992-01-01

    A method for the determination of very low concentrations of U, Th, Fe, Zn, Tl, Cd, Cu and Ag in high purity aluminium with isotope dilution mass spectrometry (IDMS) is developed using a compact and cost-efficient thermal ionization quadrupole mass spectrometer. The detection limits obtained are (in ng/g):U=0.018, Th=0.06, Fe=82, Zn=86, Tl=0.2, Cd=4, Cu=1, Ag=2.6. By this method it is possible to determine the α-emitters U and Th in aluminium down to the sub-ng/g level with good precision of 0.4-10% and 0.5-5%, respectively. The results should also be accurate because IDMS is a reliable analytical method. The dissolution of aluminium is carried out by aqua regia followed by the trace/matrix separation and the isolation of the trace elements by anion exchange chromatography (U, Th, Zn, Tl, Cd), electrodeposition (Cu, Ag) and extraction (Fe). Different aluminium samples are analysed by IDMS and the results are compared with those of other methods. (orig.)

  16. Studies on the sorption behaviours of Th(IV) and U(VI) from aqueous sulphate solutions using impregnated resin

    International Nuclear Information System (INIS)

    Khatab, A.F.; Sheta, M.E.; Mahfouz, M.G.; Tolba, A.A.

    2007-01-01

    The sorption behaviours of thorium (IV) and uranium (VI) from aqueous sulphate solutions have been studied using n-dodecylamine and tri-n-octylamine (TOA) dissolved in benzene and impregnated onto amberlite XAD-4 (styrene-divinyl benzene copolymer). The sorption behaviours were evaluated as a function of free acidity, salting out effect, ph value, equilibrium time, V/m ratio, initial metal ion concentration, loaded amine concentration and sorption temperature. The equilibrium time for Th(IV) and U(VI) sorption from aqueous sulphate solution was found to be 90 and 60 minutes, respectively. The sorption of Th(IV) was quantitatively at ph range 3.7-4.3 and at 4.3-5.2 for U(VI). The sorption capacity of the impregnated resin was determined by batch method and it was found to be 0.031 and 0.033 mmol/g for Th(IV) and U(VI), respectively. Elution of Th(IV) from thorium-loaded impregnated resin was quantitatively achieved by using 2 mol/l HNO 3 and by using 0.1 mol/l HCl for U(VI)

  17. Characterization of bauxite residue (red mud) for 235U, 238U, 232Th and 40K using neutron activation analysis and the radiation dose levels as modeled by MCNP.

    Science.gov (United States)

    Landsberger, S; Sharp, A; Wang, S; Pontikes, Y; Tkaczyk, A H

    2017-07-01

    This study employs thermal and epithermal neutron activation analysis (NAA) to quantitatively and specifically determine absorption dose rates to various body parts from uranium, thorium and potassium. Specifically, a case study of bauxite residue (red mud) from an industrial facility was used to demonstrate the feasibility of the NAA approach for radiological safety assessment, using small sample sizes to ascertain the activities of 235 U, 238 U, 232 Th and 40 K. This proof-of-concept was shown to produce reliable results and a similar approach could be used for quantitative assessment of other samples with possible radiological significance. 238 U and 232 Th were determined by epithermal and thermal neutron activation analysis, respectively. 235 U was determined based on the known isotopic ratio of 238 U/ 235 U. 40 K was also determined using epithermal neutron activation analysis to measure total potassium content and then subtracting its isotopic contribution. Furthermore, the work demonstrates the application of Monte Carlo Neutral-Particle (MCNP) simulations to estimate the radiation dose from large quantities of red mud, to assure the safety of humans and the surrounding environment. Phantoms were employed to observe the dose distribution throughout the human body demonstrating radiation effects on each individual organ. Copyright © 2016 Elsevier Ltd. All rights reserved.

  18. A study on physical characteristics of supercritical light - water reactor loaded with (232U-238Th-238U) oxide fuel

    International Nuclear Information System (INIS)

    Kulikov, E. G.; Shmelev, A. N.; Apse, V. A.; Kulikov, G. G.

    2007-01-01

    The attractiveness of using (U-Th)-fuel in supercritical light water reactor is considered. The dilution of 2 33U in 2 38U is proposed with the purpose of increasing non-proliferation of this fissile isotope. Comparison of different fuel compositions is accomplished from the point of view of fissile isotope breeding and achieved burn-up; parasitic neutron absorption cross-sections are also compared. It is analyzed the impact for neutron balance of both cladding materials: zirconium alloy and stainless steel

  19. Evaluation of an interlaboratory comparison of the chemical assay of U, Th, oxide coated particles

    International Nuclear Information System (INIS)

    Tamberg, T.; Thiele, D.; Brodda, B.G.

    1981-09-01

    The prototype reactor THTR in Schmehausen (Germany, F.R.) burns a (Th,U)O 2 nuclear fuel using 93% enriched uranium. This material is particularly Safeguards sensitive. It was therefore desirable for the Safeguards Analytical Laboratory (SAL) and other laboratories of the Agency Network to collect experience and test their performance in the analysis of such materials. Support was requested from the ''Joint Programme between the IAEA and the Federal Republic of Germany for the Development of Safeguards Techniques'' to perform, as a first step, an interlaboratory comparison of the chemical assay of U and Th in pyrocarbon-coated BISO-type fuel particles. Such an intercomparison was organized under the auspices of the Institut fuer Chemische Technologie (ICT) of the Kernforschungsanlage Juelich GmbH (KFA). SAL prepared a statistical evaluation of the results which was discussed in Vienna in June 1980. The objective of the project was to define the state of the art in the chemical assay of U-Th fuels and the analytical requirements for the sampling of materials of major interest to Agency Safeguards at present

  20. Th, U and trace elements determination in Egyiptian Lake sediments by INAA and laser fluorimetry

    International Nuclear Information System (INIS)

    Ismail, S.S.; Grass, F.; Ghods, A.

    1995-01-01

    A study was undertaken to determine element concentrations in Aswan High Dam Lake sediments. Sediment samples were collected from 40 to 500 km upstream of the dam to follow the sedimentation process and the distribution of Th, U and the trace elements in the lake. INAA was applied for the determination of Sm, Ce, Lu, Th, Cr, Yb, Au, Hf, Ba, Nd, Cs, Tb, Sc, Rb, Fe, Zn, Co, Eu, and Sb, while Laser Fluorimetry was applied for U determination. The accuracy and the reproducibility of the techniques were tested with IAEA standard materials (SL, Soil-7). The U values ranged from 4 ppm to 18 ppm, Th values were between 2 and 10 ppm, and showed a very good correlation with the rare earth elements and Fe. The distribution of most of the elements in the lake follows the same trend as the distribution of the clays in the sediments. Ba showed a negative correlation with most of the elements under investigation. (author) 5 refs.; 4 figs.; 7 tabs

  1. 230Th, 232Th and 238U determinations in phosphoric acid fertilizer and process products by ICP-MS

    International Nuclear Information System (INIS)

    Nascimento, Marcos R.L. do; Guerreiro, Luisa M.R.; Bonifacio, Rodrigo L.; Taddei, Maria H.T.

    2015-01-01

    Through processing of Santa Quiteria-CE mine phosphate rock, Brazil has established a project for production of phosphoric acid fertilizer and uranium as a by-product. Under leaching conditions of phosphate rock with sulfuric acid, which is the common route for preparing phosphoric acid fertilizer, a large part of uranium, thorium and their decay products naturally present in the rock are solubilized. In order to assess the contamination potential in phosphoric acid and others process products, this paper describes a previous precipitation and direct methods for routine analysis of thorium and uranium isotopes by ICP-MS. In all samples, 230 Th, 232 Th and 238 U were directly determined after dilution, except 230 Th in phosphoric acid loaded with uranium sample, which to overcome equipment contamination effect, was determined after its separation by oxalate precipitation using lanthanum as a carrier. The results obtained by the proposed method by ICP-MS, were in good agreement when compared to alpha spectrometry for 230 Th, and ICP-OES and spectrophotometry with arsenazo III for elementary uranium and thorium determinations. (author)

  2. Geophysical interpretation of U, Th, and rare earth element mineralization of the Bokan Mountain peralkaline granite complex, Prince of Wales Island, southeast Alaska

    Science.gov (United States)

    McCafferty, Anne E.; Stoeser, Douglas B.; Van Gosen, Bradley S.

    2014-01-01

    A prospectivity map for rare earth element (REE) mineralization at the Bokan Mountain peralkaline granite complex, Prince of Wales Island, southeastern Alaska, was calculated from high-resolution airborne gamma-ray data. The map displays areas with similar radioelement concentrations as those over the Dotson REE-vein-dike system, which is characterized by moderately high %K, eU, and eTh (%K, percent potassium; eU, equivalent parts per million uranium; and eTh, equivalent parts per million thorium). Gamma-ray concentrations of rocks that share a similar range as those over the Dotson zone are inferred to locate high concentrations of REE-bearing minerals. An approximately 1300-m-long prospective tract corresponds to shallowly exposed locations of the Dotson zone. Prospective areas of REE mineralization also occur in continuous swaths along the outer edge of the pluton, over known but undeveloped REE occurrences, and within discrete regions in the older Paleozoic country rocks. Detailed mineralogical examinations of samples from the Dotson zone provide a means to understand the possible causes of the airborne Th and U anomalies and their relation to REE minerals. Thorium is sited primarily in thorite. Uranium also occurs in thorite and in a complex suite of ±Ti±Nb±Y oxide minerals, which include fergusonite, polycrase, and aeschynite. These oxides, along with Y-silicates, are the chief heavy REE (HREE)-bearing minerals. Hence, the eU anomalies, in particular, may indicate other occurrences of similar HREE-enrichment. Uranium and Th chemistry along the Dotson zone showed elevated U and total REEs east of the Camp Creek fault, which suggested the potential for increased HREEs based on their association with U-oxide minerals. A uranium prospectivity map, based on signatures present over the Ross-Adams mine area, was characterized by extremely high radioelement values. Known uranium deposits were identified in the U-prospectivity map, but the largest tract occurs

  3. The effect of dissolved hydrogen on the dissolution of {sup 233}U doped UO{sub 2}(s) high burn-up spent fuel and MOX fuel

    Energy Technology Data Exchange (ETDEWEB)

    Carbol, P. [Inst. for Transuranium Elements, Karlsruhe (Germany); Spahiu, K. (ed.) [and others

    2005-03-01

    In this report the results of the experimental work carried out in a large EU-research project (SFS, 2001-2004) on spent fuel stability in the presence of various amounts of near field hydrogen are presented. Studies of the dissolution of {sup 233}U doped UO{sub 2}(s) simulating 'old' spent fuel were carried out as static leaching tests, autoclave tests with various hydrogen concentrations and electrochemical tests. The results of the leaching behaviour of a high burn-up spent fuel pellet in 5 M NaCl solutions in the presence of 3.2 bar H{sub 2} pressure and of MOX fuel in dilute synthetic groundwater under 53 bar H{sub 2} pressure are also presented. In all the experimental studies carried out in this project, a considerable effect of hydrogen in the dissolution rates of radioactive materials was observed. The experimental results obtained in this project with a-doped UO{sub 2}, high burn-up spent fuel and MOX fuel together with literature data give a reliable background to use fractional alteration/dissolution rates for spent fuel of the order of 10{sup -6}/yr - 10{sup -8}/yr with a recommended value of 4x10{sup -7}/yr for dissolved hydrogen concentrations above 10{sup -3} M and Fe(II) concentrations typical for European repository concepts. Finally, based on a review of the experimental data and available literature data, potential mechanisms of the hydrogen effect are also discussed. The work reported in this document was performed as part of the Project SFS of the European Commission 5th Framework Programme under contract no FIKW-CT-2001-20192 SFS. It represents the deliverable D10 of the experimental work package 'Key experiments using a-doped UO{sub 2} and real spent fuel', coordinated by SKB with the participation of ITU, FZK-INE, ENRESA, CIEMAT, ARMINES-SUBATECH and SKB.

  4. The effect of dissolved hydrogen on the dissolution of {sup 233}U doped UO{sub 2}(s) high burn-up spent fuel and MOX fuel

    Energy Technology Data Exchange (ETDEWEB)

    Carbol, P [Inst. for Transuranium Elements, Karlsruhe (Germany); Spahiu, K [and others

    2005-03-01

    In this report the results of the experimental work carried out in a large EU-research project (SFS, 2001-2004) on spent fuel stability in the presence of various amounts of near field hydrogen are presented. Studies of the dissolution of {sup 233}U doped UO{sub 2}(s) simulating 'old' spent fuel were carried out as static leaching tests, autoclave tests with various hydrogen concentrations and electrochemical tests. The results of the leaching behaviour of a high burn-up spent fuel pellet in 5 M NaCl solutions in the presence of 3.2 bar H{sub 2} pressure and of MOX fuel in dilute synthetic groundwater under 53 bar H{sub 2} pressure are also presented. In all the experimental studies carried out in this project, a considerable effect of hydrogen in the dissolution rates of radioactive materials was observed. The experimental results obtained in this project with a-doped UO{sub 2}, high burn-up spent fuel and MOX fuel together with literature data give a reliable background to use fractional alteration/dissolution rates for spent fuel of the order of 10{sup -6}/yr - 10{sup -8}/yr with a recommended value of 4x10{sup -7}/yr for dissolved hydrogen concentrations above 10{sup -3} M and Fe(II) concentrations typical for European repository concepts. Finally, based on a review of the experimental data and available literature data, potential mechanisms of the hydrogen effect are also discussed. The work reported in this document was performed as part of the Project SFS of the European Commission 5th Framework Programme under contract no FIKW-CT-2001-20192 SFS. It represents the deliverable D10 of the experimental work package 'Key experiments using a-doped UO{sub 2} and real spent fuel', coordinated by SKB with the participation of ITU, FZK-INE, ENRESA, CIEMAT, ARMINES-SUBATECH and SKB.

  5. Investigate the capability of INAA absolute method to determine the concentrations of 238U and 232Th in rock samples

    International Nuclear Information System (INIS)

    Alnour, I.A.

    2014-01-01

    This work aimed to study the capability of INAA absolute method in determining the elemental concentration of 238 U and 232 Th in the rock samples. The INAA absolute method was implemented in PUSPATI TRIGA Mark II research reactor, Malaysian Nuclear Agency (NM). The accuracy of INAA absolute method was performed by analyzing the IAEA certified reference material (CRM) Soil-7. The analytical results showed the deviations between experimental and certified values were mostly less than 10 % with Z-score in most cases less than 1. In general, the results of analysed CRM Soil-7 show a good agreement between certified and experimental results which mean that the INAA absolute method can be used accurately for elemental analysis of uranium and thorium in various types of samples. The concentration of 238 U and 232 Th ranged from 1.77 to 24.25 and 0.88 to 95.50 ppm respectively. The highest value of 238 U and 232 Th was recorded for granite rock sample G17 of 238 U and sample G9 of 232 Th, whereas the lower value was 1.77 ppm of 238 U recorded in sandstone rock and 0.88 ppm of 232 Th for gabbro. Moreover, a comparison of the 238 U and 232 Th results obtained by the INAA absolute method shows an acceptable level of consistency with those obtained by the INAA relative method. (author)

  6. Disequilibria in the disintegration series of U and Th and chemical parameters in thermal spring waters from the Tatun volcanic area (Taiwan)

    International Nuclear Information System (INIS)

    Lin Chunchih; Chu Tiehchi; Huang Yufen

    2003-01-01

    The activity concentrations of 238 U, 234 U, 230 Th, 226 Ra, 232 Th, and 228 Th in thermal spring waters in the Tatun volcanic area were determined. Parameters including acidity, Cl - and SO 4 2- concentrations in spring waters at the sampling sites have been investigated to allow interpretation of the migration of the radionuclides, and to elucidate the influence of these parameters on the variations of radionuclide contents. Radioactive disequilibria were found in uranium and thorium series in thermal spring waters. The contents of uranium and thorium decreased with increasing pH. The ratios of 230 Th/ 234 U, 226 Ra/ 230 Th and 228 Th/ 232 Th show significant disequilibria. The 226 Ra/ 230 Th ratio (0.60-34.8) decreased with the Cl - or SO 4 2- concentration. All 228 Th/ 232 Th ratios (1.01-9.49) deviated from unity due to the co-precipitation of 228 Ra with barium and lead sulfate. (orig.)

  7. Power density effect on feasibility of water cooled thorium breeder reactor

    International Nuclear Information System (INIS)

    Sidik, Permana; Takaki, Naoyuki; Sekimoto, Hiroshi

    2008-01-01

    Breeding is made possible by the high value of neutron regeneration ratio η for 233 U in thermal energy region. The reactor is fueled by 233 U-Th oxide and it has used the light water as moderator. Some characteristics such as spectrum, η value, criticality, breeding performance and number density are evaluated. Several power densities are evaluated in order to analyze its effect to the breeding performance. The η value of fissile 233 U obtains higher value than 2 which may satisfy the breeding capability especially for thermal reactor for all investigated MFR. The increasing enrichment and decreasing conversion ratio are more significant for MFR 233 U enrichment. Number density of 233 Pa decreases significantly with decreasing power density which leads the reactor has better breeding performance because lower capture rate of 233 Pa. (author)

  8. Quantification of {sup 232}Th, {sup 234}U, {sup 235}U and {sup 238}U in river mollusks by magnetic sector mass spectrometry with inductively coupled plasma source (Icp-SFMS); Cuantificacion de {sup 232}Th, {sup 234}U, {sup 235}U y {sup 238}U en moluscos de rios por espectrometria de masas de sector magnetico con fuente de plasma acoplado inductivamente (ICP-SFMS)

    Energy Technology Data Exchange (ETDEWEB)

    Arevalo R, D. L.; Hernandez M, H.; Romero G, E. T.; Lara A, N. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Alfaro de la T, M. C., E-mail: arevalo0591@hotmail.com [Universidad Autonoma de San Luis Potosi, Dr. Salvador Nava s/n, Zona Universitaria, 78290 San Luis Potosi, SLP (Mexico)

    2016-09-15

    The present work deals with the methodology established for the quantification of {sup 232}Th, {sup 234}U, {sup 238}U and {sup 235}U in the shell of gastropod mollusks collected in the rivers Valles, Coy and Axtla of San Luis Potosi, Mexico, which belong to the Panuco River basin; these rivers have as main source of pollution the discharge of municipal sewage, waste from small industries, agricultural and cattle residues and from natural sources. Conventional methods for measuring radio-nuclides are confronted with certain conditions related to the requirement in measurement, basically in the characterization that is related to the concepts of precision and accuracy. The analysis of the gastropod mollusk shell was performed by the Icp-SFMS technique; the main advantages of this technique lie in the isotope quantification capacity, the high precision and the low limits of detection, in this study are very important because these elements are in concentrations between ppb and ppt. This technique allowed the analysis of the samples having a complex matrix by the presence CaCO{sub 3} minimizing the interferences thanks to the ionization efficiency of the Ar plasma. For the species Pachychilus monachus were found concentrations of {sup 232}Th of 0.16-5.37 μg/g and of total U of 0.101-4.081 μg/g being this species where the highest values of total U were found. For Thiara (melanoids) tuberculata the lowest values were found among the different species ({sup 232}Th 0.61-3.61 μg/g and total U 0.006-0.042 μg/g), for Pachychilus suturalis, values of {sup 232}Th of 0.58-6.4 μg/g and for Pachychilus sp. were found between 0.26-7.62 μg/g and for total U values between 0.28-3.33 μg/g. The method offers several advantages: speed, good precision, low values of quantification limits and high sensitivity in the measurement of radio-nuclides and heavy metals. (Author)

  9. Precise Th/U-dating of small and heavily coated samples of deep sea corals

    Science.gov (United States)

    Lomitschka, Michael; Mangini, Augusto

    1999-07-01

    Marine carbonate skeletons like deep-sea corals are frequently coated with iron and manganese oxides/hydroxides which adsorb additional thorium and uranium out of the sea water. A new cleaning procedure has been developed to reduce this contamination. In this further cleaning step a solution of Na 2EDTA (Na 2H 2T B) and ascorbic acid is used which composition is optimised especially for samples of 20 mg of weight. It was first tested on aliquots of a reef-building coral which had been artificially contaminated with powdered ferromanganese nodule. Applied on heavily contaminated deep-sea corals (scleractinia), it reduced excess 230Th by another order of magnitude in addition to usual cleaning procedures. The measurement of at least three fractions of different contamination, together with an additional standard correction for contaminated carbonates results in Th/U-ages corrected for the authigenic component. A good agreement between Th/U- and 14C-ages can be achieved even for extremely coated corals.

  10. Calculating the mass distribution of heavy nucleus fission product by neutrons

    International Nuclear Information System (INIS)

    Gudkov, A.N.; Koldobskij, A.B.; Kolobashkin, V.M.; Semenova, E.V.

    1981-01-01

    The technique of calculating the fission product mass yields by neutrons which are necessary for performing nucleus physical calculations in designing nuclear reactor cores is considered. The technique is based on the approximation of fission product mass distribution over the whole mass range by five Gauss functions. New analytical expressions for determining energy weights of used gaussians are proposed. The results of comparison of experimental data with calculated values for fission product mass obtained for reference processes in the capacity of which the fission reactions are chosen: 233 U, 235 U fission by thermal neutrons, 232 Th, 233 U, 235 U, 238 U by fission spectrum neutrons and 14 MeV neutrons and for 232 Th fission reactions by 11 MeV neutrons and 238 U by 7.7 MeV neutrons. On the basis of the analysis of results obtained the conclusion is drawn on a good agreement of fission product mass yield calculation values obtained using recommended values of mass distribution parameters with experimental data [ru

  11. 232-Th and 238-U radioactive contaminations of sediments along the South China Sea of East Coast Peninsular Malaysia by INAA

    International Nuclear Information System (INIS)

    Khadijeh Rezaee Ebrahim Saraee; Elias Saion; Naghavi, K.

    2009-01-01

    The concentrations of 232 Th and 238 U were determined in 30 sediment samples, collected from the South China Sea bed of the east coast peninsular Malaysia. The samples were dried in an oven for 10 days at 65 C and stored in polyethylene bottles for future analysis. INAA used to analyse surficial sediments. The 232 Th and 238 U activities in surficial sediments determined. These activities increased from 14.02 to 30.17 Bq Kg -1 and from 47.49 to 113.69 Bq Kg -1 for 232 Th and 238 U , respectively. Results show diffuse distribution of 232 Th and 238 U actinides in studied area. The Th/Sc and U/Sc ratios of the suricial sediments have same distribution and the highest their values are in station EC12. The radioactive contamination of the investigated sector is compared with the results obtained of other coastal sediments of Malaysia. Results show activity contamination in coastal sediments of the South China Sea is lower than the Straits of Malacca and is the same with the Straits of Johor. (Author)

  12. Influence of moderator to fuel ratio (MFR) on burning thorium in a subcritical assembly

    International Nuclear Information System (INIS)

    Wojciechowski, Andrzej

    2014-01-01

    The conversion ratio (CR) of Th-232 to U-233 calculation results for a subcritical reactor assembly is presented as a function of MFR, burnup, power density (PD) and fissile concentration. The calculated model is based on subcritical assembly which makes configuration of fuel rods and volumes of moderator and coolant changes possible. This comfortable assembly enables investigation of CR in a thorium cycle for different value of MFR. Additionally, the calculation results of U-233 saturation concentration are explained by mathematical model. The value of MFR main influences the saturation concentration of U-233 and fissile and the fissile concentration dependence of CR. The saturation value of CR is included in the range CR ∈ (0.911, 0.966) and is a slowly increasing function of MFR. The calculations were done with a MCNPX 2.7 code

  13. Preserving Samples and Their Scientific Integrity — Insights into MSR from the Astromaterials Acquisition and Curation Office at NASA Johnson Space Center

    Science.gov (United States)

    Calaway, M. J.; Regberg, A. B.; Mitchell, J. L.; Fries, M. D.; Zeigler, R. A.; McCubbin, F. M.; Harrington, A. D.

    2018-04-01

    Rigorous collection of samples for contamination knowledge, the information gained from the characterization of reference materials and witness plates in concurrence with sample return, is essential for MSR mission success.

  14. A Reassessment of U-Th and 14C Ages for Late-Glacial High-Frequency Hydrological Events at Searles Lake, California

    Science.gov (United States)

    Lin, J.C.; Broecker, W.S.; Hemming, S.R.; Hajdas, I.; Anderson, Robert F.; Smith, G.I.; Kelley, M.; Bonani, G.

    1998-01-01

    U-Th isochron ages of tufas formed on shorelines suggest that the last pluvial event in Lake Lahontan and Searles Lake was synchronous at about 16,500 cal yr B.P. (equivalent to a radiocarbon age of between 14,000 and 13,500 yr B.P.), whereas the timing of this pluvial event determined by radiocarbon dating is on the order of 1000 yr younger. The timing of seven distinct periods of near desiccation in Searles Lake during late-glacial time has been reinvestigated for U-Th age determination by mass spectrometry. U-Th dating of evaporite layers in the interbedded mud and salt unit called the Lower Salt in Searles Lake was hampered by the uncertainty in assessing the initial 230Th/232Th of the samples. The resulting ages, corrected by a conservative range of initial 230Th/ 232Th ratios, suggest close correlation of the abrupt changes recorded in Greenland ice cores (Dansgaard-Oeschger events) and wet-dry conditions in Searles Lake between 35,000 and 24,000 Cal yr B.P. ?? 1998 University of Washington.

  15. Associating Physical and Chemical Properties to Evaluate Buffer Materials by Th and U Sorption

    Energy Technology Data Exchange (ETDEWEB)

    Jan, Yi-Lin; Chen, Tzu-Yun; Cheng, Hwai-Ping; Hsu, Chun-Nan; Tseng, Chia-Liang; Wei,Yuan-Yaw; Yang, Jen-Yan; Ke, Cheng-Hsiung; Chuang, Jui-Tang; Teng, Shi-Ping

    2003-02-27

    The physical and chemical properties of buffer materials to be used for a radwaste disposal repository should be evaluated prior to use. In a conventional approach, independent studies of physical and/or chemical characteristics are conducted. This study investigated the relationship between the plastic index (PI) and distribution ratio (Rd) of buffer materials composed of varying ratios of quartz sand and bentonite. Thorium (Th) and Uranium (U) were the nuclides of interest, and both synthetic groundwater and seawater were used as the liquid phases to simulate conditions representative of deep geological disposal within an island. Atterberg tests were used to determine PI values, and batch sorption experiments were employed to measure Rd values. The results show that Th reached maximum sorption behavior when the bentonite content exceeded 30 % of the mixture. Contrariwise, the sorption of U increased linearly with bentonite content, up to bentonite contents of 100%, and this correlation was present regardless of the liquid phase used. A further result is that U has a better additivity with respect to Rd than Th in both synthetic groundwater and synthetic seawater. These results will allow a determination of more effective buffer material composition, and improved estimates of the overall Rd of the buffer material mixture from the Rd of each mineral component.

  16. Design study of Thorium-232 and Protactinium-231 based fuel for long life BWR

    Energy Technology Data Exchange (ETDEWEB)

    Trianti, N.; Su' ud, Z.; Riyana, E. S. [Nuclear Physics and Biophysics Research Division Department of Physics - Institut Teknologi Bandung (ITB) Jalan Ganeca 10 Bandung 40132 (Indonesia)

    2012-06-06

    A preliminary design study for the utilization of thorium added with {sup 231}Pa based fuel on BWR type reactor has been performed. In the previous research utilization of fuel based Thorium-232 and Uranium-233 show 10 years operation time with maximum excess-reactivity about 4.075% dk/k. To increase reactor operation time and reduce excess-reactivity below 1% dk/k, Protactinium (Pa-231) is used as Burnable Poison. Protactinium-231 has very interesting neutronic properties, which enable the core to reduce initial excess-reactivity and simultaneously increase production of {sup 233}U to {sup 231}Pa in burn-up process. Optimizations of the content of {sup 231}Pa in the core enables the BWR core to sustain long period of operation time with reasonable burn-up reactivity swing. Based on the optimization of fuel element composition (Th and Pa) in various moderation ratio we can get reactor core with longer operation time, 20 {approx} 30 years operation without fuel shuffling or refuelling, with average power densities maximum of about 35 watt/cc, and maximum excess-reactivity 0.56% dk/k.

  17. Alternative fuel cycles and non-proliferation aspects

    International Nuclear Information System (INIS)

    Kessler, G.

    1980-10-01

    The most important physical characteristics of the U/Pu and the Th/U fuel cycles and the technical data of the most significant converter reactors operating with Th/U fuel are outlined in the report. Near breeders as well as breeders with a thermal neutron spectrum are briefly discussed, and the potential of breeders with fast neutron spectra in the Th/U fuel is outlined. The essential criteria for the comparison of the alternative fuel cycles with the reference Pu/U cycle are the consumption of natural uranium, the numbers of U-233 producing and U-233 consuming converter reactors and the amounts of fission material transported and handled within the fuel cycle (reprocessing, refabrication). Although the alternative U/Th fuel cycles are feasible with some advantages and some disadvantages as compared to the reference U/Pu cycle, not much experience has so far been gathered with pilot plants of the fuel cycle. The respective status in reprocessing, refabrication and waste disposal is briefly discussed. Finally, a comparison of the risk potential inherent in secular storage is presented and questions of resistance to proliferation and of safeguards of the U/Th cycle are discussed

  18. 12 CFR Appendix A to Part 233 - Model Notice

    Science.gov (United States)

    2010-01-01

    ... FUNDING OF UNLAWFUL INTERNET GAMBLING (REGULATION GG) Part 233, App. A Appendix A to Part 233—Model Notice... 12 Banks and Banking 3 2010-01-01 2010-01-01 false Model Notice A Appendix A to Part 233 Banks and... that your institution processed payments through our facilities for Internet gambling transactions...

  19. 40 CFR 86.233-94-86.234-94 - [Reserved

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 18 2010-07-01 2010-07-01 false [Reserved] 86.233-94-86.234-94 Section 86.233-94-86.234-94 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR... New Medium-Duty Passenger Vehicles; Cold Temperature Test Procedures §§ 86.233-94—86.234-94 [Reserved] ...

  20. 27 CFR 24.233 - Addition of spirits to wine.

    Science.gov (United States)

    2010-04-01

    ... wine. 24.233 Section 24.233 Alcohol, Tobacco Products and Firearms ALCOHOL AND TOBACCO TAX AND TRADE BUREAU, DEPARTMENT OF THE TREASURY LIQUORS WINE Spirits § 24.233 Addition of spirits to wine. (a) Prior to the addition of spirits. Wine will be placed in tanks approved for the addition of spirits. The...

  1. Measurement of Fragment Mass Distributions in Neutron-induced Fission of 238U and 232Th at Intermediate Energies

    International Nuclear Information System (INIS)

    Simutkin, V.D.

    2008-01-01

    Conceptual analysis of accelerator-driven systems assumes extensive use of nuclear data on neutron-induced reactions at intermediate energies. In particular, information about the fission fragment yields from the 238 U(n,f) and 232 Th(n,f) reactions is of particular interest at neutron energies from 10 to 200 MeV. However, there is a lack of such data for both 238 U and 232 Th. Up to now, the intermediate energy measurements have been performed for 238 U only, and there are no data for the 232 Th(n,f) reaction. The aim of the work is to provide such data. Fission fragment mass distributions for the 232 Th(n,f) and 238 U(n,f) reactions have been measured for the incident neutron energies 32.8 MeV, 45.3 MeV and 59.9 MeV. The experiments have been performed at the neutron beam facility of the Universite Catholique de Louvain, Belgium. A multi-section Frisch-gridded ionization chamber has been used as a fission fragment detector. The data obtained have been interpreted in terms of the multimodal random neck-rupture model (MMRNRM). (authors)

  2. Impact on breeding rate of different Molten Salt reactor core structures

    International Nuclear Information System (INIS)

    Wang Haiwei; Mei Longwei; Cai Xiangzhou; Chen Jingen; Guo Wei; Jiang Dazhen

    2013-01-01

    Background: Molten Salt Reactor (MSR) has several advantages over the other Generation IV reactor. Referred to the French CNRS research and compared to the fast reactor, super epithermal neutron spectrum reactor type is slightly lower and beading rate reaches 1.002. Purpose: The aim is to explore the best conversion zone layout scheme in the super epithermal neutron spectrum reactor. This study can make nuclear fuel as one way to solve the energy problems of mankind in future. Methods: Firstly, SCALE program is used for molten salt reactor graphite channel, molten salt core structure, control rods, graphite reflector and layer cladding structure. And the SMART modules are used to record the important actinides isotopes and their related reaction values of each reaction channel. Secondly, the thorium-uranium conversion rate is calculated. Finally, the better molten salt reactor core optimum layout scheme is studied comparing with various beading rates. Results: Breading zone layout scheme has an important influence on the breading rate of MSR. Central graphite channels in the core can get higher neutron flux irradiation. And more 233 Th can convert to 233 Pa, which then undergoes beta decay to become 233 U. The graphite in the breading zone gets much lower neutron flux irradiation, so the life span of this graphite can be much longer than that of others. Because neutron flux irradiation in the uranium molten salt graphite has nearly 10 times higher than the graphite in the breading zone, it has great impact on the thorium-uranium conversion rates. For the super epithermal neutron spectrum molten salt reactors, double salt design cannot get higher thorium-uranium conversion rates. The single molten salt can get the same thorium-uranium conversion rate, meanwhile it can greatly extend the life of graphite in the core. Conclusions: From the analysis of calculation results, Blanket breeding area in different locations in the core can change the breeding rates of thorium

  3. 40 CFR 233.12 - Attorney General's statement.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 24 2010-07-01 2010-07-01 false Attorney General's statement. 233.12... STATE PROGRAM REGULATIONS Program Approval § 233.12 Attorney General's statement. (a) Any State that seeks to administer a program under this part shall submit a statement from the State Attorney General...

  4. Atomistic simulations of nanocrystalline U0.5Th0.5O2 solid solution under uniaxial tension

    Directory of Open Access Journals (Sweden)

    Hongxing Xiao

    2017-12-01

    Full Text Available Molecular dynamics simulations were performed to investigate the uniaxial tensile properties of nanocrystalline U0.5Th0.5O2 solid solution with the Born–Mayer–Huggins potential. The results indicated that the elastic modulus increased linearly with the density relative to a single crystal, but decreased with increasing temperature. The simulated nanocrystalline U0.5Th0.5O2 exhibited a breakdown in the Hall–Petch relation with mean grain size varying from 3.0 nm to 18.0 nm. Moreover, the elastic modulus of U1-yThyO2 solid solutions with different content of thorium at 300 K was also studied and the results accorded well with the experimental data available in the literature. In addition, the fracture mode of nanocrystalline U0.5Th0.5O2 was inclined to be ductile because the fracture behavior was preceded by some moderate amount of plastic deformation, which is different from what has been seen earlier in simulations of pure UO2.

  5. Bioaccessibility of U, Th and Pb in particulate matter from an abandoned uranium mine

    Science.gov (United States)

    Millward, Geoffrey; Foulkes, Michael; Henderson, Sam; Blake, William

    2016-04-01

    Currently, there are approximately 150 uranium mines in Europe at various stages of either operation, development, decommissioning, restoration or abandonment (wise-uranium.com). The particulate matter comprising the mounds of waste rock and mill tailings poses a risk to human health through the inadvertent ingestion of particles contaminated with uranium and thorium, and their decay products, which exposes recipients to the dual toxicity of heavy elements and their radioactive emissions. We investigated the bioaccessibility of 238U, 232Th and 206,214,210Pb in particulate samples taken from a contaminated, abandoned uranium mine in South West England. Sampling included a mine shaft, dressing floor and waste heap, as well as soils from a field used for grazing. The contaminants were extracted using the in-vitro Unified Bioaccessibility Research Group of Europe Method (UBM) in order to mimic the digestion processes in the human stomach (STOM) and the combined stomach and gastrointestinal tract (STOM+INT). Analyses of concentrations of U, Th and Pb in the extracts were by ICP-MS and the activity concentrations of radionuclides were determined on the same particles, before and after extraction, using gamma spectroscopy. 'Total' concentrations of U, Th and Pb for all samples were in the range 57 to 16,200, 0.28 to 3.8 and 69 to 4750 mg kg-1, respectively. For U and Pb the concentrations in the STOM fraction were lower than the total and STOM+INT fractions were even lower. However, for Th the STOM+INT fractions were higher than the STOM due to the presence of Th carbonate species within the gastrointestinal fluid. Activity concentrations for 214Pb and 210Pb, including total, STOM and STOM+INT, were in the range 180 to samples were 39% and 8% in the STOM and STOM+INT, respectively, whereas the respective BAFs for 232Th were 3% and 9%. For stable 206Pb the STOM and STOM+INT BAFs were 16% and 3% for the most contaminated samples, whereas those from the field had 44% in the

  6. Unlipidated Outer Membrane Protein Omp16 (U-Omp16) from Brucella spp. as Nasal Adjuvant Induces a Th1 Immune Response and Modulates the Th2 Allergic Response to Cow’s Milk Proteins

    Science.gov (United States)

    Ibañez, Andrés E.; Smaldini, Paola; Coria, Lorena M.; Delpino, María V.; Pacífico, Lucila G. G.; Oliveira, Sergio C.; Risso, Gabriela S.; Pasquevich, Karina A.; Fossati, Carlos Alberto; Giambartolomei, Guillermo H.; Docena, Guillermo H.; Cassataro, Juliana

    2013-01-01

    The discovery of novel mucosal adjuvants will help to develop new formulations to control infectious and allergic diseases. In this work we demonstrate that U-Omp16 from Brucella spp. delivered by the nasal route (i.n.) induced an inflammatory immune response in bronchoalveolar lavage (BAL) and lung tissues. Nasal co-administration of U-Omp16 with the model antigen (Ag) ovalbumin (OVA) increased the amount of Ag in lung tissues and induced OVA-specific systemic IgG and T helper (Th) 1 immune responses. The usefulness of U-Omp16 was also assessed in a mouse model of food allergy. U-Omp16 i.n. administration during sensitization ameliorated the hypersensitivity responses of sensitized mice upon oral exposure to Cow’s Milk Protein (CMP), decreased clinical signs, reduced anti-CMP IgE serum antibodies and modulated the Th2 response in favor of Th1 immunity. Thus, U-Omp16 could be used as a broad Th1 mucosal adjuvant for different Ag formulations. PMID:23861971

  7. A new technique for thick source alpha counting determination of U and Th

    CERN Document Server

    Michael, C T

    2000-01-01

    A new technique for the calculation of U and Th concentration is presented based on the alpha particle spectrum taken from a thick sample by using a silicon detector. Four approaches to the analysis of the experimental data are presented, one being an improvement on the known pairs technique. By the proposed technique it is possible to calculate the concentrations of certain daughter nuclides in the two series, or the sum of the activity concentrations of others. This allows the detection of secular disequilibrium in our samples. This technique also has the advantage of being more accurate and provides the opportunity to cross-check the results derived from the different approaches.

  8. Fused salt power reactor study: Minutes of discussion meeting No. 2

    International Nuclear Information System (INIS)

    Alexander, L. G.

    1956-01-01

    Remarks made by participants in a 1956 meeting are sketched. Economics was a major concern. Significant topics included development of a new alloy for use in the heat exchanger, conversion ratios in a U-233 breeder, the effects of ThF 4 on corrosion, and means of producing various transmutation products other than U-233.

  9. 238U series isotopes and 232Th in carbonates and black shales from the Lesser Himalaya: implications to dissolved uranium abundances in Ganga-Indus source waters

    International Nuclear Information System (INIS)

    Singh, S.K.; Dalai, Tarun K.; Krishnaswami, S.

    2003-01-01

    238 U and 232 Th concentrations and the extent of 238 U- 234 U- 230 Th radioactive equilibrium have been measured in a suite of Precambrian carbonates and black shales from the Lesser Himalaya. These measurements were made to determine their abundances in these deposits, their contributions to dissolved uranium budget of the headwaters of the Ganga and the Indus in the Himalaya and to assess the impact of weathering on 238 U- 234 U- 230 Th radioactive equilibrium in them. 238 U concentrations in Precambrian carbonates range from 0.06 to 2.07 μg g -1 . The 'mean' U/Ca in these carbonates is 2.9 ng U mg -1 Ca. This ratio, coupled with the assumption that all Ca in the Ganga-Indus headwaters is of carbonate origin and that U and Ca behave conservatively in rivers after their release from carbonates, provides an upper limit on the U contribution from these carbonates, to be a few percent of dissolved uranium in rivers. There are, however, a few streams with low uranium concentrations, for which the carbonate contribution could be much higher. These results suggest that Precambrian carbonates make only minor contributions to the uranium budget of the Ganga-Indus headwaters in the Himalaya on a basin wide scale, however, they could be important for particular streams. Similar estimates of silicate contribution to uranium budget of these rivers using U/Na in silicates and Na* (Na corrected for cyclic and halite contributions) in river waters show that silicates can contribute significantly (∼40% on average) to their U balance. If, however, much of the uranium in these silicates is associated with weathering resistant minerals, then the estimated silicate uranium component would be upper limits. Uranium concentration in black shales averages about 37 μg g -1 . Based on this concentration, supply of U from at least ∼50 mg of black shales per liter of river water is needed to balance the average river water U concentration, 1.7 μg L -1 in the Ganga-Indus headwaters

  10. 232Th and 238U neutron emission cross section calculations and analysis of experimental data

    International Nuclear Information System (INIS)

    Tel, E.

    2004-01-01

    In this study, pre-equilibrium neutron-emission spectra produced by (n,xn) reactions on nuclei 2 32Th and 2 38U have been calculated. Angle-integrated cross sections in neutron induced reactions on targets 2 32Th and 2 38U have been calculated at the bombarding energies up to 18 MeV. We have investigated multiple pre-equilibrium matrix element constant from internal transition for 2 32Th (n,xn) neutron emission spectra. In the calculations, the geometry dependent hybrid model and the cascade exciton model including the effects of pre-equilibrium have been used. In addition, we have described how multiple pre-equilibrium emissions can be included in the Feshbach-Kerman-Koonin (FKK) fully quantum-mechanical theory. By analyzing (n,xn) reaction on 232 T h and 2 38U, with the incident energy from 2 Me V to 18 Me V, the importance of multiple pre-equilibrium emission can be seen cleady. All calculated results have been compared with experimental data. The obtained results have been discussed and compared with the available experimental data and found agreement with each other

  11. Cooling rates and the depth of detachment faulting at oceanic core complexes: Evidence from zircon Pb/U and (U-Th)/He ages

    Science.gov (United States)

    Grimes, Craig B.; Cheadle, Michael J.; John, Barbara E.; Reiners, P.W.; Wooden, J.L.

    2011-01-01

    Oceanic detachment faulting represents a distinct mode of seafloor spreading at slow spreading mid-ocean ridges, but many questions persist about the thermal evolution and depth of faulting. We present new Pb/U and (U-Th)/He zircon ages and combine them with magnetic anomaly ages to define the cooling histories of gabbroic crust exposed by oceanic detachment faults at three sites along the Mid-Atlantic Ridge (Ocean Drilling Program (ODP) holes 1270D and 1275D near the 15??20???N Transform, and Atlantis Massif at 30??N). Closure temperatures for the Pb/U (???800??C-850??C) and (U-Th)/He (???210??C) isotopic systems in zircon bracket acquisition of magnetic remanence, collectively providing a temperature-time history during faulting. Results indicate cooling to ???200??C in 0.3-0.5 Myr after zircon crystallization, recording time-averaged cooling rates of ???1000??C- 2000??C/Myr. Assuming the footwalls were denuded along single continuous faults, differences in Pb/U and (U-Th)/He zircon ages together with independently determined slip rates allow the distance between the ???850??C and ???200??C isotherms along the fault plane to be estimated. Calculated distances are 8.4 ?? 4.2 km and 5.0 2.1 km from holes 1275D and 1270D and 8.4 ?? 1.4 km at Atlantis Massif. Estimating an initial subsurface fault dip of 50 and a depth of 1.5 km to the 200??C isotherm leads to the prediction that the ???850??C isotherm lies ???5-7 km below seafloor at the time of faulting. These depth estimates for active fault systems are consistent with depths of microseismicity observed beneath the hypothesized detachment fault at the TAG hydrothermal field and high-temperature fault rocks recovered from many oceanic detachment faults. Copyright 2011 by the American Geophysical Union.

  12. U.S. Army Reserve 88th Readiness Division Finds Big Savings

    Energy Technology Data Exchange (ETDEWEB)

    None

    2018-02-20

    Fact sheet features lighting work done for the U.S. Army Reserve 88th Readiness Division, which was recognized in two 2017 Interior Lighting Campaign exemplary recognition categories. The troffer lighting upgrade projects at the two recognized sites are expected to save more than 246,000 kWh annually or roughly enough electricity to run 23 homes for a year.

  13. Extreme fractionation of 234U 238U and 230Th 234U in spring waters, sediments, and fossils at the Pomme de Terre Valley, southwestern Missouri

    Science.gov (United States)

    Szabo, B. J.

    1982-01-01

    Isotopic fractionation as great as 1600% exists between 234U and 238U in spring waters, sediments, and fossils in the Pomme de Terre Valley, southwestern Missouri. The activity ratios of 234U 238U in five springs range from 7.2 to 16 in water which has been discharged for at least the past 30,000 years. The anomalies in 234U 238U ratio in deep water have potential usefulness in hydrologic investigations in southern Missouri. Clayey units overlying the spring bog sediments of Trolinger Spring are enriched in 230Th relative to their parent 234U by as much as 720%. The results indicate that both preferential displacement via alpha recoil ejection and the preferential emplacement via recoiling and physical entrapment are significant processes that are occurring in the geologic environment. ?? 1982.

  14. A novel hash based least significant bit (2-3-3) image steganography in spatial domain

    OpenAIRE

    Manjula, G. R.; Danti, Ajit

    2015-01-01

    This paper presents a novel 2-3-3 LSB insertion method. The image steganography takes the advantage of human eye limitation. It uses color image as cover media for embedding secret message.The important quality of a steganographic system is to be less distortive while increasing the size of the secret message. In this paper a method is proposed to embed a color secret image into a color cover image. A 2-3-3 LSB insertion method has been used for image steganography. Experimental results show ...

  15. Crystal chemistry and thermal behavior of La doped (U, Th)O2

    Science.gov (United States)

    Keskar, Meera; Shelke, Geeta P.; Shafeeq, Muhammed; Krishnan, K.; Sali, S. K.; Kannan, S.

    2017-12-01

    X-ray diffraction, chemical and thermal studies of [(U0.2Th0.8)1-yLay]O2+x (LUTL) and [(U0.3Th0.7)1-yLay]O2+x (UTL); compounds (where y ≤ 0.4) were carried out. These compounds were synthesized by gel combustion method followed by heating in reduced atmospheres at 1673 K. To co-relate lattice parameters with metal and oxygen concentrations, reduced oxides were heated in Ar, CO2 and air atmospheres. Retention of FCC phase was confirmed in all mixed oxides with y ≤ 0.4. The cubic lattice parameters could be expressed in a linear equation of x and y as: a (Ǻ) = 5.5709 - 0.187 x + 0.032 y; [x Oxidation studies and simple ionic model calculations suggested that uranium is predominantly present as a mixture of +5 and + 6 states when La/U ratio ∼2. Oxidation kinetics of mixed oxides was studied by non-isothermal method using thermogravimetry and was found to be a diffusion controlled reaction. High temperature X-ray diffraction studies of LUTL and UTL mixed oxides showed positive thermal expansion in the temperature range of 298-1273 K and % expansion increases with La concentration.

  16. Thorium-based fuel cycles: Reassessment of fuel economics and proliferation risk

    Energy Technology Data Exchange (ETDEWEB)

    Serfontein, Dawid E., E-mail: Dawid.Serfontein@nwu.ac.za [Senior Lecturer at the School of Mechanical and Nuclear Engineering, North West University (PUK-Campus), PRIVATE BAG X6001, Internal Post Box 360, Potchefstroom 2520 (South Africa); Mulder, Eben J. [Professor at the School of Mechanical and Nuclear Engineering, North West University (South Africa)

    2014-05-01

    At current consumption and current prices, the proven reserves for natural uranium will last only about 100 years. However, the more abundant thorium, burned in breeder reactors, such as large High Temperature Gas-Cooled Reactors, and followed by chemical reprocessing of the spent fuel, could stretch the 100 years for uranium supply to 15,000 years. Thorium-based fuel cycles are also viewed as more proliferation resistant compared to uranium. However, several barriers to entry caused all countries, except India and Russia, to abandon their short term plans for thorium reactor projects, in favour of uranium/plutonium fuel cycles. In this article, based on the theory of resonance integrals and original analysis of fast fission cross sections, the breeding potential of {sup 232}Th is compared to that of {sup 238}U. From a review of the literature, the fuel economy of thorium-based fuel cycles is compared to that of natural uranium-based cycles. This is combined with a technical assessment of the proliferation resistance of thorium-based fuel cycles, based on a review of the literature. Natural uranium is currently so cheap that it contributes only about 10% of the cost of nuclear electricity. Chemical reprocessing is also very expensive. Therefore conservation of natural uranium by means of the introduction of thorium into the fuel is not yet cost effective and will only break even once the price of natural uranium were to increase from the current level of about $70/pound yellow cake to above about $200/pound. However, since fuel costs constitutes only a small fraction of the total cost of nuclear electricity, employing reprocessing in a thorium cycle, for the sake of its strategic benefits, may still be a financially viable option. The most important source of the proliferation resistance of {sup 232}Th/{sup 233}U fuel cycles is denaturisation of the {sup 233}U in the spent fuel by {sup 232}U, for which the highly radioactive decay chain potentially poses a large

  17. High resolution photofission measurements in 238U and 232Th. Final report

    International Nuclear Information System (INIS)

    Lancman, H.

    1985-12-01

    A novel technique for measuring the photofission cross section with very high photon energy resolution has been developed. The photons are obtained from selected resonances in the (p,γ) reaction on various light nuclei. The photon energy resolution approaches 200 eV in favorable cases. The photon energy spread at each (p,γ) resonance is approx.20 keV on the average. Measurements of the photo-fission cross sections of 232 Th and 238 U have been carried out in the energy range from 5.8 to 12 MeV. Intermediate structure has been found in both nuclei at excitation energies around 6 MeV. Various properties of this structure, such as average areas of resonances, their spacing, width, and the underlying bakground, as well as the experimental fission probability averaged over the intermediate structure have been found to agree with theoretical predictions based on a double-humped fission barrier. In the case of 232 Th, the feature of this barrier, a rather high first hump and a deep secondary well, are quite different from those predicted by current theoretial barrier calculations. 13 refs., 4 figs., 3 tabs

  18. Speciation of An(IV) (Pu, Np, U and Th) in citrate media

    Energy Technology Data Exchange (ETDEWEB)

    Bonin, L.; Moisy, P. [CEA Valrho, DEN/DRCP/SCPS/LCA, Bagnols sur Ceze (France); Cote, G. [Ecole Nationale Superieure de Chimie, 75 - Paris (France)

    2008-07-01

    The acquisition of additional data concerning actinide decorporation is planned under the French nuclear and environmental toxicology program. Citric acid is a fundamental biological constituent found at relatively high concentrations in blood plasma and is involved in many biological processes. Spectrophotometry was the primary method used to determine the speciation of An(IV)-citrate systems (Th, U, Np, Pu). Complexation phenomena were identified, especially the formation of complexes with 1:1 and 1:2 stoichiometries. The corresponding conditional constants were calculated according to the experimental conditions. Depending on the acid-base form of the ligand (H{sub 2}Cit{sup -}, HCit{sup 2-} or Cit{sup 3-}) the apparent stability constants were also calculated and compared for various tetravalent actinides. (orig.)

  19. A comparison of U/Th and rapid-screen 14C dates from Line Island fossil corals

    Science.gov (United States)

    Grothe, Pamela R.; Cobb, Kim M.; Bush, Shari L.; Cheng, Hai; Santos, Guaciara M.; Southon, John R.; Lawrence Edwards, R.; Deocampo, Daniel M.; Sayani, Hussein R.

    2016-03-01

    Time-consuming and expensive radiometric dating techniques limit the number of dates available to construct absolute chronologies for high-resolution paleoclimate reconstructions. A recently developed rapid-screen 14C dating technique reduces sample preparation time and per sample costs by 90%, but its accuracy has not yet been tested on shallow-water corals. In this study, we test the rapid-screen 14C dating technique on shallow-water corals by comparing 44 rapid-screen 14C dates to both high-precision 14C dates and U/Th dates from mid- to late-Holocene fossil corals collected from the central tropical Pacific (2-4°N, 157-160°W). Our results show that 42 rapid-screen 14C and U/Th dates agree within uncertainties, confirming closed-system behavior and ensuring chronological accuracy. However, two samples that grew ˜6500 years ago have calibrated 14C ages ˜1000 years younger than the corresponding U/Th ages, consistent with diagenetic alteration as indicated by the presence of 15-23% calcite. Mass balance calculations confirm that the observed dating discrepancies are consistent with 14C addition and U removal, both of which occur during diagenetic calcite recrystallization. Under the assumption that aragonite-to-calcite replacement is linear through time, we estimate the samples' true ages using the measured 14C and U/Th dates and percent calcite values. Results illustrate that the rapid-screen 14C dates of Holocene-aged fossil corals are accurate for samples with less than 2% calcite. Application of this rapid-screen 14C method to the fossil coral rubble fields from Kiritimati Island reveal significant chronological clustering of fossil coral across the landscape, with older ages farther from the water's edge.

  20. Optimization of binary breeder reactor. 2. Preliminary base for control analysis and fuel management

    International Nuclear Information System (INIS)

    Dias, A.F.; Nascimento, J.A. do; Ishiguro, Y.

    1985-01-01

    Neutronic calculations to verify the reactivity effects, of sodium voids and Doppler, with the variation of the composition of parasitic absorbers were done. A LMFBR type reactor loaded with mixed fuel, (U 233 -Th 232 )O 2 in the internal core and (U 238 -Pu 239 )O 2 in external core, was considered. In reactivity calculations the EXPANDA and CITATION computer codes were utilized. Buckling effects and importance of determination of the spatial selfshielding factors were analysed. (M.C.K.) [pt