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Sample records for u-235 60-100 450-500

  1. 31 CFR 540.315 - Uranium-235 (U235).

    Science.gov (United States)

    2010-07-01

    ... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false Uranium-235 (U235). 540.315 Section... FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.315 Uranium-235 (U235). The term uranium-235 or U235 means the fissile...

  2. Heterogeneity in the 238U/235U Ratios of Angrites.

    Science.gov (United States)

    Tissot, F.; Dauphas, N.; Grove, T. L.

    2016-12-01

    Angrites are differentiated meteorites of basaltic composition, of either volcanic or plutonic origin, that display minimal post-crystallization alteration, metamorphism, shock or impact brecciation. Because quenched angrites cooled very rapidly, all radiochronometric systems closed simultaneously in these samples. Quenched angrites are thus often used as anchors for cross-calibrating short-lived dating methods (e.g., 26Al-26Mg) and the absolute dating techniques (e.g, Pb-Pb). Due to the constancy of the 238U/235U ratio in natural samples, Pb-Pb ages have long been calculated using a "consensus" 238U/235U ratio, but the discovery of resolvable variations in the 238U/235U ratio of natural samples, means that the U isotopic composition of the material to date also has to be determined in order to obtain high-precision Pb-Pb ages. We set out (a) to measure at high-precision the 238U/235U ratio of a large array of angrites to correct their Pb-Pb ages, and (b) to identify whether all angrites have a similar U isotopic composition, and, if not, what were the processes responsible for this variability. Recently, Brennecka & Wadhwa (2012) suggested that the angrite-parent body had a homogeneous 238U/235U ratio. They reached this conclusion partly because they propagated the uncertainties of the U isotopic composition of the various U double spikes that they used onto the final 238U/235U ratio the sample. Because this error is systematic (i.e., it affects all samples similarly), differences in the δ238U values of samples corrected by the same double spike are better known than one would be led to believe if uncertainties on the spike composition are propagated. At the conference, we will present the results of the high-precision U isotope analyses for six angrite samples: NWA 4590, NWA 4801, NWA 6291, Angra dos Reis, D'Orbigny, and Sahara 99555. We will show that there is some heterogeneity in the δ238U values of the angrites and will discuss the possible processes by

  3. Ternary Fission of U235 by Resonance Neutrons

    International Nuclear Information System (INIS)

    Kvitek, I.; Popov, Ju.P.; Rjabov, Ju.V.

    1965-01-01

    Recently a number of papers have appeared indicating considerable variations in the ratio of the ternary-fission cross-section to the binary-fission cross-section of U 235 on transition from one neutron resonance to another. However, such variations have not been discovered in U 233 and Pu 239 . The paper reports investigations of the ternary fission of U 235 by neutrons with an energy of 0.1 to 30 eV. Unlike other investigators of the ternary fission of U 235 , we identified the ternary-fission event by the coincidence of one of the fission fragments with a light long-range particle. This made it passible to separate ternary fissions from the possible contribution of the (n, α)reaction. The measurements were performed at the fast pulsed reactor of the Joint Institute for Nuclear Research by the time-of-flight method. A flight length of 100 m was used, giving a resolution of 0.6 μs/m. Gas scintillation counters filled with xenon at a pressure of 2 atm were used to record the fission fragments and the light long-range particle. A layer of enriched U 235 ∼2 mg/cm 2 thick and ∼300 cm 2 in area was applied to an aluminium foil 20-fim thick. The scintillations from the fission fragments were recorded in the gas volume on one side of the foil and those from the light long-range particles in that on the other. In order to assess the background (e.g . coincidences of the pulse from a fragment with that from a fission gamma quantum or a proton from the (n, p) reaction in the aluminium foil), a measurement was carried out in which the volume recording the long-range particle was shielded with a supplementary aluminium filter 1-mm thick. The results obtained indicate the absence of the considerable variations in the ratio between the ternary-and binary- fission cross-sections for U 235 that have been noted by other authors. Measurements showed no irregularity in the ratio of the cross-sections in the energy range 0.1 to 0.2 eV. The paper discusses the possible effect of

  4. Critical mass experiment using U-235 foils and lucite plates

    International Nuclear Information System (INIS)

    Sanchez, R.; Butterfield, K.; Kimpland, R.; Jaegers, P.

    1998-01-01

    The main objective of this experiment was to show how the multiplication of the system increases as moderated material is placed between highly enriched uranium foils. In addition, this experiment served to demonstrate the hand-stacking techniques, and approach to criticality by remote operation. This experiment was designed by Tom McLaughlin in the mid seventies as part of the criticality safety course that is taught at Los Alamos Critical Experiment Facility (LACEF). The W-U-235 ratio for this experiment was 215 which is where the minimum critical mass for this configuration occurs

  5. Isotopic analysis of uranium hexafluoride highly enriched in U-235

    International Nuclear Information System (INIS)

    Chaussy, L.; Boyer, R.

    1968-01-01

    Isotopic analysis of uranium in the form of the hexafluoride by mass-spectrometry gives gross results which are not very accurate. Using a linear interpolation method applied to two standards it is possible to correct for this inaccuracy as long as the isotopic concentrations are less than about 10 per cent in U-235. Above this level, the interpolations formula overestimates the results, especially if the enrichment of the analyzed samples is higher than 1.3 with respect to the standards. A formula is proposed for correcting the interpolation equation and for the extending its field of application to high values of the enrichment (≅2) and of the concentration. It is shown that by using this correction the results obtained have an accuracy which depends practically only on that of the standards, taking into account the dispersion in the measurements. (authors) [fr

  6. Computer program FPIP-REV calculates fission product inventory for U-235 fission

    Science.gov (United States)

    Brown, W. S.; Call, D. W.

    1967-01-01

    Computer program calculates fission product inventories and source strengths associated with the operation of U-235 fueled nuclear power reactor. It utilizes a fission-product nuclide library of 254 nuclides, and calculates the time dependent behavior of the fission product nuclides formed by fissioning of U-235.

  7. Should we ignore U-235 series contribution to dose?

    International Nuclear Information System (INIS)

    Beaugelin-Seiller, Karine; Goulet, Richard; Mihok, Steve; Beresford, Nicholas A.

    2016-01-01

    Environmental Risk Assessment (ERA) methodology for radioactive substances is an important regulatory tool for assessing the safety of licensed nuclear facilities for wildlife, and the environment as a whole. ERAs are therefore expected to be both fit for purpose and conservative. When uranium isotopes are assessed, there are many radioactive decay products which could be considered. However, risk assessors usually assume 235 U and its daughters contribute negligibly to radiological dose. The validity of this assumption has not been tested: what might the 235 U family contribution be and how does the estimate depend on the assumptions applied? In this paper we address this question by considering aquatic wildlife in Canadian lakes exposed to historic uranium mining practices. A full theoretical approach was used, in parallel to a more realistic assessment based on measurements of several elements of the U decay chains. The 235 U family contribution varied between about 4% and 75% of the total dose rate depending on the assumptions of the equilibrium state of the decay chains. Hence, ignoring the 235 U series will not result in conservative dose assessments for wildlife. These arguments provide a strong case for more in situ measurements of the important members of the 235 U chain and for its consideration in dose assessments. - Highlights: • Realistic ecological risk assessment infers a complete inventory of radionuclides. • U-235 family may not be minor when assessing total dose rates experienced by biota. • There is a need to investigate the real state of equilibrium decay of U chains. • There is a need to improve the capacity to measure all elements of the U decay chains.

  8. Limitations on the precision of 238U/235U measurements and implications for environmental monitoring

    International Nuclear Information System (INIS)

    Russ III, G.P.

    1997-01-01

    The ability to determine the isotopic composition of uranium in environmental samples is an important component of the International Atomic Energy Agency's (IAEA) safeguards program, and variations in the isotopic ratio 238 U/ 235 U provide the most direct evidence of isotopic enrichment activities. The interpretation of observed variations in 238 U/ 235 U depends on the ability to distinguish enrichment from instrumental biases and any variations occurring in the environment but not related to enrichment activities. Instrumental biases that have historically limited the accuracy of 238 U/ 235 U determinations can be eliminated by the use of the 233 U/ 236 U double-spike technique. With this technique, it is possible to determine the 238 U/ 235 U in samples to an accuracy equal to the precision of the measurement, ca. 0.1% for a few 10's of nanograms of uranium. Given an accurate determination of 238 U/ 235 U, positive identification of enrichment activities depends on the observed value being outside the range of 238 U/ 235 U's expected as a result of natural or environmental variations. Analyses of a suite of soil samples showed no variation beyond 0.2% in 238 U/ 235 U

  9. Theoretical studies aiming at the IEA-R1 reactor core conversion from high U-235 enrichment to low U-235 enrichment

    International Nuclear Information System (INIS)

    Frajndlich, R.

    1982-01-01

    The research reactors, of which the fuel elements are of MTR type, functions presently, almost in their majority with high U-235 enrichment. The fear that those fuel elements might generate a considerabLe proliferation of nuclear weapons rendered almost mandatory the conversion of highly enriched fuel elements to a low U-235 enrichment. As the IEA-R1 reactor of IPEN is operating with highly enriched fuel elements a study aiming at this conversion was done. The problems related to the conversion and the results obtained, demonstrated the technical viabilty for its realization. (E.G.) [pt

  10. Reimiep 87. An interlaboratory U-235 enrichment determination by gamma measurement on solid UF6 sample

    International Nuclear Information System (INIS)

    Aparo, M.; Cresti, P.

    1988-01-01

    Gamma spectroscopy technique, based on the measurement of U 235 186 KeV flux, is now currently used for the determination of Uranium enrichment in different material of nuclear fuel cycle, namely: Uranium metallic, UO 2 pellets, UF 6 liquid or solid. The present paper describes the use of such a technique and the obtained results in determining the U 235 /U atomic isotopic abundance on a certified UF 6 solid sample. The measurements have been carried out in the frame work of the partecipation to the ''UF 6 Interlaboratory Measurements Evaluation Programme'' organized by CBNM/Geel with the support of the ESARDA (European Safeguards Research and Development Association)

  11. 14.2 MeV neutron induced U-235 fission cross section measurement

    International Nuclear Information System (INIS)

    Li Jingwen; Shen Guanren; Ye Zongyuan; Li Anli; Zhou Shuhua; Sun Zhongfan; Wu Jingxia; Huang Tanzi

    1986-01-01

    The cross section of U-235 fission induced by 14.2 MeV neutrons was measured by the time correlated associated particle method. The result obtained is (2.078+-0.040) barn. Comparison with other author's is also given. (author)

  12. Fuel cycle cost comparison of choices in U-235 recycle in the HTGR

    International Nuclear Information System (INIS)

    Rothstein, M.P.

    1976-07-01

    An analysis of alternative options for the recycle of discharged makeup U-235 (''residual'' makeup) in HTGRs shows that the three-particle system which has been the reference plan remains optimal. This result considers both the resource utilization and the handling costs attendant to the alternative strategies (primarily in the recycle facility and in waste disposal). Furthermore, this result appears to be true under all forseeable economic conditions. A simple risk assessment indicates that recycle cost (including reprocessing, refabrication, and related waste disposal) would have to double or triple in order for the alternative U-235 recycle schemes to become attractive. This induces some degree of confidence in the choice of staying with the reference cycle in spite of the large degree of uncertainty over recycle and its costs

  13. Determination of U-235 quantity in fresh fuel elements by neutron coincidence collar technique

    International Nuclear Information System (INIS)

    Almeida, M.C.M. de; Almeida, S.G. de; Marzo, M.A.S.; Moita, L.P.M.

    1990-01-01

    The U-235 quantity per lenght of fresh fuel assemblies of the Angra-I first recharge was determined by Neutron Coincidence Collar technique (N.C.C.). This technique is well-founded in fresh fuel assemblies activation by thermal neutrons from AmLi source to generate U-235 fission neutrons. These neutrons are detected by coincidence method in polyethylene structure where 18 He-3 detectors were placed. The coincidence counting results, in active mode (AmLi), showed 0,7% to standard deviation and equal to 1,49% to mass in 1000s of counting. The accuracies of different calibration methods were evaluated and compared. The results showed that the operator declared values are consistent. This evaluation was part of technical-exchange program between Safeguards Laboratory from C.N.E.N. and Los Alamos National Lab., United States. (author)

  14. Activation Doppler Measurements on U 238 and U 235 in Some Fast Reactor Spectra

    Energy Technology Data Exchange (ETDEWEB)

    Tiren, L I; Gustafsson, I

    1968-03-15

    Measurements of the Doppler effect in U-238 capture and U-235 fission have been made by means of the activation technique in three different neutron spectra in the fast critical assembly FR0. The experiments involved the irradiation of thin uranium metal foils or oxide disks, which were heated in a small oven located at the core centre. The measurements on U-238 were extended to 1780 deg K and on U-235 to 1470 deg K. A core region surrounding the oven was homogenized in order to facilitate the interpretation of results. The reaction rates in the uranium samples were detected by gamma counting. The experimental method was checked with regard to systematic errors by irradiations in a thermal spectrum. The data obtained for U-238 capture were corrected for the effect of neutron collisions in the oven wall, and were extrapolated to zero sample thickness. In the softest spectrum (core 5) a Doppler effect (relative increase in capture rate) of 0.260 {+-} 0.018 was obtained on heating from 343 to 1780 deg K, and in the hardest spectrum (core 3) the corresponding value was 0.030 {+-} 0.003. An appreciable Doppler effect in U-235 fission was obtained only in the softest spectrum, in which the measured increase in fission rate on heating from 320 to 1470 deg K was 0.007 {+-} 0.003.

  15. Isotopic analysis of uranium hexafluoride highly enriched in U-235; Analyse isotopique de l'hexafluorure d'uranium fortement enrichi en U 235

    Energy Technology Data Exchange (ETDEWEB)

    Chaussy, L; Boyer, R [Commissariat a l' Energie Atomique, Pierrelatte (France). Centre d' Etudes Nucleaires

    1968-07-01

    Isotopic analysis of uranium in the form of the hexafluoride by mass-spectrometry gives gross results which are not very accurate. Using a linear interpolation method applied to two standards it is possible to correct for this inaccuracy as long as the isotopic concentrations are less than about 10 per cent in U-235. Above this level, the interpolations formula overestimates the results, especially if the enrichment of the analyzed samples is higher than 1.3 with respect to the standards. A formula is proposed for correcting the interpolation equation and for the extending its field of application to high values of the enrichment ({approx_equal}2) and of the concentration. It is shown that by using this correction the results obtained have an accuracy which depends practically only on that of the standards, taking into account the dispersion in the measurements. (authors) [French] L'analyse isotopique de l'uranium sous forme d'hexafluorure, par spectrometrie de masse, fournit des resultats bruts entaches d'inexactitude. Une methode d'interpolation lineaire entre deux etalons permet de corriger cette inexactitude, tant que les concentrations isotopiques sont inferieures a 10 pour cent en U-235 environ. Au-dessus de cette valeur, la formule d'interpolation surestime les resultats, notamment si l'enrichissement des echantillons analyses par rapport aux etalons est superieur a 1,3. On propose une formule de correction de l'equation d'interpolation qui etend son domaine d'application jusqu'a des valeurs elevees d'enrichissement ({approx_equal}2) et de concentration. On montre experimentalement que par cette correction, les resultats atteignent, a la precision des mesures, une exactitude qui ne depend pratiquement plus que de celles des etalons. (auteurs)

  16. 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

  17. Effective cross sections of U-235 and Au in a TRIGA-type reactor core

    International Nuclear Information System (INIS)

    Harasawa, S.; Auu, G.A.

    1992-01-01

    The dependence of effective cross sections of gold and uranium for neutron spectrum in Rikkyo University Reactor (TRIGA Mark- II, RUR) fuel cell was studied using computer calculations. The dependence of thermal neutron spectrum with temperature was also investigated. The effective cross section of gold in water of the fuel cell at 32degC was 90.3 barn and the fission cross section of U-235, 483 barn. These two values are similar to the cross sections for neutron energy of 0.034 eV. (author)

  18. Computer simulation of the natural U 238 and U 235 radioactive series decay

    International Nuclear Information System (INIS)

    Barna, A.; Oncescu, M.

    1980-01-01

    The principles of the computer simulation of a radionuclide decay - its decay scheme adoption and codification -, and the adoption principle of a radionuclide chain in a series are applied to the natural U 238 and U 235 series radionuclide decay computer simulation. Using the computer simulation data of these two series adopted chains, the decay characteristic quantities of the series radionuclides, the gamma spectra and the basic characteristics of each of these series are determined and compared with the experimental values given in the literature. (author)

  19. The Effect of Early Diagenesis on the 238U/235U Ratio of Platform Carbonates.

    Science.gov (United States)

    Tissot, F.; Chen, C.; Go, B. M.; Naziemiec, M.; Healy, G.; Swart, P. K.; Dauphas, N.

    2017-12-01

    In the past 15 years, the so-called non-traditional stable isotopes systems (e.g., Mg, Fe, Mo, U) have emerged as powerful tracers of both high-T and low-T geochemical processes (e.g., [1]). Of particular interest for paleoredox studies is the ratio of "stable" isotopes of U (238U/235U), which has the potential to track the global extent of oceanic anoxia (e.g., [2, 3]). Indeed, in the modern ocean, U exists in two main oxidation states, soluble U6+ and insoluble U4+, and has a mean residence time of 400 kyr ([4]), much longer than the global ocean mixing time (1-2 kyr). As such the salinity-normalized ocean is homogeneous with regards to both U concentrations and isotopes (δ238USW = -0.392±0.005 ‰, [2]). The value of δ238USW at any given time is therefore the balance between U input to the ocean, mainly from rivers, and U removal, mostly into biogenic carbonates, anoxic/euxinic sediments and suboxic/hypoxic sediments (e.g., [2, 5]). Because the 238U/235U ratio of the past ocean cannot be measured directly, it has to be estimated from the measurement of the 238U/235U ratio of a sedimentary rock and assuming a constant fractionation factor. Carbonates appear as a promising record since they span most of Earth's history, and the δ238U values of modern primary carbonate precipitates and well-preserved fossil aragonitic coral up to 600 ka are indistinguishable from that of seawater (e.g., [2, 6, 7]). Yet, the effect of secondary processes on the δ238U values of non-coral carbonates, which represent the bulk of the rock record, has only been studied in a handful of shallow samples (down to 40cm, [6]) and remains poorly understood. To investigate the effect of early diagenesis on the 238U/235U ratio of carbonates on the 30kyr to 1Myr timescale, we measured δ13C, δ18O, and δ238U in samples from a 220m long drill core from the Bahamas carbonate platform. In order to separate lattice bound U from secondary U we developed a leaching protocol applicable to carbonate

  20. Measurement of mass distribution of U-235 fission products in the intermediate neutron region

    International Nuclear Information System (INIS)

    Nakagomi, Yoshihiro; Kobayashi, Shohei; Yamamoto, Shuji; Kanno, Ikuo; Wakabayashi, Hiroaki.

    1982-01-01

    The mass distribution and the momentum distribution of U-235 fission products in the intermediate neutron region were measured by using a combination system of the Yayoi intermediate neutron column and an electron linear accelerator. The double energy measurement method was applied. A fission chamber, which consists of an enriched uranium target and two Si surface barrier detectors, was used for the measurement of the neutrons with energy above 1.3 eV. The linear accelerator was operated at the repetition rate of 100 Hz and the pulse width of 10 ns. The data obtained by the two-dimensional pulse height analysis were analyzed by the Schmitt's method. The preliminary results of the mass distribution and the momentum distribution of fission fragments were obtained. (Kato, T.)

  1. 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

  2. Measurements of Pu239:U235 fission ratio using foils at temperatures up to 400 deg. C

    Energy Technology Data Exchange (ETDEWEB)

    Carter, D H; Puckett, B J; Richards, A E [General Reactor Physics Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

    1964-05-15

    The paper describes the use of activation foils for the measurement of Pu239:U235 fission ratios in subcritical lattices at temperatures up to 390 deg C. Counting techniques and the method of analysis of the results are described in detail and the results are compared with fission chamber measurements. (author) 4 refs., 6 figs., 7 tabs.

  3. Method of fault diagnosis in nozzle cascades for U-235 enrichment

    International Nuclear Information System (INIS)

    Schuette, R.; Steinhaus, H.

    1978-09-01

    In a separation nozzle cascade for enrichment of the light uranium isotope U-235 some 450 stages are connected in series. For optimum separation performance of such a plant the design values of the nozzle inlet pressure, of the UF 6 concentration of the UF 6 -cut and the cut of the light additional gas must be matched in all separation stages. Also the feed stream, the product stream, and the tails stream have to be controlled according to the cascade design values. Since it is not possible to measure the cuts directly, these values are calculated on the basis of the material flow balances of the cascade using the pressure values and the UF 6 concentration measurements in each stage, these data being supplemented by concentration measurements in the light and heavy fractions of selected stages. This approach requires the use of a digital computer for processing some 1500 readings to calculate the 2500 plant parameters defining the plant state. This study describes a method of diagnosing the major faults to be expected in a separation nozzle cascade. It is based on the fact that the fault profiles are characterized sufficiently well by maximum values and values to identify the cause of a fault and localize the point where it occurs by means of simple relations between these six values and of their relative positions. The performance of the method has been tested in experiments in the ten-stage pilot plant. For use in commercial separation nozzle cascades the range of performance and the special mode of implementation can be derived from the characteristics of the plant components (separation nozzles, compressors, control valves). The methodological approach in this fault diagnosis also provides the basis for computer aided control procedures to raise a separation cascade from any steady plant condition to its set point operation. (orig./HP) 891 HP [de

  4. Derivation of decay heat benchmarks for U235 and Pu239 by a least squares fit to measured data

    International Nuclear Information System (INIS)

    Tobias, A.

    1989-05-01

    A least squares technique used by previous authors has been applied to an extended set of available decay heat measurements for both U235 and Pu239 to yield simultaneous fits to the corresponding beta, gamma and total decay heat. The analysis takes account of both systematic and statistical uncertainties, including correlations, via calculations which use covariance matrices constructed for the measured data. The results of the analysis are given in the form of beta, gamma and total decay heat estimates following fission pulses and a range of irradiation times in both U235 and Pu239. These decay heat estimates are considered to form a consistent set of benchmarks for use in the assessment of summation calculations. (author)

  5. Recovery of enriched Uranium (20% U-235) from wastes obtained in the preparation of fuel elements for argonaut type reactors

    International Nuclear Information System (INIS)

    Uriarte, A.; Ramos, L.; Estrada, J.; del Val, J. L.

    1962-01-01

    Results obtained with the two following installations for recovering enriched uranium (20% U-235) from wastes obtained in the preparation of fuel elements for Argonaut type reactors are presented. Ion exchange unit to recover uranium form mother liquors resulting from the precipitation ammonium diuranate (ADU) from UO 2 F 2 solutions. Uranium recovery unit from solid wastes from the process of manufacture of fuel elements, consisting of a) waste dissolution, and b) extraction with 10% (v/v) TBP. (Author) 9 refs

  6. 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

  7. Using 238U/235U ratios to understand the formation and oxidation of reduced uranium solids in naturally reduced zones

    Science.gov (United States)

    Jemison, N.; Johnson, T. M.; Druhan, J. L.; Davis, J. A.

    2016-12-01

    Uranium occurs in groundwater primarily as soluble and mobile U(VI), which can be reduced to immobile U(IV), often observed in sediments as uraninite. Numerous U(VI)-contaminated sites, such as the DOE field site in Rifle, CO, contain naturally reduced zones (NRZ's) that have relatively high concentrations of organic matter. Reduction of heavy metals occurs within NRZ's, producing elevated concentrations of iron sulfides and U(IV). Slow, natural oxidation of U(IV) from NRZ's may prolong U(VI) contamination of groundwater. The reduction of U(VI) produces U(IV) with a higher 238U/235U ratio. Samples from two NRZ sediment cores recovered from the Rifle site revealed that the outer fringes of the NRZ contain U(IV) with a high 238U/235U ratio, while lower values are observed in the center . We suggest that as aqueous U(VI) was reduced in the NRZ, it was driven to lower 238U/235U values, such that U(IV) formed in the core of the NRZ reflects a lower 238U/235U. Two oxidation experiments were conducted by injecting groundwater containing between 14.9 and 21.2 mg/L dissolved O2 as an oxidant into the NRZ. The oxidation of U(IV) from this NRZ increased aqueous U(VI) concentrations and caused a shift to higher 238U/235U in groundwater as U(IV) was oxidized primarily on the outer fringes of the NRZ. In total these observations suggest that the stability of solid phase uranium is governed by coupled reaction and transport processes. To better understand various reactive transport scenarios we developed a model for the formation and oxidation of NRZ's utilizing the reactive transport software CrunchTope. These simulations suggest that the development of isotopically heterogeneous U(IV) within NRZ's is largely controlled by permeability of the NRZ and the U(VI) reduction rate. Oxidation of U(IV) from the NRZ's is constrained by the oxidation rate of U(IV) as well as iron sulfides, which can prevent oxidation of U(IV) by scavenging dissolved oxygen.

  8. Preliminary study for the transport of the fuel rods of U235 enriched to 1.8 per cent

    International Nuclear Information System (INIS)

    Cardenas, H.; Perez, A.

    1998-01-01

    Transport of 1,8% U235 enriched fuel rods needs both the evaluation of the radiological risk and considerations about criticality aspects. Issues as diverse production characteristics, storage facilities in the source of origin an economical aspects have to be added to the radiological and nuclear considerations. Transport of those rods through national territory must comply with the Argentine Regulatory authority's regulations, based on the Safety Series No. 6, (ed. 1985) -as amended 1990- IAEA. Safety criteria are exposed, taking into account the amount of material to be transported, container characteristics, packaging type and expedition conditions. (author)

  9. The application of the Harwell neutron absorptiometer to the analysis of U-235 in nuclear fuel components

    International Nuclear Information System (INIS)

    Jones, T.L.; Watson, J.; Taylor, T.A.H.

    1979-05-01

    This paper describes the application of the Harwell Neutron Absorptiometer to routine analysis of the U-235 content of fuel element inserts manufactured at the Dounreay Nuclear Power Development Establishment for the use in Materials Testing Reactors. The instrument response, which is principally dependent on the 235 U closely follows a logarithmic relationship. Neutron attenuation due to the aluminium matrix and the presence of 238 U is less than 2% of the total attenuation. The absorptiometer can be used to estimate the weight of 235 U in a single insert with a total error in the range 1 to 1.6%. (author)

  10. Recovery of enriched Uranium (20% U-235) from wastes obtained in the preparation of fuel elements for argonaut type reactors

    Energy Technology Data Exchange (ETDEWEB)

    Uriarte, A; Ramos, L; Estrada, J; Val, J L. del

    1962-07-01

    Results obtained with the two following installations for recovering enriched uranium (20% U-235) from wastes obtained in the preparation of fuel elements for Argonaut type reactors are presented. Ion exchange unit to recover uranium form mother liquors resulting from the precipitation ammonium diuranate (ADU) from UO{sub 2}F{sub 2} solutions. Uranium recovery unit from solid wastes from the process of manufacture of fuel elements, consisting of a) waste dissolution, and b) extraction with 10% (v/v) TBP. (Author) 9 refs.

  11. Simultaneous measurement of fission fragments and prompt neutrons for thermal neutron-induced fission of U-235

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-03-01

    Simultaneous measurement of fission fragments and prompt neutrons following the thermal neutron induced fission of U-235 has been performed in order to obtain the neutron multiplicity (v) and its emission energy ({eta}) against the specified mass (m{sup *}) and the total kinetic energy (TKE). The obtained value of -dv/dTKE(m{sup *}) showed a saw-tooth distribution. The average neutron energy <{eta}>(m{sup *}) had a distribution with a reflection symmetry around the half mass division. The measurement also gave the level density parameters of the specified fragment, a(m{sup *}), and this parameters showed a saw-tooth trend too. The analysis by a phenomenological description of this parameters including the shell and collective effects suggested the existence of a collective motion of the fission fragments. (author)

  12. Reactive transport of uranium in a groundwater bioreduction study: Insights from high-temporal resolution 238U/235U data

    Science.gov (United States)

    Shiel, A. E.; Johnson, T. M.; Lundstrom, C. C.; Laubach, P. G.; Long, P. E.; Williams, K. H.

    2016-08-01

    We conducted a detailed investigation of U isotopes in conjunction with a broad geochemical investigation during field-scale biostimulation and desorption experiments. This investigation was carried out in the uranium-contaminated alluvial aquifer of the Rifle field research site. In this well-characterized setting, a more comprehensive understanding of U isotope geochemistry is possible. Our results indicate that U isotope fractionation is consistently observed across multiple experiments at the Rifle site. Microbially-mediated reduction is suggested to account for most or all of the observed fractionation as abiotic reduction has been demonstrated to impart much smaller, often near-zero, isotopic fractionation or isotopic fractionation in the opposite direction. Data from some time intervals are consistent with a simple model for transport and U(VI) reduction, where the fractionation factor (ε = +0.65‰ to +0.85‰) is consistent with experimental studies. However, during other time intervals the observed patterns in our data indicate the importance of other processes in governing U concentrations and 238U/235U ratios. For instance, we demonstrate that departures from Rayleigh behavior in groundwater systems arise from the presence of adsorbed species. We also show that isotope data are sensitive to the onset of oxidation after biostimulation ends, even in the case where reduction continues to remove contaminant uranium downstream. Our study and the described conceptual model support the use of 238U/235U ratios as a tool for evaluating the efficacy of biostimulation and potentially other remedial strategies employed at Rifle and other uranium-contaminated sites.

  13. Experiments to determine the rate of beta energy release following fission of Pu239 andU235 in a fast reactor

    International Nuclear Information System (INIS)

    Murphy, M.F.; Taylor, W.H.; Sweet, D.W.; March, M.R.

    1979-02-01

    Measurements have been made of the rate of beta energy release from Pu239 and U235 fission fragments over a period of 107 seconds following a 105 second irradiation in the zero-power fast reactor Zebra. Results are compared with predictions using the UKFPDD-1 decay data file and two different sets of fission product yield data. (author)

  14. Investigating Uranium Mobility Using Stable Isotope Partitioning of 238U/235U and a Reactive Transport Model

    Science.gov (United States)

    Bizjack, M.; Johnson, T. M.; Druhan, J. L.; Shiel, A. E.

    2015-12-01

    We report a numerical reactive transport model which explicitly incorporates the effectively stable isotopes of uranium (U) and the factors that influence their partitioning in bioactive systems. The model reproduces trends observed in U isotope ratios and concentration measurements from a field experiment, thereby improving interpretations of U isotope ratios as a tracer for U reactive transport. A major factor contributing to U storage and transport is its redox state, which is commonly influenced by the availability of organic carbon to support metal-reducing microbial communities. Both laboratory and field experiments have demonstrated that biogenic reduction of U(VI) fractionates the stable isotope ratio 238U/235U, producing an isotopically heavy solid U(IV) product. It has also been shown that other common reactive transport processes involving U do not fractionate isotopes to a consistently measurable level, which suggests the capacity to quantify the extent of bioreduction occurring in groundwater containing U using 238U/235U ratios. A recent study of a U bioremediation experiment at the Rifle IFRC site (Colorado, USA) applied Rayleigh distillation models to quantify U stable isotope fractionation observed during acetate amendment. The application of these simplified models were fit to the observations only by invoking a "memory-effect," or a constant source of low-concentration, unfractionated U(VI). In order to more accurately interpret the measured U isotope ratios, we present a multi-component reactive transport model using the CrunchTope software. This approach is capable of quantifying the cycling and partitioning of individual U isotopes through a realistic network of transport and reaction pathways including reduction, oxidation, and microbial growth. The model incorporates physical heterogeneity of the aquifer sediments through zones of decreased permeability, which replicate the observed bromide tracer, major ion chemistry, U concentration, and U

  15. Effect of U-238 and U-235 cross sections on nuclear characteristics of fast and thermal reactors

    Energy Technology Data Exchange (ETDEWEB)

    Akie, Hiroshi; Takano, Hideki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kaneko, Kunio

    1997-03-01

    Benchmark calculation has been made for fast and thermal reactors by using ENDF/B-VI release 2(ENDF/B-VI.2) and JENDL-3.2 nuclear data. Effective multiplication factors (k{sub eff}s) calculated for fast reactors calculated with ENDF/B-VI.2 becomes about 1% larger than the results with JENDL-3.2. The difference in k{sub eff} is caused mainly from the difference in inelastic scattering cross section of U-238. In all thermal benchmark cores, ENDF/B-VI.2 gives smaller multiplication factors than JENDL-3.2. In U-235 cores, the difference is about 0.3%dk and it becomes about 0.6% in TCA U cores. The difference in U-238 data is also important in thermal reactors, while there are found 0.1-0.3% different v values of U isotopes in thermal energy between ENDF/B-VI.2 and JENDL-3.2. (author)

  16. Neutronics and thermalhydraulics characteristics of the CANDU core fueled with slightly enriched uranium 0.9% U235

    International Nuclear Information System (INIS)

    Raica, V.; Sindile, A.

    1999-01-01

    The interest concerning the slightly enriched uranium (SEU) fuel cycle is due to the possibility to adapt (to convert) the current reactor design using natural uranium fuel to this cycle. Preliminary evaluations based on discharged fuel burnup estimates versus enrichment and on Canadian experience in fuel irradiation suggest that for a 0.93% U-235 enrichment no design modifications are required, not even for the fuel bundle. The purpose of this paper is to resume the results of the studies carried on in order to clarify this problem. The calculation methodology used in reactor physics and thermal-hydraulics analyses that were performed adapted and developed the AECL suggested methodology. In order to prove the possibility to use the SEU 0.93% without any design modification, all the main elements from the CANDU Reactor Physics Design Manual were studied. Also, some thermal-hydraulics analyses were performed to ensure that the operating and safety parameters were respected. The estimations sustain the assumption that the current reactor and fuel bundle design is compatible to the using of the SEU 0.93% fuel. (author)

  17. 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.

  18. 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.

  19. Coulex fission of 234U, 235U, 237Np, and 238Np studied within the SOFIA experimental program

    International Nuclear Information System (INIS)

    Martin, Julie-Fiona

    2014-01-01

    SOFIA (Studies On FIssion with Aladin) is an experimental project which aims at systematically measuring the fission fragments' isotopic yields as well as their total kinetic energy, for a wide variety of fissioning nuclei. The PhD work presented in this dissertation takes part in the SOFIA project, and covers the fission of nuclei in the region of the actinides: 234 U, 235 U, 237 Np and 238 Np. The experiment is led at the heavy-ion accelerator GSI in Darmstadt, Germany. This facility provides intense relativistic primary beam of 238 U. A fragmentation reaction of the primary beam permits to create a secondary beam of radioactive ions, some of which the fission is studied. The ions of the secondary beam are sorted and identified through the FR-S (Fragment Separator), a high resolution recoil spectrometer which is tuned to select the ions of interest.The selected - fissile - ions then fly further to Cave-C, an experimental area where the fission experiment itself takes place. At the entrance of the cave, the secondary beam is excited by Coulomb interaction when flying through an target; the de-excitation process involves low-energy fission. Both fission fragments fly forward in the laboratory frame, due to the relativistic boost inferred from the fissioning nucleus.A complete recoil spectrometer has been designed and built by the SOFIA collaboration in the path of the fission fragments, around the existing ALADIN magnet. The identification of the fragments is performed by means of energy loss, time of flight and deviation in the magnet measurements. Both fission fragments are fully (in mass and charge) and simultaneously identified.This document reports on the analysis performed for (1) the identification of the fissioning system, (2) the identification of both fission fragments, on an event-by-event basis, and (3) the extraction of fission observables: yields, TKE, total prompt neutron multiplicity. These results, concerning the actinides, are discussed, and

  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. 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

  2. Assay of Uranium Isotopic Ratios 234U/238U, 235U/238U in Bottom Sediment Samples Using Destructive and Non Destructive Techniques (Nasser Lake)

    International Nuclear Information System (INIS)

    Agha, A.R.; El-Mongy, S.A.; Kandel, A.E.

    2011-01-01

    Nasser Lake is the greatest man-made lake in the World. It is considered as the main source of water where the Nile water is impounded behind the Aswan high dam.. Uranium has three naturally occurring isotopes 234 U, 235 U and 238 U with isotopic abundance 0.00548, 0.7200 and 99.2745 atom percent. Dissolved uranium in the lake is primary due to weathering process. Monitoring of the isotopic ratios of uranium is used as a good indicator to trace and evaluate the origin and activities associated with any variation of uranium in the lake environment. The main objective of the present study is to clarify any potential variation of natural uranium 234 U/ 238 U, 235 U/ 238 U ratios in sediment samples of Nasser Lake by using destructive alpha and non destructive gamma- techniques. The results show that the uranium isotopic activity ratios are very close to the natural values. This study can also be used for radiological protection and safety evaluation purposes.

  3. Fuel cycle cost, reactor physics and fuel manufacturing considerations for Erbia-bearing PWR fuel with > 5 wt% U-235 content

    Energy Technology Data Exchange (ETDEWEB)

    Franceschini, F.; Lahoda, E. J.; Kucukboyaci, V. N. [Westinghouse Electric Co. LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2012-07-01

    The efforts to reduce fuel cycle cost have driven LWR fuel close to the licensed limit in fuel fissile content, 5.0 wt% U-235 enrichment, and the acceptable duty on current Zr-based cladding. An increase in the fuel enrichment beyond the 5 wt% limit, while certainly possible, entails costly investment in infrastructure and licensing. As a possible way to offset some of these costs, the addition of small amounts of Erbia to the UO{sub 2} powder with >5 wt% U-235 has been proposed, so that its initial reactivity is reduced to that of licensed fuel and most modifications to the existing facilities and equipment could be avoided. This paper discusses the potentialities of such a fuel on the US market from a vendor's perspective. An analysis of the in-core behavior and fuel cycle performance of a typical 4-loop PWR with 18 and 24-month operating cycles has been conducted, with the aim of quantifying the potential economic advantage and other operational benefits of this concept. Subsequently, the implications on fuel manufacturing and storage are discussed. While this concept has certainly good potential, a compelling case for its short-term introduction as PWR fuel for the US market could not be determined. (authors)

  4. Comparative studies for determining U-235/U-238 relation in solutions of natural and depleted uranium using gamma spectrometry and neutron activation analysis

    International Nuclear Information System (INIS)

    Cassorla F, V.; Valle M, L.; Pena V, L.

    1988-01-01

    Two experimental methods were developed for determining U-235/U-238 ratio in uranium solutions. The isotopic was measured by high resolution ratio gamma-ray spectrometry (G.S.) and neutron activation analysis (N.A.A.). The precision obtained was similar for both methods, but better sensitivity was obtained by N.A.A. The accuracy in both cases was stablished by comparison with samples previously analyzed by mass spectrometry, the results were satisfactory for both techniques. Studies involving the influence of the nitric acid concentration on the isotopic ratio measurement, also were done. In addition, computer programs for faster data reduction were developped, in the case of N.A.A. (author)

  5. Establishment of an authenticated physical standard for gamma spectrometric determination of the U-235 content of MTR fuel and evaluation of measurement procedures

    International Nuclear Information System (INIS)

    Fleck, C.M.

    1979-12-01

    Measurements of U-235 content in a standard MTR fuel element were carried out, using scintillation and semi-conductor spectrometers. Three different types of measurement were carried out: a) Comparison of different primary standards among one another and with single fuel plates. b) Calibration of the MTR fuel element as an authenticated physical standard. c) Evaluation of over all errors in assay measurements on MTR fuel elements. The error of the whole assay measurement will be approximately 0.9%. The Uranium distribution in the single fuel plates is the original source of error. In the case of equal Uranium contents in all fuel plates of one fuel assembly, the error of assay measurements would be about 0.3% relative to the primary standards

  6. 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

  7. Studies on the separation of rare earth elements and the nuclear decay properties of short lived rare-earth nuclides in U-235 fission products

    International Nuclear Information System (INIS)

    Ohyoshi, Akira; Ohyoshi, Emiko.

    1980-01-01

    The effect of a complex-forming agent, with which rare earths consecutively form the complexes, on the separation of a pair of adjacent rare earths by electromigration has been investigated. The relation between the separation factor for two complexes and the ligand-ion concentration was examined in the separation of La-Ce and Ce-Pr pairs with nitrilotriacetic acid. Rare earths were able to be isolated rapidly at the optimum ligand-ion concentration in lower one, and this method was applied to study the nuclear decay properties of the short lived isotopes of La, Ce, Pr, Nd and Yt formed in the fission of U-235. This method permits the direct measurement of the decay of La-144 without the interference from the radiation of other fission products. The gamma-ray spectrum of La-144 was measured with a high resolution Ge(Li) detector, and the gamma-transition was observed. From the decay plots of two strong photopeaks, the half-life of La-144 was determined. In the case of Ce fraction, the photopeaks assigned to respective isotopes were observed. In the studies on the decay properties of Pr-148 and Pr-149, the decay plot of the strong photopeak showed good linearity, and the accurate half-life of Pr-148 was determined. Similarly, the half-life of Pr-149 was longer than the previously reported value. (Kako, I.)

  8. Research and design calculation of multipurpose critical assembly using moderated light water and low enriched fuel from 1.6 to 5.0% U-235

    International Nuclear Information System (INIS)

    Nguyen Kien Cuong; Vo Doan Hai Dang; Luong Ba Vien; Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Minh Tuan; Nguyen Manh Hung; Pham Quang Huy; Tran Quoc Duong; Tran Tri Vien

    2015-01-01

    Basing on the idea in ??using fuel of nuclear power plants such as PWR (AP-1000) and VVER-1000 with light water as moderation, design calculation of critical assembly was performed to confirm the possibility of using these fuels. Designed critical assembly has simple structure consisting of low enriched fuel from 1.6% to 5% U-235; water has functions as cooling, biological protection and control. Critical assembly is operated at nominal power 100 W with fuel pitch about 2.0 cm. Applications of the critical assembly are quite abundant in basic research, education and training with low investment cost compare with research reactor and easy in operation. So critical assembly can be used for university or training centre for nuclear engineering training. Main objectives of the project are: design calculation in neutronics, thermal hydraulics and safety analysis for critical configuration benchmarks using low enriched fuel; design in mechanical and auxiliary systems for critical assembly; determine technical specifications and estimate construction, installation cost of critical assembly. The process of design, fabrication, installation and construction of critical assembly will be considered with different implementation phases and localization capabilities in installation of critical assembly is highly feasibility. Cost estimation of construction and installation of critical assembly was implemented and showed that investment cost for critical assembly is much lower than research reactor and most of components, systems of critical assembly can be localized with current technique quality of the country. (author)

  9. Determination of the activity of the uranium isotopes U-234, U-235 and U-238 in environmental samples by alpha spectrometry

    International Nuclear Information System (INIS)

    Kromphorn, G.

    1996-02-01

    Different materials containing urandium are regularly investigated in the Laboratory for Environmental Radioactivity of the Physikalisch-Technische Bundesanstalt (PTB) with respect to the activity of the uranium isotopes ( 234 U, 235 U, and 238 U). Moreover for reasons of quality assurance, the PTB takes part in international comparisons where also uranium contents are to be determined in environmental samples and in the framework of which reference materials can be certified. Finally in national comparisons the PTB has the task to determine values of the specific activity for the different isotopes which can play the role of nominal (orientation) values. The single steps of uranium analyses are described after a compilation of the most important data of the uranium isotopes contained in natural uranium: The use of 232 U as tracer, the chemical separation analytics, the production of α-sources and the measuring methods. Analyses of a soil sample and a waste water sample with respect to their specific uranium activity have been chosen as examples of a practical application. (orig.) [de

  10. Distribution of equilibrium burnup for an homogeneous core with fuel elements of slightly enriched uranium (0.85% U-235) at Atucha I nuclear power plant

    International Nuclear Information System (INIS)

    Sidelnik, J.I.; Perez, R.A.; Salom, G.F.

    1987-01-01

    At Atucha I, the present fuel management with natural uranium comprises three burnup areas and one irradiation path, sometimes performing four steps in the reactor core, according to the requirements. The discharge burnup is 6.0 Mw d/kg U for a waste reactivity of 6.5 m k and a heavy water purity of 99.75%. This is a preliminary study to obtain the distribution of equilibrium burnup of an homogeneous core with slightly enriched uranium (0.85% by weight U-235), using the time-averaged method implemented in the code PUMA and a representative model of one third of core and fixed rod position. It was found a strategy of three areas and two paths that agrees with the present limits of channel power and specific power in fuel rod. The discharge burnup obtained is 11.6 Mw d/kg U. This strategy is calculated with the same method and a full core representation model is used to verify the obtained results. (Author)

  11. Uranium ((234)U, (235)U and (238)U) contamination of the environment surrounding phosphogypsum waste heap in Wiślinka (northern Poland).

    Science.gov (United States)

    Olszewski, Grzegorz; Boryło, Alicja; Skwarzec, Bogdan

    2015-08-01

    The aim of this work was to determine the uranium concentration ((234)U, (235)U and (238)U) and values of the activity ratio (234)U/(238)U in soil samples collected near phosphogypsum waste heap in Wiślinka (northern Poland). On the basis of the studies it was found that the values of the (234)U/(238)U activity ratio in the analyzed soils collected in the vicinity of phosphogypsum dump in Wiślinka are in most cases close to one and indicate the phosphogypsum origin of the analyzed nuclides. The obtained results of uranium concentrations are however much lower than in previous years before closing of the phosphogypsum stockpile. After this process and covering the phosphogypsum stockpile in Wiślinka with sewage sludge, phosphogypsum particles are successfully immobilized. In the light of the results the use of phosphate fertilizers seems to be a major problem. Prolonged and heavy rains can cause leaching accumulated uranium isotopes in the phosphogypsum stockpile, which will be washed into the Martwa Wisła and on the fields in the immediate vicinity of this storage. Copyright © 2015 Elsevier Ltd. All rights reserved.

  12. 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

  13. Distinct 238U/235U ratios and REE patterns in plutonic and volcanic angrites: Geochronologic implications and evidence for U isotope fractionation during magmatic processes

    Science.gov (United States)

    Tissot, François L. H.; Dauphas, Nicolas; Grove, Timothy L.

    2017-09-01

    Angrites are differentiated meteorites that formed between 4 and 11 Myr after Solar System formation, when several short-lived nuclides (e.g., 26Al-26Mg, 53Mn-53Cr, 182Hf-182W) were still alive. As such, angrites are prime anchors to tie the relative chronology inferred from these short-lived radionuclides to the absolute Pb-Pb clock. The discovery of variable U isotopic composition (at the sub-permil level) calls for a revision of Pb-Pb ages calculated using an ;assumed; constant 238U/235U ratio (i.e., Pb-Pb ages published before 2009-2010). In this paper, we report high-precision U isotope measurement for six angrite samples (NWA 4590, NWA 4801, NWA 6291, Angra dos Reis, D'Orbigny, and Sahara 99555) using multi-collector inductively coupled plasma mass-spectrometry and the IRMM-3636 U double-spike. The age corrections range from -0.17 to -1.20 Myr depending on the samples. After correction, concordance between the revised Pb-Pb and Hf-W and Mn-Cr ages of plutonic and quenched angrites is good, and the initial (53Mn/55Mn)0 ratio in the Early Solar System (ESS) is recalculated as being (7 ± 1) × 10-6 at the formation of the Solar System (the error bar incorporates uncertainty in the absolute age of Calcium, Aluminum-rich inclusions - CAIs). An uncertainty remains as to whether the Al-Mg and Pb-Pb systems agree in large part due to uncertainties in the Pb-Pb age of CAIs. A systematic difference is found in the U isotopic compositions of quenched and plutonic angrites of +0.17‰. A difference is also found between the rare earth element (REE) patterns of these two angrite subgroups. The δ238U values are consistent with fractionation during magmatic evolution of the angrite parent melt. Stable U isotope fractionation due to a change in the coordination environment of U during incorporation into pyroxene could be responsible for such a fractionation. In this context, Pb-Pb ages derived from pyroxenes fraction should be corrected using the U isotope composition

  14. Ground water contamination with (238)U, (234)U, (235)U, (226)Ra and (210)Pb from past uranium mining: cove wash, Arizona.

    Science.gov (United States)

    Dias da Cunha, Kenya Moore; Henderson, Helenes; Thomson, Bruce M; Hecht, Adam A

    2014-06-01

    The objectives of the study are to present a critical review of the (238)U, (234)U, (235)U, (226)Ra and (210)Pb levels in water samples from the EPA studies (U.S. EPA in Abandoned uranium mines and the Navajo Nation: Red Valley chapter screening assessment report. Region 9 Superfund Program, San Francisco, 2004, Abandoned uranium mines and the Navajo Nation: Northern aum region screening assessment report. Region 9 Superfund Program, San Francisco, 2006, Health and environmental impacts of uranium contamination, 5-year plan. Region 9 Superfund Program, San Franciso, 2008) and the dose assessment for the population due to ingestion of water containing (238)U and (234)U. The water quality data were taken from Sect. "Data analysis" of the published report, titled Abandoned Uranium Mines Project Arizona, New Mexico, Utah-Navajo Lands 1994-2000, Project Atlas. Total uranium concentration was above the maximum concentration level for drinking water (7.410-1 Bq/L) in 19 % of the water samples, while (238)U and (234)U concentrations were above in 14 and 17 % of the water samples, respectively. (226)Ra and (210)Pb concentrations in water samples were in the range of 3.7 × 10(-1) to 5.55 × 102 Bq/L and 1.11 to 4.33 × 102 Bq/L, respectively. For only two samples, the (226)Ra concentrations exceeded the MCL for total Ra for drinking water (0.185 Bq/L). However, the (210)Pb/(226)Ra ratios varied from 0.11 to 47.00, and ratios above 1.00 were observed in 71 % of the samples. Secular equilibrium of the natural uranium series was not observed in the data record for most of the water samples. Moreover, the (235)U/(total)U mass ratios ranged from 0.06 to 5.9 %, and the natural mass ratio of (235)U to (total)U (0.72 %) was observed in only 16 % of the water samples, ratios above or below the natural ratio could not be explained based on data reported by U.S. EPA. In addition, statistical evaluations showed no correlations among the distribution of the radionuclide concentrations

  15. 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)

  16. Decay scheme of the U235

    International Nuclear Information System (INIS)

    Gaeta, R.

    1965-01-01

    A study of the Th 2 31 excited levels from the alpha decay of the U 2 35, is carried out. The alpha particle spectrum was measured by means of a semiconductor counter spectrometer with an effective resolution of 18 keV. Nineteen new lines were identified. The gamma-ray spectrum was measured with thin samples of U 2 35, free from decay products, and in such geometrical conditions, that most of the interference effects were eliminated. The gamma-gamma coincidence spectra have made easier a better knowledge of the transition between the several levels. (Author) 110 refs

  17. Decay heat measurement of U-235

    International Nuclear Information System (INIS)

    Baumung, K.

    1976-01-01

    The calorimeter and the transport mechanism for the fuel samples was designed and is under construction now. Calculations of the heat-source distributions for different 235U-contents led to an optimal enrichment of the UO 2 -samples which minimizes the effects of the bad heat conductivity of the oxide on temperature measurement. Monte-Carlo-calculations of the γ-leakage-spectra yielded data which allow, from the γ-energy-flow measurements, to calculate the total γ-energy loss as well as the portions of the β- and γ-heating. (orig.) [de

  18. Catalytic generation of methane at 60-100 °C and 0.1-300 MPa from source rocks containing kerogen Types I, II, and III

    Science.gov (United States)

    Wei, Lin; Schimmelmann, Arndt; Mastalerz, Maria; Lahann, Richard W.; Sauer, Peter E.; Drobniak, Agnieszka; Strąpoć, Dariusz; Mango, Frank D.

    2018-06-01

    water and thermogenic methane in hydrous pyrolysis experiments above 300 °C, the hydrogen isotopic composition of added water exerted limited influence on the δ2H value of methane generated catalytically at low temperatures. We hypothesize that the catalytic sites responsible for methanogenesis are located in hydrophobic microenvironments with limited access to water. The δ13CCH4 values of methane generated catalytically at 60-100 °C range from ∼-57.6 to -41.4‰ and are thus similar to typical thermogenic methane (δ13CCH4 >-50‰) and microbially generated methane (methane can diagnose the low-temperature regime of catalytic methanogenesis. Furthermore, testing of freshly cored anoxic rocks is needed to determine whether the use of archived, oxygen-exposed rocks in geochemical maturation/catalysis studies introduces artifacts in experimental hydrocarbon yields.

  19. 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

  20. Nuclear Excitation by Electronic Transition of U-235

    Energy Technology Data Exchange (ETDEWEB)

    Chodash, Perry Adam [Univ. of California, Berkeley, CA (United States)

    2015-07-14

    Nuclear excitation by electronic transition (NEET) is a rare nuclear excitation that is theorized to occur in numerous isotopes. One isotope in particular, 235U, has been studied several times over the past 40 years and NEET of 235U has never been conclusively observed. These past experiments generated con icting results with some experiments claiming to observe NEET of 235U and others setting limits for the NEET rate. This dissertation discusses the latest attempt to measure NEET of 235U. If NEET of 235U were to occur, 235mU would be created. 235mU decays by internal conversion with a decay energy of 76 eV and a half-life of 26 minutes. A pulsed Nd:YAG laser operating at 1064 nm with a pulse energy of 789 mJ and a pulse width of 9 ns was used to generate a uranium plasma. The plasma was captured on a catcher plate and electrons emitted from the catcher plate were accelerated and focused onto a microchannel plate detector. A decay of 26 minutes would suggest the creation of 235mU and the possibility that NEET occurred. However, measurements performed using a variety of uranium targets spanning depleted uranium up to 99.4% enriched uranium did not observe a 26 minute decay. Numerous other decays were observed with half-lives ranging from minutes up to hundreds of minutes. While NEET of 235U was not observed during this experiment, an upper limit for the NEET rate of 235U was determined. In addition, explanations for the con icting results from previous experiments are given. Based on the results of this experiment and the previous experiments looking for NEET of 235U, it is likely that NEET of 235U has never been observed.

  1. A review of U-235 decay heat measurements and calculations

    International Nuclear Information System (INIS)

    Walker, W.H.

    1979-08-01

    Recent scintillator measurements of fission product decay β and γ power, and calorimetric measurements of their sum are analyzed to obtain estimates of E sub(β) and E sub(γ), the β and γ components of the delayed energy per fission in a reactor. Calculations using the ENDF/B-4 fission product file are compared to the measured results and used to estimate the contributions to E sub(β) and E sub(γ) for decay times greater than 10 5 s. A value of E sub(ν), the anti-neutrino component, consistent with the measured component is also calculated. It is found that the decay heat measured in two calorimetric experiments (the sum of the β and γ components) is about 15 percent greater than the separately-measured energies (averages of five β and two γ measurements). Thus, depending on normalization, E sub(β) and E sub(γ) can vary widely. After all experimental uncertainties are taken into account the range of possible values has as lower limits the values calculated using ENDF/B-4, with upper limits about 40 percent greater. (author)

  2. Phase relationship in AL-Cu-Sc alloys at 450-500 deg C

    International Nuclear Information System (INIS)

    Kharakterova, M.L.

    1991-01-01

    Al-Cu-Sc alloys containing up to 40% Cu and up to 6% Sc at 450 deg C and 500 deg C are studied using light microscopy, X-ray-spectral microanalysis, X-ray diffraction analysis, scanning electron microscopy, measurement of microhardness and electric resistance. It is determined, that in equilibrium with aluminium solid solution under the given temperature ther are Al 3 Sc, CuAl 2 phases of the respective binary systems and W (ScCu 6.6-4 Al 5.4-8 ) ternary phase. Isothermal cross sections of Al-Cu-Sc system at 450 and 500 deg C are plotted. Microhardness of equilibrium phases is measured. Combined solubility of copper and scandium in aluminium is determined

  3. Decay scheme of the U{sup 2}35; Esquema de desintegracion del U-235

    Energy Technology Data Exchange (ETDEWEB)

    Gaeta, R

    1965-07-01

    A study of the Th{sup 2}31 excited levels from the alpha decay of the U{sup 2}35, is carried out. The alpha particle spectrum was measured by means of a semiconductor counter spectrometer with an effective resolution of 18 keV. Nineteen new lines were identified. The gamma-ray spectrum was measured with thin samples of U{sup 2}35, free from decay products, and in such geometrical conditions, that most of the interference effects were eliminated. The gamma-gamma coincidence spectra have made easier a better knowledge of the transition between the several levels. (Author) 110 refs.

  4. Determination of U235 enrichment from nuclear fuel by neutronic activation

    International Nuclear Information System (INIS)

    Almeida, M.C.M. de.

    1988-01-01

    The enrichment of 235 U in UO 2 pellets samples through the instrumental neutron activation analysis method (I.N.A.A.) was determined. By high resolution gamma-ray spectrometry (H.R.G.S.), from analysis of isotopic ratios between fission products peaks from 235 U and 239 Np different energies peaks from 238 U, the enrichment was achieved. The 'Boatstrap' statistics technique for the analytical results, which is based in shaping results of an unknown distribution to the Gaussian distribution by B replications in interested statistics such as: the mean and its standard error, was introduced. (M.J.C.) [pt

  5. U-235 Holdup Measurements in the 321-M Lathe HEPA Banks

    International Nuclear Information System (INIS)

    Salaymeh, S.R.

    2002-01-01

    The Analytical Development Section of Savannah River Technology Center (SRTC) was requested by the Facilities Decommissioning Division (FDD) to determine the holdup of enriched uranium in the 321-M facility as part of an overall deactivation project of the facility. The results of the holdup assays are essential for determining compliance with the Waste Acceptance Criteria, Material Control and Accountability, and to meet criticality safety controls. This report covers holdup measurements of uranium residue in six high efficiency particulate air (HEPA) filter banks of the A-lathe and B-lathe exhaust systems of the 321-M facility. This report discusses the non-destructive assay measurements, assumptions, calculations, and results of the uranium holdup in these six items

  6. Distribution of U-235 and U-238 in food(II)

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Yong Woo; Lee, Jeong Min; Oh, Hyun Kyun; Han, Man Joong [KORTIC, Taejon (Korea, Republic of)

    2004-02-15

    To accomplish the objective of this work, analyses of uranium concentration in 100 food samples (fruit and vegetables) were carried out. The surveillance results on uranium isotope concentrations in food of Korea will be used as a basis of estimating the annual effective dose to the public by uranium and a baseline data to establish a national recommend value for the protection from radiation of food.

  7. Cumulative fission yield of Ce-148 produced by thermal-neutron fission of U-235

    International Nuclear Information System (INIS)

    Hasan, A.A.

    1984-12-01

    Cumulative fission yield of 148 cesium isotopes and some other fission products produced by thermal-neutron fission of 235 uranium is determined by Germanium/Lithium spectroscopic methods. The measuremets were done at Tsing-Hua open pool reactor using 3 to 4 mg of 93.15% enriched 235 uranium samples. Gamma rays are assigned to the responsible fission products by matching gamma rays energies and half lives. Fission rate is calculated by fission track method. Cumulative fission yields of 148 cesium, 90 krypton, 130 iodine, 144 lanthanum, 89 krypton, 136 xenon, 137 xenon and 140 cesium are calculated. This values are compared with previously predicted values and showed good agreement. 21 Ref

  8. Average cross section measurements in U-235 fission neutron spectrum for some threshold reactions

    International Nuclear Information System (INIS)

    Maidana, N.L.

    1993-01-01

    The average cross section in the 235 U fission spectrum has been measured by the activation technique, for the following thresholds reactions: 115 In(n,n') 115m In, 232 Th(n,f) P.F., 46 , 47 , 48 Ti(n,p) 46,47 , 48 Sc, 55 Mn(n,2 n) 54 Mn, 51 V(n,α) 48 Sc, 90 Zr(n,2 n) 89 Zr, 93 Nb(n,2 n) 92m Nb, 58 Ni(n,2 n) 57 Ni, 24 Mg(n,p) 24 Na, 56 Fe(n,p) 56 Mn, 59 Co(n,α) 56 Mn and 63 Cu(n,α) 60 Co. The activation foils were irradiated close (∼ 4 mm) to the core of the IEA-R1 research reactor in the IPEN-CNEN/SP. The reactor was operated at 2 MW yielding a fast neutron flux around 5 x 10 12 n.cm -2 . s -1 . The neutron flux density was monitored by activation reactions with well known averaged cross sections and with effective thresholds above 1 MeV. The foil activities were measured in a calibrated HPGe spectrometer. The neutron spectrum has been calculated using the SAIPS unfolding system applied to the activation data. A detailed error analysis was performed using the covariance matrix methodology. The results were compared with those from other authors. (author)

  9. Beta decay heat following U-235, U-238 and Pu-239 neutron fission

    Science.gov (United States)

    Li, Shengjie

    1997-09-01

    This is an experimental study of beta-particle decay heat from 235U, 239Pu and 238U aggregate fission products over delay times 0.4-40,000 seconds. The experimental results below 2s for 235U and 239Pu, and below 20s for 238U, are the first such results reported. The experiments were conducted at the UMASS Lowell 5.5-MV Van de Graaff accelerator and 1-MW swimming-pool research reactor. Thermalized neutrons from the 7Li(p,n)7Be reaction induced fission in 238U and 239Pu, and fast neutrons produced in the reactor initiated fission in 238U. A helium-jet/tape-transport system rapidly transferred fission fragments from a fission chamber to a low background counting area. Delay times after fission were selected by varying the tape speed or the position of the spray point relative to the beta spectrometer that employed a thin-scintillator-disk gating technique to separate beta-particles from accompanying gamma-rays. Beta and gamma sources were both used in energy calibration. Based on low-energy(energies 0-10 MeV. Measured beta spectra were unfolded for their energy distributions by the program FERD, and then compared to other measurements and summation calculations based on ENDF/B-VI fission-product data performed on the LANL Cray computer. Measurements of the beta activity as a function of decay time furnished a relative normalization. Results for the beta decay heat are presented and compared with other experimental data and the summation calculations.

  10. 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.

  11. Study of Photon Strength Functions of Actinides: the case of U-235, Np-238 and Pu-241

    CERN Document Server

    Guerrero, C; Cano-Ott, D; Martinez, T; Mendoza, E; Villamarin, D; 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; Andriamonje, S; Calviani, M; Chiaveri, E; Gonzalez-Romero, E; Kadi, Y; Vicente, M C; Vlachoudis, V; 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; 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; 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; Heil, M; 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

    The decay from excited levels in medium and heavy nuclei can be described in a statistical approach by means of Photon Strength Functions and Level Density distributions combined with the theory of the compound. The study of electromagnetic cascades following neutron capture by means of high efficiency detectors has been shown to be well suited for probing the properties of the Photon Strength Function of heavy (high level density) and/or radioactive (high background) nuclei. In this work we have investigated for the first time the validity of the recommended PSF for actinides, in particular 235U, 238Np and 241Pu. Our study includes the search for resonance structures in the PSF below Sn and draws conclusions regarding their existence and their characteristics in terms of energy, width and electromagnetic nature.

  12. Development and testing of computer aided start-up procedures for U-235 enrichment separation nozzle cascades

    International Nuclear Information System (INIS)

    Hein, H.; Schuette, R.; Steinhaus, H.

    1978-09-01

    In industrial scale separation nozzle cascades measured data acquisition and process control will be performed by on-line computer systems. To develop the necessary measurement and monitoring techniques, on-line control of the ten stage pilot plant by the WANG 2200 miniature computer system was tested. On-line data processing at the same time has reduced the number of measurements because a large number of operating data which cannot be measured directly can be obtained from a reduced number of measurements by calculation from the material balances in the gas dynamics network of the enrichment cascade. This is performed on-line by means of a cascade monitoring code which produces diagrams of the deviations of major operating parameters from their set values and thus offers to cascade operators an improved picture of the present status of a cascade, thereby facilitating the detection of causes of defects. The graphic representation chosen for this purpose and the performance of the system in practical operation is explained by an example of cascade adjustment and by a few selected examples of cascade fault conditions. Calculation of the quantities derived at all stages from the pressure (po) and concentration (No) readings requires that the individual flow characteristics of the separation nozzles are known. Alos the compressor characteristics have to be determined. Since the monitoring computer in calculating the cascade status utilizes these component characteristics in current enrichment operation, routinely establishing and plotting these two characteristics is also attributed to the computer system (start-up procedures). (orig.) 891 HP [de

  13. Study of U235 neutron fission spectrum by the knowledge of cross sections average over that spectrum

    International Nuclear Information System (INIS)

    Suarez, P.M.

    1997-01-01

    A literature search of cross sections averaged over the fission neutron spectrum confirms inconsistencies between calculated and experimental values for high threshold reactions. Since, in this case, calculated averaged cross sections are systematically lower than measured values, it is concluded that the representations used to carry out these calculations underestimate the number of neutrons in the high energy region of the spectrum. A careful measurement of the averaged cross section for the 45 Sc(n,2n) 44g Sc and 45 Sc(n,2n) 44m Sc high threshold reactions had been performed in the RA-6 Neutron Activation Analysis Laboratory after carefully checking that the neutron flux at the core position where the samples were being irradiated was indeed an undisturbed fission spectrum. The experimental values are greater than those calculated with either, Watt type representations or the one based on the Madland and Nix model for the prompt fission spectrum. In many areas of nuclear engineering, like validation of nuclear data, reactor calculations, applied nuclear physics, shielding design, etc., it is of great practical importance to have a representation for the neutron flux that can be expressed in a closed analytical form and that agrees with experimental results, specially for the most widely fissile nuclide, 235 U. The results of the calculations mentioned above lead us to propose an analytical form for the 235 U fission neutron spectrum that better agrees with experimental results in the whole energy spectrum. We propose two different forms; both are a modification of the Watt-type form that has been adopted within the ENDF/B-V files. One of the new analytical representations is defined in two regions: below 9.5 MeV it is exactly the same formula as that used within the ENDF/B-V files, above this energy the parameters of this formula are changed. The other proposed analytical representation is expressed by a single formula in the whole energy range. These two new representations were checked against several thest which showed that they reproduce experimental results, even for high threshold reactions, much more faithfully than representations based on either macroscopic (Watt and Maxwellian type) or microscopic (Madland and Nix) models do. A new method that takes into account the correlation matrix and covariance calculations is presented for the treatment of uncertainties in all pertinent calculations involved. New methods for the critical evaluation of nuclear data are also presented. (author). 97 refs., figs., tabs

  14. A new method to measure the U-235 content in fresh LWR fuel assemblies via fast-neutron passive self-interrogation

    Science.gov (United States)

    Menlove, Howard; Belian, Anthony; Geist, William; Rael, Carlos

    2018-01-01

    The purpose of this paper is to provide a solution to a decades old safeguards problem in the verification of the fissile concentration in fresh light water reactor (LWR) fuel assemblies. The problem is that the burnable poison (e.g. Gd2O3) addition to the fuel rods decreases the active neutron assay for the fuel assemblies. This paper presents a new innovative method for the verification of the 235U linear mass density in fresh LEU fuel assemblies that is insensitive to the burnable poison content. The technique makes use of the 238U atoms in the fuel rods to self-interrogate the 235U mass. The innovation for the new approach is that the 238U spontaneous fission (SF) neutrons from the rods induces fission reactions (IF) in the 235U that are time correlated with the SF source neutrons. Thus, the coincidence gate counting rate benefits from both the nu-bar of the 238U SF (2.07) and the 235U IF (2.44) for a fraction of the IF reactions. Whereas, the 238U SF background has no time-correlation boost. The higher the detection efficiency, the higher the correlated boost because background neutron counts from the SF are being converted to signal doubles. This time-correlation in the IF signal increases signal/background ratio that provides a good precision for the net signal from the 235U mass. The hard neutron energy spectrum makes the technique insensitive to the burnable poison loading where a Cd or Gd liner on the detector walls is used to prevent thermal-neutron reflection back into the fuel assembly from the detector. We have named the system the fast-neutron passive collar (FNPC).

  15. Determination of the isotope U-235 in uranium hexafluoride by gas mass spectrometry: results of an interlaboratory experiment performed in 1975

    International Nuclear Information System (INIS)

    Duerr, W.; Grossgut, W.; Beyrich, W.

    1977-02-01

    Samples of UF 6 with a 235 U content of about 0.4, 0.7 and 3% were measured with 10 gas mass spectrometers in 8 European laboratories. Identical reference materials were used with 235 U abundances deviating less than 6% from those of the samples and known with an accuracy better than +- 0.15%. By statistical evaluation of the data, errors of about 0.1% were calculated for the determination of the ratio of ratios 235 U: 238 U (sample)/ 235 U: 238 U (reference) with increasing tendency for 235 U abundances below the natural range. (orig./HP) [de

  16. Adjustment of a direct method for the determination of man body burden in Pu-239 on by X-ray detection of U-235

    International Nuclear Information System (INIS)

    Boulay, P.

    1968-04-01

    The use of Pu-239 on a larger scale sets a problem about the contamination measurement by aerosol at lung level. A method of direct measurement of Pu-239 lung burden is possible, thanks to the use of a large area window proportional counter. A counter of such pattern, has been especially carried out for this purpose. The adjustment of the apparatus allows an adequate sensibility to detect a contamination at the maximum permissible body burden level. Besides, a method for individual 'internal calibration', with a plutonium mock: the protactinium-233, is reported. (author) [fr

  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. 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

  19. Cleanup levels for Am-241, Pu-239, U-234, U-235 and U-238 in soils at the Rocky Flats Environmental Technology Site

    International Nuclear Information System (INIS)

    Roberts, R.; Colby, B.; Brooks, L.; Slaten, S.

    1997-01-01

    This presentation briefly outlines a cleanup program at a Rocky Flats site through viewgraphs and an executive summary. Exposure pathway analyses to be performed are identified, and decontamination levels are listed for open space and office worker exposure areas. The executive summary very briefly describes the technical approach, RESRAD computer code to be used for analyses, recommendations for exposure levels, and application of action levels to multiple radionuclide contamination. Determination of action levels for surface and subsurface soils, based on radiation doses, is discussed. 1 tab

  20. Influence of silver concentration in Ag-x(Sb0.40S0.60)(100-x) thin amorphous films on photoinduced crystallization

    Czech Academy of Sciences Publication Activity Database

    Gutwirth, J.; Wágner, T.; Bezdička, Petr; Kotulanová, Eva; Vlček, Milan; Kasap, S. O.; Frumar, M.

    2007-01-01

    Roč. 9, č. 10 (2007), s. 3064-3071 ISSN 1454-4164 R&D Projects: GA MŠk LC523; GA ČR GA203/06/1368 Institutional research plan: CEZ:AV0Z40320502; CEZ:AV0Z40500505 Keywords : Ag-Sb-S films * amorphous * photoinduced crystallization Subject RIV: CA - Inorganic Chemistry Impact factor: 0.827, year: 2007

  1. Geochemical and multi-isotopic (87Sr/86Sr, 143Nd/144Nd, 238U/235U) perspectives of sediment sources, depositional conditions, and diagenesis of the Marcellus Shale, Appalachian Basin, USA

    Science.gov (United States)

    Phan, Thai T.; Gardiner, James B.; Capo, Rosemary C.; Stewart, Brian W.

    2018-02-01

    We investigate sediment sources, depositional conditions and diagenetic processes affecting the Middle Devonian Marcellus Shale in the Appalachian Basin, eastern USA, a major target of natural gas exploration. Multiple proxies, including trace metal contents, rare earth elements (REE), the Sm-Nd and Rb-Sr isotope systems, and U isotopes were applied to whole rock digestions and sequentially extracted fractions of the Marcellus shale and adjacent units from two locations in the Appalachian Basin. The narrow range of εNd values (from -7.8 to -6.4 at 390 Ma) is consistent with derivation of the clastic sedimentary component of the Marcellus Shale from a well-mixed source of fluvial and eolian material of the Grenville orogenic belt, and indicate minimal post-depositional alteration of the Sm-Nd system. While silicate minerals host >80% of the REE in the shale, data from sequentially extracted fractions reflect post-depositional modifications at the mineralogical scale, which is not observed in whole rock REE patterns. Limestone units thought to have formed under open ocean (oxic) conditions have δ238U values and REE patterns consistent with modern seawater. The δ238U values in whole rock shale and authigenic phases are greater than those of modern seawater and the upper crust. The δ238U values of reduced phases (the oxidizable fraction consisting of organics and sulfide minerals) are ∼0.6‰ greater than that of modern seawater. Bulk shale and carbonate cement extracted from the shale have similar δ238U values, and are greater than δ238U values of adjacent limestone units. We suggest these trends are due to the accumulation of chemically and, more likely, biologically reduced U from anoxic to euxinic bottom water as well as the influence of diagenetic reactions between pore fluids and surrounding sediment and organic matter during diagenesis and catagenesis.

  2. Adjustment of a direct method for the determination of man body burden in Pu-239 on by X-ray detection of U-235; Mise au point d'une methode directe de determination de la charge corporelle en plutonium 239 chez l'homme par detection X de l'uranium 235

    Energy Technology Data Exchange (ETDEWEB)

    Boulay, P [Commissariat a l' Energie Atomique, Bruyeres-le-Chatel (France). Centre d' Etudes

    1968-04-01

    The use of Pu-239 on a larger scale sets a problem about the contamination measurement by aerosol at lung level. A method of direct measurement of Pu-239 lung burden is possible, thanks to the use of a large area window proportional counter. A counter of such pattern, has been especially carried out for this purpose. The adjustment of the apparatus allows an adequate sensibility to detect a contamination at the maximum permissible body burden level. Besides, a method for individual 'internal calibration', with a plutonium mock: the protactinium-233, is reported. (author) [French] L'utilisation a une echelle de plus en plus large du plutonium-239 pose un probleme de la mesure de la contamination par aerosol au niveau du poumon. Une methode de mesure directe de la charge pulmonaire en plutonium-239 est possible grace a l'utilisation d'un compteur proportionnel a fenetre de grande surface. Un compteur de ce type a specialement ete realise dans ce but. La mise au point de l'appareillage permet une sensibilite suffisante pour deceler une contamination au niveau de la Q.M.A (quantite maximale admissible). D'autre part, une methode 'd'etalonnage interne' de l'individu a l'aide d'un simulateur de plutonium, le protactinium-233, est decrite. (auteur)

  3. Choice and utilization of slightly enriched uranium fuel for high performance research reactors

    International Nuclear Information System (INIS)

    Cerles, J.M.; Schwartz, J.P.

    1978-01-01

    Problems relating to the replacement of highly enriched (90% or 93% U 235 ) uranium fuel: by moderately enriched (20% or 40% in U 235 ) metallic uranium fuel and slightly enriched (3% or 8% in U 235 ) uranium oxide fuel are discussed

  4. 78 FR 60928 - Request To Amend a License To Export High-Enriched Uranium

    Science.gov (United States)

    2013-10-02

    ... NUCLEAR REGULATORY COMMISSION Request To Amend a License To Export High-Enriched Uranium Pursuant... manufacture HEU The Netherlands. National Nuclear Security Uranium uranium (17.1 targets in France... export from 9.4 kg of U-235 contained in 10.1 kg uranium to a new cumulative total of 17.1 kg of U-235...

  5. 78 FR 72123 - Request To Amend a License to Export High-Enriched Uranium

    Science.gov (United States)

    2013-12-02

    ... NUCLEAR REGULATORY COMMISSION Request To Amend a License to Export High-Enriched Uranium Pursuant... manufacture HEU targets in Belgium. National Nuclear Security Uranium (HEU) uranium France for irradiation in... 5.8 kg of U- 235 contained in 6.2 kg uranium to a new cumulative total of 12.615 kg of U-235...

  6. A report on the first loading of ZENITH

    International Nuclear Information System (INIS)

    Absalom, R.M.; Barclay, F.R.; Cameron, I.R.; Della Loggia, V.E.; Hage, W.; Kinchin, G.H.; Sanders, J.E.; Tiren, L.I.; Wilson, D.J.

    1960-12-01

    The loading to criticality of the first core of the zero energy reactor ZENITH, and the results of the subsequent experimental programme are described. The composition of the reactor core corresponds to graphite: U235 and Th232: U235 atomic ratios of 3843 and 9.51 respectively. (author)

  7. An interim report on the second loading of Zenith. Preliminary report

    Energy Technology Data Exchange (ETDEWEB)

    Absalom, R M; Barclay, F R; Cameron, I R; Collins, A G; Della Loggia, V E; Freemantle, R G; Hage, W; Kinchin, G H; Sanders, J E; Tiren, L I; Wilson, D J

    1961-02-15

    The loading to criticality of the second core of the high-temperature zero-energy reactor ZENITH and the results of the subsequent experimental programme are described. The composition of the reactor core corresponds to graphite: U235 and TH232 : U235 atomic ratios of 2107 and 4.75 respectively.

  8. A report on the first loading of ZENITH

    Energy Technology Data Exchange (ETDEWEB)

    Absalom, R M; Barclay, F R; Cameron, I R; Della Loggia, V E; Hage, W; Kinchin, G H; Sanders, J E; Tiren, L I; Wilson, D J [Industrial Power Reactors Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

    1960-12-15

    The loading to criticality of the first core of the zero energy reactor ZENITH, and the results of the subsequent experimental programme are described. The composition of the reactor core corresponds to graphite: U235 and Th232: U235 atomic ratios of 3843 and 9.51 respectively. (author)

  9. Determination of uranium-235 by differential gamma spectrometry

    International Nuclear Information System (INIS)

    Suner, A.A.; La Gamma de Batistoni, A.M.G.; Botbol, J.

    1974-12-01

    A method for the determination of U-235 contained in solutions of uranium, by gamma spectrometry with Ge(Li) detector is described. Ra-226 is coprecipitated in BaSO 4 . The activity at 186 keV is measured, substracted by the corresponding of a standard. The detection limit is 1% of increment of U-235 over the standard. (author)

  10. Low cost solar dryer for fish

    African Journals Online (AJOL)

    Administrator

    Department of Electrical and Other Energy Sources, College of Agricultural Engineering and Technology, DBSKKV, ... The average solar energy ranged between 450 - 500 ..... classification and selection of dryers, In Handbook of industrial.

  11. 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)

  12. Absolute measurements of the alpha-gamma emitters activities by a sum-coincidence method

    International Nuclear Information System (INIS)

    Tobias, C.C.B.

    1982-01-01

    The absolute activity of U-235 contained in a UO 2 sample, using a sum-coincidence circuit which selected only the alpha particles which were simultaneous with the well known 184 Kev gamma radiation from Th-231. The alpha particles were detected by ZnS(Ag) scintillator specially designed to show its maximun efficiency for U-235 alpha particles, whereas the gamma radiation was detected by NaI(Tl) scintillation detector. The values obtained for the half-life of U-235 was compared with data from various observers using different experimental techniques. (Author) [pt

  13. Self-Consistent Criteria for Evaluation of Neutron Interaction

    International Nuclear Information System (INIS)

    Henry, H.F.; Newlon, C.E.; Knight, J.R.

    2007-01-01

    New safe interaction criteria for containers of fissionable materials handled at the Oak Ridge Gaseous Diffusion Plant have been developed on the basis of an interaction theory using the basic concepts of a safe solid angle subtended by interacting containers, and the multiplication factor as determined by two-group theory for an individually safe containers The calculated results agree satisfactorily with experimental data obtained with identical interacting units involving both cylinders and slabs containing highly enriched uranium, the core compositions of which were varied between H/U-235 atomic ratios of 44.3 and 337. The application of the derived interaction criteria to items containing material with low moderation or low U-235 assay, and to containers for which nuclear safety is dependent upon control of the U-235 mass or U-235 concentration is discussed.

  14. Analysis of uncertainty of the uranium safeguard in CORAL-I; Analisis de incertidumbre de la Salvaguardia del Uranio en Coral-I

    Energy Technology Data Exchange (ETDEWEB)

    Blazquez, J.; Barrado, J. M.

    2013-07-01

    In this work is described the experimental procedure safeguard of U-235, it is possible to measure the enrichment and self-absorption coefficient of gamma discs using a Nal scintillation detector. Details are in the uncertainty analysis.

  15. Single-level resonance parameters fit nuclear cross-sections

    Science.gov (United States)

    Drawbaugh, D. W.; Gibson, G.; Miller, M.; Page, S. L.

    1970-01-01

    Least squares analyses of experimental differential cross-section data for the U-235 nucleus have yielded single level Breit-Wigner resonance parameters that fit, simultaneously, three nuclear cross sections of capture, fission, and total.

  16. On uncertainties and fluctuations of averaged neutron cross sections in unresolved resonance energy region for 235U, 238U, 239Pu

    International Nuclear Information System (INIS)

    Van'kov, A.A.; Blokhin, A.I.; Manokhin, V.N.; Kravchenko, I.V.

    1985-01-01

    This paper analyses the reasons for the differences which exist between group-averaged evaluated cross-section data from different evaluated data files for U235, U238 and Pu239 in the unresolved resonance energy region. (author)

  17. FY08 Annual Report for Nuclear Resonance Fluorescence Imaging

    Energy Technology Data Exchange (ETDEWEB)

    Warren, Glen A.; Caggiano, Joseph A.

    2009-01-06

    FY08 annual report for project the "Nuclear Resonance Fluorescence Imaging" project. Reviews accomplishments of last 3 years, including U-235 signature search, comparison of different photon sources, and examination of NRF measurements using monochromatic photon source.

  18. The possible use of cermet fuel in the DIDO and PLUTO heavy-water research reactors

    International Nuclear Information System (INIS)

    Kennedy, T.D.A.

    1981-08-01

    As part of a study of the feasibility of using low-enrichment fuels in DIDO and PLUTO reactors the heat transfer and safety aspects involved in replacing the present U/AL-alloy (75% w/w U 235 ) fuel plates with U/AL-cermet (20% w/w U 235 ) plates, having the same outside dimensions to retain the same hydraulic characteristics, have been investigated. (U.K.)

  19. Reactivity and reaction rate studies on the fourth loading of ZENITH

    International Nuclear Information System (INIS)

    Cameron, I.R.; Freemantle, R.G.; Reed, D.L.; Wilson, D.J.

    1963-08-01

    The determination of the excess reactivity, control rod worths, prompt neutron lifetime, flux fine structure, and reaction rates of various nuclides for the fourth loading of the heated zero energy reactor ZENITH is described. The core contains 7.76 kg of U235, giving a carbon/U235 atom ratio of 7578, and forms the most dilute of the range studied. Comparisons of the experimental results with calculations using multigroup diffusion codes are presented. (author)

  20. A report on core 5 of ZENITH

    International Nuclear Information System (INIS)

    Barclay, F.R.; Cameron, I.R.; Freemantle, R.G.; Reed, D.L.; Wilson, D.J.

    1964-06-01

    The determination of the excess reactivity, control rod worths, flux fine structure, temperature coefficients, differential spectra and reaction rates of various nuclides for the fifth loading of the heated zero energy reactor ZENITH is described. The core contained 32.6 Kgm of U235, giving a carbon/U235 atomic ratio of 1072, and formed the least moderated of the range studied. Comparisons of the experimental results with calculations using multigroup diffusion codes are presented. (author)

  1. Reactivity and reaction rate studies on the fourth loading of ZENITH

    Energy Technology Data Exchange (ETDEWEB)

    Cameron, I R; Freemantle, R G; Reed, D L; Wilson, D J [General Reactor Physics Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

    1963-08-15

    The determination of the excess reactivity, control rod worths, prompt neutron lifetime, flux fine structure, and reaction rates of various nuclides for the fourth loading of the heated zero energy reactor ZENITH is described. The core contains 7.76 kg of U235, giving a carbon/U235 atom ratio of 7578, and forms the most dilute of the range studied. Comparisons of the experimental results with calculations using multigroup diffusion codes are presented. (author)

  2. 78 FR 52170 - Proposed Subsequent Arrangement

    Science.gov (United States)

    2013-08-22

    ... 700,888 g of U.S.-origin natural uranium oxide material, containing 5,005 g of the isotope U-235 (0.71% enrichment) and 18,699,112 g of U.S.-origin enriched uranium oxide material, containing 654,386 g of the isotope U-235 (less than five percent enrichment) in the form of uranium fuel fabrication scrap from...

  3. Semiempirical method to determine the uranium isotopic compositions

    International Nuclear Information System (INIS)

    Tegas Sutondo

    2008-01-01

    In a nuclear reactor design calculation, some variations of U 235 enrichment are commonly needed. This will affect the isotopic compositions of the 3 main uranium isotopes i.e. U 234 , U 235 and U 238 for the respective enrichment. Due to the limited compositions data available, it is urgent to make an approximate way that can be used to determine the compositions of the 3 isotopes, for the desired enrichments. This paper presents the theoretical background used for constructing a semi empirical formula to estimate the composition of the 3 uranium isotopes as a function of U 235 enrichment, obtained based on the measurement data available. Based on the available data, and the lack of compositions data within the enrichment range between 3.5 % and around 12 %, it is concluded that 2 separate linear equations i.e. for ≤ 3.5 % and ≥ 3.5 % might be needed for U 235 isotope. For the U 234 isotope, a polynomial equation of 4 th order is well suited to be used for the whole range of enrichment between 0.711 % and 20 %, whilst for higher enrichment (> 20 %), a power function seems to give a better approach. The composition of U 238 can then be determined from the U 235 and U 234 composition at the desired enrichment of U 235 . (author)

  4. Data base for a CANDU-PHW operating on the thorium cycle

    International Nuclear Information System (INIS)

    1979-07-01

    This report, prepared for INFCE, gives data for an extrapolated 1000 MW(e) CANDU-PHW design operating on various thorium cycles. In all these cycles thorium is the main fertile component of the fuel and all assume recycling of the uranium component. In the reference cycle, the requirements for externally supplied fissile material are met using U-235, with the feed adjusted to provide a fuel burnup of approximately 30,000 MW.d/t(U). Two versions of the reference cycle are treated. In one, the U-235 is supplied in a highly enriched form (93% U-235 in uranium); in the other, the U-235 is supplied at a lower enrichment, such that the uranium present in the feed fuel is ''denatured''. The effects of varying the fuel burnup and the recycle delay time are discussed for the reference cases. Data are also given for thorium cycles using plutonium instead of U-235 to meet requirements for externally supplied fissile material. The special case of ''self sufficient equilibrium thorium cycles'', which require no external source of fissile material for equilibrium operation, is also treated

  5. Data base for a CANDU-PHW operating on the thorium cycle

    International Nuclear Information System (INIS)

    1979-07-01

    This report, prepared for INFCE, gives data for an extrapolated 1000 MW(e) CANDU-PHW design operating on various thorium cycles. In the reference cycle, the requirements for externally supplied fissile material are met using U-235, with the feed adjusted to provide a fuel burnup of approximately 30 000 MW.d/t(U). Two versions of the reference cycle are treated. In one, the U-235 is supplied in a highly enriched form (93 percent U-235 in uranium); in the other, the U-235 is supplied at a lower enrichment, such that the uranium present in the feed fuel is 'denatured'. The effects of varying the fuel burnup and the recycle delay time are discussed. Data are also given for thorium cycles using plutonium instead of U-235 to meet requirements for externally-supplied fissile material. The special case of 'self-sufficient equilibrium thorium cycles', which require no external source of fissile material for equilibrium operation, is also treated. (author)

  6. Isotope separation utilizing Zeeman compensated magnetic extraction

    International Nuclear Information System (INIS)

    Forsen, H.K.

    1978-01-01

    A vapor flow of elemental uranium is directed into a region where narrow band, tuned laser radiation is repeatedly applied to provide at least two energy step selective ionization of the U 235 isotope in the vapor flow. A magnetic field is applied in the region of the ionized U 235 which creates a Lorentz force on the moving ions directing them toward one of a plurality of collection plates placed generally parallel to the vapor flow to permit collection of the U 235 particles in substantially enriched proportions as compared to the concentration in the vapor flow generally. To prevent a broadening of the absorption lines for both the U 235 and U 238 isotopes in the vapor flow from the applied magnetic field and thus prevent substantial reduction in the selectivity of the excitation and ionization, the magnetic field is preferably applied in a time varying magnitude which is phased with respect to the repetitive application of laser radiation to provide a relatively low field strength and corresponding small Zeeman splitting during selective excitation and ionization of the U 235 particles

  7. Anatomy of a field trial: Wood-based biochar and compost influences a Pacific Northwest soil

    Science.gov (United States)

    Biochar land application research in elevated rainfall areas (980 millimeters of annual rainfall) of the U.S. Pacific Northwest is lacking. A proof-of-concept field study examined the effects of spruce-pine-fir wood chip biochar (slow pyrolysis; 450-500 degrees Celsius; 35 megagrams per hectare), d...

  8. Kinetics and mechanisms of the oxide film growth on the surface of α-Fe in transitional domains

    International Nuclear Information System (INIS)

    Mukhambetov, D.G.; Berber, N.N.; Kargin, D.B.; Chalaya, O.V.

    2003-01-01

    The object of this work was to study the kinetics of the α-Fe surface oxidation with prevailing cubic texture at temperatures of 450-500 deg. C. The basic conformity to natural laws and mechanisms of the two-phase thin oxide films grows are determined. (author)

  9. Modern Electromagnetic Scattering

    Science.gov (United States)

    2013-08-10

    homeworks. The latter was an awesome preparation for the frustration that research brings, and the former turned sorrow into blissfulness ; • Dr...ra n ru n tim e 100 150 200 250 300 350 400 450 500 0 10 20 30 40 50 0 5 10 15 20 25 30 Figure 2.2: The ratio of total computational time required to

  10. COMPARATIVE ANALYSIS OF STRUCTURAL CHANGES IN U-MO DISPERSED FUEL OF FULL-SIZE FUEL ELEMENTS AND MINI-RODS IRRADIATED IN THE MIR REACTOR

    Directory of Open Access Journals (Sweden)

    ALEKSEY. L. IZHUTOV

    2013-12-01

    The full-size fuel rods were irradiated up to an average burnup of ∼ 60%235U; the mini-rods were irradiated to an average burnup of ∼ 85%235U. The presented data show a significant increase of the void fraction in the U-Mo alloy as the U-235 burnup rises from ∼ 40% up to ∼ 85%. The effect of irradiation test conditions and U-235 burnup were analyzed with regard to the formation of an interaction layer between the matrix and fuel particles as well as generation of porosity in the U-Mo alloy. Shown here are changes in distribution of U fission products as the U-235 burnup increases from ∼ 40% up to ∼ 85%.

  11. Fission physics experiments at the time-of-flight spectrometer GNEIS in Gatchina (PNPI)

    International Nuclear Information System (INIS)

    Shcherbakov, O.A.

    1994-01-01

    The outline of and fission physics experiments at the Gatchina neutron spectrometer GNEIS based on the 1 GeV PNPI proton synchrotron are presented. The prefission gamma-ray spectrum of the (n, gamma f) reaction were investigated. The capture gamma-ray spectra for 721.6 eV and 1211.4 eV resonances in U-238 were measured and the nature of the 721.6 eV resonance in U-238 were examined. The forward-backward asymmetry in slow neutron fission of U-235 and energy dependence of the forward-backward and instrumental asymmetry coefficients were obtained. Fission cross section ratios for Th-232 to U-235 and for U-238 to U-235 in the energy range up to 200 MeV were measured. The results of the cross section ratios agreed well with those of Behrens et al. and Difilippo et al. (T.H.)

  12. Experience with the RE fuel transition at the Studsvik R2 reactor

    International Nuclear Information System (INIS)

    Pazsit, I.; Saltvedt, K.

    1991-01-01

    Irradiation of 7 LEU fuel elements is underway in the Studsvik R2 reactor. Four of these have 490 g U-235, and three 320 g U-235 loading, and the enrichment is 19.7% for all of them. The irradiation of LEU fuel started in 1987. The heavier elements have burnup figures 67% (CERCA), 50% (B and W), 47% (NUKEM) and 19% (B and W). One of the lighter elements has reached a burnup of 65%. To support the whole-core conversion process, reactor physical calculations were performed to see if a one-step conversion is possible with a suitable fuel management strategy such that all HEU fuel is burned up. The calculations show that it is possible to perform such a conversion with fuel elements containing 400 g U-235. (orig.)

  13. Burnup credit applications in a high-capacity truck cask

    International Nuclear Information System (INIS)

    Boshoven, J.K.

    1993-01-01

    The use of burnup credit in the criticality safety analysis of the GA-4 Cask increases the cask's capacity from three spent fuel assemblies to four, resulting in reduced public and occupational risk and reduced life cycle costs. GA's criticality calculations for burnup credit, including the associated uncertainties and analytical bias, establish the minimum burnup required as a function of initial enrichment to maintain K eff ≤ 0.95 under any conceivable condition. The minimum burnup requirement as a function of initial enrichment has been determined to be 15,000 MWd/MTU for 3.5 wt% U-235 fuel, 20,000 MWd/MTU for 4.0 wt% U-235 fuel and 25,000 MWd/MTU for 4.5 wt% U-235 fuel. The minimum burnup requirement as a function of enrichment is well below the typical burnup levels seen in the current and projected spent fuel inventory. (J.P.N.)

  14. NEACRP thermal fission product benchmark

    International Nuclear Information System (INIS)

    Halsall, M.J.; Taubman, C.J.

    1989-09-01

    The objective of the thermal fission product benchmark was to compare the range of fission product data in use at the present time. A simple homogeneous problem was set with 200 atoms H/1 atom U235, to be burnt up to 1000 days and then decay for 1000 days. The problem was repeated with 200 atoms H/1 atom Pu239, 20 atoms H/1 atom U235 and 20 atoms H/1 atom Pu239. There were ten participants and the submissions received are detailed in this report. (author)

  15. ENDF/B-5 Standards Data Library (including modifications made in 1986). Summary of contents and documentation

    International Nuclear Information System (INIS)

    DayDay, N.; Lemmel, H.D.

    1986-01-01

    This document summarizes the contents and documentation of the ENDF/B-5 Standards Data Library (EN5-ST) released in September 1979. The library contains complete evaluations for all significant neutron reactions in the energy range 10 -5 eV to 20 MeV for H-1, He-3, Li-6, B-10, C-12, Au-197 and U-235 isotopes. In 1986 the files for C-12, Au-197 and U-235 were slightly modified. The entire library or selective retrievals from it can be obtained free of charge from the IAEA Nuclear Data Section. (author)

  16. Preliminary study on direct recycling of spent PWR fuel in PWR system

    International Nuclear Information System (INIS)

    Waris, Abdul; Nuha; Novitriana; Kurniadi, Rizal; Su'ud, Zaki

    2012-01-01

    Preliminary study on direct recycling of PWR spent fuel to support SUPEL (Straight Utilization of sPEnt LWR fuel in LWR system) scenario has been conducted. Several spent PWR fuel compositions in loaded PWR fuel has been evaluated to obtain the criticality of reactor. The reactor can achieve it criticality for U-235 enrichment in the loaded fresh fuel is at least 4.0 a% with the minimum fraction of the spent fuel in the core is 15.0 %. The neutron spectra become harder with the escalating of U-235 enrichment in the loaded fresh fuel as well as the amount of the spent fuel in the core.

  17. Procedure and device for separating isotopes of high mass

    International Nuclear Information System (INIS)

    Eerkens, J.W.

    1977-01-01

    The invention refers to isotope separation and to selectively photon-induced energy transfer from an isotope molecule containing the isotope to be separated as well as to a chemical reaction with a reactive agent in order to produce a chemical compound containing atoms of the desired isotope. For example, in the most preferable form of the invention, gaseous UF 6 is contained in a mixture of U 235 F 6 and U 238 F 6 molecules in a reaction chamber. A chemically reactive substance, which for U 235 separation may be gaseous HCl according to the invention, is also introduced into the reaction chamber. (HK) [de

  18. Long-term management and use of depleted uranium

    International Nuclear Information System (INIS)

    Max, A.

    2001-01-01

    The products resulting from the process of enrichment of natural uranium, or reprocessed uranium, are enriched uranium products as the light fraction and depleted uranium (uranium tails) as the heavy fraction. If the source material is natural uranium, the mass ratios of uranium products and uranium tails can be derived relatively easily from the required enrichment level of the uranium product (product assay (% of U-235)) and the selected depletion level of the uranium tails (tails assay (% of U-235)). The paper discusses among other aspects the dependence of the tails mass on the required enrichment level of the relevant uranium product, for various tails assays. (orig./CB) [de

  19. Neutronics of a mixed-flow gas-core reactor

    International Nuclear Information System (INIS)

    Soran, P.D.; Hansen, G.E.

    1977-11-01

    The study was made to investigate the neutronic feasibility of a mixed-flow gas-core reactor. Three reactor concepts were studied: four- and seven-cell radial reactors and a seven-cell scallop reactor. The reactors were fueled with UF 6 (either U-233 or U-235) and various parameters were varied. A four-cell reactor is not practical nor is the U-235 fueled seven-cell radial reactor; however, the 7-cell U-233 radial and scallop reactors can satisfy all design criteria. The mixed flow gas core reactor is a very attractive reactor concept and warrants further investigation

  20. Measurement of Alpha Emitters Concentration in Imported Cigarettes

    International Nuclear Information System (INIS)

    Nasser Allah, Z.K.; Musa, W.A.; AL-Rawi, A.A.S.

    2011-01-01

    The aime of this study was to measured the alpha emitters concentration of (15) different kinds of imported cigarettes. the nuclear reaction used U-235(n, f) obtained by the bombardment of U-235 with thermal neutrons from (Am B e)neutron source with thermal flux of(5*10 3 n.cm -2 .s -1 ). The Results obtained showed the values of the Uranium concentration, and varies from (0.041 ppm) in five stares kind to (2.374ppm) in Machbeth (chocolate) 100's kind. All the result obtained are within the limit levels as given by UNSCAR data

  1. ENDF/B-5 Standards Data Library (including modifications made in 1986). Summary of contents and documentation

    Energy Technology Data Exchange (ETDEWEB)

    DayDay, N; Lemmel, H D

    1986-05-01

    This document summarizes the contents and documentation of the ENDF/B-5 Standards Data Library (EN5-ST) released in September 1979. The library contains complete evaluations for all significant neutron reactions in the energy range 10{sup -5}eV to 20 MeV for H-1, He-3, Li-6, B-10, C-12, Au-197 and U-235 isotopes. In 1986 the files for C-12, Au-197 and U-235 were slightly modified. The entire library or selective retrievals from it can be obtained free of charge from the IAEA Nuclear Data Section. (author) Refs, figs, tabs

  2. Dosimetry results for Big Ten and related benchmarks

    International Nuclear Information System (INIS)

    Hansen, G.E.; Gilliam, D.M.

    1980-01-01

    Measured average reaction cross sections for the Big Ten central flux spectrum are given together with calculated values based on the U.S. Evaluated Nuclear Data File ENDF/B-IV. Central reactivity coefficients for 233 U, 235 U, 239 Pu, 6 Li and 10 B are given to check consistency of bias between measured and calculated reaction cross sections for these isotopes. Spectral indexes for the Los Alamos 233 U, 235 U and 239 Pu metal critical assemblies are updated, utilizing the Big Ten measurements and interassembly calibrations, and their implications for inelastic scattering are reiterated

  3. The moderization of the Leopard library

    International Nuclear Information System (INIS)

    Claro, L.H.; Cunha Menezes Filho, A. da.

    1983-07-01

    The LEOPARD library is updated and tested for typical PWR unit cells with enrichment ranging from 1.0 to 4.%(W/o) and H 2 O:U ratio varying from 1.0 to 10.0. A reasonably good agreement with experimental values for some spectral indices is obtained if the fission cross section of U-235 is reduced by .6% in the thermal range and by 20% in the epithermal range; the epithermal capture cross section for U-238 is increased by about 20% and the number of neutrons per fission in the thermal range of U-235 is increased by .8%. (Author) [pt

  4. Prevention of criticality accidents. Fuel elements storage; Prevencion de accidentes de criticidad. Almacenamiento de elementos combustibles

    Energy Technology Data Exchange (ETDEWEB)

    Canavese, S I; Capadona, N M

    1991-12-31

    Before the need to store fuel elements of the plate type MTR (Materials Testing Reactors), produced with enriched uranium at 20% in U235 for research reactors, it requires the design of a deposit for this purpose, which will give intrinsic security at a great extent and no complaints regarding its construction, is required. (Author). [Espanol] Partiendo de la necesidad de almacenar elementos combustibles tipo placa MTR (Materials Testing Reactors), producidos con uranio enriquecido al 20% en U235 para reactores de investigacion, se requiere el diseno de un deposito para tal fin que brinde esencialmente un alto grado de seguridad intrinseca y que no ofrezca complicaciones en cuanto a su construccion. (Autor).

  5. Prompt neutron energy spectrum for the spontaneous fission of Cf-252

    International Nuclear Information System (INIS)

    Blinov, M.V.; Boykov, G.S.; Vitenko, V.A.

    1985-06-01

    The prompt neutron spectrum for the spontaneous fission of Cf-252 has been measured in 0.01-10 MeV region by the time-of-flight technique using a fast ionization chamber with U-235 layers as the neutron detector. Numerical data for the spectrum are presented, with an error file. (author)

  6. Environmental site description for a Uranium Atomic Vapor Laser Isotope Separation (U-AVLIS) production plant at the Paducah Gaseous Diffusion Plant site

    Energy Technology Data Exchange (ETDEWEB)

    Marmer, G.J.; Dunn, C.P.; Moeller, K.L.; Pfingston, J.M.; Policastro, A.J.; Yuen, C.R.; Cleland, J.H. (ed.)

    1991-09-01

    Uranium enrichment in the United States has utilized a diffusion process to preferentially enrich the U-235 isotope in the uranium product. The U-AVLIS process is based on electrostatic extraction of photoionized U-235 atoms from an atomic vapor stream created by electron-beam vaporization of uranium metal alloy. The U-235 atoms are ionized when precisely tuned laser light -- of appropriate power, spectral, and temporal characteristics -- illuminates the uranium vapor and selectively photoionizes the U-235 isotope. A programmatic document for use in screening DOE site to locate a U-AVLIS production plant was developed and implemented in two parts. The first part consisted of a series of screening analyses, based on exclusionary and other criteria, that identified a reasonable number of candidate sites. These sites were subjected to a more rigorous and detailed comparative analysis for the purpose of developing a short list of reasonable alternative sites for later environmental examination. This environmental site description (ESD) provides a detailed description of the PGDP site and vicinity suitable for use in an environmental impact statement (EIS). The report is based on existing literature, data collected at the site, and information collected by Argonne National Laboratory (ANL) staff during a site visit. 65 refs., 15 tabs.

  7. Determination of the isotopic composition of natural and slightly enriched uranium by alpha-spectrometry

    DEFF Research Database (Denmark)

    Kunzendorf, Helmar

    1968-01-01

    Determinations of the isotope contents of 238U, 235U and 234U in five uranium samples containing 0–5 at% 235U were carried out on the basis of a least-squares fit of the α-spectra from the samples, measured with a semiconductor detector, to the theoretically expected α-spectra. With a simple source...

  8. Environmental site description for a Uranium Atomic Vapor Laser Isotope Separation (U-AVLIS) production plant at the Paducah Gaseous Diffusion Plant site

    International Nuclear Information System (INIS)

    Marmer, G.J.; Dunn, C.P.; Moeller, K.L.; Pfingston, J.M.; Policastro, A.J.; Yuen, C.R.; Cleland, J.H.

    1991-09-01

    Uranium enrichment in the United States has utilized a diffusion process to preferentially enrich the U-235 isotope in the uranium product. The U-AVLIS process is based on electrostatic extraction of photoionized U-235 atoms from an atomic vapor stream created by electron-beam vaporization of uranium metal alloy. The U-235 atoms are ionized when precisely tuned laser light -- of appropriate power, spectral, and temporal characteristics -- illuminates the uranium vapor and selectively photoionizes the U-235 isotope. A programmatic document for use in screening DOE site to locate a U-AVLIS production plant was developed and implemented in two parts. The first part consisted of a series of screening analyses, based on exclusionary and other criteria, that identified a reasonable number of candidate sites. These sites were subjected to a more rigorous and detailed comparative analysis for the purpose of developing a short list of reasonable alternative sites for later environmental examination. This environmental site description (ESD) provides a detailed description of the PGDP site and vicinity suitable for use in an environmental impact statement (EIS). The report is based on existing literature, data collected at the site, and information collected by Argonne National Laboratory (ANL) staff during a site visit. 65 refs., 15 tabs

  9. 75 FR 6223 - Application For a License To Export High-Enriched Uranium

    Science.gov (United States)

    2010-02-08

    ... NUCLEAR REGULATORY COMMISSION Application For a License To Export High-Enriched Uranium Pursuant to 10 CFR 110.70(c) ``Public notice of receipt of an application,'' please take notice that the..., Uranium (93.35%). uranium (16.3 targets for December 28, 2009, XSNM3623, kilograms U-235). irradiation in...

  10. 75 FR 7525 - Application for a License To Export High-Enriched Uranium

    Science.gov (United States)

    2010-02-19

    ... NUCLEAR REGULATORY COMMISSION Application for a License To Export High-Enriched Uranium Pursuant to 10 CFR 110.70(c) ``Public notice of receipt of an application,'' please take notice that the..., February 2, Uranium (93.35%). uranium (87.3 elements in 2010, February 2, 2010, kilograms U-235). France...

  11. An ideal cascade for uranium 235 enrichment by centrifuge jet nozzle process

    International Nuclear Information System (INIS)

    Santos, E.C. dos.

    1981-01-01

    The design of an ideal cascade for the process of isotope separation by centrifugation for the U 235 enrichment, is presented. A selection of building materials used in fabrication of isotope separation plants, showing the importance of aluminium, due the bauxite mines in Northern Brazil, is done. (M.C.K.) [pt

  12. Temperature coefficients of reactivity in the fourth loading of ZENITH

    International Nuclear Information System (INIS)

    Caro Manso, R.; Freemantle, R.G.; Rogers, J.D.

    1962-10-01

    Measurements have been made of the temperature coefficients of reactivity associated with the core plus end reflectors and the side reflector of the fourth core loading of ZENITH, which had a carbon/U235 atomic ratio of 7788 and no other absorber. (author)

  13. 10 CFR Appendix B to Part 30 - Quantities 1 of Licensed Material Requiring Labeling

    Science.gov (United States)

    2010-01-01

    ... .1 1 Based on alpha disintegration rate of Th-232, Th-230 and their daughter products. 2 Based on alpha disintegration rate of U-238, U-234, and U-235. Note: For purposes of § 20.303, where there is...

  14. The study of prompt and delayed muon induced fission III. The ratios of prompt to delayed fission yields

    NARCIS (Netherlands)

    Rösel, Ch.; Hänscheid, H.; Hartfiel, J.; Mutius, von R.; Achard van Enschut, d' J.F.M.; David, P; Janszen, H.; Johansson, T.; Konijn, J.; Krogulski, T.; Laat, de C.T.A.M.; Paganetti, H.; Petitjean, C.; Polikanov, S.M.; Reist, H.W.; Risse, F.; Schaller, L.A.; Schrieder, W.; Sinha, A.K.; Taal, A.; Theobald, J.P.; Tibell, G.; Trautmann, N.

    1993-01-01

    The ratios of prompt to delayed fission yields for the isotopes U-233, U-234, U-235, U-236, U-238, Np-237, Pu-242, and Pu-244 and the fission probabilities relative to each other have been investigated experimentally. Using the value of the total fission probability for Np-237 the absolute

  15. Temperature coefficients of reactivity in the fourth loading of ZENITH

    Energy Technology Data Exchange (ETDEWEB)

    Caro Manso, R; Freemantle, R G; Rogers, J D [Graphite Reactor Physics Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

    1962-10-15

    Measurements have been made of the temperature coefficients of reactivity associated with the core plus end reflectors and the side reflector of the fourth core loading of ZENITH, which had a carbon/U235 atomic ratio of 7788 and no other absorber. (author)

  16. Atomic and molecular resonance ionization

    International Nuclear Information System (INIS)

    Botter, R.; Petit, A.

    1990-01-01

    Published in summary form only the paper recalls the principle of resonance photoionization, transition probability, selectivity and critical parameters. Examples of applications are briefly treated: Trace analysis by resonance ionization mass spectroscopy for detection of Fe in Zr F 4 for fabrication of optical fibers and laser isotopic separation of U 235 and Gd 157 [fr

  17. Uranium isotopic determination by alpha spectroscopy

    International Nuclear Information System (INIS)

    Acena, M.; Garcoa-Torano, E.

    1979-01-01

    A method for alpha-spectrometry determinations of uranium isotopes, using surface barrier detectors, is described. This method is based in the shape similarity of the most intense line groups for the nuclides 234 U, 235 U, 236 U and 238 U. The method yields analytical results sufficiently accurate in samples with 235 U contents lower than 25% in atoms. (author)

  18. Role of thorium in the industry advantage of atomic energy

    International Nuclear Information System (INIS)

    Souza Santos, M.D. de; Goldemberg, J.; Lopes, J.L.

    1985-01-01

    Based in the utilization of others fossil substances, such as plutonium and uranium 233, produzed through the thorium and natural uranium (238), it is discussed the relative merits of alternative processes: to produce U233 on Pu 239 to substitute the initial load of U235. (M.C.K.) [pt

  19. Generation of one energy group cross section library with MC2 computer code

    International Nuclear Information System (INIS)

    Cunha Menezes Filho, A. da; Souza, A.L. de.

    1982-01-01

    One group temperature dependent cross sections are generated via MC 2 for Pu-242, Ni-58, Fe-56, U-235, U-238, Pu-239, Pu-240, Pu-241, Be-9 e Th-232. The influence of the buckling and the weighting functions is studied throught calculations of an important integral parameter: the critical radius. (author) [pt

  20. Extension of the Th-232 burnup chain in the WIMSD/4 program library

    International Nuclear Information System (INIS)

    Caldeira, A.D.

    1991-07-01

    The Th-232 burnup chain was extended through U-236, in the WIMSD/4 program library. The evolution of the values of k i nf and U-235 number density, as function of time, for the modified TRX1 problem, calculated with the new library, shows an improvement in the results when compared with LEOPARD program. (author)

  1. Analysis of benchmark critical experiments with ENDF/B-VI data sets

    International Nuclear Information System (INIS)

    Hardy, J. Jr.; Kahler, A.C.

    1991-01-01

    Several clean critical experiments were analyzed with ENDF/B-VI data to assess the adequacy of the data for U 235 , U 238 and oxygen. These experiments were (1) a set of homogeneous U 235 -H 2 O assemblies spanning a wide range of hydrogen/uranium ratio, and (2) TRX-1, a simple, H 2 O-moderated Bettis lattice of slightly-enriched uranium metal rods. The analyses used the Monte Carlo program RCP01, with explicit three-dimensional geometry and detailed representation of cross sections. For the homogeneous criticals, calculated k crit values for large, thermal assemblies show good agreement with experiment. This supports the evaluated thermal criticality parameters for U 235 . However, for assemblies with smaller H/U ratios, k crit values increase significantly with increasing leakage and flux-spectrum hardness. These trends suggest that leakage is underpredicted and that the resonance eta of the ENDF/B-VI U 235 is too large. For TRX-1, reasonably good agreement is found with measured lattice parameters (reaction-rate ratios). Of primary interest is rho28, the ratio of above-thermal to thermal U 238 capture. Calculated rho28 is 2.3 (± 1.7) % above measurement, suggesting that U 238 resonance capture remains slightly overpredicted with ENDF/B-VI. However, agreement is better than observed with earlier versions of ENDF/B

  2. Quantities of uranium-235 buried in disposal boxes, 1985--1991

    International Nuclear Information System (INIS)

    Cook, J.R.

    1991-01-01

    IWT was asked by J. R. Schornhorst of NPSR to determine the distribution of the quantity of enriched uranium per disposal box (B-25) of the years 1985--1991 to provide input to an uptake of the E Area Safety Analysis. This information was considered important since the issue of criticality is an important concern in safety analyses. Information found in the COBRA data base shows no disposal containers exceeded 100 grams of U-235. The COBRA data base was queried in a two-step process. First a short program in the NATURAL language was used to retrieve all records beginning with January 1983 having a Burial Code of less than 4, indicating low-level waste disposed in trenches. These records were then passed to a temporary storage file and read into a program written in Statistical Analysis System (SAS) language. SAS was used to eliminate waste from the Naval Fuel Facility, which will not operate in the future, and to sort the records in order of increasing amounts of U-235. The SAS procedure FREQ was then used to produce a cumulative frequency distribution of grams of U-235. A total of 53,198 packages were disposed of during this time period, 277 of which contained U-235. The programs used and resulting output are attached as Appendix I

  3. Development of long-life neutron detectors for the prototype heavy water reactor 'Fugen'

    International Nuclear Information System (INIS)

    Ohteru, Shigeru; Shirayama, Shimpey.

    1981-01-01

    The development of long-life neutron detectors as the flux monitors for the prototype heavy water reactor has been made. Three kinds of neutron monitors, namely start-up monitor (SUM), power up monitor (PUM) and local power monitor (LPM), are provided. The LPM consists of 4 ion chamber type neutron detectors and a guide tube of power calibration monitor (PCM). This is useful for reactor control and fuel soundness monitor. The improvement of the neutron detectors was made for the operation under high neutron flux and gamma-ray heating. For the long-life operation, U-234 was mixed into U-235 for the conversion in the detectors. The ratio of U-234 to U-235 is 3 to 1. The PCM is also an ion chamber type detector with U-235. The mixing ratio of U-234 to U-235 was determined by a test with the JMTR. The characteristic performance was also investigated by the JMTR. After the completion of Fugen, various tests on the long-life detectors were performed with Fugen. It was hard to test the output linearity of the detectors with a large scale reactor. Therefore, it was tested that the operation range of the detectors is within the linear region of detector output. The voltage-current characteristics and the correlation of output current and saturation current were measured. The variation of the neutron sensitivity of the detectors with the cumulative dose was also studied. (Kato, T.)

  4. Neutrons as Party Animals: An Analogy for Understanding Heavy-Element Fissility

    Science.gov (United States)

    Reed, B. Cameron

    2012-01-01

    I teach a general education class on the history of nuclear physics and the Manhattan Project. About halfway through the course we come to the discovery of fission and Niels Bohr's insight that it is the rare isotope of uranium, U-235, which fissions under slow-neutron bombardment as opposed to the much more common U-238 isotope. As an…

  5. Utilization of thorium in PWR type reactors

    International Nuclear Information System (INIS)

    Correa, F.

    1977-01-01

    Uranium 235 consumption is comparatively evaluated with thorium cycle for a PWR type reactor. Modifications are only made in fuels components. U-235 consumption is pratically unchanged in both cycles. Some good results are promised to the mixed U-238/Th-232 fuel cycle in 1/1 proportion [pt

  6. ICPP digital wide area portal monitor

    International Nuclear Information System (INIS)

    Nichols, C.E.

    1978-01-01

    A portal montior is described which is used to check personnel for contamination. The monitor is extremely sensitive. In addition to personnel contamination control, it serves as an extremely effective SNM detector capable of detecting U-235 in half the quantities specified in Federal Standards

  7. Mo-99 production by fission and future projections

    International Nuclear Information System (INIS)

    Carranza, E.C.; Novello, A.; Bronca, M.; Cestau, D.; Bavaro, R.; Centurion, R.; Bravo, C.; Bronca, P.; Gualda, E.; Fraguas, F.; Giomi, A.; Ivaldi, L.

    2012-01-01

    Description of the I-131 and Mo-99 production process: The process starts with the irradiation of uranium-aluminum mini plates in the RA-3, Argentinean Reactor No.3, Ezeiza Atomic Center. In a nuclear reactor there is a constant flow of neutrons and when a neutron with proper energy impacts on a nucleus of U-235, it is absorbed at the same time generate an unstable configuration nuclear. For this reason, the nucleus formed is fission, getting two different atoms. Approximately 6% of the fissions produce Mo-99 and 3% produce I-131; the percentage remaining corresponds to formation of atoms without interest for use in medicine. In conclusion, the objective of the process developed in the Fission Plant, is starting from uranium mini plates, separate the Mo-99 and I-131 generated, the remaining elements formed. - Evolution of Mo-99 Production in the last 10 years: The Fission Mo-99 Plant Production begins routine production of Mo-99 in 1985, using targets made of uranium enriched at 90% U-235. In the 1990s, global concern regarding the use of highly enriched uranium, due to non-proliferation issues, caused the interruption of supply of nuclear material (HEU enriched at 90% of U-235). Following this, Argentina developed target based on low-enriched uranium (less than 20% U-235), becoming in 2002 the first country in the world to produce Mo-99 with LEU targets. From 2002 to date, the activity produced of Mo-99 has been tripled annually (author)

  8. Natural radionuclides in the environment and problems of uranium mining

    International Nuclear Information System (INIS)

    Bowie, S.H.U.

    1981-01-01

    The subject is discussed under the headings: introduction (U-238, U-235, Th-232, K-40, and their decay products); distribution of radionuclides; α, β and γ radiation; uranium in rocks; uranium in soil and water; uranium mining (hazards of uranium and radon during mining and in tailings); assessment of risk. (U.K.)

  9. INDL/A - Suppl. 86/5. IAEA Nuclear Data Library for evaluated neutron reaction data of Actinides

    Energy Technology Data Exchange (ETDEWEB)

    Goulo, V G; Lemmel, H D; McLaughlin, P M [eds.

    1986-05-01

    This supplement to INDL/A contains five evaluations by V.A. Konshin on U-235 and Pu-239, 240, 241, 242. They are available upon request on magnetic tape, free of charge, from the IAEA Nuclear Data Section. (author) Refs, graphs

  10. Data base for a CANDU-PHW operating on a once-through, slightly enriched uranium cycle (AECL-6594)

    International Nuclear Information System (INIS)

    1979-07-01

    This report, prepared for INFCE, gives data for an extrapolated 1000 MW(e) CANDU-PHW design operating on a once-through fuel cycle with a feed fuel of slightly enriched uranium - 1.2 weight % U-235 in uranium. The effects of varying fuel enrichment, maximum channel power, and economic parameters are also discussed

  11. Transient neutrons flux behaviour in a spherical reactor core

    International Nuclear Information System (INIS)

    Souza, A.W.A. de.

    1978-11-01

    This work studies the transient neutron flux in a fast reactor of spherical geometry. The burning of U 235 nuclei is equated and two kinds of reflector were studied. The numeric solutions are then compared with the results for those reflectors. (author) [pt

  12. 77 FR 35366 - Proposed Subsequent Arrangement

    Science.gov (United States)

    2012-06-13

    ... arrangement concerns the retransfer of 6,672,212 g of U.S.-origin enriched uranium fuel fabrications scrap, containing 233,977 g of the isotope U-235 (less than five percent enrichment), from Nuclear Fuel Industries... material, which is currently located at Nuclear Fuels Industries, Ltd. in Japan, will be transferred to...

  13. 76 FR 2892 - Proposed Subsequent Arrangement

    Science.gov (United States)

    2011-01-18

    ... concerns the retransfer of 2,536 g of U.S.-origin uranium (11 g U-235) and 27 g of plutonium contained in eight spent fuel rod segments from Nippon Nuclear Fuel Development Co., Ltd in Oarai-machi, Ibaraki-ken... Nippon Nuclear Fuel Development Co., Ltd, will be transferred to Studsvik Nuclear AB for: (a) Pre...

  14. Prevention of criticality accidents. Fuel elements storage

    International Nuclear Information System (INIS)

    Canavese, S.I.; Capadona, N.M.

    1990-01-01

    Before the need to store fuel elements of the plate type MTR (Materials Testing Reactors), produced with enriched uranium at 20% in U235 for research reactors, it requires the design of a deposit for this purpose, which will give intrinsic security at a great extent and no complaints regarding its construction, is required. (Author) [es

  15. INDL/A - Suppl. 86/5. IAEA Nuclear Data Library for evaluated neutron reaction data of Actinides

    International Nuclear Information System (INIS)

    Goulo, V.G.; Lemmel, H.D.; McLaughlin, P.M.

    1986-01-01

    This supplement to INDL/A contains five evaluations by V.A. Konshin on U-235 and Pu-239, 240, 241, 242. They are available upon request on magnetic tape, free of charge, from the IAEA Nuclear Data Section. (author)

  16. Cost analysis and economic comparison for alternative fuel cycles in the heavy water cooled canadian reactor (CANDU)

    International Nuclear Information System (INIS)

    Yilmaz, S.

    2000-01-01

    Three main options in a CANDU fuel cycle involve use of: (1) natural uranium (0.711 weight percent U-235) fuel, (2) slightly enriched uranium (1.2 weight percent U-235) fuel, and (3) recovered uranium (0.83 weight percent U-235) fuel from light water reactor spent fuel. ORIGEN-2 computer code was used to identify composition of the spent fuel for each option, including the standard LWR fuel (3.3 weight percent U-235). Uranium and plutonium credit calculations were performed using ORIGEN-2 output. WIMSD-5 computer code was used to determine maximum discharge burnup values for each case. For the 3 cycles selected (natural uranium, slightly enriched uranium, recovered uranium), levelized fuel cycle cost calculations are performed over the reactor lifetime of 40 years, using unit process costs obtained from literature. Components of the fuel cycle costs are U purchase, conversion, enrichment, fabrication, SF storage, SF disposal, and reprocessing where applicable. Cost parameters whose effects on the fuel cycle cost are to be investigated are escalation ratio, discount rate and SF storage time. Cost estimations were carried out using specially developed computer programs. Share of each cost component on the total cost was determined and sensitivity analysis was performed in order to show how a change in a main cost component affects the fuel cycle cost. The main objective of this study has been to find out the most economical option for CANDU fuel cycle by changing unit prices and cost parameters

  17. CONHOR. Code system for determination of power distribution and burnup for the HOR reactor. Version 1.0.. User`s manual

    Energy Technology Data Exchange (ETDEWEB)

    Serov, I V; Hoogenboom, J E

    1993-07-01

    The main calculational tool is the CITATION code. CITATION is used for both static and burnup calculations. The pointwise flux density and power distributions obtained from these calculations are used to obtain the values of the desired quantities at the beginning of a burnup cycle. To obtain the most trustful values of the desired quantities CONHOR employs experimental information together with the CITATION calculated flux distributions. Axially averaged foil activation rates are obtained based on both CITATION pointwise flux density distributions and measured foil activity counts. These two sets of activation rates are called the distributions of auxiliary quantities and are compared with each other in order to pick up the corrections to the U-235 number densities in fuel containing elements. The methodical corrections to the calculational auxiliary quantities are obtained on this basis as well. They are used to obtain the methodical corrections to the desired quantities. The corrected desired quantities are the recommended ones. The correction procedure requires the knowledge of the sensitivity coefficients of the average foil activation rates with respect to the U-235 number densities (through the text of this manual U-235 is denoted also and especially in the input-output description sections as a BUrning-COrrected material, or `BuCo` material). These sensitivity coefficients are calculated by the CONHOR SENS module. CITATION is employed to perform the calculations with perturbed values of U-235 number densities. Burnup calculations can be performed being based on either corrected or uncorrected U-235 number densities. Through the text of this manual XXXX means a 4-symbol identification of the burnup cycle to be studied. XX-1 and XX+1 mean correspondingly the previous and the following cycles. (orig./HP).

  18. CONHOR. Code system for determination of power distribution and burnup for the HOR reactor. Version 1.0.. User's manual

    International Nuclear Information System (INIS)

    Serov, I.V.; Hoogenboom, J.E.

    1993-07-01

    The main calculational tool is the CITATION code. CITATION is used for both static and burnup calculations. The pointwise flux density and power distributions obtained from these calculations are used to obtain the values of the desired quantities at the beginning of a burnup cycle. To obtain the most trustful values of the desired quantities CONHOR employs experimental information together with the CITATION calculated flux distributions. Axially averaged foil activation rates are obtained based on both CITATION pointwise flux density distributions and measured foil activity counts. These two sets of activation rates are called the distributions of auxiliary quantities and are compared with each other in order to pick up the corrections to the U-235 number densities in fuel containing elements. The methodical corrections to the calculational auxiliary quantities are obtained on this basis as well. They are used to obtain the methodical corrections to the desired quantities. The corrected desired quantities are the recommended ones. The correction procedure requires the knowledge of the sensitivity coefficients of the average foil activation rates with respect to the U-235 number densities (through the text of this manual U-235 is denoted also and especially in the input-output description sections as a BUrning-COrrected material, or 'BuCo' material). These sensitivity coefficients are calculated by the CONHOR SENS module. CITATION is employed to perform the calculations with perturbed values of U-235 number densities. Burnup calculations can be performed being based on either corrected or uncorrected U-235 number densities. Through the text of this manual XXXX means a 4-symbol identification of the burnup cycle to be studied. XX-1 and XX+1 mean correspondingly the previous and the following cycles. (orig./HP)

  19. The Distance Dependence of Regional Phase Discriminants

    Science.gov (United States)

    1992-11-09

    600. 650. Ranqe fkml 19 RC - fenni obs 110. -rT T 105. 100. 95. 90. 85. 80. 75. 70. 65. S60. 00 55. S 50. 1- 45. 40. 35. 30. 25. 20. 15.L 10. 5. 0...Z -5.r. I...... , I . . I . 1.’ 111111, 111 LIL I J 50. 100. 150. 200. 250. 300. 350. 400. 450. 500. 550 600. 650 Range(kin] 20 TC - fenni obs 110...450. 500. 550. 600. 650. R;nnp fLNer] P&S - fenni 110. 105. 100. 95. 90. 85. 80. 75. 70.2 65. S60. 00 55.L c so 1- 45. 40. 35. 30. 25. 20

  20. Five-membered rings as diazo components in optical data storage devices: An ab initio investigation of the lowest singlet excitation energies

    DEFF Research Database (Denmark)

    Åstrand, P.-O.; Sommer-Larsen, P.; Hvilsted, Søren

    2000-01-01

    been investigated as diazo components for a potential use in optical das storage materials. It is found that the diazo compounds with two heterocyclic five-membered rings have pi --> pi* excitation energies corresponding to laser wavelengths in the region 450-500 nm whereas one five-membered ring...... and a phenyl group as diazo components results in wavelengths in the region 400-335 nm. (C) 2000 Published by Elsevier Science B.V....

  1. The changes of ADI structure during high temperature annealing

    OpenAIRE

    A. Krzyńska; M. Kaczorowski

    2010-01-01

    The results of structure investigations of ADI during it was annealing at elevated temperature are presented. Ductile iron austempered at temperature 325oC was then isothermally annealed 360 minutes at temperature 400, 450, 500 and 550oC. The structure investigations showed that annealing at these temperatures caused substantial structure changes and thus essential hardness decrease, which is most useful property of ADI from point of view its practical application. Degradation advance of the ...

  2. Pyrolysis of rice husk and corn stalk in auger reactor:Part 1. Characterization of char and gas at various temperatures

    OpenAIRE

    Yu, Yang; Yang, Yang; Cheng, Zhicai; Blanco, Paula H.; Liu, Ronghou; Bridgwater, A.V.; Cai, Junmeng

    2016-01-01

    In this study, rice husk and corn stalk have been pyrolyzed in an auger pyrolysis reactor at pyrolysis temperatures of 350, 400, 450, 500, 550, and 600 °C in order to investigate the effect of the pyrolysis temperature on the pyrolysis performance of the reactor and physicochemical properties of pyrolysis products (this paper focuses on char and gas). The results have shown that the pyrolysis temperature significantly affects the mass yields and properties of the pyrolysis products. The mass ...

  3. Uranium isotope ratios of Muonionalusta troilite and complications for the absolute age of the IVA iron meteorite core

    Science.gov (United States)

    Brennecka, Gregory A.; Amelin, Yuri; Kleine, Thorsten

    2018-05-01

    The crystallization ages of planetary crustal material (given by basaltic meteorites) and planetary cores (given by iron meteorites) provide fiducial marks for the progress of planetary formation, and thus, the absolute ages of these objects fundamentally direct our knowledge and understanding of planet formation and evolution. The lone precise absolute age of planetary core material was previously obtained on troilite inclusions from the IVA iron meteorite Muonionalusta. This previously reported Pb-Pb age of 4565.3 ± 0.1 Ma-assuming a 238U/235U =137.88-only post-dated the start of the Solar System by approximately 2-3 million years, and mandated fast cooling of planetary core material. Since an accurate Pb-Pb age requires a known 238U/235U of the sample, we have measured both 238U/235U and Pb isotopic compositions of troilite inclusions from Muonionalusta. The measured 238U/235U of the samples range from ∼137.84 to as low as ∼137.22, however based on Pb and U systematics, terrestrial contamination appears pervasive and has affected samples to various extents for Pb and U. The cause of the relative 235U excess in one sample does not appear to be from terrestrial contamination or the decay of short-lived 247Cm, but is more likely from fractionation of U isotopes during metal-silicate separation during core formation, exacerbated by the extreme U depletion in the planetary core. Due to limited Pb isotopic variation and terrestrial disturbance, no samples of this study produced useful age information; however the clear divergence from the previously assumed 238U/235U of any troilite in Muonionalusta introduces substantial uncertainty to the previously reported absolute age of the sample without knowledge of the 238U/235U of the sample. Uncertainties associated with U isotope heterogeneity do not allow for definition of a robust age of solidification and cooling for the IVA core. However, one sample of this work-paired with previous work using short

  4. A device for uranium series leaching from glass fiber in HEPA filter

    International Nuclear Information System (INIS)

    Gye-Nam Kim; Hye-Min Park; Wang-Kyu Choi; Jei-Kwon Moon

    2012-01-01

    For the disposal of a high efficiency particulate air (HEPA) glass filter into the environment, the glass fiber should be leached to lower its radioactive concentration to the clearance level. To derive an optimum method for the removal of uranium series from a HEPA glass fiber, five methods were applied in this study. That is, chemical leaching by a 4.0 M HNO 3 -0.1 M Ce(IV) solution, chemical leaching by a 5 wt% NaOH solution, chemical leaching by a 0.5 M H 2 O 2 -1.0 M Na 2 CO 3 solution, chemical consecutive chemical leaching by a 4.0 M HNO 3 solution, and repeated chemical leaching by a 4.0 M HNO 3 solution were used to remove the uranium series. The residual radioactivity concentrations of 238 U, 235 U, 226 Ra, and 234 Th in glass after leaching for 5 h by the 4.0 M HNO 3 -0.1 M Ce(IV) solution were 2.1, 0.3, 1.1, and 1.2 Bq/g. The residual radioactivity concentrations of 238 U, 235 U, 226 Ra, and 234 Th in glass after leaching for 36 h by 4.0 M HNO 3 -0.1 M Ce(IV) solution were 76.9, 3.4, 63.7, and 71.9 Bq/g. The residual radioactivity concentrations of 238 U, 235 U, 226 Ra, and 234 Th in glass after leaching for 8 h by a 0.5 M H 2 O 2 -1.0 M Na 2 CO 3 solution were 8.9, 0.0, 1.91, and 6.4 Bq/g. The residual radioactivity concentrations of 238 U, 235 U, 226 Ra, and 234 Th in glass after consecutive leaching for 8 h by the 4.0 M HNO 3 solution were 2.08, 0.12, 1.55, and 2.0 Bq/g. The residual radioactivity concentrations of 238 U, 235 U, 226 Ra, and 234 Th in glass after three repetitions of leaching for 3 h by the 4.0 M HNO 3 solution were 0.02, 0.02, 0.29, and 0.26 Bq/g. Meanwhile, the removal efficiencies of 238 U, 235 U, 226 Ra, and 234 Th from the waste solution after its precipitation-filtration treatment with NaOH and alum for reuse of the 4.0 M HNO 3 waste solution were 100, 100, 93.3, and 100%. (author)

  5. Remarks on the influence of enrichment reduction on fuel cycle costs

    International Nuclear Information System (INIS)

    Krull, W.

    1985-01-01

    The cost factors influencing the fuel cycle cost analysis for research reactors are discussed in detail with special emphasis on fuel element fabrication costs, burnup and reprocessing costs. Two different aspects for the conversion from HEU to LEU are considered: plus 14% U-235 weight per LEU fuel element and plus ca. 50 % U-235 weight per LEU fuel element. The cost factors and these conversion aspects were taken for calculating the changes in fuel cycle costs for the three different meat materials U 3 O 8 , U 3 Si 2 and U 3 Si. The results of these calculations can be summarized as following: - if in the HEU case the fuel loading and the burnup of a fuel element is low there will be some economic advantages in the LEU case; - if in the HEU case the fuel loading and the burnup of a fuel element is high there will be economic disadvantages in the LEU case. (author)

  6. Development of a “Fission-proxy” Method for the Measurement of 14-MeV Neutron Fission Yields at CAMS

    Energy Technology Data Exchange (ETDEWEB)

    Gharibyan, Narek [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2016-10-25

    Relative fission yield measurements were made for 50 fission products from 25.6±0.5 MeV alpha-induced fission of Th-232. Quantitative comparison of these experimentally measured fission yields with the evaluated fission yields from 14-MeV neutron-induced fission of U-235 demonstrates the feasibility of the proposed fission-proxy method. This new technique, based on the Bohr-independence hypothesis, permits the measurement of fission yields from an alternate reaction pathway (Th-232 + 25.6 MeV α → U-236* vs. U-235 + 14-MeV n → U-236*) given that the fission process associated with the same compound nucleus is independent of its formation. Other suitable systems that can potentially be investigated in this manner include (but are not limited to) Pu-239 and U-237.

  7. Contribution to the study of the influences of the excitation energy on the characteristics of the fission process

    International Nuclear Information System (INIS)

    Wagemans, C.

    1979-01-01

    Neutron induced and spontaneous fission with neutron energies from 10 -2 to 2.10 5 eV have been studied. Thermal neutron induced fission measurements in Pa 231 , Th 232 , Np 237 , U 233 , U 235 , Pu 239 and Pu 241 are reported. Energy and mass distributions of heavy fission fragments due to the spontaneous fission of Pu 240 are compared to the results obtained by thermal neutron fission of Pu 239 ; the events observed with U 236 , Pu 240 , Pa 232 and Np 238 are explained by the Bohr theory of fission channels. Ternary fission phenomena of U 233 , U 235 , Pu 239 , Pa 231 and Np 237 induced by thermal neutrons are explained and compared to models of Carjan and Feather. (MDC)

  8. Limits on uranium and thorium bulk content in GERDA Phase I detectors

    Science.gov (United States)

    GERDA Collaboration; Agostini, M.; Allardt, M.; Bakalyarov, A. M.; Balata, M.; Barabanov, I.; Baudis, L.; Bauer, C.; Becerici-Schmidt, N.; Bellotti, E.; Belogurov, S.; Belyaev, S. T.; Benato, G.; Bettini, A.; Bezrukov, L.; Bode, T.; Borowicz, D.; Brudanin, V.; Brugnera, R.; Caldwell, A.; Cattadori, C.; Chernogorov, A.; D'Andrea, V.; Demidova, E. V.; di Vacri, A.; Domula, A.; Doroshkevich, E.; Egorov, V.; Falkenstein, R.; Fedorova, O.; Freund, K.; Frodyma, N.; Gangapshev, A.; Garfagnini, A.; Grabmayr, P.; Gurentsov, V.; Gusev, K.; Hakemüller, J.; Hegai, A.; Heisel, M.; Hemmer, S.; Hofmann, W.; Hult, M.; Inzhechik, L. V.; Janicskó Csáthy, J.; Jochum, J.; Junker, M.; Kazalov, V.; Kihm, T.; Kirpichnikov, I. V.; Kirsch, A.; Kish, A.; Klimenko, A.; Kneißl, R.; Knöpfle, K. T.; Kochetov, O.; Kornoukhov, V. N.; Kuzminov, V. V.; Laubenstein, M.; Lazzaro, A.; Lebedev, V. I.; Lehnert, B.; Liao, H. Y.; Lindner, M.; Lippi, I.; Lubashevskiy, A.; Lubsandorzhiev, B.; Lutter, G.; Macolino, C.; Majorovits, B.; Maneschg, W.; Medinaceli, E.; Mingazheva, R.; Misiaszek, M.; Moseev, P.; Nemchenok, I.; Palioselitis, D.; Panas, K.; Pandola, L.; Pelczar, K.; Pullia, A.; Riboldi, S.; Rumyantseva, N.; Sada, C.; Salamida, F.; Salathe, M.; Schmitt, C.; Schneider, B.; Schönert, S.; Schreiner, J.; Schütz, A.-K.; Schulz, O.; Schwingenheuer, B.; Selivanenko, O.; Shevchik, E.; Shirchenko, M.; Simgen, H.; Smolnikov, A.; Stanco, L.; Stepaniuk, M.; Vanhoefer, L.; Vasenko, A. A.; Veresnikova, A.; von Sturm, K.; Wagner, V.; Walter, M.; Wegmann, A.; Wester, T.; Wiesinger, C.; Wojcik, M.; Yanovich, E.; Zhitnikov, I.; Zhukov, S. V.; Zinatulina, D.; Zuber, K.; Zuzel, G.

    2017-05-01

    Internal contaminations of 238U, 235U and 232Th in the bulk of high purity germanium detectors are potential backgrounds for experiments searching for neutrinoless double beta decay of 76Ge. The data from GERDA Phase I have been analyzed for alpha events from the decay chain of these contaminations by looking for full decay chains and for time correlations between successive decays in the same detector. No candidate events for a full chain have been found. Upper limits on the activities in the range of a few nBq/kg for 226Ra, 227Ac and 228Th, the long-lived daughter nuclides of 238U, 235U and 232Th, respectively, have been derived. With these upper limits a background index in the energy region of interest from 226Ra and 228Th contamination is estimated which satisfies the prerequisites of a future ton scale germanium double beta decay experiment.

  9. Processing device for gaseous waste containing uranium hexafluoride

    International Nuclear Information System (INIS)

    Hirosawa, Jun-ichi.

    1985-01-01

    Purpose: To enable to detect the inactivation of chemical traps thereby reduce the amount of adsorbents. Constitution: Two chemical traps are disposed in series and γ-detector for detecting γ-rays generated from U-235 in hexafluoride is disposed to the outer surface of a pipeway connecting these two chemical traps. Further, chemical traps are adapted to be swtichable between the first stage and the second stage thereof by the ON-OFF operation of a valve. Then, by determining γ-rays from U-235 at the pipeway downstream from the gas exit of the chemical traps, the counted value for the γ-rays is substantially at the background level so long as the chemical trap has an adsorbing performance for uranium hexafluoride. Then, since the γ-ray counted value is increased at the step upon inactivation of the chemical trap, the inactivation of the trap can be detected. (Yoshino, Y.)

  10. A neutron method for NDA analysis in the SAPPHIRE Project

    International Nuclear Information System (INIS)

    Lewis, K.D.

    1995-01-01

    The implementation of Project SAPPHIRE, the top secret mission to the Republic of Kazakhstan to recover weapons grade nuclear materials, consisted of four major elements: (1) the re-packing of fissile material from Kazakh containers into suitable US containers; (2) nondestructive analyses (NDA) to quantify the U-235 content of each container for Nuclear Criticality Safety and compliance purposes; (3) the packaging of the fissile material containers into 6M/2R drums, which are internationally approved for shipping fissile material; and (4) the shipping or transport of the recovered fissile material to the United States. This paper discusses the development and application of a passive neutron counting technique used in the NDA phase of SAPPHIRE operations to analyze uranium/beryllium (U/Be) alloys and compounds for U-235 content

  11. Waste arisings from a high-temperature reactor with a uranium-thorium fuel cycle

    International Nuclear Information System (INIS)

    1979-09-01

    This paper presents an equilibrium-recycle condition flow sheet for a high-temperature gas-cooled reactor (HTR) fuel cycle which uses thorium and high-enriched uranium (93% U-235) as makeup fuel. INFCE Working Group 7 defined percentage losses to various waste streams are used to adjust the heavy-element mass flows per gigawatt-year of electricity generated. Thorium and bred U-233 are recycled following Thorex reprocessing. Fissile U-235 is recycled one time following Purex reprocessing and then is discarded to waste. Plutonium and other transuranics are discarded to waste. Included are estimates of volume, radioactivity, and heavy-element content of wastes arising from HTR fuel element fabrication; HTR operation, maintenance, and decommissioning; and reprocessing spent fuel where the waste is unique to the HTR fuel cycle

  12. Evaluation of Pu sample oscillations in CESAR

    Energy Technology Data Exchange (ETDEWEB)

    Brunet, M.

    1974-10-15

    A set of 12 plutonium samples of various compositions were oscillated in CESAR in 1973. Comparisons were made to the oscillated reactivity effect of a known specimen of U-235 and boron where the detector signals were corrected against a background signal based on comparison to the motion of a control rod in the central location of critical assembly. An equivalent sample method was tested first for various samples of U-235 and boron to establish a means of correction in the detector response. Inferred plutonium reaction rates in the experiments were compared to transport theory calculations using the APOLLO code. Addition effort is needed to reconcile differences in measured and calculated results requiring both chemical analyses of the plutonium isotopes in the samples and improved cross sections for the plutonium isotopes.

  13. Neutronic analysis for the fission Mo99 production by irradiation of leu targets in TRIGA 14 MW reactor

    International Nuclear Information System (INIS)

    Dulugeac, S. D.; Mladin, M.; Budriman, A. G.

    2013-01-01

    Molybdenum production can be a solution for the future in the utilization of the Romanian TRIGA, taking into account the international market supply needs. Generally, two different techniques are available for Mo 99 production for use in medical Tc 99 generation.The first one is based on neutron irradiation of molybdenum targets of natural isotopic composition or enriched in Mo 98 . In a second process, Mo 99 is obtained as a result of the neutron induced fission of U 235 according to U 235 (n,f) Mo 99 . The objectives of the paper are related to Mo 99 production as a result of fission. Neutron physics parameters are determined and presented, such as: thermal flux axial distribution for the critical reactor at 10 MW inside the irradiation location; reactivity introduced by three Uranium foil containers; neutron fluxes and fission rates in the Uranium foils; released and deposited power in the Uranium foils; Mo 99 activity in the Uranium foils. (authors)

  14. Theoretical analysis of nuclear reactors (Phase I), I-V, Part IV, Nuclear fuel depletion

    International Nuclear Information System (INIS)

    Pop-Jordanov, J.

    1962-07-01

    Nuclear fuel depletion is analyzed in order to estimate the qualitative and quantitative fuel property changes during irradiation and the influence of changes on the reactivity during long-term reactor operation. The changes of fuel properties are described by changes of neutron absorption and fission cross sections. Part one of this report covers the economic significance of fuel burnup and the review of fuel isotopic changes during depletion. Pat two contains the analysis of the U 235 chain, analytical expressions for the concentrations of U 235 , U 236 and Np 237 as a function of burnup. Part three contains the analysis of neutron spectrum influence on the Westcott method for calculating the cross sections. Part four contains the calculation method applied on Calder Hall type reactor. The results were obtained by applying ZUSE-22 R digital computer

  15. History of the use of uranium

    International Nuclear Information System (INIS)

    Miettinen, J.K.

    1989-01-01

    Uranium was found 200 years ago, though the first use for it - in colouring glass yellow, orange or green - was only found 40 years later. When its radioactivity was discovered in 1896, interest in research into uranium increased and for a brief period it was used for improving the ductility of steel. The isolation of radium from uranium ore in 1904 caused a boom for uranium mining for radium. It found use in healing skin cancer, for various 'health' preparations like radon-containing water, and for making self-luminous paints. The discovery of fission 50 years ago increased the use of uranium into large industrial-scale applications. For fission weapons highly enriched U-235 and Pu-239 were needed. Today the main use is for uranium enriched to about 3 per cent U-235 for light water power reactors. Other important uses are for submarines, icebreakers and satellites

  16. Measuring Uranium Decay Rates for Advancement of Nuclear Forensics and Geochronology

    Energy Technology Data Exchange (ETDEWEB)

    Parsons-Davis, Tashi [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2017-10-31

    Radioisotopic dating techniques are highly valuable tools for understanding the history of physical and chemical processes in materials related to planetary sciences and nuclear forensics, and rely on accurate knowledge of decay constants and their uncertainties. The decay constants of U-238 and U-235 are particularly important to Earth science, and often the measured values with lowest reported uncertainties are applied, although they have not been independently verified with similar precision. New direct measurements of the decay constants of U-238, Th-234, U-235, and U-234 were completed, using a range of analytical approaches. An overarching goal of the project was to ensure the quality of results, including metrological traceability to facilitate implementation across diverse disciplines. This report presents preliminary results of these experiments, as a few final measurements and calculations are still in progress.

  17. Estimation of the self-attenuation correction factor for gamma rays emission from nuclear materials

    International Nuclear Information System (INIS)

    Badawy, A.; El-Gammal, W.A.

    2001-01-01

    This work presents an investigation of the self-attenuation of gamma-rays emission from nuclear materials (NMs) for measuring the U-235 enrichment, U-235 mass content and isotopic composition of NMs by non-destructive assay technique [NDA]. The measurements then would not need the use of suitable NM Standards which may not be available in many situations. The self-attenuation correction factor (F) may be estimated by the use of the linear attenuation factor of the assayed sample, the geometrical configuration of the assay set-up and the position of the assayed sample relative to the detector. A developed mathematical analysis makes use of specific parameters which affect the estimation of the self-attenuation of the source-detector system which emits passive gamma-rays at certain prominent signatures

  18. Reduced enrichment program for the FRM-II, status 2004

    International Nuclear Information System (INIS)

    Roehrmoser, A.; Petry, W.; Boening, K; Wieschalla, N.

    2005-01-01

    The new research reactor FRM-II of the Technische Universitaet Muenchen (TUM) has been designed to provide a maximal thermal neutron flux at mere 20 MW power. The single element design uses silicide fuel of densities 3.0 and 1.5 g/cm 3 of highly enriched uranium (HEU, 93 % U-235). With the nuclear license, that was granted in May 2003, a condition was imposed to reduce the enrichment of FRM-II to medium enriched uranium (MEU) with not more than 50 % U-235 until the end of the year 2010. The TUM has established an international working group to meet this target. This paper presents the backgrounds and the results and plannings for the first of three 2 1/2 year periods to reach the conversion in time. (author)

  19. Liquid Thermal Diffusion during the Manhattan Project

    Science.gov (United States)

    Cameron Reed, B.

    2011-06-01

    On the basis of Manhattan Engineer District documents, a little known Naval Research Laboratory report of 1946, and other sources, I construct a more complete history of the liquid-thermal-diffusion method of uranium enrichment during World War II than is presented in official histories of the Manhattan Project. This method was developed by Philip Abelson (1913-2004) and put into operation at the rapidly-constructed S-50 plant at Oak Ridge, Tennessee, which was responsible for the first stage of uranium enrichment, from 0.72% to 0.85% U-235, producing nearly 45,000 pounds of enriched U-235 by July 1945 at a cost of just under 20 million. I review the history, design, politics, construction, and operation of the S-50 liquid-thermal-diffusion plant.

  20. Radiometric determination of uranium and its decay products found in uranium ores

    International Nuclear Information System (INIS)

    Alencar, D.M. de.

    1982-01-01

    Uranium is analysed by three different methods based on gamma spectroscopy. The first method evaluates the 205 KeV photopeak emitted in the U-235 decay. It consits of a direct way of measuring 235 U, is applicable to the fuel element control and its inherent uncertainty is 13%. The second method assumes that uranium is in secular equilibrium in the sample and thus, uses the 242 KeV peak of Ra-226 and also the ratio between the 186 and 242 KeV peak areas. The third method analyses the contributions of U-235 and Ra-226 for the 186 KeV peak area; its error is negligible and its uncertainty is of 3%. (C.L.B.) [pt

  1. A Comparison of Theoretical and Experimental Values of the Activation Doppler Effect in Some Fast Reactor Spectra

    International Nuclear Information System (INIS)

    Haeggblom, H.; Tiren, L.I.

    1968-08-01

    Results of activation Doppler measurements on the U 238 (n,γ) and U 235 (n, fission) reactions in the FR0 and MSCA fast critical assemblies have been compared with theoretical values. The study covers neutron spectra with median fission energies from 50 to 240 keV. The calculated Doppler effect in U 238 in the FR0 cores is 20 - 35 % lower than the measured values. The sensitivity of the theoretical result with regard to changes in cross sections and neutron spectrum has been studied. The theoretical value for U 235 (FR0 core 5) is 4 times higher than the measured one. The report includes a brief description of the DORIX-2 method of calculating effective resonance cross sections appropriate to activation Doppler measurements. References to the cross section data used for the comparisons are also given

  2. Delayed photoneutrons of the of the Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Ngo Quang Huy; Ha Van Thong; Vu Hai Long; Ngo Phu Khang; Nguyen Nhi Dien; Pham Van Lam; Huynh Dong Phuong; Luong Ba Vien; Le Vinh Vinh

    1994-01-01

    Time spectrum of delayed neutrons of the Dalat nuclear research reactor is measured and analyzed. It corresponds to a shut-down neutron fluxes of about 10 5 /10 8 n/cm 2 /sec after 100 hours continuous reactor operation at steady power level of 500 kW. Data processing of experimental time neutron spectrum gives 16 exponents, of which 10, resulting from photoneutrons due to (γ,n) reactions on beryllium used inside the reactor core, are obtained by using successive exponential stripping fitting method. For the Dalat reactor, the effective delayed photoneutron fraction relative to the total effective delayed neutron fraction is β B e eff =0.49%β eff for a beryllium weight relative to U 235 fuel of m B e/m U = 8.5. This result is acceptable in comparison to those obtained for other Be-U 235 media. (author). 5 refs., 2 figs., 4 tabs

  3. Energy released in fission

    International Nuclear Information System (INIS)

    James, M.F.

    1969-05-01

    The effective energy released in and following the fission of U-235, Pu-239 and Pu-241 by thermal neutrons, and of U-238 by fission spectrum neutrons, is discussed. The recommended values are: U-235 ... 192.9 ± 0.5 MeV/fission; U-238 ... 193.9 ± 0.8 MeV/fission; Pu-239 ... 198.5 ± 0.8 MeV/fission; Pu-241 ... 200.3 ± 0.8 MeV/fission. These values include all contributions except from antineutrinos and very long-lived fission products. The detailed contributions are discussed, and inconsistencies in the experimental data are pointed out. In Appendix A, the contribution to the total useful energy release in a reactor from reactions other than fission are discussed briefly, and in Appendix B there is a discussion of the variations in effective energy from fission with incident neutron energy. (author)

  4. Standard format and content for the physical protection section of a license application (for facilities other than nuclear power plants)

    International Nuclear Information System (INIS)

    1976-06-01

    The document presented has been prepared as an aid to uniformity and completeness in the preparation and review of the physical protection section of license applications. It is applicable to fuel reprocessing plants, fuel manufacturing plants, SNM tranportation, or other special nuclear material operations involving the possession and use of uranium 235 (contained in uranium enriched to 20 percent or more in the U-235 isotope), uranium 233, or plutonium alone or in any combination in a quantity of 5000 grams or more computed by the formula: grams = (grams contained U-235) + 2.5 (grams U-233 + grams plutonium). The document is not intended to be used for nuclear power plants. The information specified is the minimum needed for a license application. Additional information may be required for completion of the staff review of a particular application

  5. Barium 139 as Fission Indicator

    Energy Technology Data Exchange (ETDEWEB)

    Broda, E.

    1943-07-01

    This report is based on a measurement performed at the Cavendish Laboratory (Cambridge) by E. Broda in December 1943 where a technique has been worked out for measuring the fission density in a uranium containing medium in relative units by determining the amount of a suitable fission product formed. Generally a given fission product will be formed in natural uranium by slow neutron fission of U235 or by fast neutron fission of either U235 or U238. It is intended to translate the relative units into absolute units by comparison of the Ba yield with the indication of UF6 fission chamber in the same medium. This has to be done separately for fast and slow neutron fission as the yields may be different. Another application of the technique developed is the measurement of thermal neutron density in an uraniferous medium without using a detector subject to variations of sensitivity according to the properties of the medium. (nowak)

  6. The prospect of uranium nitride (UN) and mixed nitride fuel (UN-PuN) for pressurized water reactor

    Science.gov (United States)

    Syarifah, Ratna Dewi; Suud, Zaki

    2015-09-01

    Design study of small Pressurized Water Reactors (PWRs) core loaded with uranium nitride fuel (UN) and mixed nitride fuel (UN-PuN), Pa-231 as burnable poison, and Americium has been performed. Pa-231 known as actinide material, have large capture cross section and can be converted into fissile material that can be utilized to reduce excess reactivity. Americium is one of minor actinides with long half life. The objective of adding americium is to decrease nuclear spent fuel in the world. The neutronic analysis results show that mixed nitride fuel have k-inf greater than uranium nitride fuel. It is caused by the addition of Pu-239 in mixed nitride fuel. In fuel fraction analysis, for uranium nitride fuel, the optimum volume fractions are 45% fuel fraction, 10% cladding and 45% moderator. In case of UN-PuN fuel, the optimum volume fractions are 30% fuel fraction, 10% cladding and 60% coolant/ moderator. The addition of Pa-231 as burnable poison for UN fuel, enrichment U-235 5%, with Pa-231 1.6% has k-inf more than one and excess reactivity of 14.45%. And for mixed nitride fuel, the lowest value of reactivity swing is when enrichment (U-235+Pu) 8% with Pa-231 0.4%, the excess reactivity value 13,76%. The fuel pin analyze for the addition of Americium, the excess reactivity value is lower than before, because Americium absorb the neutron. For UN fuel, enrichment U-235 8%, Pa-231 1.6% and Am 0.5%, the excess reactivity is 4.86%. And for mixed nitride fuel, when enrichment (U-235+Pu) 13%, Pa-231 0.4% and Am 0.1%, the excess reactivity is 11.94%. For core configuration, it is better to use heterogeneous than homogeneous core configuration, because the radial power distribution is better.

  7. Calculation of the fissile mass of a graphite moderated critical assembly using 93% enriched uranium

    International Nuclear Information System (INIS)

    Correa, F.; Marzo, M.A.S.; Collussi, I.; Ferreira, A.C.A.

    1976-01-01

    The critical mass of uranium has been calculated for a graphite moderated set fueled with 93% enriched uranium to be mounted on the Instituto de Energia Atomica split table Zero Power Reactor. The core composition was optimized to permit the maximum number of configurations to be studied. Analysis of three core compositions shows that 8 Kg of uranium enriched to 93% - U-235 (by weight) and 100 Kg of thorium would be sufficient for criticality experiments [pt

  8. Results of research and development works of the Institute for Nuclear Engineering in 1982

    International Nuclear Information System (INIS)

    1983-02-01

    Report on the following works: a) enrichment of U-235 according to the nozzle enrichment process, b) physical advancement of that process, c) separative element testing and uranium hexafluoride technology for that process, d) development and testing of components for that process, e) plant development to this and f) basic research for the particle injection for fusion experiments and -reactors as well as g) investigations on molecular and cluster beams. Listing of the publications and reports performed in 1982. (PW) [de

  9. Determination of radioactivity in petroleum products and wastes

    International Nuclear Information System (INIS)

    Hrichi, Hajer

    2009-01-01

    At this end engineering study, we determined the activities of gamma- emitting radionuclides belonging to the families of 238 U, 235 U, 232 Th and 40K in the petroleum products and wastes of the refinery S.T.I.R. The activities of radionuclides which exceed that of crude oil prove that it's a technologically enhanced natural radioactivity since several chemical products were injected during the refining process. (Author)

  10. Otto Hahn - Research and responsibility. Conflicts of a scientist; Otto Hahn - Forschung und Verantwortung. Konflikte eines Wissenschaftlers

    Energy Technology Data Exchange (ETDEWEB)

    Hoffmann, K.

    2005-07-01

    The life of Otto Hahn is documented and the time where science started in the mysterious field of radioactivity. The main steps: Youth, studies, first practical experiences, research at Berlin university, first world war, success for atomic researchers, national socialism - night over Germany, fission of uranium atom, menace with the atomic bomb of Hitler, the American super explosive U235, hunting on atomic researchers, diplomacy with atomic bombs, in conflict with conscience and policy, against nuclear arm tests and atomic arm race. (GL)

  11. Otto Hahn - Research and responsibility. Conflicts of a scientist

    International Nuclear Information System (INIS)

    Hoffmann, K.

    2005-01-01

    The life of Otto Hahn is documented and the time where science started in the mysterious field of radioactivity. The main steps: Youth, studies, first practical experiences, research at Berlin university, first world war, success for atomic researchers, national socialism - night over Germany, fission of uranium atom, menace with the atomic bomb of Hitler, the American super explosive U235, hunting on atomic researchers, diplomacy with atomic bombs, in conflict with conscience and policy, against nuclear arm tests and atomic arm race. (GL)

  12. Controllability studies for an advanced CANDU boiling light water reactor

    International Nuclear Information System (INIS)

    Lepp, R.M.; Hinds, H.W.

    1976-12-01

    Bulk controllability studies carried out as part of a conceptual design study of a 1200 MWe CANDU boiling-light-water reactor fuelled with U 235 - or Pu-enriched uranium oxide are outlined. The concept, the various models developed for its simulation on a hybrid computer and the perturbations used to test system controllability, are described. The results show that this concept will have better bulk controllability than similar CANDU-BLW reactors fuelled with natural uranium. (author)

  13. Modeling of the water gap in BWR fuel elements using SCALE/TRITON; Modellierung des Wasserspalts bei SWR-BE mit SCALE/TRITON

    Energy Technology Data Exchange (ETDEWEB)

    Tittelbach, S.; Chernykh, M. [WTI Wissenschaftlich-Technische Ingenieurberatung GmbH, Juelich (Germany)

    2012-11-01

    The authors show that an adequate modeling of the water gap in BWR fuel element models using the code TRITON requires an explicit consideration of the Dancoff factors. The analysis of three modeling options reveals that considering the moderating effects of the water gap coolant for the peripheral fuel elements the resulting deviations of the U-235 and Pu-239 concentrations are significantly reduced. The increased temporal calculation efforts are justified with respect to the burnup credits for criticality safety analyses.

  14. Simulation of the TRIGA-ININ reactor using EXT-2, in R-θ RΘ and temperature of 20 Centigrade

    International Nuclear Information System (INIS)

    Aguilar H, F.

    1983-10-01

    The simulation of the TRIGA-ININ reactor, it was beginning considering the more simple case (follower bars equal to fuel elements, cell bar transitory with vacuum, etc.), this it left tuning as the obtained results were observed and it was studied the literature with respect to this reactor, in the following step the followers are considered as standard elements but with 32 grams of U-235 and so forth until reaching to the configuration that is considered definitive. (Author)

  15. (e,e'f) coincidence experiments for fission decay of giant resonances in 235,238U

    International Nuclear Information System (INIS)

    Weber, T.; Heil, R.D.; Kneissl, U.; Pecho, W.; Wilke, W.; Emrich, H.J.; Kihm, T.; Knoepfle, K.T.

    1988-01-01

    Extending previous work on 238 U, 235 U(e,e'f) coincidence data were taken at 4 momentum transfers yielding both E1, E2/E0 and E3 form factors and the respective multipole strength distributions in the giant resonance region of 238 U (4 x x /Γ a is obtained as a function of excitation energy for separated multipoles. The giant E2 resonance exhibits an increased symmetric fission contribution compared to E1 and E3 resonances. (orig.)

  16. The prospect of uranium nitride (UN) and mixed nitride fuel (UN-PuN) for pressurized water reactor

    International Nuclear Information System (INIS)

    Syarifah, Ratna Dewi; Suud, Zaki

    2015-01-01

    Design study of small Pressurized Water Reactors (PWRs) core loaded with uranium nitride fuel (UN) and mixed nitride fuel (UN-PuN), Pa-231 as burnable poison, and Americium has been performed. Pa-231 known as actinide material, have large capture cross section and can be converted into fissile material that can be utilized to reduce excess reactivity. Americium is one of minor actinides with long half life. The objective of adding americium is to decrease nuclear spent fuel in the world. The neutronic analysis results show that mixed nitride fuel have k-inf greater than uranium nitride fuel. It is caused by the addition of Pu-239 in mixed nitride fuel. In fuel fraction analysis, for uranium nitride fuel, the optimum volume fractions are 45% fuel fraction, 10% cladding and 45% moderator. In case of UN-PuN fuel, the optimum volume fractions are 30% fuel fraction, 10% cladding and 60% coolant/ moderator. The addition of Pa-231 as burnable poison for UN fuel, enrichment U-235 5%, with Pa-231 1.6% has k-inf more than one and excess reactivity of 14.45%. And for mixed nitride fuel, the lowest value of reactivity swing is when enrichment (U-235+Pu) 8% with Pa-231 0.4%, the excess reactivity value 13,76%. The fuel pin analyze for the addition of Americium, the excess reactivity value is lower than before, because Americium absorb the neutron. For UN fuel, enrichment U-235 8%, Pa-231 1.6% and Am 0.5%, the excess reactivity is 4.86%. And for mixed nitride fuel, when enrichment (U-235+Pu) 13%, Pa-231 0.4% and Am 0.1%, the excess reactivity is 11.94%. For core configuration, it is better to use heterogeneous than homogeneous core configuration, because the radial power distribution is better

  17. Stream sediment sampling and analysis. Final report

    International Nuclear Information System (INIS)

    Means, J.L.; Voris, P.V.; Headington, G.L.

    1986-04-01

    The objectives were to sample and analyze sediments from upstream and downstream locations (relative to the Goodyear Atomic plant site) of three streams for selected pollutants. The three streams sampled were the Scioto River, Big Beaver Creek, and Big Run Creek. Sediment samples were analyzed for EPA's 129 priority pollutants (Clean Water Act) as well as isotopic uranium ( 234 U, 235 U, and 238 U) and technetium-99

  18. Observation on the irradiation behavior of U-Mo alloy dispersion fuel

    International Nuclear Information System (INIS)

    Hofman, Gerard L.; Meyer, Mitchell K.; Park, Jong-Man

    2000-01-01

    Initial results from the postirradiation examination of high-density dispersion fuel test RERTR-3 are discussed. The U-Mo alloy fuels in this test were irradiated to 40% U-235 burnup at temperature ranging from 140 0 C to 240 0 C. Temperature has a significant effect on overall swelling of the test plates. The magnitude of the swelling appears acceptable and no unstable irradiation behavior is evident. (author)

  19. Diode laser excited optogalvanic spectroscopy of glow discharges

    International Nuclear Information System (INIS)

    Barshick, C. M.; Shaw, R. W.; Jennings, L. W.; Post-Zwicker, A.; Young, J. P.; Ramsey, J. M.

    1997-01-01

    The development of diode-laser-excited isotopically-selective optogalvanic spectroscopy (OGS) of uranium metal, oxide and fluoride in a glow discharge (GD) is presented. The technique is useful for determining 235 U/( 235 U+ 238 U) isotope ratios in these samples. The precision and accuracy of this determination is evaluated, and a study of experimental parameters pertaining to optimization of the measurement is discussed. Application of GD-OGS to other f-transition elements is also described

  20. Diode laser excited optogalvanic spectroscopy of glow discharges

    International Nuclear Information System (INIS)

    Barshick, C.M.; Shaw, R.W.; Post-Zwicker, A., Young, J.P.; Ramsey, J.M.

    1996-01-01

    The development of diode-laser-excited isotopically-selective optogalvanic spectroscopy (OGS) of uranium metal, oxide and fluoride in a glow discharge (GD) is presented. The technique is useful for determining isotopic ratios of 235 U/( 235 U + 238 U) in the above samples. The precision and accuracy of this determination is evaluated, and a study of experimental parameters pertaining to optimization of he measurement is discussed. Application of the GD-OGS to other f-transition elements is also described

  1. Giant atoms for the production of nuclear fuel

    International Nuclear Information System (INIS)

    Fahr, H.J.

    1980-01-01

    Neutral atoms can be blown up to the size of a football, if the electrons of the atomic shell are appropriately excited by photons just below the ionization energy. Such atoms, called Rydberg atoms, behave very differently to the usual ones in multiple respects. The fact that they can very easily be ionized by conventional electrostatic fields is being investigated as a method of isotope separation, for instance in the very important case of U 235 and U 238 . (orig.) [de

  2. Simulation of the TRIGA-ININ reactor using EXT-2, in R-{theta} R{theta} and temperature of 20 Centigrade; Simulacion del reactor TRIGA-ININ utilizando EXT-2, en geometria R-{theta} y una temperatura de 20 Centigrados

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar H, F [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    1983-10-15

    The simulation of the TRIGA-ININ reactor, it was beginning considering the more simple case (follower bars equal to fuel elements, cell bar transitory with vacuum, etc.), this it left tuning as the obtained results were observed and it was studied the literature with respect to this reactor, in the following step the followers are considered as standard elements but with 32 grams of U-235 and so forth until reaching to the configuration that is considered definitive. (Author)

  3. Post-irradiation analysis of low enriched U-Mo/Al dispersions fuel miniplate tests, RERTR 4 and 5

    International Nuclear Information System (INIS)

    Hofman, G.L.; Finlay, M.R.; Kim, Y.S.

    2005-01-01

    Interpretation of the post irradiation data of U-Mo/Al dispersion fuel mini plates irradiated in the Advanced Test Reactor to a maximum U-235 burn up of 80% are presented. The analyses addresses fuel swelling and porosity formation as these fuel performance issues relate to fuel fabrication and irradiation parameters. Specifically, mechanisms involved in the formation of porosity observed in the U-Mo/Al interaction phase are discussed and, means of mitigating or eliminating this irradiation phenomenon are offered. (author)

  4. Minutes of Technical Division Steering Committee Meeting, September 13, 1955 -- Savannah River Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Evans, L.C.

    1955-09-16

    The Steering Committee approved 8 studies related to separation processes, analytical chemistry, waste handling, and recycle development. Safety and security issues were discussed. Appendices detail the financial status of the Technical Division and estimated man months for development studies approved for the Purex Process, tritium separations, thorium recycle, U-235 separations, and 100-, 200-, and 300-Area studies in analytical chemistry development. The status of 25 other Technical Division studies are listed along with their budget.

  5. QUEIMAP: a computer routine for punctual analysis of a nuclear fuel depletion with accumulation of fission fragments

    International Nuclear Information System (INIS)

    Couto, R.T.

    1990-01-01

    QUEIMAP is a computer routine for burnup calculation, composed of five FORTRAN-77 subroutines. Its objective is to solve depletion equation of four radionuclides conversion chain, U238, U235, Th232, as well as fission fragments equations. In this paper the burnup is considered punctual and evolutioned under cross section. It presents the solution algorithms employed by QUEIMAP, the validation of its results and the way of use it. (M.I.)

  6. Local tissue distribution of fissile nuclides

    International Nuclear Information System (INIS)

    Smith, J.M.

    1981-01-01

    Conventional tissue-section autoradiography of alpha-emitting actinide elements may require prohibitively long exposure times. Neutron-induced or fission-track autoradiography can be used for fissile nuclides such as 233 U, 235 U, and 239 Pu to circumvent this difficulty. The detection limit for these nuclides is about 4 x 10 -13 (weight fraction). This paper describes a specific technique for determining their microdistribution with histologically stained tissue sections

  7. The text of the agreement of 18 September 1987 between Chile and the Agency for the application of safeguards to nuclear material supplied from the People's Republic of China

    International Nuclear Information System (INIS)

    1988-03-01

    The text of the Agreement between the Government of the Republic of Chile and the Agency for the application of safeguards to nuclear material in the form of UF 6 , enriched to 20% in the isotope U-235, supplied from the People's Republic of China for the fabrication of MTR-type fuel elements for Lo Aguirre research reactor, is reproduced. The agreement entered into force on 18 September 1987

  8. Management and inspection of integrity of spent fuel from IRT MEPhI research reactor

    International Nuclear Information System (INIS)

    Aden, V.G.; Bulkin, S.Y.; Sokolov, A.V.; Bushuev, A.V.; Redkin, A.F.; Portnov, A.A.

    2002-01-01

    The information on wet storage and dry storage of the spent nuclear fuel (SNF) of the IRT MEPhI reactor and experience from SNF shipment for reprocessing are presented. The procedure and a facility for nondestructive inspection of local power density fields and the burnup of fuel assemblies based on studying the γ-activity of some fission products generated in U 235 and procedure for inspection of the fuel element cladding leak tightness are described. (author)

  9. Pre-Emption and Precedent: The Significance of Iraq (1981) and Syria (2007) for an Israeli Response to an Iranian Nuclear Threat

    Science.gov (United States)

    2011-06-10

    served as a liaison with the French to ensure only U235, enriched uranium, was sent to Iraq. While Meshad was unwilling to work with undercover Mossad...supported Hossein Mousavi, Ahmadinejad‘s primary opposition candidate in the June 2010 presidential election.167 According to The Economist , Ali Mohammadi...hoflr20&div=16&id=&page= (accessed 14 May 2011). Economist . ―Who killed the professor? New light is being cast on the strange death of an Iranian

  10. Environmental-impact appraisal related to special nuclear materials. License No. SNM-696; Docket No. 70-734

    International Nuclear Information System (INIS)

    1983-06-01

    This Environmental Impact Appraisal is issued by the US Nuclear Regulatory Commission in response to an application by GA Technologies, Inc., (GA) for renewal of Special Nuclear Material (SNM) License No. SNM-696 covering plant operations at San Diego, California. The proposed action provides for continuing research, development, and production activities involving SNM, uranium enriched in the U-235 and U-233 isotopes, and plutonium

  11. An automated solution enrichment system for uranium analysis

    International Nuclear Information System (INIS)

    Jones, S.A.; Sparks, R.; Sampson, T.; Parker, J.; Horley, E.; Kelly, T.

    1993-01-01

    An automated Solution Enrichment system (SES) for analysis of Uranium and U-235 isotopes in process samples has been developed through a joint effort between Los Alamos National Laboratory and Martin Marietta Energy systems, Portsmouth Gaseous Diffusion Plant. This device features an advanced robotics system which in conjuction with stabilized passive gamma-ray and X-ray fluorescence detectors provides for rapid, non-destructive analyses of process samples for improved special nuclear material accountability and process control

  12. Comparison of measured and calculated burn-up of AVR-Fuel-Elements

    Energy Technology Data Exchange (ETDEWEB)

    Wagemann, R.

    1974-03-15

    Burn-up comparisons are made for small batches of three types of AVR fuel elements using a coupled EREBUS-MUPO neutronic analysis compared against test results from both nondestructive gamma-ray measurements of cesium-137 activity and destructive mass spectrometry measurements of the ratio of U-233 to U-235. The comparisons are relatively good for average burn-up and reasonably good for burn-up distributions.

  13. Gamma spectrometry analysis of fertilizers used in Costa Rica

    International Nuclear Information System (INIS)

    Loria, L.G.; Jimenez, R.; Badilla, M.

    2002-01-01

    Using the low level gamma spectrometry technique, the specific activity of natural and artificial isotopes present in national consumption fertilizers, supplied by local dealers, was quantified. The most outstanding isotope found in some of the fertilizers is U 235 , with an specific activity higher than the expected in this kind of product, which might mean that this isotope comes from imported raw material. (Author) [es

  14. Integral test of JENDL-3.3 for fast reactors

    International Nuclear Information System (INIS)

    Chiba, Gou

    2003-01-01

    An integral test of JENDL-3.3 was performed for fast reactors. Various types of fast reactors were analyzed. Calculation values of the nuclear characteristics were greatly especially affected by the revisions of the cross sections of U-235 capture and elastic scattering reactions. The C/E values were improved for ZPPR cross where plutonium is mainly fueled, but not for BFS cores where uranium is mainly fueled. (author)

  15. Analysis for the radionuclides of the natural uranium and thorium decay chains with special reference to uranium mine tailings

    International Nuclear Information System (INIS)

    Lowson, R.T.; Short, S.A.

    1986-08-01

    A detailed review is made of the experimental techniques that are available, or are in the process of development, for the determination of 238 U, 235 U, 234 U, 231 Pa, 232 Th, 230 Th, 228 Th, 228 Ra, 226 Ra, 223 Ra, 210 Po and 210 Pb. These products of the uranium and thorium decay chains are found in uranium mine tailings. Reference is also made to a procedure for the selective phase extraction of mineral phases from uranium mine tailings

  16. Comparative evaluation of group constants from UKNDL and the BNAB-70 system

    International Nuclear Information System (INIS)

    Bobkov, Yu.G.; Kolesov, V.E.; Krivtsov, A.S.; Manokhin, V.N.; Solov'ev, N.A.; Usachev, L.N.

    1976-01-01

    The comparison is made between the 26-group constants BNAB-70 with similar constants obtained from the evaluated UNKDL data. The data are compared by the capture and fission cross-section of Pu-239, U-235, U-238, the capture cross-section of Fe-56 and absorption of B-10 within an energy range from 100 eV to 10 MeV

  17. 76 FR 30326 - Proposed Subsequent Arrangement

    Science.gov (United States)

    2011-05-25

    ... content of 28.276 kg (24.541 ekg) of U.S.-origin highly enriched uranium (HEU) (26.342 kg U-235) and 0.0048 g of plutonium contained in three HEU driver fuels that have been irradiated in the YAYOI nuclear... three HEU driver fuels from the core of YAYOI to be cut and de-cladded in the reactor room into...

  18. Quarterly Report 24.1.3.1

    Energy Technology Data Exchange (ETDEWEB)

    Shusterman, Jennifer A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2018-01-17

    We are measuring freshly separated uranium samples using modern list mode (event-by-event) electronics with high resolution HPGe detectors to study the in-growth behaviors of uranium daughters’ gamma-rays. These data will show how we can use gamma-ray spectroscopy to determine the separation date for processed uranium. With this knowledge, one can obtain proper uranium isotope ratios using standard safeguards accountability software such as U-235 or MGAU.

  19. Effect of local burn-up variation on computed mean nuclide concentrations

    International Nuclear Information System (INIS)

    Moeller, W.

    1982-01-01

    Mean concentrations of U-235, U-236, U-238, Pu-239, Pu-240, Pu-241 and Pu-242 in some volume areas of WWER-440 fuel assemblies have been calculated from corresponding burn-up microdistribution data and compared with those calculated from burn-up mean values. Differences occurring were below 3% for the uranium nuclides but, at low burn-ups, considerable for Pu-241 and Pu-242. (author)

  20. Nuclear Britain

    Energy Technology Data Exchange (ETDEWEB)

    Bunyard, P

    1981-01-01

    The subject is dealt with in chapters, entitled: (preface) the new resistance; (1) the weapons connection; (2) the AGR - a bungled programme; (3) reactor choice - a fearful dilemma; (4) the deadly atom (uranium - mining, processing, enrichment of U-235, political and international aspects); (5) the fast reactor - how safe; (6) radiation risk and cancer; (7) atomic waste - the industry's Achilles Heel; (8) accidents - the ever-present threat; (9) the growing opposition; (10) the nuclear state.

  1. Reduction of fuel enrichment for research reactors built-up in accordance with Russian (Soviet) projects

    International Nuclear Information System (INIS)

    Aleksandrov, A.B.; Enin, A.A.; Tkachyov, A.A.

    2001-01-01

    In accordance with the Russian program of reduced enrichment for research and test reactors (RERTR) built-up in accordance with Russian (Soviet) projects, AO 'NCCP' performs works on FA fabrication with reduced enrichment fuel. The main trends and results of performed works on research reactors FEs and FAs based on UO 2 and U-9%Mo fuel with U 235 19.7% enrichment are described. (author)

  2. Behaviour of the reactivity for BWR fuel cells; Comportamiento de la reactividad para celdas de combustible BWR

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez, J. A.; Alonso, G.; Delfin, A.; Vargas, S. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Del Valle G, E., E-mail: galonso@inin.gob.mx [IPN, Escuela Superior de Fisica y Matematicas, U. P. Adolfo Lopez Mateos, Col. Lindavista, 07738 Mexico D. F. (Mexico)

    2011-11-15

    In this work the behaviour of the reactivity of a fuel assembly type BWR was studied, the objective is to obtain some expressions that consider the average enrichment of U-235 and the gadolinium concentration like a function of the fuel cells burnt. Also, the applicability of the lineal reactivity model was analyzed for fuel cells type BWR. The analysis was carried out with the CASMO-4 code. (Author)

  3. Comparative Analysis of Structural Changes In U-Mo Dispersed Fuel of Full-Size Fuel Elements And Mini-Rods Irradiated In The MIR Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Izhutov, Aleksey L.; Iakovlev, Valeriy V.; Novoselov, Andrey E. and others

    2013-12-15

    The paper summarizes the irradiation test and post-irradiation examination (PIE) data for the U-Mo low-enriched fuel that was irradiated in the MIR reactor under the RERTR Program. The PIE data were analyzed for both full-size fuel rods and mini-rods with atomized powder dispersed in Al matrix as well as with additions of 2%, 5% and 13% of silicon in the matrix and ZrN protective coating on the fuel particles. The full-size fuel rods were irradiated up to an average burnup of ∼ 60%{sup 235}U; the mini-rods were irradiated to an average burnup of ∼ 85%{sup 235}U. The presented data show a significant increase of the void fraction in the U-Mo alloy as the U-235 burnup rises from ∼ 40% up to ∼ 85%. The effect of irradiation test conditions and U-235 burnup were analyzed with regard to the formation of an interaction layer between the matrix and fuel particles as well as generation of porosity in the U-Mo alloy. Shown here are changes in distribution of U fission products as the U-235 burnup increases from ∼ 40% up to ∼ 85%.

  4. Determination of enrichment of recycle uranium fuels for different burnup values

    International Nuclear Information System (INIS)

    Zabunoglu, Okan H.

    2008-01-01

    Uranium (U) recovered from spent LWR fuels by reprocessing, which contains small amounts of U-236, is to be enriched before being re-irradiated as the recycle U. During the enrichment of recovered U in U-235, the mass fraction of U-236 also increases. Since the existence of U-236 in the recycle U has a negative effect on neutron economy, a greater enrichment of U-235 in the recycle U is required for reaching the same burnup as can be reached by the fresh U fuel. Two burnup values play the most important role in determining the enrichment of recycle U: (1) discharge burnup of spent fuel from which the recycle U is obtained and (2) desired discharge burnup of the recycle U fuel. A step-by-step procedure for calculating the enrichment of the recycle U as a function of these two burnup values is introduced. The computer codes MONTEBURNS and ORIGEN-S are made use of and a three-component (U-235, U-236, U-238) enrichment scheme is applied for calculating the amount of U-236 in producing the recycle U from the recovered U. As was aimed, the resulting expression is simple enough for quick/hand calculations of the enrichment of the recycle U for any given discharge burnup of spent fuel and for any desired discharge burnup of the recycle U fuel, most accurately within the range of 33,000-50,000 MWd/tonU

  5. Monte Carlo analyses of TRX slightly enriched uranium-H2O critical experiments with ENDF/B-IV and related data sets (AWBA Development Program)

    International Nuclear Information System (INIS)

    Hardy, J. Jr.

    1977-12-01

    Four H 2 O-moderated, slightly-enriched-uranium critical experiments were analyzed by Monte Carlo methods with ENDF/B-IV data. These were simple metal-rod lattices comprising Cross Section Evaluation Working Group thermal reactor benchmarks TRX-1 through TRX-4. Generally good agreement with experiment was obtained for calculated integral parameters: the epi-thermal/thermal ratio of U238 capture (rho 28 ) and of U235 fission (delta 25 ), the ratio of U238 capture to U235 fission (CR*), and the ratio of U238 fission to U235 fission (delta 28 ). Full-core Monte Carlo calculations for two lattices showed good agreement with cell Monte Carlo-plus-multigroup P/sub l/ leakage corrections. Newly measured parameters for the low energy resonances of U238 significantly improved rho 28 . In comparison with other CSEWG analyses, the strong correlation between K/sub eff/ and rho 28 suggests that U238 resonance capture is the major problem encountered in analyzing these lattices

  6. Separation of uranium and plutonium isotopes for measurement by multi collector inductively coupled plasma mass spectroscopy

    International Nuclear Information System (INIS)

    Martinelli, R.E.; Hamilton, T.F.; Kehl, S.R.; Williams, R.W.

    2009-01-01

    Uranium (U) and plutonium (Pu) isotopes in coral soils, contaminated by nuclear weapons testing in the northern Marshall Islands, were isolated by ion-exchange chromatography and analyzed by mass spectrometry. The soil samples were spiked with 233 U and 242 Pu tracers, dissolved in minerals acids, and U and Pu isotopes isolated and purified on commercially available ion-exchange columns. The ion-exchange technique employed a TEVA R column coupled to a UTEVA R column. U and Pu isotope fractions were then further isolated using separate elution schemes, and the purified fractions containing U and Pu isotopes analyzed sequentially using multi-collector inductively coupled plasma mass spectrometer (MCICP-MS). High precision measurements of 234 U/ 235 U, 238 U/ 235 U, 236 U/ 235 U, and 240 Pu/ 239 Pu in soil samples were attained using the described methodology and instrumentation, and provide a basis for conducting more detailed assessments of the behavior and transfer of uranium and plutonium in the environment. (author)

  7. The generation of denatured reactor plutonium by different options of the fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Broeders, C.H.M.; Kessler, G. [Inst. for Neutron Physics and Reactor Technology, Research Center Karlsruhe (Germany)

    2006-11-15

    Denatured (proliferation resistant) reactor plutonium can be generated in a number of different fuel cycle options. First denatured reactor plutonium can be obtained if, instead of low enriched U-235 PWR fuel, re-enriched U-235/U-236 from reprocessed uranium is used (fuel type A). Also the envisaged existing 2,500 t of reactor plutonium (being generated world wide up to the year 2010), mostly stored in intermediate fuel storage facilities at present, could be converted during a transition phase into denatured reactor plutonium by the options fuel type B and D. Denatured reactor plutonium could have the same safeguards standard as present low enriched (<20% U-235) LWR fuel. It could be incinerated by recycling once or twice in PWRs and subsequently by multi-recycling in FRs (CAPRA type or IFRs). Once denatured, such reactor plutonium could remain denatured during multiple recycling. In a PWR, e.g., denatured reactor plutonium could be destroyed at a rate of about 250 kg/GWey. While denatured reactor plutonium could be recycled and incinerated under relieved IAEA safeguards, neptunium would still have to be monitored by the IAEA in future for all cases in which considerable amounts of neptunium are produced. (orig.)

  8. Supply of low enriched (LEU) and highly enriched uranium (HEU) for research reactors

    International Nuclear Information System (INIS)

    Mueller, H.

    1997-01-01

    Enriched uranium for research reactors in the form of LEU /= low enriched uranium at 19.75% U-235) and HEU (= highly enriched uranium at 90 to 93% U-235) was and is - due to its high U-235 enrichment - a political fuel other than enriched uranium for power reactors. The sufficient availability of LEU and HEU is a vital question for research reactors, especially in Europe, in order to perform their peaceful research reactor programs. In the past the USA were in the Western hemisphere sole supplier of LEU and HEU. Today the USA have de facto stopped the supply of LEU and HEU, for HEU mainly due to political reasons. This paper deals, among others, with the present availability of LEU and HEU for European research reactors and touches the following topics: - historical US supplies, - influence of the RERTR-program, - characteristics of LEU and HEU, - military HEU enters the civil market, -what is the supply situation for LEU and HEU today? - outlook for safe supplies of LEU and HEU. (author)

  9. Oklo natural reactor

    International Nuclear Information System (INIS)

    Fujii, Isao

    1985-01-01

    In 1954, Professor Kazuo, Kuroda of Arkansas University in USA published the possibility that spontaneously generated natural nuclear reactors existed in prehistoric age. In 1972, 18 years after that, Commissariat a l'Energie Atomique published that in the Oklo uranium deposit in Gabon, Africa, a natural nuclear reactor was found. This fact was immediately informed to the whole world, but in Japan, its details have not necessarily been well known. The chance of investigating into this fact and visiting the Oklo deposit by the favor of COMUF, the owner of the Oklo deposit, was given, therefore, the state of the natural reactors, which has been known so far, is reported. At present, 12 natural reactors have been found in the vicinity of the Oklo deposit. The natural reactors were generated spontaneously in uranium deposits about 1.7 billion years ago when the isotopic abundance of U-235 was 3 %, and the chain reaction started naturally. When the concentration of U-235 lowered, the reaction stopped naturally. The abnormality in the U-235 abundance in natural uranium was found, and the cause was pursued. The evidence of the existence of natural reactors was shown. (Kako, I.)

  10. Criticality parameters for uranyl nitrate or plutonium nitrate systems in tributyl phosphate/kerosine and water

    International Nuclear Information System (INIS)

    Weber, W.

    1985-01-01

    This report presents the calculated values of smallest critical masses and volumina and neutron physical parameters for uranyl nitrate (3, 4, 5% U-235) or plutonium nitrate (5% Pu-240), each in a 30 per cent solution of tributyl phosphate (TBP)/kerosine. For the corresponding nitrate-water solutions, newly calculated results are presented together with a revised solution density model. A comparison of the data shows to what extent the criticality of nitrate-TBP/kerosine systems can be assessed on the basis of nitrate-water parameters, revealing that such data can be applied to uranyl nitrate/water systems, taking into account that the smallest critical mass of uranyl nitrate-TBP/kerosine systems, up to a 5 p.c. U-235 enrichment, is by 4.5 p.c. at the most smaller than that of UNH-water solutions. Plutonium nitrate (5% Pu-240) in the TBP/kerosine solution will have a smallest critical mass of up to 7 p.c. smaller, as compared with the water data. The suitability of the computing methods and cross-sections used is verified by recalculating experiments carried out to determine the lowest critical enrichment of uranyl nitrate. The calculated results are well in agreement with experimental data. The lowest critical enrichment is calculated to be 2.10 p.c. in the isotope U-235. (orig.) [de

  11. Comparative Analysis of Structural Changes In U-Mo Dispersed Fuel of Full-Size Fuel Elements And Mini-Rods Irradiated In The MIR Reactor

    International Nuclear Information System (INIS)

    Izhutov, Aleksey L.; Iakovlev, Valeriy V.; Novoselov, Andrey E. and others

    2013-01-01

    The paper summarizes the irradiation test and post-irradiation examination (PIE) data for the U-Mo low-enriched fuel that was irradiated in the MIR reactor under the RERTR Program. The PIE data were analyzed for both full-size fuel rods and mini-rods with atomized powder dispersed in Al matrix as well as with additions of 2%, 5% and 13% of silicon in the matrix and ZrN protective coating on the fuel particles. The full-size fuel rods were irradiated up to an average burnup of ∼ 60% 235 U; the mini-rods were irradiated to an average burnup of ∼ 85% 235 U. The presented data show a significant increase of the void fraction in the U-Mo alloy as the U-235 burnup rises from ∼ 40% up to ∼ 85%. The effect of irradiation test conditions and U-235 burnup were analyzed with regard to the formation of an interaction layer between the matrix and fuel particles as well as generation of porosity in the U-Mo alloy. Shown here are changes in distribution of U fission products as the U-235 burnup increases from ∼ 40% up to ∼ 85%

  12. Uranium Isotopes in Calcium Carbonate: A Possible Proxy for Paleo-pH and Carbonate Ion Concentration?

    Science.gov (United States)

    Chen, X.; Romaniello, S. J.; Herrmann, A. D.; Wasylenki, L. E.; Anbar, A. D.

    2015-12-01

    Natural variations of 238U/235U in marine carbonates are being explored as a paleoredox proxy. However, in order for this proxy to be robust, it is important to understand how pH and alkalinity affect the fractionation of 238U/235U during coprecipitation with calcite and aragonite. Recent work suggests that the U/Ca ratio of foraminiferal calcite may vary with seawater [CO32-] concentration due to changes in U speciation[1]. Here we explore analogous isotopic consequences in inorganic laboratory co-precipitation experiments. Uranium coprecipitation experiments with calcite and aragonite were performed at pH 8.5 ± 0.1 and 7.5 ± 0.1 using a constant addition method [2]. Dissolved U in the remaining solution was periodically collected throughout the experiments. Samples were purified with UTEVA resin and 238U/235U was determined using a 233U-236U double-spike and MC-ICP-MS, attaining a precision of ± 0.10 ‰ [3]. Small but resolvable U isotope fractionation was observed in aragonite experiments at pH ~8.5, preferentially enriching heavier U isotopes in the solid phase. 238U/235U of the dissolved U in these experiments can be fit by Rayleigh fractionation curves with fractionation factors of 1.00002 - 1.00009. In contrast, no resolvable U isotope fractionation was detected in an aragonite experiment at pH ~7.5 or in calcite experiments at either pH. Equilibrium isotope fractionation among dissolved U species is the most likely mechanism driving these isotope effects. Our quantitative model of this process assumes that charged U species are preferentially incorporated into CaCO3 relative to the neutral U species Ca2UO2(CO3)3(aq), which we hypothesize to have a lighter equilibrium U isotope composition than the charged U species. According to this model, the magnitude of U isotope fractionation should scale with the fraction of the neutral U species in the solution, in agreement with our experimental results. These findings suggest that U isotope variations in

  13. Elevated-temperature tensile properties of 2 1/4 Cr-1 Mo steel irradiated in the EBR-II, AD-2 experiment

    International Nuclear Information System (INIS)

    Klueh, R.L.; Vitek, J.M.

    1984-01-01

    The effect of irradiated on the tensile properties of 2 1/4 Cr-1 Mo steel was determined for specimens irradiation in EBR-II at 390 to 550 0 C. Unirradiated control specimens and specimens aged for 5000 h at the irradiation temperatures were also tested. Irradiation to approximately 9 dpa at 390 0 C increased the strength and decreased the ductility compared with the unirradiated and aged specimens. Softening occurred in samples irradiated and tested at 450, 500, and 550 0 C

  14. Repairing fuel for reinsertion

    International Nuclear Information System (INIS)

    Krukshenk, A.

    1986-01-01

    Eqiupment for nuclear reactor fuel assembly repairing produced by Westinghouse and Brawn Bovery companies is described. Repair of failed fuel assemblies replacement of defect fuel elements gives a noticeable economical effect. Thus if the cost of a new fuel assembly is 450-500 thousand dollars, the replacement of one fuel element in it costs approximately 40-60 thousand dollars. In simple cases repairing includes either removal of failed fuel elements from a fuel assembly and its reinsertion with the rest of fuel elements into the reactor core (reactor refueling), or replacement of unfailed fuel elements from one fuel assembly to a new one (fuel assembly overhaul and reconditioning)

  15. Active control of multi-modal propagation of tonal noise in

    DEFF Research Database (Denmark)

    Laugesen, Søren

    1996-01-01

    The active control of tonal noise propagating in ducts at frequencies where many modes are able to propagate is considered. The final objective of the work reported is to cancel the prominent 450-500 Hz blade passing frequency of the rotary suction fans found in chimney stacks of power stations...... laboratory. The study of the full control system, which comprises thirty secondary sources and thirty-two error sensors, has revealed a number of new results, in particular with respect to the positioning of the sources and sensors along the length of the duct. Eventually, a reduction of the radiated power...

  16. Dynamic Diffraction Studies on the Crystallization, Phase Transformation, and Activation Energies in Anodized Titania Nanotubes

    OpenAIRE

    Hani Albetran; Victor Vega; Victor M. Prida; It-Meng Low

    2018-01-01

    The influence of calcination time on the phase transformation and crystallization kinetics of anodized titania nanotube arrays was studied using in-situ isothermal and non-isothermal synchrotron radiation diffraction from room temperature to 900 °C. Anatase first crystallized at 400 °C, while rutile crystallized at 550 °C. Isothermal heating of the anodized titania nanotubes by an increase in the calcination time at 400, 450, 500, 550, 600, and 650 °C resulted in a slight reduction in anatase...

  17. Accumulation and release of implanted hydrogen from blisters in Si during the thermal treatment

    International Nuclear Information System (INIS)

    Aleksandrov, P.A.; Baranova, E.K.; Baranova, I.V.; Budaragin, V.V.; Litvinov, V.L.

    2004-01-01

    The processes of accumulation of ion implanted hydrogen in blisters in silicon and its release during the thermal treatment at 350-1020 deg C have been studied by optical techniques. It is established that accumulation of gaseous hydrogen inside blisters takes place at temperatures lower than ∼ 450-500 deg C and is accompanied by the growth of blisters thickness and deformation of their caps. At higher temperatures hydrogen leaves cavities and dissolves in silicon. Due to internal pressure dropping the elasticity deformed top layer partially relaxes, and the blister thickness decreases. Etching of the surface layer reveals the agglomerations of small voids ( [ru

  18. Intragranular ferrite morphologies in medium carbon vanadium-microalloyed steel

    Directory of Open Access Journals (Sweden)

    Fadel A.

    2013-01-01

    Full Text Available The aim of this work was to determine TTT diagram of medium carbon V-N micro-alloyed steel with emphasis on the development of intragranular ferrite morphologies. The isothermal treatment was carried out at 350, 400, 450, 500, 550 and 600°C. These treatments were interrupted at different times in order to analyze the evolution of the microstructure. Metallographic evaluation was done using optical and scanning electron microscopy (SEM. The results show that at high temperatures (≥ 500°C polygonal intragranulary nucleated ferrite idiomorphs, combined with grain boundary ferrite and pearlite were produced and followed by an incomplete transformation phenomenon. At intermediate temperatures (450, 500°C an interloced acicular ferrite (AF microstructure is produced, and at low temperatures (400, 350°C the sheave of parallel acicular ferrite plates, similar to bainitic sheaves but intragranularly nucleated were observed. In addition to sheaf type acicular ferrite, the grain boundary nucleated bainitic sheaves are observed. [Projekat Ministartsva nauke Republike Srbije, br. OI174004

  19. Effect of composition of chlorophyll and ruthenium dyes mixture (hybrid) on the dye-sensitized solar cell performance

    Science.gov (United States)

    Pratiwi, D. D.; Nurosyid, F.; Kusumandari; Supriyanto, A.; Suryana, R.

    2018-03-01

    The fabrication of dye-sensitized solar cell (DSSC) has been conducted by varying the composition of natural dye from moss chlorophyll (Bryophyte) and synthesis dye from ruthenium complex N719. The sandwich structure of DSSC consists of the working electrode using TiO2, dye, electrolyte, and counter electrode using carbon. The composition of chlorophyll and synthesis dyes mixture were 100% and 0%, 80% and 20%, 60% and 40%, 40% and 60%, and 20% and 80%. The UV-Vis absorption spectra of moss chlorophyll showed the first peak in the wavelength range of 450-500 nm and the second peak at wavelength of 650-700 nm. The peak value of absorbance at wavelengths of 450-500 nm was 6.1004 and at wavelengths of 650-700 nm was 3.5835. The IPCE characteristic curves showed the absorption peak of photon for DSSCs occurred at wavelength of 550-650 nm. It considered that photon in this wavelength can contribute dominantly to produce the optimum electrons. The I-V characteristics of DSSCs with composition of chlorophyll and synthesis dyes mixture of 100% and 0%, 80% and 20%, 60% and 40%, 40% and 60%, and 20% and 80% resulted the efficiency of 0.0022; 0.0194; 0.0239; 0.0342; and 0.0414, respectively. It suggested that the addition of a little composition of the ruthenium complex dye into moss chlorophyll dye can increase the efficiency significantly.

  20. Analysis of reaction cross-section production in neutron induced fission reactions on uranium isotope using computer code COMPLET.

    Science.gov (United States)

    Asres, Yihunie Hibstie; Mathuthu, Manny; Birhane, Marelgn Derso

    2018-04-22

    This study provides current evidence about cross-section production processes in the theoretical and experimental results of neutron induced reaction of uranium isotope on projectile energy range of 1-100 MeV in order to improve the reliability of nuclear stimulation. In such fission reactions of 235 U within nuclear reactors, much amount of energy would be released as a product that able to satisfy the needs of energy to the world wide without polluting processes as compared to other sources. The main objective of this work is to transform a related knowledge in the neutron-induced fission reactions on 235 U through describing, analyzing and interpreting the theoretical results of the cross sections obtained from computer code COMPLET by comparing with the experimental data obtained from EXFOR. The cross section value of 235 U(n,2n) 234 U, 235 U(n,3n) 233 U, 235 U(n,γ) 236 U, 235 U(n,f) are obtained using computer code COMPLET and the corresponding experimental values were browsed by EXFOR, IAEA. The theoretical results are compared with the experimental data taken from EXFOR Data Bank. Computer code COMPLET has been used for the analysis with the same set of input parameters and the graphs were plotted by the help of spreadsheet & Origin-8 software. The quantification of uncertainties stemming from both experimental data and computer code calculation plays a significant role in the final evaluated results. The calculated results for total cross sections were compared with the experimental data taken from EXFOR in the literature, and good agreement was found between the experimental and theoretical data. This comparison of the calculated data was analyzed and interpreted with tabulation and graphical descriptions, and the results were briefly discussed within the text of this research work. Copyright © 2018 The Authors. Published by Elsevier Ltd.. All rights reserved.

  1. Improved Technique for the Determination of Uranium Minor Isotopes Concentrations in Microparticles by Using Secondary Ion Mass-Spectrometer in Multicollection Mode

    International Nuclear Information System (INIS)

    Aleshin, M.; Elantyev, I.; Stebelkov, Y.

    2015-01-01

    Traditional method of the analysis implies simultaneous measuring of secondary ion currents of isotopes 234U + , 235U + , 238U + , ions with mass 236 amu (236U + and 235UH + ) and hydride ions 238UH + by using mass-spectrometer Cameca IMS1280 in multicollection mode. Calculating of uranium isotopic composition is performed using the results of 40 successive measurements of those currents (cycles). Duration of each measurement is 8 s. Small amounts of uranium minor isotopes are limitation for precise determination of their concentrations. To prevent the damage of the secondary ions detector the intensity of ion current should be no more than 5 x 10 5 s -1 . This limitation does not allow setting a higher primary ion current for the increasing of minor uranium isotopes ions emission because of the signal of ions 238U + gets too high. New technique is developed to improve the accuracy of determination of uranium minor isotopes concentrations. Process of measurement is divided on two steps. First step is a measurement of ion currents during 20 cycles by five detectors. The second step implies the elimination of ions 238U + hitting to the detector and 10 times increasing of primary ion current. The ratio 235U/238U is calculated from the first step results, so uncertainty of determination of this value is 1.4 times bigger than with duration of 40 cycles of the measurement. The ratios 234U/235U and 236U/235U are calculated during the second step. This technique allows to determine content of 234U and 236U with 3 and 5 times less uncertainties respectively, but with different degree of the sputtering particles. Moreover the duration of each cycle was set less (1 second) to use data more efficient. The technique accordingly with every second counting provides uncertainty of determination 236U concentration 4 times less than traditional method at the same degree of sputtering particles. (author)

  2. Evaluation and compilation of fission product yields 1993

    International Nuclear Information System (INIS)

    England, T.R.; Rider, B.F.

    1995-01-01

    This document is the latest in a series of compilations of fission yield data. Fission yield measurements reported in the open literature and calculated charge distributions have been used to produce a recommended set of yields for the fission products. The original data with reference sources, and the recommended yields axe presented in tabular form. These include many nuclides which fission by neutrons at several energies. These energies include thermal energies (T), fission spectrum energies (F), 14 meV High Energy (H or HE), and spontaneous fission (S), in six sets of ten each. Set A includes U235T, U235F, U235HE, U238F, U238HE, Pu239T, Pu239F, Pu241T, U233T, Th232F. Set B includes U233F, U233HE, U236F, Pu239H, Pu240F, Pu241F, Pu242F, Th232H, Np237F, Cf252S. Set C includes U234F, U237F, Pu240H, U234HE, U236HE, Pu238F, Am241F, Am243F, Np238F, Cm242F. Set D includes Th227T, Th229T, Pa231F, Am241T, Am241H, Am242MT, Cm245T, Cf249T, Cf251T, Es254T. Set E includes Cf250S, Cm244S, Cm248S, Es253S, Fm254S, Fm255T, Fm256S, Np237H, U232T, U238S. Set F includes Cm243T, Cm246S, Cm243F, Cm244F, Cm246F, Cm248F, Pu242H, Np237T, Pu240T, and Pu242T to complete fission product yield evaluations for 60 fissioning systems in all. This report also serves as the primary documentation for the second evaluation of yields in ENDF/B-VI released in 1993

  3. Evaluation and compilation of fission product yields 1993

    Energy Technology Data Exchange (ETDEWEB)

    England, T.R.; Rider, B.F.

    1995-12-31

    This document is the latest in a series of compilations of fission yield data. Fission yield measurements reported in the open literature and calculated charge distributions have been used to produce a recommended set of yields for the fission products. The original data with reference sources, and the recommended yields axe presented in tabular form. These include many nuclides which fission by neutrons at several energies. These energies include thermal energies (T), fission spectrum energies (F), 14 meV High Energy (H or HE), and spontaneous fission (S), in six sets of ten each. Set A includes U235T, U235F, U235HE, U238F, U238HE, Pu239T, Pu239F, Pu241T, U233T, Th232F. Set B includes U233F, U233HE, U236F, Pu239H, Pu240F, Pu241F, Pu242F, Th232H, Np237F, Cf252S. Set C includes U234F, U237F, Pu240H, U234HE, U236HE, Pu238F, Am241F, Am243F, Np238F, Cm242F. Set D includes Th227T, Th229T, Pa231F, Am241T, Am241H, Am242MT, Cm245T, Cf249T, Cf251T, Es254T. Set E includes Cf250S, Cm244S, Cm248S, Es253S, Fm254S, Fm255T, Fm256S, Np237H, U232T, U238S. Set F includes Cm243T, Cm246S, Cm243F, Cm244F, Cm246F, Cm248F, Pu242H, Np237T, Pu240T, and Pu242T to complete fission product yield evaluations for 60 fissioning systems in all. This report also serves as the primary documentation for the second evaluation of yields in ENDF/B-VI released in 1993.

  4. Chemical processing of HTR fuels applying either THOREX or PUREX flow sheets

    Energy Technology Data Exchange (ETDEWEB)

    Zimmer, E; Merz, E [Kernforschungsanlage, Juelich GmbH, Institut fuer Chemische Technologie der Nuklearen Entsorgung, Juelich (Germany)

    1985-07-01

    Two fuel cycles are considered for utilization in high temperature gas-cooled reactors (HTRs): the high-enriched thorium-uranium (HEU 93% U-235) and the low-enriched uranium (LEU 8-12% U-235) fuel concept. For both fuel compositions suitable reprocessing procedures are required which are capable to separate the actinides thorium, uranium and plutonium from fission products and from each other. In any case, the processes under consideration utilize Tri-n-butylphosphate (TBP) together with a straight-chain paraffinic diluent (C{sub 8}-C{sub 14}, to day usually dodecane) as extractant in an aqueous nitrate system; most commonly, the related processes are known by the acronyms PUREX and THOREX. The PUREX process has become the reprocessing procedure quite generally used for all fuel types containing natural, slightly or highly enriched uranium together with lower or higher contents of plutonium. The THOREX process on the other hand has been developed to separate thorium, uranium and fission products from thorium based irradiated fuel. Generally, the utilization of the thorium fuel cycle is most attractive for High Temperature Reactors. On the other hand, the strong recommendation of INFCE to abandon the use of high-enriched uranium for nuclear energy applications virtually rules out the thorium fuel cycle, since economic utilization of thorium as a fertile material requires the use of high-enriched U-235. Thus, it was decided in the Federal Republic of Germany to switch over, at least for the foreseeable future, to the low enrichment uranium-plutonium fuel cycle, well aware of its economic shortcomings. In this paper various THOREX flowsheets as well as a PUREX variant suitable for LEU fuel reprocessing are described. Both processes have in common that the main stream is always presented by the fertile material, that means thorium and U-238, respectively.

  5. Depleted uranium - influence on the health and environment; Ochudobneny uran - vplyva na zdravie a zivotne prostredie

    Energy Technology Data Exchange (ETDEWEB)

    Rosskopfova, O [Katedra jadrovej chemie, Prirodovedecka fakulta, Univerzita Komenskeho, Bratislava (Slovakia)

    2002-07-01

    The uranium as radioactive element occurs in low concentrations in all components of environment. In the sample of natural uranium the isotope U-235 has the highest share (99.27 weight per cent). Chemical toxicity of uranium is comparable with toxicity of the elements like As and Pb. Depleted uranium is adjoining product in the production of enriched uranium, which is required in the production of the nuclear fuel and in the production of material used in nuclear arms. It mainly includes isotope U-238, and the content of isotopes U-235 and U-234 is sharply lowered. According to NRC depleted uranium is defined like uranium, in which percentile share of isotope U-235 is less than 0.711 weight percent. The activity of depleted uranium from viewpoint of external irradiation does not represent higher risk. Much higher risk for man represent the neurotoxic effects of uranium, which can get into human body by inhaling of dispersed particles, of contaminated dust and aerosols from atmosphere or by consumption of contaminated foodstuffs and water. Basic dangerous of irradiation by depleted uranium are mainly aerosols, which increase the probability of occurrence of lung cancer. The next dangerous is the damage of another organs like kidneys, liver and bones, where these aerosols are transported by blood like oxides from the lungs. In the environmental parts because of presence of natural uranium the depleted uranium is difficultly identifiably by standard detective methods. Thus it is necessary to use suitable radiochemical separative methods in combination with suitable detective method. (author)

  6. Neutron measurement method for the detection of transuranic elements in the nuclear fuel cycle; Neutronenmessverfahren fuer den Nachweis von Transuranen im Kernbrennstoff-Kreislauf

    Energy Technology Data Exchange (ETDEWEB)

    Sokcic-Kostic, Marina; Schultheis, Roland [NUKEM Technologies GmbH, Alzenau (Germany)

    2014-04-15

    By handling and storing burned-down fuel elements it is duty to measure the existing nuclear fuel content. For the criticality analysis of the interim storage, for instance, it is imperative to know the nuclear fuel inventory in order to give a detailed description on the safety of storage to the supervisory authority. Among other things, it is necessary to consider that the possibility of mixing up stored fuel elements in the fuel pool was not able to be excluded. Due to high penetration of matter and its origin from decay or spontaneous fission of transuranic elements neutron verification methods are suited best for the proof of fission material as long as it has been burned-down beforehand. If fission chambers are additionally used as detectors, measurements can be even carried out in environments with high gamma levels. A highly improved measuring quality can be achieved, by comparing measurement results with the results of computer-aided simulations such as e.g. burn-up programs or MCNP- calculations. Hereby the impact on the measurement result by special marginal conditions of the measuring environment (e.g. addition of boric acid into the water of the fuel pool) can be estimated and thus revised. It is shown, that the passive neutron measurement is much easier to manage as an active measurement. As restriction it is to be considered, that measurements refer essentially to transuranic elements. Uranium such as U-235, however, is difficult to detect. For fuel elements applies, that the creation of transuranic elements is directly linked to the burn-up of U-235. Hence direct conclusions to the burn-up of U-235 can be drawn, by measuring transuranic content. (orig.)

  7. Cooperation on Improved Isotopic Identification and Analysis Software for Portable, Electrically Cooled High-Resolution Gamma Spectrometry Systems Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Dreyer, Jonathan G. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Wang, Tzu-Fang [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Vo, Duc T. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Funk, Pierre F. [Inst. for Radiological Protection and Nuclear Safety (IRSN), Fontenay-aux-Roses (France); Weber, Anne-Laure [Inst. for Radiological Protection and Nuclear Safety (IRSN), Fontenay-aux-Roses (France)

    2017-07-20

    Under a 2006 agreement between the Department of Energy (DOE) of the United States of America and the Institut de Radioprotection et de Sûreté Nucléaire (IRSN) of France, the National Nuclear Security Administration (NNSA) within DOE and IRSN initiated a collaboration to improve isotopic identification and analysis of nuclear material [i.e., plutonium (Pu) and uranium (U)]. The specific aim of the collaborative project was to develop new versions of two types of isotopic identification and analysis software: (1) the fixed-energy response-function analysis for multiple energies (FRAM) codes and (2) multi-group analysis (MGA) codes. The project is entitled Action Sheet 4 – Cooperation on Improved Isotopic Identification and Analysis Software for Portable, Electrically Cooled, High-Resolution Gamma Spectrometry Systems (Action Sheet 4). FRAM and MGA/U235HI are software codes used to analyze isotopic ratios of U and Pu. FRAM is an application that uses parameter sets for the analysis of U or Pu. MGA and U235HI are two separate applications that analyze Pu or U, respectively. They have traditionally been used by safeguards practitioners to analyze gamma spectra acquired with high-resolution gamma spectrometry (HRGS) systems that are cooled by liquid nitrogen. However, it was discovered that these analysis programs were not as accurate when used on spectra acquired with a newer generation of more portable, electrically cooled HRGS (ECHRGS) systems. In response to this need, DOE/NNSA and IRSN collaborated to update the FRAM and U235HI codes to improve their performance with newer ECHRGS systems. Lawrence Livermore National Laboratory (LLNL) and Los Alamos National Laboratory (LANL) performed this work for DOE/NNSA.

  8. Review of international solutions to NEACRP benchmark BWR lattice cell problems

    International Nuclear Information System (INIS)

    Halsall, M.J.

    1977-12-01

    This paper summarises international solutions to a set of BWR benchmark problems. The problems, posed as an activity sponsored by the Nuclear Energy Agency Committee on Reactor Physics, were as follows: 9-pin supercell with central burnable poison pin, mini-BWR with 4 pin-cells and water gaps and control rod cruciform, full 7 x 7 pin BWR lattice cell with differential U 235 enrichment, and full 8 x 8 pin BWR lattice cell with water-hole, Pu-loading, burnable poison, and homogenised cruciform control rod. Solutions have been contributed by Denmark, Japan, Sweden, Switzerland and the UK. (author)

  9. LEU fuel element produced by the Egyptian fuel manufacturing pilot plant

    International Nuclear Information System (INIS)

    Zidan, W.I.

    2000-01-01

    The Egyptian Fuel Manufacturing Pilot Plant, FMPP, is a Material Testing Reactor type (MTR) fuel element facility, for producing the specified fuel elements required for the Egyptian Second Research Reactor, ETRR-2. The plant uses uranium hexafluoride (UF 6 , 19.75% U 235 by wt) as a raw material which is processed through a series of the manufacturing, inspection and test plan to produce the final specified fuel elements. Radiological safety aspects during design, construction, operation, and all reasonably accepted steps should be taken to prevent or reduce the chance of accidents occurrence. (author)

  10. Natural radioisotopes. The ''atomic clock'' for the age determination of rocks and archeological discoveries; Natuerliche Radioisotope. Die ''Atomuhr'' fuer die Bestimmung des absoluten Alters von Gesteinen und archaeologischen Funden

    Energy Technology Data Exchange (ETDEWEB)

    Heuel-Fabianek, Burkhard [Forschungszentrum Juelich (Germany)

    2017-04-01

    The contribution describes the fundamentals of radiometric age determination based on natural radionuclides. Organic (carbon containing) materials can be dated up to an age of 60.000 years using C-14. The methods used for radiometric dating of rocks and minerals include the radioactive decay series of U-238, U-235, Th -232, but also the beta decay of Rb-87 to Sr-87 or K-40 to Ar-40. The absolute age of rocks is not necessarily identical with the radiometric dating result, since geological processes could influence the radionuclide ratio.

  11. α spectrum analysis technology research on uranium in environmental water

    International Nuclear Information System (INIS)

    Qiu, Yongmei; Yang, Yong; Ma, Junge

    2009-04-01

    In order to measure the nuclide abundance ratio of uranium in environmental water, the method of '717 anion exchanging resin' is discussed. The dis- traction circuit is determined by 717 anion exchange leaching curve, recovery ratio of anion exchaging, recovery ratio of former disposal and recovery ratio of electrodeposit. The circuit has good result in distracting and enriching uranium by using '717 anion exchanging resin', the resolution of uranium in the spectrum is perfect. The activities and the nuclide abundance ratios of 238 U, 235 U, 234 U in the different reach of some location of INPC have been gained. (authors)

  12. Sensitivity and uncertainty analysis for fission product decay heat calculations

    International Nuclear Information System (INIS)

    Rebah, J.; Lee, Y.K.; Nimal, J.C.; Nimal, B.; Luneville, L.; Duchemin, B.

    1994-01-01

    The calculated uncertainty in decay heat due to the uncertainty in basic nuclear data given in the CEA86 Library, is presented. Uncertainties in summation calculation arise from several sources: fission product yields, half-lives and average decay energies. The correlation between basic data is taken into account. The uncertainty analysis were obtained for thermal-neutron-induced fission of U235 and Pu239 in the case of burst fission and irradiation time. The calculated decay heat in this study is compared with experimental results and with new calculation using the JEF2 Library. (from authors) 6 figs., 19 refs

  13. Prompt neutron fission spectrum mean energies for the fissile nuclides and 252Cf

    International Nuclear Information System (INIS)

    Holden, N.E.

    1985-01-01

    The international standard for a neutron spectrum is that produced from the spontaneous fission of 252 Cf, while the thermal neutron induced fission neutron spectra for the four fissile nuclides, 233 U, 235 U, 239 Pu, and 241 Pu are of interest from the standpoint of nuclear reactors. The average neutron energies of these spectra are tabulated. The individual measurements are recorded with the neutron energy range measured, the method of detection as well as the average neutron energy for each author. Also tabulated are the measurements of the ratio of mean energies for pairs of fission neutron spectra. 75 refs., 9 tabs

  14. Fission fragment yields and total kinetic energy release in neutron-induced fission of235,238U,and239Pu

    Science.gov (United States)

    Tovesson, F.; Duke, D.; Geppert-Kleinrath, V.; Manning, B.; Mayorov, D.; Mosby, S.; Schmitt, K.

    2018-03-01

    Different aspects of the nuclear fission process have been studied at Los Alamos Neutron Science Center (LANSCE) using various instruments and experimental techniques. Properties of the fragments emitted in fission have been investigated using Frisch-grid ionization chambers, a Time Projection Chamber (TPC), and the SPIDER instrument which employs the 2v-2E method. These instruments and experimental techniques have been used to determine fission product mass yields, the energy dependent total kinetic energy (TKE) release, and anisotropy in neutron-induced fission of U-235, U-238 and Pu-239.

  15. Neutron cross sections for uranium-235 (ENDF/B-IV Release 3)

    International Nuclear Information System (INIS)

    Lubitz, C.R.

    1996-09-01

    The resonance parameters in ENDF6 (Release 2) U235 were adjusted to make the average capture and fission cross sections below 900 eV agree with selected differential capture and fission measurements. The measurements chosen were the higher of the credible capture measurements and the lower of the fission results, yielding a higher epithermal alpha. In addition, the 2200 m/s cross sections were adjusted to obtain agreement with the integral value of K1. As a result, criticality calculations for thermal benchmarks, and agreement with a variety of integral parameters, are improved

  16. The hazards of nuclear power: myth or reality

    International Nuclear Information System (INIS)

    Franklin, N.L.

    1985-01-01

    The subject is discussed under the headings: introduction (social problems of modern large scale technology); anxieties concerning nuclear power (biological radiation effects; background radiation; correlation between dose and cancer mortality; difficulty of arriving at reliable conclusions); discharge policy at Sellafield reprocessing plant (radioactive waste discharged to the sea - 'critical groups' - reduction of discharge and cost); reactor accidents (coolant failure and effects; protective systems; discussion of cases; improving safety, and the cost); proliferation of nuclear weapons (U235 - description of enrichment methods; tritium for thermonuclear weapons; Pu; attempts to prevent proliferation); consequences (uncertainties; activities against nuclear power; radioactive waste disposal; social and political aspects). (U.K.)

  17. Disposal criticality analysis for the ceramic waste form from the ANL electrometallurgical treatment process - Internal configurations

    International Nuclear Information System (INIS)

    Lell, R. M.; Agrawal, R.; Morris, E. E.

    2000-01-01

    Criticality safety issues for disposal of the ANL ceramic waste were examined for configurations within the waste package. Co-disposal of ceramic waste and DOE spent fuel is discussed briefly; co-disposal of ANL ceramic and metal wastes is examined in detail. Calculations indicate that no significant potential for criticality exists until essentially all of the important neutron absorbers are flushed from the degraded ceramic waste. Even if all of the neutron absorbers are removed from the ceramic waste rubble, the package remains far subcritical if the blended salts used in ceramic waste production have an initial U-235 enrichment below 40%

  18. Uranium isotope ratio measurements in field settings

    International Nuclear Information System (INIS)

    Shaw, R.W.; Barshick, C.M.; Young, J.P.; Ramsey, J.M.

    1997-01-01

    The authors have developed a technique for uranium isotope ratio measurements of powder samples in field settings. Such a method will be invaluable for environmental studies, radioactive waste operations, and decommissioning and decontamination operations. Immediate field data can help guide an ongoing sampling campaign. The measurement encompasses glow discharge sputtering from pressed sample hollow cathodes, high resolution laser spectroscopy using conveniently tunable diode lasers, and optogalvanic detection. At 10% 235 U enrichment and above, the measurement precision for 235 U/( 235 U+ 238 U) isotope ratios was ±3%; it declined to ±15% for 0.3% (i.e., depleted) samples. A prototype instrument was constructed and is described

  19. Calculation of multigroup constants in WIMS format with programs fedgroup and flange and comparison of the results obtained using different evaluated libraries

    International Nuclear Information System (INIS)

    Trkov, A.; Budnar, M.; Copic, M.; Perdan, A.; Ravnik, M.

    1982-01-01

    Multigroup constants for 1-H-1, 92-U-235, and 92-U-238 have been calculated. Averaged cross-sections and other constants have been prepared in the WIMS 69-group format. Comparison has been made between group constants obtained with several evaluated libraries (KEDAK-3 1975, 1979, ENDF/B-4, ENDF/B-5) and the WIMS-D library. Observed differences are most pronounced in the resonance and fast region. From test runs on fuel cell with the WIMS program it can be deduced that these differences affect the fewgroup constants significantly. (author)

  20. PARR-2: reactor description and experiments

    International Nuclear Information System (INIS)

    Wyne, M.F.; Meghji, J.H.

    1990-12-01

    PARR-2 is a miniature neutron source reactor (MNSR) research reactor has been designed at the rate of 27 kW. Reactor assembly comprises of peaking characteristics with a self limiting flux. In this report reactor description with its assembly and instrumentation control system has been explained. The reactor engineering and physics experiments which can be performed on this reactor are explained in this report. PARR-2 is fueled with HEU fuel pins which are about 90% enriched in U-235. Specific requirements for the safety of the reactor, its building and the personnel, normal instrumentation as required in an industrial environment is sufficient. (A.B.)

  1. Preliminary development of thermal nuclear cell homogenization code

    International Nuclear Information System (INIS)

    Su'ud, Z.; Shafii, M. A.; Yudha, S. P.; Waris, A.; Rijal, K.

    2012-01-01

    Nuclear fuel cell homogenization for thermal reactors usually include three main parts, i.e., fast energy resonance part which usually adopt narrow resonance approximation to treat the resonance, low (intermediate) energy region in which the resonance can not be treated accurately using NR approximation and therefore we should use intermediate resonance treatment, and thermal energy region (very low) in which the effect of thermal must be treated properly. In n this study the application of the intermediate resonance approximation treatment for low energy nuclear resonance is discussed. The method is iterative based. As a sample the method is applied in U-235 low lying resonance and the result is presented and discussed.

  2. Modeling of the CTEx subcritical unit using MCNPX code

    International Nuclear Information System (INIS)

    Santos, Avelino; Silva, Ademir X. da; Rebello, Wilson F.; Cunha, Victor L. Lassance

    2011-01-01

    The present work aims at simulating the subcritical unit of Army Technology Center (CTEx) namely ARGUS pile (subcritical uranium-graphite arrangement) by using the computational code MCNPX. Once such modeling is finished, it could be used in k-effective calculations for systems using natural uranium as fuel, for instance. ARGUS is a subcritical assembly which uses reactor-grade graphite as moderator of fission neutrons and metallic uranium fuel rods with aluminum cladding. The pile is driven by an Am-Be spontaneous neutron source. In order to achieve a higher value for k eff , a higher concentration of U235 can be proposed, provided it safely remains below one. (author)

  3. Calculation of neutronic parameters of IEA-R1 reactor and purpose of a new configuration

    International Nuclear Information System (INIS)

    Kosaka, N.; Fanaro, L.C.C.B.; Yamaguchi, M.

    1989-01-01

    The program for reducing the fuel enrichment of the IEA-R1 reactor considers fuel plates containing U308-AL with 19,9% of U-235. The geometry of the new 18 fuel plate fuel elements has been kept the same. This work describes the calculation methods utilized at IPEN-CNEN/SP and some neutronic parameters of the present configuration of IEA-R1 as well as for a new configuration porposed with a new LEU fuel element are shown. (author) [pt

  4. Method to separate isotopes

    International Nuclear Information System (INIS)

    Coenen, H.; Neuschuetz, D.

    1980-01-01

    An extraction by means of supercritical pure gases (e.g. CD 2 ) or gas mixtures is proposed to separate isotopes, especially H/D, Li-6/Li-7, and U-235/U-238, and water or benzol is used as entrainer. The extraction shall be carried out at pressure rates of about 350 bar above the critical pressure with the temperature being by up to 100 0 C above the critical temperature. A NaCl-solution and the change of the isotopic ratio Cl-35/Cl-37 are investigated for example purpose. (UWI) [de

  5. The Oklo reactors

    International Nuclear Information System (INIS)

    Skytte Jensen, B.

    1982-01-01

    The Oklo reactors comprise up to nine 235-U depleted zones in an uranium ore in the Republic of Gabon in West Africa. The depletion in fissile U-235 has been proved to have caused by nuclear chain reactions. The study of the Oklo phenomenon indicates that very efficient retardation mechanisms may operate in nature - at least under special conditions. A closer study of these processes ought to be made to establish the limitations to their occurrence. The Oklo sandstone formation today would probably be considered unacceptable as a host rock for a repository. (EG)

  6. Minimum critical masses for uranium at the Portsmouth Gaseous Diffusion Plant

    International Nuclear Information System (INIS)

    Tayloe, R.W. Jr.; Davis, T.C.

    1994-06-01

    This report presents a tabulation of safe masses and minimum critical masses for uranium (U). These minimum critical mass and safe mass tables were obtained by interpolating between the values reported in the literature to obtain values as a function of enrichment within the 1.5 percent to 100 percent range. Equivalent mass values for uranium-235 (U 235 ), uranium hexafluoride (UF 6 ), and uranyl fluoride (UO 2 F 2 ) have been generated from the safe mass and minimum critical masses for uranium

  7. Analysis of Doppler effect measurement in FCA cores using JENDL-3.2 library

    International Nuclear Information System (INIS)

    Okajima, Shigeaki

    1996-01-01

    For the evaluation of the calculation accuracy of the 238 U Doppler effect using JENDL-3.2 library, the previously measured Doppler reactivity worths in the FCA were systematically analyzed. In the analysis the Doppler reactivity worth was calculated by a first order perturbation theory. The calculated results were compared with those using JENDL-3.1 library. The JENDL-3.2 calculation in MOX fuel mock-up cores agrees well with the experimental values within the experimental error. In U-235/Pu fuel cores, the JENDL-3.2 calculation gives 12-15% larger Doppler reactivity worths than the JENDL-3.1 calculation. (author)

  8. Nuclear data for actinide production and depletion calculations

    International Nuclear Information System (INIS)

    Benjamin, R.W.

    1978-01-01

    The status of nuclear cross section data required for actinide depletion calculations in thermal reactors is summarized, and recommendations are made for future work. The primary fertile and fissile nuclides ( 232 Th, 233 U, 235 U, 238 U, and 239 Pu) are not reviewed. Nuclear data for the transactinium mass region are, with few exceptions, reasonably complete and adequate for current thermal-reactor depletion calculations. There is a real need, however, for well-documented reactor production studies to use as benchmarks for data testing. 3 figures, 6 tables

  9. Chapter 3: Exponential experiments on graphite-moderated lattices fuelled with near-natural uranium metal rods

    International Nuclear Information System (INIS)

    McCulloch, D.B.; Clarke, W.G.; Ashworth, F.P.O.; Hoskins, T.A.

    1963-01-01

    Exponential experiments have been carried out on graphite lattices fuelled by 1.2 in. diameter uranium metal rods at three near-natural U 235 compositions, 0.6 Co, 1.3 Co and 1.6 Co. The results, together with those already existing from earlier exponential or critical measurements on these and similar natural uranium rods, have been correlated with the theory of Syrett (1961) and also with the modified form of this theory given in Vol.1, Ch. 7. (author)

  10. Test calculations of photoneutrons emission from surface of uranium sphere irradiated by 28 MeV electrons

    International Nuclear Information System (INIS)

    Blokhin, A.I.; Degtyarev, I.I.

    2002-01-01

    In this paper the results of physical verification for the BOFOD photonuclear data files are reported, available for the uranium isotopes U 235 , U 238 . These results were compared with calculated data by the parameterization driven model of photonuclear reaction and experimental data. Experimental data of photoneutron yields from surface of uranium sphere irradiated by 28 MeV electrons are used for a verification. Both calculations have been carried out with the RTS and T general purpose Monte Carlo code with detailed electron-photon-nucleon transport simulation using the ENDF/B-VI and EPDL evaluated data libraries

  11. ENDF/B-V 7 Standards Data File (EN5-ST Library)

    International Nuclear Information System (INIS)

    DayDay, N.; Lemmel, H.D.

    1980-10-01

    This document summarizes the contents and documentation of the ENDF/B-V 7 Standards Data File (EN5-ST Library) released in September 1979. The library contains complete evaluations for all significant neutron reactions in the energy range 10 -5 eV to 20 MeV for H-1, He-3, Li-6, B-10, C-12, Au-197 and U-235 isotopes. The entire library or selective retrievals from it can be obtained free of charge from the IAEA Nuclear Data Section. (author)

  12. Hamor-2: a computer code for LWR inventory calculation

    International Nuclear Information System (INIS)

    Guimaraes, L.N.F.; Marzo, M.A.S.

    1985-01-01

    A method for calculating the accuracy inventory of LWR reactors is presented. This method uses the Hamor-2 computer code. Hamor-2 is obtained from the coupling of two other computer codes Hammer-Techion and Origen-2 for testing Hamor-2, its results were compared to concentration values measured from activides of two PWR reactors; Kernkraftwerk Obrighein (KWO) and H.B. Robinson (HBR). These actinides are U 235 , U 236 , U 238 , Pu 239 , Pu 241 and PU 242 . The computer code Hammor-2 shows better results than the computer code Origem-2, when both are compared with experimental results. (E.G.) [pt

  13. Uranium enrichment

    International Nuclear Information System (INIS)

    Mohrhauer, H.

    1982-01-01

    The separation of uranium isotopes in order to enrich the fuel for light water reactors with the light isotope U-235 is an important part of the nuclear fuel cycle. After the basic principals of isotope separation the gaseous diffusion and the centrifuge process are explained. Both these techniques are employed on an industrial scale. In addition a short review is given on other enrichment techniques which have been demonstrated at least on a laboratory scale. After some remarks on the present situation on the enrichment market the progress in the development and the industrial exploitation of the gas centrifuge process by the trinational Urenco-Centec organisation is presented. (orig.)

  14. A certified reference material for radionuclides in the water sample from Irish Sea (IAEA-443)

    DEFF Research Database (Denmark)

    Pham, M.K.; Betti, M.; Povinec, P.P.

    2011-01-01

    A new certified reference material (CRM) for radionuclides in sea water from the Irish sea (IAEA-443) is described and the results of the certification process are presented. Ten radionuclides (3H, 40K, 90Sr, 137Cs, 234U, 235U, 238U, 238Pu, 239+240Pu and 241Am) have been certified, and information...... values on massic activities with 95% confidence intervals are given for four radionuclides (230Th, 232Th, 239Pu and 240Pu). Results for less frequently reported radionuclides (99Tc, 228Th, 237Np and 241Pu) are also reported. The CRM can be used for quality assurance/quality control of the analysis...

  15. Certified reference material for radionuclides in fish flesh sample IAEA-414 (mixed fish from the Irish Sea and North Sea)

    DEFF Research Database (Denmark)

    Pham, M.K.; Sanchez-Cabeza, J.A.; Povinec, P.P.

    2006-01-01

    A certified reference material (CRM) for radionuclides in fish sample IAEA-414 (mixed fish from the Irish Sea and North Seas) is described and the results of the certification process are presented. Nine radionuclides (K-40, Cs-137, Th-232, U-234, U-235, U-238, Pu-238, Pu239+240 and Am-241) were...... ratios are also included. The CRM can be used for quality assurance/quality control of the analysis of radionuclides in fish sample, for the development and validation of analytical methods and for training purposes. The material is available from IAEA, Vienna, in 100 g units. (c) 2006 Elsevier Ltd. All...

  16. Research on magnet replicas and the very incomplete Meissner effect. Final technical report, January 15, 1991--January 15, 1994

    International Nuclear Information System (INIS)

    Weinstein, R.

    1994-08-01

    A wide variety of temperature/processing changes and high energy irradiation has substantially increased B t,max . This plus variation of the HTS mix to Y 1.7 Ba 2 Cu 3 O 7 Pt 0.01 U 235 (25 ppM) has increased trapped field by a factor of 8 and promises an additional factor of 2. The original goal, to produce 10,000 Gauss of trapped field, has been exceeded by a factor of 7. Various applications, including motors, generators, levitating bearings, flywheels, magnetic bumpers, and MHD propulsion are in progress at various labs and industries

  17. Diode laser excited optogalvanic spectroscopy of glow discharges

    International Nuclear Information System (INIS)

    Barshick, C.M.; Shaw, R.W.; Jennings, L.W.; Post-Zwicker, A.; Young, J.P.; Ramsey, J.M.

    1997-01-01

    The development of diode-laser-excited isotopically-selective optogalvanic spectroscopy (OGS) of uranium metal, oxide and fluoride in a glow discharge (GD) is presented. The technique is useful for determining 235 U/( 235 U+ 238 U) isotope ratios in these samples. The precision and accuracy of this determination is evaluated, and a study of experimental parameters pertaining to optimization of the measurement is discussed. Application of GD-OGS to other f-transition elements is also described. copyright 1997 American Institute of Physics

  18. Uranium content of petroleum by Fission track technique

    International Nuclear Information System (INIS)

    Paschaa, A.S.; Mafra, O.Y.; Oliveira, C.A.N.; Pinto, L.R.

    1982-01-01

    This paper examines the feasibility of the fission track registration technique to investigate the natural uranium concentration in petroleum. The application is briefly described and the results obtained indicate the presence of uranium concentrations in samples of Brazilian petroleum which are over the detection limit of the fission track technique. The irradiations were performed by using fluxes with predominance of thermal neutrons, which have a fission cross-section for U 235 equal to 579 barns. Since the neutron fluxes were not comp sed exclusively of thermal neutrons, fissions from fast neutrons would also be taken into account for U 238 and Th 232

  19. Power maximization of a spheric reflected reactor with optimized fuel distribution

    International Nuclear Information System (INIS)

    Reade, Joamar Rodrigues Vincent

    1979-01-01

    The maximum power of a spheric reflected reactor was determined using the theory of optimal control. The control variable employed was the fuel distribution, in accordance to constraints on the power density and on the concentration fuel. It was considered a thermal reactor with a fixed radius. The reactor was fuelled with U-235 and moderated with light water. The nuclear reactor was described by a diffusion theory model. The analytical solution was obtained for both two and four groups of energy and a FORTRAN program was developed to obtain the numerical results. (author)

  20. Classification of soil samples according to their geographic origin using gamma-ray spectrometry and principal component analysis

    International Nuclear Information System (INIS)

    Dragovic, Snezana; Onjia, Antonije

    2006-01-01

    A principal component analysis (PCA) was used for classification of soil samples from different locations in Serbia and Montenegro. Based on activities of radionuclides ( 226 Ra, 238 U, 235 U, 4 K, 134 Cs, 137 Cs, 232 Th and 7 Be) detected by gamma-ray spectrometry, the classification of soils according to their geographical origin was performed. Application of PCA to our experimental data resulted in satisfactory classification rate (86.0% correctly classified samples). The obtained results indicate that gamma-ray spectrometry in conjunction with PCA is a viable tool for soil classification

  1. Startup of Torrey Pines Mark III and Puerto Rico Nuclear Center reactors with TRIGA-FLIP fuel

    Energy Technology Data Exchange (ETDEWEB)

    Chesworth, R. H. [Gulf E and ES, San Diego, CA (United States)

    1972-07-01

    This paper discusses the characteristics of TRIGA FLIP cores in two different geometries: the normal TRIGA single-rod geometry as typified by the installation in the Torrey Pines Mark III reactor; and the four-rod cluster geometry as typified by the conversion core installed in the Puerto Rico Nuclear Center reactor at Mayaguez. In both reactors the fuel is 8-1/2 wt % uranium, 70% enriched in U-235. The hydrogen to zirconium atom ratio is 1.5 to 1.65 and the cladding material is stainless steel. The basic neutronic characteristics of the fuel in both reactor installations are briefly discussed.

  2. Analytical and numerical comparisons of the α parameter obtained from several expressions versus reactivity

    International Nuclear Information System (INIS)

    Minguez, E.

    1983-01-01

    A review of the analytical expressions between α-parameter and reactivity (rho) has been done in this article. Finally, several lineal approximations have been obtained for two important points of values. The analytical expansion has been made using only one family of precursors from delayed neutrons; diffusion equation in one energy groups; neutron flux separability with known spatial function. Numerical results have been obtained using typical data from fast and thermal reactors; considering and homogeneous media of Pu 239 in the first case, and U-235 in the second one. (author)

  3. Non-destructive isotopic uranium assay by multiple delayed neutron measurements

    International Nuclear Information System (INIS)

    Papadopoulos, N.N.; Tsagas, N.F.

    1991-01-01

    The high accuracy and precision required in nuclear safeguards measurements can be achieved by an improved neutron activation technique based on multiple delayed fission neutron counting under various experimental conditions. For the necessary ultrahigh counting statistics required, cyclic activation of multiple subsamples has been applied. The home-made automated flexible analytical system with neutron flux and spectrum differentiation by irradiation position adjustment and cadmium screening, permits the non-destructive determination of the U235 abundance and the total U element concentration needed in nuclear safeguards sample analysis, with a high throughout and a low operational cost. Careful experimental optimization led to considerable improvement of the results

  4. Uranium isotopes in El hamraween harbour sediments

    International Nuclear Information System (INIS)

    Salahel Din, K.

    2009-01-01

    Isotopes of uranium in marine sediments collected from El Hamraween harbour and Ras El-Bhar areas on the Egyptian coast of the Red Sea have been studied using radiochemical separation procedures and alpha-particle spectrometry. Activity concentrations of 238 U, 235 U, 234 U were calculated. The activities observed indicating the enhancement of radioactivity level in El Hamraween harbor area due to the activities of phosphate shipment operation. Secular equilibrium between 234 U and 238 U was found in the analyzed samples. The average activity ratio of 235 U/ 238 U was close to the value 0.046 for uranium in nature

  5. Preliminary studies for the upgrading of the ININ TRIGA MARK III reactor

    International Nuclear Information System (INIS)

    Aguilar, J.V.

    1984-01-01

    An increase on the demand of services has led to the -consideration of reactor upgrading, for which a project has been proposed. This project includes the modification of the reactor core, using FLIP type (Erbium loaded, 70% U-235 enriched) fuel elements and the upgrading of the control - console which implies a careful analysis of the Safety -considerations involved in the system modification. In this paper, a brief description is given of: The modifications intended; Pre-operational tests for the system components to be modified; The safety report needed for the licensing of the upgraded reactor; The project management and evaluation

  6. Analysis of di-calcic phosphate from the Erkimia plant in Flix, Catalonia (Spain)

    International Nuclear Information System (INIS)

    2006-01-01

    As an investigation showed that solid wastes released in the Ebra River between 1978 and 1988 from the Erkimia factory displayed high natural radionuclides contents, this document reports analysis performed on the di-calcic phosphate produced by this factory (from Moroccan phosphorite). It indicates and comments the results obtained by gamma spectrometry for some radionuclides belonging to the U 238 and U 235 chains (uranium 238, thorium 230 and 234, radium 226, lead 210, Uranium 235). It comments the associated potential radiological risks, and gives some recommendations to limit these risks

  7. Nuclear design report for Yonggwang nuclear power plant unit 2 cycle 7

    International Nuclear Information System (INIS)

    Zee, Sung Kyun; Choi, Gyoo Hwan; Lee, Ki Bog; Park, Sang Yoon

    1993-02-01

    This report presents nuclear design calculations for Cycle 7 of Yonggwang Unit 2. Information is given on fuel loading, power density distributions, reactivity coefficients, control rod worths and operational limits. In addition, the report contains all necessary data for the startup tests including predicted values for the comparison with the measured data. The reload consists of 64 KOFA's enriched by nominally 3.70 w/o U235. Among the KOFA's, 40 fuel assemblies contain gadolinia rods. The fuel assemblies in the core are arranged in a low leakage loading pattern. The cycle length of Cycle 7 amounts to 367 EFPD corresponding to a cycle burnup of 14770 MWD/MTU. (Author)

  8. Hazardous waste operational plan for site 300

    International Nuclear Information System (INIS)

    Roberts, R.S.

    1982-01-01

    This plan outlines the procedures and operations used at LLNL's Site 300 for the management of the hazardous waste generated. This waste consists primarily of depleted uranium (a by-product of U-235 enrichment), beryllium, small quantities of analytical chemicals, industrial type waste such as solvents, cleaning acids, photographic chemicals, etc., and explosives. This plan details the operations generating this waste, the proper handling of this material and the procedures used to treat or dispose of the hazardous waste. A considerable amount of information found in this plan was extracted from the Site 300 Safety and Operational Manual written by Site 300 Facility personnel and the Hazards Control Department

  9. TRIGA high wt -% LEU fuel development program. Final report

    International Nuclear Information System (INIS)

    West, G.B.

    1980-07-01

    The principal purpose of this work was to investigate the characteristics of TRIGA fuel where the contained U-235 was in a relatively high weight percent (wt %) of LEU (low enriched uranium - enrichment of less than 20%) rather than a relatively low weight percent of HEU (high enriched uranium). Fuel with up to 45 wt % U was fabricated and found to be acceptable after metallurgical examinations, fission product retention tests and physical property examinations. Design and safety analysis studies also indicated acceptable prompt negative temperature coefficient and core lifetime characteristics for these fuels

  10. Migration modelling of different plutonium chemical forms through porous media

    International Nuclear Information System (INIS)

    Saltelli, A.

    1979-01-01

    Two solutions of the migration equations are described. The first relates to the transport equations for the decay chain Am 243→Pu 239→U 235. Numerical integration was performed in this case by a simulation code written in CSMP III language and plutonium is considered to be all in the same chemical form. The second case relates to the problem of Pu speciation and migration. The decay chain Pu 240→U 236 is considered and numerical integration is performed by a modified version of Bo code COLUMN. Pseudo first order reactions are supposed to act between Pu states to maintain equilibrium during the migration

  11. Activity report on research and development work 1980 of the institute for Nuclear Process Technology

    International Nuclear Information System (INIS)

    1981-02-01

    Within the framework of guaranteeing supplies of nuclear fuel for the generation of nuclear energy, the KfK developed the separation nozzle method for the enrichment of U-235. It is based on partial separation of this uranium isotope by a deflected jet from a mixture of UF 6 /hydrogen. This method is now being applied on a technical scale for the first time in Brazil within the framework of an agreement concluded between the Federal Republic of Germany and Brazil and approved by the International Atomic Energy Agency. (orig./EF) [de

  12. Study Of Thorium As A Nuclear Fuel.

    Directory of Open Access Journals (Sweden)

    Prakash Humane

    2017-10-01

    Full Text Available Conventional fuel sources for power generation are to be replacing by nuclear power sources like nuclear fuel Uranium. But Uranium-235 is the only fissile fuel which is in 0.72 found in nature as an isotope of Uranium-238. U-238 is abundant in nature which is not fissile while U-239 by alpha decay naturally converted to Uranium- 235. For accompanying this nuclear fuel there is another nuclear fuel Thorium is present in nature is abundant can be used as nuclear fuel and is as much as safe and portable like U-235.

  13. Flux stability and power control in the Soviet RBMK-1000 reactors

    International Nuclear Information System (INIS)

    Meriwether, G.H.; McNeece, J.P.

    1993-08-01

    As a result of the Chernobyl accident, the Soviets have studied and implemented various design changes to improve the safety of the RBMK reactors. The safety measurements include modifications of the control rod configuration, fuel enrichment increase from 2.0 to 2.4 weight percent U-235, and installation of additional supplemental absorbers. The purpose of this study is to investigate the effects of increased fuel enrichment, different control rod positions, and supplemental absorber loadings on reactivity control, power distribution within the large RBMK core, and relative stability against power oscillations

  14. Uranium-series dating of Quaternary deposits

    International Nuclear Information System (INIS)

    Schwarcz, H.; Gascoyne, M.

    1984-01-01

    In view of the interest in the problem of time scales in geomorphology it is fortunate that there exists a number of geochronometers applicable to the measurement of the age of such young deposits. This paper is specifically devoted to those which arise from the disequilibrium between the daughter isotopes of U-238 and U-235, and their respective parents. The authors describe applications to Quaternary continental deposits that can give information about climatic change (travertine, lacrustine limestones, pedogenic carbonates, detrinal sediments, volcanic rocks). (Auth.)

  15. Fissility of actinide nuclei induced by 60-130 MeV photons

    International Nuclear Information System (INIS)

    Morcelle, Viviane; Tavares, Odilon A.P.

    2004-06-01

    Nuclear fissilities obtained from recent photofission reaction cross section measurements carried out at Saskatchewan Accelerator Laboratory (Saskatoon, Canada) in the energy range 60-130 MeV for 232 Th, 233 U, 235 U, 238 U, and 237 Np nuclei have been analysed in a systematic way. To this aim, a semiempirical approach has been developed based on the quasi-deuteron nuclear photoabsorption model followed by the process of competition between neutron evaporation and fission for the excited nucleus. The study reproduces satisfactorily well the increasing trend of nuclear fissility with parameter Z 2 =A. (author)

  16. Burn-up calculations for a thorium HTR with one and with two types of fuel particle

    Energy Technology Data Exchange (ETDEWEB)

    Griggs, C. F.

    1975-06-15

    Cell burn-up calculations have been made on a thorium pin-cell operating with one or with two types of particle. With one particle, the input thorium and uranium are mixed prior to irradiation and all discharged uranium is recycled. With two particles, the fuel is kept in two streams and only the uranium generated from thorium is recycled. The two models are found to give similar power generations from a given initial U-235 input. The choice between the two types of particle is probably not determined by reactor physics considerations but by the value of the fuel credits and by the cost of fuel fabrication and reprocessing.

  17. Energy dependence of the neutron multiplicity P/sub nu/ in fast neutron induced fission of /sup 235,238/U and 239Pu

    International Nuclear Information System (INIS)

    Zucker, M.S.; Holden, N.E.

    1986-01-01

    Certain applications require knowledge of the higher moments of the neutron multiplicity probability. It can be shown that the second factorial moment is proportional to the fission rate in the sample, and that the third factorial moment can be of use in disentangling spontaneous fission from induced fission. Using a source of unpublished work in which neutron multiplicities were derived for the fast neutron induced fission of U-235, U-238, and Pu-239, the multiplicity probability has been calculated as a function of neutron energy for the energy range 0 to 10 MeV

  18. Photoneutron and Photonuclear Cross Sections According to Packed cluster Model

    International Nuclear Information System (INIS)

    El-Mekkawi, L.S.; El-Bakty, O.M.

    1998-01-01

    Photonuclear gross sections have been estimated for 232 Th, 237 Np, 239 Pu, 233 U, 234 U, 235 U, 238 U in the energy range from threshold up to 20 MeV, by perturbation balance in Packed Cluster. The Packed Cluster (gamma, f) and (gamma, n) cross sections require complete absence of any (gamma,2n) or (gamma,nf) cross sections for 233 U and 234 U as in experiment. It also explains the early (gamma,n) and gamma,nf) reactions in 235 U

  19. Physics experiment on the Dragon reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hunt, C.

    1974-10-15

    The paper describes a set of DRAGON experiments planned to measure burn-up effects in DRAGON irradiated fuel. Irradiated fuel elements from DRAGON are to be subjected to reactivity measurements in the HECTOR experimental reactor to infer the residual U235 content followed by isotopic analyses at CEA laboratories in 1975. Fast neutron damage to DRAGON graphite is compared to fast neutron dose measurements using Ni58 (n,p) Co58 activation wires in both DRAGON and the DIDO MTR. Gamma scanning of irradiated fuel elements are used to compare axial power profiles to those derived from two-dimensional and three-dimensional calculations of the DRAGON reactor.

  20. Fission-energy release for 16 fissioning nuclides. Final report

    International Nuclear Information System (INIS)

    Sher, R.

    1981-03-01

    Results are presented of a least-squares evaluation of the components of energy release per fission in 232 Th, 233 U, 235 U, 238 U, 239 Pu, and 241 Pu. For completeness, older (1978) results based on systematics are presented for these and ten other isotopes of interest. There have been recent indications that the delayed energy components may be somewhat higher than those used previously, but the LSQ results do not seem to change significantly when modest (approx. 1 MeV) increases in the total delayed energy are included in the inputs. Additional measurements of most of the energy components are still needed to resolve remaining discrepancies

  1. Conversion ratio and consumption of fissile material in PWR reactors

    International Nuclear Information System (INIS)

    Tiba, C.

    1977-01-01

    It has been shown that the uranium resources will be insufficient for future projected demand. The many solutions to this problem are considered and, in particular, the effect of enrichment on the conversion ratio and hence total uranium comsumption is studied. The developed computacional method employs the one-group neutron diffusion theory. The model is verified by calculating typical burn-up, conversion ratio, U-235 comsumption and plutonium production values in PWR's, and comparing results with those in the published literature. The associated costs of U and U-Pu fuel cycles are also studied for various enrichment values [pt

  2. Experience from and research activities at the Otaniemi TRIGA reactor

    International Nuclear Information System (INIS)

    Bars, Bruno

    1976-01-01

    Experience from the Finnish TRIGA Reactor is reported, small changes and improvements in the control console of the Fir-1 reactor have been made. A minicomputer based data collecting system is planned and installed. It will be used for collecting data from operation and radiation monitors including the new isotope laboratory, and also simultaneously smaller experiments such as control rod calibration. A minicomputer is used for on-line reactor noise studies. The automatic uranium analyzer has a maximum sensitivity of 0.03 μg U 235 and 1.2 Th 232 . The system is now used at a sampling rate of about one sample per minute. (author)

  3. Intercomparison of radionuclide measurements in marine sediment sample IAEA-368

    International Nuclear Information System (INIS)

    Ballestra, S.; Lopez, J.J.; Gastaud, J.; Parsi, P.; Vas, D.; Noshkin, V.

    1991-08-01

    The results of an intercomparison exercise on a Pacific Ocean sediment sample, IAEA-368, designed for the determination of artificial and natural radionuclides levels, are reported. The data from 89 laboratories representing 37 countries have been evaluated. The following are the recommended values, with confidence intervals, for 60 Co, 155 Eu, 210 7Pb, 226 Ra, 238 U, 238 Pu and 239+240 Pu (Reference date: 1 January 1990). Information values for 40 K, 90 Sr, 137 Cs, 228 Th, 230 Th, 232 Th, 234 U, 235 U and 241 Am are also reported. (author)

  4. Fundamentals of pressurized water reactors

    International Nuclear Information System (INIS)

    Murray, L.

    1982-01-01

    In many countries, the pressurized water reactor (PWR) is the most widely used, even though it requires enrichment of the uranium to about 3% in U-235 and the moderator-coolant must be maintained at a high pressure, about 2200 pounds per square inch. Our objective in this series of seven lectures is to describe the design and operating characteristics of the PWR system, discuss the reactor physics methods used to evaluate performance, examine the way fuel is consumed and produced, study the instrumentation system, review the physics measurements made during initial startup of the reactor, and outline the administrative aspects of starting up a reactor and operating it safely and effectively

  5. The conversion of the DIDO-type reactor FRJ-2. Studies and conclusions

    International Nuclear Information System (INIS)

    Stroemich, A.; Siebertz, Ch.; Wickert, M.

    1985-01-01

    For the FRJ-2 (23 MW) of the KFA-Juelich the conversion from HEU- to LEU-fuel was investigated. Before starting the conversion calculations our methods were qualified for the application to heavy water moderated research reactors. A combination of LEU-elements with two different U-235 loadings of 180 g and 225 g was found as suitable for conversion. With these LEU-elements a working core and a transition phase was calculated. The change of the mechanical fuel element design was taken into account. (author)

  6. Process and device for uranium isotope separation and application for the manufacture of chemical compounds or for the separation of gaseous mixtures otherwise difficult to separate

    International Nuclear Information System (INIS)

    Gregorius, K.; Janner, K.; Kersting, A.; Schuster, E.; Niemann, H.J.

    1987-01-01

    The U235/U238 isotope separation is done by laser excitation with Ur 6 as the initial gaseous material. This has HBr added as the partner for a chemical reaction, preferably in the ratio of 1:10. In order to increase the selectivity and yield, the two partners in the reaction are cooled by adiabatic expansion to below 100 K before irradiation. This makes the absorption bands narrower. The excitation occurs in the Q branch of the rotation vibration spectrum. (DG) [de

  7. Heavy water reactors on the denatured thorium cycles

    International Nuclear Information System (INIS)

    1978-05-01

    This paper presents preliminary technical and economic data to INFCE on the denatured U-233/Thorium fuel cycle for use in early comparisons of alternate nuclear systems. The once-through uranium fuel cycle is discussed in a companion paper. In presenting this preliminary information at this time, it is recognized that there are several other denatured thorium fuel cycles of potential interest, such as the U-235/thorium cycle which could be implemented at an earlier date. Information on these alternate cycles is currently being developed, and will be provided to INFCE when available

  8. The development of the nuclear physics in Latvia II. The building of the Research Nuclear Reactor IRT

    International Nuclear Information System (INIS)

    Ulmanis, U.

    2004-01-01

    Nuclear research reactor IRT of the Academy of Sciences was built near Riga in Salaspils. IRT is pool aqueous - aqueous reactor with nuclear fuel U-235 contained elements, located in the core at a depth of ∼ 7 m under distilled water. Ten horizontal and 10-15 vertical experimental channels are employed in experimental research with the use of neutron fluxes. For the research with gamma rays is constructed radiation loop facility with liquid In-Ga-SN solid solution as intensive gamma-ray sources. Main activities of IRT are to conduct research in nuclear spectroscopy, neutron activation analysis, neutron diffraction and radiation physics, chemistry and biology. (authors)

  9. Basic criticality relations for gas core design

    International Nuclear Information System (INIS)

    Tanner, J.E.

    1992-01-01

    Minimum critical fissile concentrations are calculated for U-233, U-235, Pu-239, and Am-242m mixed homogeneously with hydrogen at temperatures to 15,000K. Minimum critical masses of the same mixtures in a 1000 liter sphere are also calculated. It is shown that propellent efficiencies of a gas core fizzler engine using Am-242m as fuel would exceed those in a solid core engine as small as 1000L operating at 100 atmospheres pressure. The same would be true for Pu-239 and possibly U-233 at pressures of 1000 atm. or at larger volumes

  10. Toward JENDL-3.3: comments on the problems of JENDL-3.2 for the next version developments

    Energy Technology Data Exchange (ETDEWEB)

    Hasegawa, Akira [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-03-01

    Revision work for JENDL-3.2 is foreseen and the discussions about the problems of JENDL-3.2 is under going in a small advisory group of JNDC (Japanese Nuclear Data Committee) since this April. In this article main discussion points in that group are presented. Problems from evaluators` point of view, problem from users` point of view and request from anticipated users are summarized. Some discussions on U-235 cross sections in the thermal and epi-thermal energy range are made in relation of the new evaluation of resolved resonance parameters performed by Leal/Derrien. (author)

  11. Effect of Ionizing Radiation on Luminous Bacteria Cells

    International Nuclear Information System (INIS)

    Kudryasheva, N.; Rozhko, T.; Alexandrova, M.; Vasyunkina, E.; Arkhipova, V.

    2011-01-01

    Marine luminous bacteria were used to monitor toxicity of alpha- (Am-241, U-235+238) and beta- (tritium) radionuclide solutions. Increase or inhibition of bacterial luminescence was observed under exposure to radionuclides. Radiation toxicity of Am and chemical toxicity of U were demonstrated. Effects of U were similar to those of stable heavy metals: sensitivity was about 10-5 M. Sensitivity of the bacteria to Am-241 was 300 Bq/L (10 -11 M). Inhibition of bacterial growth was observed under exposure to Am-241 and tritium. Role of peroxides and electron transfer processes in the effects of radionuclides on luminous bacteria is discussed.

  12. Development and use of the IVV-2M fuel assembly

    International Nuclear Information System (INIS)

    Aden, V.G.; Bulkin, Yu.M.; Vasenkov, V.I.

    1987-01-01

    The design and performance of a fuel assembly, intended for use in the water cooled and moderated IVV-2M research reactor and consisting of dioxide dispersed in an aluminium alloy matrix encased in an aluminium alloy can, is presented. Experimental and theoretical studies included neutron characteristics of the reactor core, thermohydraulic behavior, and thermal neutron flux distribution. Data are also presented on the specific charge in u 235, the metal to water volume ratio, the specific heat removal surface, and maximum power of the assembly

  13. International proliferation on nuclear weapons

    International Nuclear Information System (INIS)

    Hill, J.

    1977-01-01

    The subject is dealt with under the following headings: introduction; routes to proliferation (preparation of U 235 , Pu 239 , U 233 ); nuclear power fuel cycles and proliferation; the fast reactor fuel cycle; security aspects of the existing fuel cycle; the IAEA and the nuclear non-proliferation treaty. It is concluded that 'the basis for sound international control exists, and taken together with the further technical steps which will be taken to make the existing fuel cycles more robust against the diversion of materials by terrorists and the abuse of civil nuclear power programmes by governments, we have good reason to proceed now with the orderly exploitation of ...nuclear energy...'. (U.K.)

  14. The nuclear question at the start of the '80s: the breeder reactor

    International Nuclear Information System (INIS)

    Owen, R.; Svensson, B.

    1980-01-01

    The four newspaper articles and the letter cover the following matters: general introduction about breeder reactors and the situation in Swedish politics; visit to Dounreay to discuss breeder reactors (breeding, safety, plutonium production, radiation protection); PuO 2 -UO 2 mixed fuel; description of breeder reactors; efficiency in use of U-235; DFR and PFR; breeder reactors in Swedish politics (arguments for and against nuclear power in general, breeder reactors in particular); discussion of the future of nuclear power in Sweden. (U.K.)

  15. Manufacture of uranium compounds for research reactors fuel elements. Participation of the UCPP (Uranium compound production plant) in the Egyptian project; Elaboracion de compuestos de uranio para ser utilizados en elementos combustibles de reactores de investigacion. Participacion de la planta de fabricacion de compuestos de uranio (PFPU) en el proyecto Egipto

    Energy Technology Data Exchange (ETDEWEB)

    Boero, Norma L; Cinat, Enrique; Yorio, Daniel; Cincotta, Daniel; Ramella, Jose L; Bruno, Hernan R; Camacho, Esteban F; Pertossi, Fernando; Panunzio, Leonardo D; Fernandez, Carlos A; Sassone, Ariel [Comision Nacional de Energia Atomica, General San Martin (Argentina). Dept. de Combustibles Nucleares

    1999-07-01

    UCPP is an international qualified supplier of U{sub 3}O{sub 8} with up to 20 % enrichment in U-235. The characteristics of this powder are those specified for fuel plates manufacture for test reactors. This paper describes the works performed in the plant since its beginning, emphasising those undertaken during the last years. The transference of U{sub 3}O{sub 8} manufacturing technology to INVAP SE, the enterprise that installed a plant of similar characteristics in the Arabian Republic of Egypt, is especially described. (author)

  16. Manufacture of uranium compounds for research reactors fuel elements. Participation of the UCPP (Uranium compound production plant) in the Egyptian project

    International Nuclear Information System (INIS)

    Boero, Norma L.; Cinat, Enrique; Yorio, Daniel; Cincotta, Daniel; Ramella, Jose L.; Bruno, Hernan R.; Camacho, Esteban F.; Pertossi, Fernando; Panunzio, Leonardo D.; Fernandez, Carlos A.; Sassone, Ariel

    1999-01-01

    UCPP is an international qualified supplier of U 3 O 8 with up to 20 % enrichment in U-235. The characteristics of this powder are those specified for fuel plates manufacture for test reactors. This paper describes the works performed in the plant since its beginning, emphasising those undertaken during the last years. The transference of U 3 O 8 manufacturing technology to INVAP SE, the enterprise that installed a plant of similar characteristics in the Arabian Republic of Egypt, is especially described. (author)

  17. Beta activity of enriched uranium

    International Nuclear Information System (INIS)

    Nambiar, P.P.V.J.; Ramachandran, V.

    1975-01-01

    Use of enriched uranium as reactor fuel necessitates its handling in various forms. For purposes of planning and organising radiation protection measures in enriched uranium handling facilities, it is necessary to have a basic knowledge of the radiation status of enriched uranium systems. The theoretical variations in beta activity and energy with U 235 enrichment are presented. Depletion is considered separately. Beta activity build up is also studied for two specific enrichments, in respect of which experimental values for specific alpha activity are available. (author)

  18. Optogalvanic measurement of isotope shifts of doubly ionized uranium (U III) made using natural-U samples

    International Nuclear Information System (INIS)

    Piyakis, K.N.; Gagne, J.

    1989-01-01

    An efficient method of identifying 235 U III (in natural-U samples), with the help of the optogalvanic effect in a hollow-cathode discharge, is presented. The use of this method enabled us to carry out the measurement of isotope shifts and the preliminary investigation of hyperfine structures of U III. The 238 U-- 235 U shifts for the 591.313-, 586.045-, and 610.497-nm U III lines are found to be 921(3), 417(6), and 392(12) mK, respectively

  19. Verification to the RSG-GAS fuel discharge burn-up using SRAC2006 module of COREBN/HIST

    International Nuclear Information System (INIS)

    J-Susilo; T-M-Sembiring; G-R-Sunaryo; M-Imron

    2018-01-01

    For 30 years operation, some of the modifications to the RSG GAS core has been done, that are changes included the type of fuel from U 3 O 8 -Al to U 3 Si 2 -Al with the same density 2.96 gU/cc, the loading pattern of standard fuel elements/fuel control elements from 6/1 & 6/2 to 5/1 pattern, and in core fuel management calculation tool has been change from IAFUEL to BATAN-FUEL. To obtain an extension of the operating license for the next 10 years, the RSG-GAS Periodic Safety Assessment Document is need to prepared. According to the Regulatory Body Chairman Regulation No. 2 2015, RSG-GAS safety assessment should be done independently. As part of this assessment the fuel discharge burn-up must be estimated. In this research, to ensure that the misposition of fuel element in the core has not occurred, the investigation to the document operating report related the fuel placement has been done. Therefore, by using 78 th to 93 rd operation data, verify of the fuel discharge burn-up of the RSG-GAS has been performed by using SRAC2006 module of COREBN/HIST. In addition, the results of these calculations are also made comparative with the operating report data that is calculated by using BATAN-FUEL. Maximum fuel discharge burn-up (57.73 % of U-235) was verified still under permissible value determined by the regulatory body (<60 % of U-235). Maximum differences value between two computer codes was about 2.12 % of U-235 (3.80 %) that is fuel at the B-7 position. Fuel discharge burn-up of RSG-GAS showed almost the same value for each the operation cycle, range of 1.52 % of U-235. So it can be concluded that the RSG-GAS core operation over the last ten years was in good fuel management performance, in accordance with the design. BATAN-FUEL has been conformed well enough with COREBN/HIST. (author)

  20. Evaluating the effectiveness of dilution of the recovered uranium with depleted uranium and low-enriched uranium to obtain fuel for VVER reactors

    International Nuclear Information System (INIS)

    Smirnov, A Yu; Sulaberidze, G A; Dudnikov, A A; Nevinitsa, V A

    2016-01-01

    The possibility of the recovered uranium enrichment in a cascade of gas centrifuges with three feed flows (depleted uranium, low-enriched uranium, recovered uranium) with simultaneous dilution of U-232,234,236 isotopes was shown. A series of numerical experiments were performed for different content of U-235 in low-enriched uranium. It has been demonstrated that the selected combination of diluents can simultaneously reduce the cost of separative work and the consumption of natural uranium, not only with respect to the previously used multi-flow cascade schemes, but also in comparison to the standard cascade for uranium enrichment. (paper)

  1. Intercomparison of radionuclide measurements in marine sediment sample IAEA-367

    International Nuclear Information System (INIS)

    Ballestra, S.; Lopez, J.J.; Gastaud, J.; Vas, D.; Noshkin, V.

    1991-08-01

    The results of an intercomparison exercise on a Pacific Ocean sediment sample, IAEA-367, designed for the determination of artificial and natural radionuclides levels, are reported. The data from 81 laboratories representing 37 countries have been evaluated. The following are the recommended values, with confidence intervals, for 60 Co, 90 Sr, 137 Cs, 239+240 Pu (Reference date: 1 January 1990). Information values for 155 Eu, 238 Pu, 241 Am, and 241 Pu are reported. Information values for the following natural radionuclides 40 K, 226 Ra, 228 Th, 230 Th, 234 U, 235 U and 238 U are also reported. (author)

  2. Irradiation of MEU and LEU test fuel elements in DR 3

    International Nuclear Information System (INIS)

    Haack, K.

    1984-01-01

    Irradiation of three MEU and three LEU fuel elements in the Danish reactor DR 3. Thermal and fast neutron flux density scans of the core have been made and the results, related to the U235-content of each fuel element, are compared with the values from HEU fuel elements. The test elements were taken to burn-up percentages of 50-60%. Reactivity values of the test elements at charge and at discharge have been measured and the values are compared with those of HEU fuel elements. (author)

  3. Weld Development for Aluminum Fission Chamber

    Energy Technology Data Exchange (ETDEWEB)

    Cross, Carl Edward [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Martinez, Jesse Norris [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-05-16

    The Sigma welding team was approached to help fabricate a small fission chamber (roughly ½ inch dia. x ½ inch tall cylinder). These chambers are used as radiation sensors that contain small traces of radionuclides (Cf 252, U 235, and U 238) that serve to ionize gas atoms in addition to external radiation. When a voltage is applied within the chamber, the resulting ion flow can be calibrated and monitored. Aluminum has the advantage of not forming radioactive compounds when exposed to high external radiation (except from minor Na alloy content). Since aluminum has not been used before in this application, this presented an unexplored challenge.

  4. The Chemistry and Toxicology of Depleted Uranium

    OpenAIRE

    Sidney A. Katz

    2014-01-01

    Natural uranium is comprised of three radioactive isotopes: 238U, 235U, and 234U. Depleted uranium (DU) is a byproduct of the processes for the enrichment of the naturally occurring 235U isotope. The world wide stock pile contains some 1½ million tons of depleted uranium. Some of it has been used to dilute weapons grade uranium (~90% 235U) down to reactor grade uranium (~5% 235U), and some of it has been used for heavy tank armor and for the fabrication of armor-piercing bullets and missiles....

  5. ZZ THERMOS, Multigroup P0 to P5 Thermal Scattering Kernels from ENDF/B Scattering Law Data

    International Nuclear Information System (INIS)

    McCrosson, F.J.; Finch, D.R.

    1975-01-01

    1 - Description of problem or function: Number of groups: 30-group THERMOS thermal scattering kernels. Nuclides: Molecular H 2 O, Molecular D 2 O, Graphite, Polyethylene, Benzene, Zr bound in ZrHx, H bound in ZrHx, Beryllium-9, Beryllium Oxide, Uranium Dioxide. Origin: ENDF/B library. Weighting Spectrum: yes. These data are 30-group THERMOS thermal scattering kernels for P0 to P5 Legendre orders for every temperature of every material from s(alpha,beta) data stored in the ENDF/B library. These scattering kernels were generated using the FLANGE2 computer code (NESC Abstract 368). To test the kernels, the integral properties of each set of kernels were determined by a precision integration of the diffusion length equation and compared to experimental measurements of these properties. In general, the agreement was very good. Details of the methods used and results obtained are contained in the reference. The scattering kernels are organized into a two volume magnetic tape library from which they may be retrieved easily for use in any 30-group THERMOS library. The contents of the tapes are as follows - (Material: ZA/Temperatures (degrees K)): Molecular H 2 O: 100.0/296, 350, 400, 450, 500, 600, Molecular D 2 O: 101.0/296, 350, 400, 450, 500, 600, Graphite: 6000.0/296, 400, 500, 600, 700, 800, Polyethylene: 205.0/296, 350 Benzene: 106.0/296, 350, 400, 450, 500, 600, Zr bound in ZrHx: 203.0/296, 400, 500, 600, 700, 800, H bound in ZrHx: 230.0/296, 400, 500, 600, 700, 800, Beryllium-9: 4009.0/296, 400, 500, 600, 700, 800, Beryllium Oxide: 200.0/296, 400, 500, 600, 700, 800, Uranium Dioxide: 207.0/296, 400, 500, 600, 700, 800 2 - Method of solution: Kernel generation is performed by direct integration of the thermal scattering law data to obtain the differential scattering cross sections for each Legendre order. The integral parameter calculation is done by precision integration of the diffusion length equation for several moderator absorption cross sections followed by a

  6. Waveband specific transcriptional control of select genetic pathways in vertebrate skin (Xiphophorus maculatus).

    Science.gov (United States)

    Walter, Ronald B; Boswell, Mikki; Chang, Jordan; Boswell, William T; Lu, Yuan; Navarro, Kaela; Walter, Sean M; Walter, Dylan J; Salinas, Raquel; Savage, Markita

    2018-05-10

    Evolution occurred exclusively under the full spectrum of sunlight. Conscription of narrow regions of the solar spectrum by specific photoreceptors suggests a common strategy for regulation of genetic pathways. Fluorescent light (FL) does not possess the complexity of the solar spectrum and has only been in service for about 60 years. If vertebrates evolved specific genetic responses regulated by light wavelengths representing the entire solar spectrum, there may be genetic consequences to reducing the spectral complexity of light. We utilized RNA-Seq to assess changes in the transcriptional profiles of Xiphophorus maculatus skin after exposure to FL ("cool white"), or narrow wavelength regions of light between 350 and 600 nm (i.e., 50 nm or 10 nm regions, herein termed "wavebands"). Exposure to each 50 nm waveband identified sets of genes representing discrete pathways that showed waveband specific transcriptional modulation. For example, 350-400 or 450-500 nm waveband exposures resulted in opposite regulation of gene sets marking necrosis and apoptosis (i.e., 350-400 nm; necrosis suppression, apoptosis activation, while 450-500 nm; apoptosis suppression, necrosis activation). Further investigation of specific transcriptional modulation employing successive 10 nm waveband exposures between 500 and 550 nm showed; (a) greater numbers of genes may be transcriptionally modulated after 10 nm exposures, than observed for 50 nm or FL exposures, (b) the 10 nm wavebands induced gene sets showing greater functional specificity than 50 nm or FL exposures, and (c) the genetic effects of FL are primarily due to 30 nm between 500 and 530 nm. Interestingly, many genetic pathways exhibited completely opposite transcriptional effects after different waveband exposures. For example, the epidermal growth factor (EGF) pathway exhibits transcriptional suppression after FL exposure, becomes highly active after 450-500 nm waveband exposure, and again, exhibits strong

  7. [Fluorescence spectroscopic characteristics of fulvic acid from the long-term located fertilization in black soil].

    Science.gov (United States)

    Li, Yan-Ping; Wei, Dan; Zhou, Bao-Ku; Zhao, Yue; Zhang, Xi-Lin; Wei, Zi-Min; Li, Shu-Ling

    2011-10-01

    In order to investigate the effect of long-term located fertilization on soil fulvic acid (FA), in this study, four soil samples were taken from black soil with long-term located fertilization (about 30 year) in Harbin, Heilongjiang province. The fertilization treatments included control (CK), N, P and K fertilization (NPK), horse manure (OM), combination of organic manure and chemical fertilizations (MNPK). Soil FA was extracted from the samples and purified. The excitation, emission, synchronous, and three-dimensional-excitation emission matrix fluorescence spectroscopy (3DEEM) characteristics of the FA were determined. The excitation, emission and synchronous scan spectra all indicated that the main peaks of FA in the NPK treatment exhibited a significantly blue shift compared with CK, while those of MNPK, OM treatment caused a red shift to some extent. 3DEEM spectra of FA in all treatments exhibited four peaks (peak a, peak b, peak c, and peak d), compared with FA in CK, the wavelengths shift tendency of peak a, peak b, and peak c of FA 3DEEM in NPK, MNPK and OM treatments were similar to that of traditional spectra in FA. In order to provide quantitative information of FA humification degree in different treatments, we investigated the fluorescence index f450/500 (FI), area integration (A370-600 nm, A1 370-412 nm, A4 538-600 nm). Compared with CK, the f450/500, ratio of A1/A in NPK and A4/A in MNPK treatment increased by 4.62%, 6.12%, 7.22%, respectively. However, the f450/500, the ratio of A1/A in MNPK and A4/A in NPK treatment decreased by 3.86%, 15.31%, and 7.22% respectively. This indicated that NPK application gave a lower degree of FA humification, and combination of organic manure and chemical fertilizations would lead to a greater degree of FA aromatization in black soil with long-term located fertilization than CK.

  8. Optimum burnup of BAEC TRIGA research reactor

    International Nuclear Information System (INIS)

    Lyric, Zoairia Idris; Mahmood, Mohammad Sayem; Motalab, Mohammad Abdul; Khan, Jahirul Haque

    2013-01-01

    Highlights: ► Optimum loading scheme for BAEC TRIGA core is out-to-in loading with 10 fuels/cycle starting with 5 for the first reload. ► The discharge burnup ranges from 17% to 24% of U235 per fuel element for full power (3 MW) operation. ► Optimum extension of operating core life is 100 MWD per reload cycle. - Abstract: The TRIGA Mark II research reactor of BAEC (Bangladesh Atomic Energy Commission) has been operating since 1986 without any reshuffling or reloading yet. Optimum fuel burnup strategy has been investigated for the present BAEC TRIGA core, where three out-to-in loading schemes have been inspected in terms of core life extension, burnup economy and safety. In considering different schemes of fuel loading, optimization has been searched by only varying the number of fuels discharged and loaded. A cost function has been defined and evaluated based on the calculated core life and fuel load and discharge. The optimum loading scheme has been identified for the TRIGA core, the outside-to-inside fuel loading with ten fuels for each cycle starting with five fuels for the first reload. The discharge burnup has been found ranging from 17% to 24% of U235 per fuel element and optimum extension of core operating life is 100 MWD for each loading cycle. This study will contribute to the in-core fuel management of TRIGA reactor

  9. Validation of Cross Sections with Criticality Experiment and Reaction Rates: the Neptunium Case

    CERN Document Server

    Leong, L S; Audouin, L; Berthier, B; Le Naour, C; Stéphan, C; Paradela, C; Tarrío, D; Duran, I

    2014-01-01

    The Np-237 neutron-induced fission cross section has been recently measured in a large energy range (from eV to GeV) at the n\\_TOF facility at CERN. When compared to previous measurements the n\\_TOF fission cross section appears to be higher by 5-7\\% beyond the fission threshold. To check the relevance of the n\\_TOF data, we considered a criticality experiment performed at Los Alamos with a 6 kg sphere of Np-237, surrounded by uranium highly enriched in U-235 so as to approach criticality with fast neutrons. The multiplication factor k(eff) of the calculation is in better agreement with the experiment when we replace the ENDF/B-VII. 0 evaluation of the Np-237 fission cross section by the n\\_TOF data. We also explored the hypothesis of deficiencies of the inelastic cross section in U-235 which has been invoked by some authors to explain the deviation of 750 pcm. The large modification needed to reduce the deviation seems to be incompatible with existing inelastic cross section measurements. Also we show that t...

  10. Benchmark criticality experiments for fast fission configuration with high enriched nuclear fuel

    International Nuclear Information System (INIS)

    Sikorin, S.N.; Mandzik, S.G.; Polazau, S.A.; Hryharovich, T.K.; Damarad, Y.V.; Palahina, Y.A.

    2014-01-01

    Benchmark criticality experiments of fast heterogeneous configuration with high enriched uranium (HEU) nuclear fuel were performed using the 'Giacint' critical assembly of the Joint Institute for Power and Nuclear Research - Sosny (JIPNR-Sosny) of the National Academy of Sciences of Belarus. The critical assembly core comprised fuel assemblies without a casing for the 34.8 mm wrench. Fuel assemblies contain 19 fuel rods of two types. The first type is metal uranium fuel rods with 90% enrichment by U-235; the second one is dioxide uranium fuel rods with 36% enrichment by U-235. The total fuel rods length is 620 mm, and the active fuel length is 500 mm. The outer fuel rods diameter is 7 mm, the wall is 0.2 mm thick, and the fuel material diameter is 6.4 mm. The clad material is stainless steel. The side radial reflector: the inner layer of beryllium, and the outer layer of stainless steel. The top and bottom axial reflectors are of stainless steel. The analysis of the experimental results obtained from these benchmark experiments by developing detailed calculation models and performing simulations for the different experiments is presented. The sensitivity of the obtained results for the material specifications and the modeling details were examined. The analyses used the MCNP and MCU computer programs. This paper presents the experimental and analytical results. (authors)

  11. Criticality impacts on LWR fuel storage efficiency

    International Nuclear Information System (INIS)

    Napolitano, D.

    1992-01-01

    This presentation discusses the criticality impacts throughout storage of fuel onsite including new fuel storage, spent fuel storage, consolidation, and dry storage. The general principles for criticality safety are also be discussed. There is first an introduction which explains today's situation for criticality safety concerns. This is followed by a discussion of criticality safety Regulatory Guides, safety limits and fundamental principles. Design objectives for criticality safety in the 1990's include higher burnups, longer cycles, and higher enrichments which impact the criticality safety design. Criticality safety for new fuel storage, spent fuel storage, fuel consolidation, and dry storage are followed by conclusions. Today's situation is one in which the US does not reprocess, and does not have an operating MRS facility or repository. High density fuel storage rack designs of the 1980s, are filling up. Dry cask storage systems for spent fuel storage are being utilized. Enrichments continue to increase PWR fuel assemblies with enrichments of 4.5 to 5.0 weight percent U-235 and BWR fuel assemblies with enrichments of 3.25 to 3.5 weight percent U-235 are common. Criticality concerns affect the capacity and the economics of light water reactor (LWR) fuel storage arrays by dictating the spacing of fuel assemblies in a storage system, or the use of poisons or exotic materials in the storage system design

  12. Comminution of the U-10Mo by hydriding cycles innovative process

    Energy Technology Data Exchange (ETDEWEB)

    Faeda, Kelly C.M.; Santos, Ana Maria M. dos; Paula, Joao B. de; Pereira, Edilson M.; Pedrosa, Tercio A.; Lameiras, Fernando S.; Ferraz, Wilmar B., E-mail: ferrazw@cdtn.br, E-mail: kelly.faeda@prof.una.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2013-07-01

    The research, test and producing radioisotopes compact reactors were developed with the use of high levels of enriched fuel of approximately 90% of the fissile isotope U-235. Since the 80s', a policy under the context of international program RERTR (Reduced Enrichment for Research and Test Reactors) encourages the fuel replacement of the high enriched fuel by the low one of about 20 % U-235. One way to compensate the substitution for the low enrichment fuel is to employ high density metal uranium alloys. The fabrication of compact reactor fuel uses the metal matrix dispersion and, for this, uranium alloys are used in the form of powders. Despite the high densities, the metallic uranium based alloys are ductile and therefore difficult to be comminuted. Among the different comminution processes, the hydriding-dehydriding process has proved most advantageous, primarily due to their relative simplicity of processing and low manufacturing cost. In this paper, we present the results of the development of the U-10Mo alloy comminution process by the hydriding-dehydriding method on a laboratory scale. Samples of the alloy were subjected to different hydriding cycle numbers in order to verify its influence in relation to the particle size distribution of powders. Powders of different particle sizes were obtained and characterized by the physical and morphological characteristics by optical microscopy, scanning electron microscopy and X ray diffraction. The obtained results are evaluated and discussed. (author)

  13. Material control and accounting requirements for uranium enrichment facilities

    International Nuclear Information System (INIS)

    Ting, P.

    1991-01-01

    This paper reports that the U.S. Nuclear Regulatory Commission has defined material control and accounting (MC and A) requirement for low-enriched uranium enrichment plants licensed under 10 CFR parts 40 and 70. Following detailed assessment of potential safeguards issues relevant to these facilities, a new MC and A rule was developed. The primary safeguards considerations are detection of the loss of special nuclear material, detection of clandestine production of special nuclear material of low strategic significance for unauthorized use or distribution, and detection of unauthorized production of uranium enriched to ≥10 wt % U-235. The primary safeguards concerns identified were the large absolute limit of error associated with the material balance closing, the inability to shutdown some uranium enrichment technologies to perform a cleanout inventory of the process system, and the flexibility of some of these technologies to produce higher enrichments. Unauthorized production scenarios were identified for some technologies that could circumvent the detection of the production and removal of 5 kilograms of U-235 as high-enriched uranium through conventional material control and accounting programs. Safeguards techniques, including the use of production and process control information, measurements, and technical surveillance, were identified to compensate for these concerns

  14. Use of proportional gas scintillator in absolute measurements of alpha-gamma emitter activities

    International Nuclear Information System (INIS)

    Tobias, C.C.B.

    1987-01-01

    The absolute activity of U-235 contained in a U 3 O 8 sample was measured utilizing a sum-coincidence circuit which selects only the alpha particles which are simultaneous with the 143 KeV and 186 KeV gamma radiations from the Th-231 (product nucleus). The alpha particles were detected by means of a new type of a gas scintillating chamber, in which the light emitted by excitation of the gas atoms, due to the passage of a charged incoming particle, has its intensity increased by the action of an applied electric field. The gamma radiations were detected by means of a NaI(Tl) 1'' x 1 1/2'' scintillation detector. The value obtained for the half-life of U-235 was compared with the data available from various observers which used different experimental techniques. It is shown tht the results, are in excellent agreement with the best international data available on the subject and that, therefore, the sum-coincidence technique constitutes an important method for such measurements. (Author) [pt

  15. Requirements for future developments from utility point of view

    International Nuclear Information System (INIS)

    Nordloef, S.; Besch, O.A.

    1995-01-01

    On the BWR side the development of fuel has continued and during the recent years many new designs have entered the market such as ABB Atom SVEA-96, Siemens Atrium, GE-12 etc. These new designs provide new possibilities to optimize the core design and also give better utilization of the uranium. The future development work should emphasize on less susceptibility to severe secondary damage and also higher resistance to debris failure. Another utility demand is to increase the thermal margins such as dryout and PCI performance in order to avoid any restrictions during load, follow up and start up conditions. For three decades the PWR fuel design and the selection of material led to a satisfying results in the overall operational behaviour, which resulted in higher utilization of the fuel and materials. At the beginning of the seventieth, the utilities started with burnups of 30 MWd/kg u . Nowadays, burnups of 42 MWd/kg u are reached and burnups of 50 MWd/kg u as transition burnups are designed. The increase of enrichment from 3.0 w/o U-235 to 4.0 w/o U-235, the change of guide thimble and spacer material from stainless steel or inconel to zircalloy alloys and the different fuel core loading strategy from out-in-in to in-in-out improved the use of fuel and reduced the number of fresh fuel assemblies for reloads. (orig./HP)

  16. Analysis of Tank 38H (HTF-38-16-80, 81) and Tank 43H (HTF-43-16-82, 83) Samples for Support of the Enrichment Control and Corrosion Control Programs

    Energy Technology Data Exchange (ETDEWEB)

    Hay, M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-10-24

    SRNL analyzed samples from Tank 38H and Tank 43H to support ECP and CCP. The total uranium in the Tank 38H surface sample was 57.6 mg/L, while the sub-surface sample was 106 mg/L. The Tank 43H samples ranged from 50.0 to 51.9 mg/L total uranium. The U-235 percentage was consistent for all four samples at 0.62%. The total uranium and percent U-235 results appear consistent with recent Tank 38H and Tank 43H uranium measurements. The Tank 38H plutonium results show a large difference between the surface and sub-surface sample concentrations and somewhat higher concentrations than previous samples. The Pu-238 concentration is more than forty times higher in the Tank 38H sub-surface sample than the surface sample. The surface and sub-surface Tank 43H samples contain similar plutonium concentrations and are within the range of values measured on previous samples. The four samples analyzed show silicon concentrations somewhat higher than the previous sample with values ranging from 104 to 213 mg/L.

  17. HEU to LEU conversion and blending facility: Oxide blending alternative to produce LEU oxide for commercial use

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    The United States Department of Energy (DOE) is examining options for the disposition of surplus weapons-usable fissile materials and storage of all weapons-usable fissile materials. Disposition is a process of use or disposal of material that results in the material being converted to a form that is substantially and inherently more proliferation-resistant than the original form. Examining options for increasing the proliferation resistance of highly enriched uranium (HEU) is part of this effort. This document provides data to be used in the environmental impact analysis for the oxide blending HEU disposition option. This option provides for a yearly HEU throughput of 1 0 metric tons (MT) of uranium metal with an average U235 assay of 50% blended with 165 MT of natural assay triuranium octoxide (U{sub 3} O{sub 8}) per year to produce 177 MT of 4% U235 assay U{sub 3} O{sub 8}, for LWR fuel. Since HEU exists in a variety of forms and not necessarily in the form to be blended, worst case scenarios for preprocessing prior to blending will be assumed for HEU feed streams.

  18. The substitution technique applied for determining the parameters of a reactor with void

    International Nuclear Information System (INIS)

    Zdravkovic, Z.; Ivkovic, M.; Sotic, O.

    1965-12-01

    The material buckling of D 2 O-enriched uranium (2% U 235 ) system with voided fuel assembly has been determined by means of the substitution technique. The experiments were carried out in the zero power reactor RB for various lattice pitches, namely: 11.3 cm, 14 cm, and 16 cm. As a reference core the D 2 O-enriched uranium (2%U 235 ) ) system without voids has been used. The results have been analysed by the modified one-group perturbation theory using three regions in a substituted reactor (test, intermediate and reference). The perturbations in axial and radial directions due to voided fuel assembly, are also considered. In order to determine the change between diffusion coefficients in various regions (δD), additional experiments with single test fuel element were made. The lattice pitches of reference and test regions were the same. According to the algorithms given in appendices, the programmes for the ZUSE Z-23 digital computer were made. These programmes include evaluations of the change between diffusion coefficients (δD), and the axial buckling of the test lattice (author)

  19. Dating of pre-exposure times of lunar rocks and soils

    International Nuclear Information System (INIS)

    Eugster, O.

    1986-01-01

    Xenon produced by fission of uranium, thorium and plutonium has repeatedly been observed in lunar rocks and soils. In two basaltic rocks and in two soils Xe was found originating from fission of U-235 induced by neutrons which are due to the interactions of cosmic ray particles with lunar matter. Two facts lead to this conclusion: (1) fission Xe is present in excess of that expected for the U, Th, and Pu concentrations and for the gas retention age of the samples; and (2) the Xe-134/Xe-136 ratio of excess fission Xe is close to 1.25 as expected for neutron induced fission of U-235. Information on the duration of the exposure to cosmic rays was obtained from the Kr-81-Kr systematics whereas the effective shielding conditions were derived from the depth sensitive cosmogenic ratio Xe-131/Xe-126. For the four samples the exposure to cosmic rays in the lunar regolith is described by a two stage exposure model. The history of the four samples was derived in terms of duration and shielding depth of the two stages

  20. Aspects on thorium utilization in heavy water reactors

    International Nuclear Information System (INIS)

    1978-11-01

    Some of the main problems of the Th - PHWR cycles are analyzed. With respect to the burnup limitations introduced by SSET cycle conditions and the burnup sensitivities of this reference cycle, estimates are presented using an integrated neutron-heavy element balance method. A PHWR of 1 GW(e) very similar to the CANDU current design was selected. In the case of 0.5% uranium losses, 11000 MWD/tHE and 13000 MWD/tHE were considered for U-235, respectively, for the Pu initialization of the cycle, the corresponding inventory being 4 t U-235/GW(e) and 5 t Pu (with 72% fissile content) per GW(e) for one year delay time between reactor out to reactor in, 66% capacity factor, 27 MW (fission)/tHE medium specific power. The following aspects are also analyzed: Safety problems associated with low delayed neutron fraction values; High and intermediate burnup fuel elements conceptual problems; Specific problems of thorium reprocessing; Specific problems for radioactive wastes and thorium storage; U-232 content evaluations and related fuel fabrication problems

  1. Development of a computer program of fast calculation for the pre design of advanced nuclear fuel 10 x 10 for BWR type reactors; Desarrollo de un program de computo de calculo rapido para el prediseno de celdas de combustible nuclear avanzado 10 x 10 para reactores de agua en ebullicion

    Energy Technology Data Exchange (ETDEWEB)

    Perusquia, R.; Montes, J.L.; Ortiz, J.J. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: mrpc@nuclear.inin.mx

    2005-07-01

    In the National Institute of Nuclear Research (ININ) a methodology is developed to optimize the design of cells 10x10 of assemble fuels for reactors of water in boil or BWR. It was proposed a lineal calculation formula based on a coefficients matrix (of the change reason of the relative power due to changes in the enrichment of U-235) for estimate the relative powers by pin of a cell. With this it was developed the computer program of fast calculation named PreDiCeldas. The one which by means of a simple search algorithm allows to minimize the relative power peak maximum of cell or LPPF. This is achieved varying the distribution of U-235 inside the cell, maintaining in turn fixed its average enrichment. The accuracy in the estimation of the relative powers for pin is of the order from 1.9% when comparing it with results of the 'best estimate' HELIOS code. With the PreDiCeldas it was possible, at one minimum time of calculation, to re-design a reference cell diminishing the LPPF, to the beginning of the life, of 1.44 to a value of 1.31. With the cell design with low LPPF is sought to even design cycles but extensive that those reached at the moment in the BWR of the Laguna Verde Central. (Author)

  2. Local power peaking factor estimation in nuclear fuel by artificial neural networks

    International Nuclear Information System (INIS)

    Montes, Jose Luis; Francois, Juan Luis; Ortiz, Juan Jose; Martin-del-Campo, Cecilia; Perusquia, Raul

    2009-01-01

    This paper presents the training of an artificial neural network (ANN) to accurately predict, in very short time, a physical parameter used in nuclear fuel reactor optimization: the local power peaking factor (LPPF) in a typical boiling water reactor (BWR) fuel lattice. The ANN training patterns are distribution of fissile and burnable poison materials in the fuel lattice and their associated LPPF. These data were obtained by modeling the fuel lattices with a neutronic simulator: the HELIOS transport code. The combination of the pin U 235 enrichment and the Gd 2 O 3 (gadolinia) concentration, inside the 10 x 10 fuel lattice array, was encoded by three different methods. However, the only encoding method that was able to give a good prediction of the LPPF was the method which added the U 235 enrichment and the gadolinia concentration. The results show that the relative error in the estimation of the LPPF, obtained by the trained ANN, ranged from 0.022% to 0.045%, with respect to the HELIOS results

  3. Yield, utilization, storage and ultimate storage of depleted uranium

    International Nuclear Information System (INIS)

    Aumueller, L.; Hermann, J.

    1977-11-01

    More than 80% of the uranium leaving uranium enrichment plants is depleted to a residual content of about 0,25% U 235. Due to the present ineconomical further depletion to the technically possible residual content of 0,1% U 235, the so-called 'tails' are first of all stored. The quantity of stored depleted uranium in the FRG should be about 100.000 t by the year 2000. It represents a strategic reserve for future energy supply regardless of profitableness. The study analysis the conceivable possible uses for the tails quantity considered. These are, besides further depletion whose profitableness is considered, also the use as breeder material in breeder reactors and the use in the non-nuclear field. The main part of the study deals with the various storage possibilities of the depleted uranium in oxidic or fluoride form. A comparison of costs of alternative storage concepts showed a clear advantage for the storage of UF 6 in 48 inch containers already in use. The conceivable accidents in storing are analyzed and measures to reduce the consequences are discussed. Finally, the problems of ultimate storage for the remaining waste after further depletion or use are investigated and the costs arising here are also estimated. (RB) [de

  4. Burnup influence on the VVER-1000 reactor vessel neutron fluence evaluation

    International Nuclear Information System (INIS)

    Panayotov, I.; Mihaylov, N.; Ilieva, K.; Kirilova, D.; Manolova, M.

    2009-01-01

    The neutron fluence of the vessels of the reactors is determined regularly accordingly the RPV Surveillance Program of the Kozloduy NPP Unit 5 and 6 in order to assess the state of the metal vessel and their radiation damaging. The calculations are carried out by the method of discrete ordinates used in the TORT program for operated reactor cycles. An average reactor spectrum corresponding to fresh U-235 fuel is used as an input neutron source. The impact of the burn up of the fuel on the neutron fluence of VVER-1000 reactor vessel is evaluated. The calculations of isotopic concentrations of U-235 and Pu-239 corresponding to 4 years burn up were performed by the module SAS2H of the code system SCALE 4.4. Since fresh fuel or 4 years burn up fuel assembly are placed in periphery of reactor core the contribution of Pu-239 of first year burn up and of 4 years burn up is taken in consideration. Calculations of neutron fluence were performed with neutron spectrum for fresh fuel, for 1 year and for 4 years burn up fuel. Correction factors for neutron fluence at the inner surface of the reactor vessel, in 1/4 depth of the vessel and in the air behind the vessel were obtained. The correction coefficient could be used when the neutron fluence is assessed so in verification when the measured activity of ex-vessel detectors is compared with calculated ones. (authors)

  5. Investigating the spectral anomaly with different reactor antineutrino experiments

    Directory of Open Access Journals (Sweden)

    C. Buck

    2017-02-01

    Full Text Available The spectral shape of reactor antineutrinos measured in recent experiments shows anomalies in comparison to neutrino reference spectra. New precision measurements of the reactor neutrino spectra as well as more complete input in nuclear data bases are needed to resolve the observed discrepancies between models and experimental results. This article proposes the combination of experiments at reactors which are highly enriched in U235 with commercial reactors with typically lower enrichment to gain new insights into the origin of the anomalous neutrino spectrum. The presented method clarifies, if the spectral anomaly is either solely or not at all related to the predicted U235 spectrum. Considering the current improvements of the energy scale uncertainty of present-day experiments, a significance of three sigma and above can be reached. As an example, we discuss the option of a direct comparison of the measured shape in the currently running Double Chooz near detector and the upcoming Stereo experiment. A quantitative feasibility study emphasizes that a precise understanding of the energy scale systematics is a crucial prerequisite in recent and next generation experiments investigating the spectral anomaly.

  6. Analysis of Tank 38H (HTF-38-17-52, -53) and Tank 43H (HTF-43-17-54, -55) Samples for Support of the Enrichment Control and Corrosion Control Programs

    Energy Technology Data Exchange (ETDEWEB)

    Hay, M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Coleman, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Diprete, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-08-09

    SRNL analyzed samples from Tank 38H and Tank 43H to support ECP and CCP. The total uranium in the Tank 38H surface sample was 41.3 mg/L while the sub-surface sample was 43.5 mg/L. The Tank 43H samples contained total uranium concentrations of 28.5 mg/L in the surface sample and 28.1 mg/L in the sub-surface sample. The U-235 percentage ranged from 0.62% to 0.63% for the Tank 38H samples and Tank 43H samples. The total uranium and percent U-235 results in the table appear slightly lower than recent Tank 38H and Tank 43H uranium measurements. The plutonium results in the table show a large difference between the surface and sub-surface sample concentrations for Tank 38H. The Tank 43H plutonium results closely match the range of values measured on previous samples. The Cs-137 results for the Tank 38H surface and sub-surface samples show similar concentrations slightly higher than the concentrations measured in recent samples. The Cs-137 results for the two Tank 43H samples also show similar concentrations within the range of values measured on previous samples. The four samples show silicon concentrations somewhat lower than the previous samples with values ranging from 124 to 168 mg/L.

  7. Study on the HTGR axial fuel loading

    International Nuclear Information System (INIS)

    Tanaka, Ryokichi

    1981-01-01

    In the nuclear and thermal design of reactor cores, it is one of the important targets for reactor safety to flatten fuel temperature distribution as far as possible to prevent local peaking. As a macroscopic method to prevent temperature peaking, it is considered to give exponential type power output distribution in coolant flow direction, while flattening radial power output distribution. Assuming rod-shaped fuel, the distribution of fuel heat generation is given by an exponential function under constant maximum fuel temperature condition in the direction of channel. By applying this function to neutron source distribution, and in a premise that U-235 loading can be changed continuously, the preliminary investigation on no-reflector core by one-dimensional one-group consideration, and then the analytical solution of the diffusion equation for a core with reflectors by two group one-dimensional approximation were carried out. The results of these investigations revealed that the U-235 concentration required for achieving exponential type power output distribution is necessary to have large concentration gradient up to the distance equivalent to the length of a few fuel elements from the core inlet, but it is sufficient to have constant concentration in downstream fuel elements, which is 0.8 to 0.9 times as much as the average value along the channel, except for large flow rate channel. (Wakatsuki, Y.)

  8. Verification study of thorium cross section in MVP calculation of thorium based fuel core using experimental data

    International Nuclear Information System (INIS)

    Mai, V. T.; Fujii, T.; Wada, K.; Kitada, T.; Takaki, N.; Yamaguchi, A.; Watanabe, H.; Unesaki, H.

    2012-01-01

    Considering the importance of thorium data and concerning about the accuracy of Th-232 cross section library, a series of experiments of thorium critical core carried out at KUCA facility of Kyoto Univ. Research Reactor Inst. have been analyzed. The core was composed of pure thorium plates and 93% enriched uranium plates, solid polyethylene moderator with hydro to U-235 ratio of 140 and Th-232 to U-235 ratio of 15.2. Calculations of the effective multiplication factor, control rod worth, reactivity worth of Th plates have been conducted by MVP code using JENDL-4.0 library [1]. At the experiment site, after achieving the critical state with 51 fuel rods inserted inside the reactor, the measurements of the reactivity worth of control rod and thorium sample are carried out. By comparing with the experimental data, the calculation overestimates the effective multiplication factor about 0.90%. Reactivity worth of the control rods evaluation using MVP is acceptable with the maximum discrepancy about the statistical error of the measured data. The calculated results agree to the measurement ones within the difference range of 3.1% for the reactivity worth of one Th plate. From this investigation, further experiments and research on Th-232 cross section library need to be conducted to provide more reliable data for thorium based fuel core design and safety calculation. (authors)

  9. Burnup influence on the WWER-1000 reactor vessel neutron fluence evaluation

    International Nuclear Information System (INIS)

    Panayotov, I.; Mihaylov, N.; Ilieva, K.; Kirilova, D.; Manolova, M.

    2009-01-01

    The neutron fluence of the vessels of the reactors is determined regularly accordingly the RPV Surveillance Program of Kozloduy NPP Unit 5 and 6 in order to assess the state of the metal vessel and their radiation damaging. The calculations are carried out by the method of discrete ordinates used in the TORT program for operated reactor cycles. An average reactor spectrum corresponding to fresh U-235 fuel is used as an input neutron source. The impact of the burn up of the fuel on the neutron fluence of WWER-1000 reactor vessel is evaluated. The calculations of isotopic concentrations of U-235 and Pu-239 corresponding to 4 years burn up were performed by the module SAS2H of the code system SCALE 4.4. Since fresh fuel or 4 years burn up fuel assembly are placed in periphery of reactor core the contribution of Pu-239 of first year burn up and of 4 years burn up is taken in consideration. Calculations of neutron fluence were performed with neutron spectrum for fresh fuel, for 1 year and for 4 years burn up fuel. Correction factors for neutron fluence at the inner surface of the reactor vessel, in ? depth of the vessel and in the air behind the vessel were obtained. The correction coefficient could be used when the neutron fluence is assessed so in verification when the measured activity of ex-vessel detectors is compared with calculated ones. (Authors)

  10. ZZ ECN-BUBEBO, ECN-Petten Burnup Benchmark Book, Inventories, Afterheat

    International Nuclear Information System (INIS)

    Kloosterman, Jan Leen

    1999-01-01

    Description of program or function: Contains experimental benchmarks which can be used for the validation of burnup code systems and accompanied data libraries. Although the benchmarks presented here are thoroughly described in literature, it is in many cases not straightforward to retrieve unambiguously the correct input data and corresponding results from the benchmark Descriptions. Furthermore, results which can easily be measured, are sometimes difficult to calculate because of conversions to be made. Therefore, emphasis has been put to clarify the input of the benchmarks and to present the benchmark results in such a way that they can easily be calculated and compared. For more thorough Descriptions of the benchmarks themselves, the literature referred to here should be consulted. This benchmark book is divided in 11 chapters/files containing the following in text and tabular form: chapter 1: Introduction; chapter 2: Burnup Credit Criticality Benchmark Phase 1-B; chapter 3: Yankee-Rowe Core V Fuel Inventory Study; chapter 4: H.B. Robinson Unit 2 Fuel Inventory Study; chapter 5: Turkey Point Unit 3 Fuel Inventory Study; chapter 6: Turkey Point Unit 3 Afterheat Power Study; chapter 7: Dickens Benchmark on Fission Product Energy Release of U-235; chapter 8: Dickens Benchmark on Fission Product Energy Release of Pu-239; chapter 9: Yarnell Benchmark on Decay Heat Measurements of U-233; chapter 10: Yarnell Benchmark on Decay Heat Measurements of U-235; chapter 11: Yarnell Benchmark on Decay Heat Measurements of Pu-239

  11. Study of homogeneous fuel cells type 10 x 10

    International Nuclear Information System (INIS)

    Montes, J.L.; Perusquia, R.; Ortiz, J.J.; Francois, J.L.; Marquez, C.M.

    2005-01-01

    At the moment in the National Institute of Nuclear Research (ININ) are carried out studies with the purpose of to establish a methodology that allows to carry out the neutron design of fuel cells of type 10 x 10. During the initial stage of the process of cells design, starting from the data that have to do with the planned energy demand it requires to be estimated the average value of the enrichment in U 235 w/o of the one assemble. The experience has shown that the accuracy that is achieved in this estimate it depends, among other factors, of the information (e.g. concentrations of U 235 and Gd 2 O 3 ) of the cells that its are disposed in that moment. For what we consider convenient to enlarge the available information by means of a series of calculations of cell physics; and to the one same time some aspects can be studied on the parameters that define the characteristics of a fuel cell. In this work the effect of the presence of different distributions of the concentrations of the fissile material is analyzed and of burnup poisons on the reactivity parameters of the cell as well as in the peak factor of local power (LPPF-Local Power Peaking Factor). (Author)

  12. Summary of a joint US-Japan study of potential approaches to reduce the attractiveness of various nuclear materials for use in a nuclear explosive device by a terrorist group

    Energy Technology Data Exchange (ETDEWEB)

    Bathke, C.G. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM (United States); Inoue, N.; Kuno, Y.; Mihara, T.; Sagara, H. [Japan Atomic Energy Agency, 4-49 Muramatsu, Tokai-mura, Naka-gun, Ibaraki 319-1184 (Japan); Ebbinghaus, B.B. [Lawrence Livermore National Laboratory, P.O. Box L-168, Livermore, CA 94551 (United States); Murphy, J.; Dalton, D. [National Nuclear Security Administration, Department of Energy, 1000 Independence Ave, SW, Washington, DC 20585 (United States); Nagayama, Y. [Ministry of Education, Culture, Sports, Science and Technology, 3-2-2 Kasumigaseki, Chiyoda-ku, Tokyo 100-8959 (Japan)

    2013-07-01

    This paper summarizes the results of a joint US-Japan study to establish a mutual understanding, through scientific-based study, of potential approaches to reduce the attractiveness of various nuclear materials for use in a terrorist nuclear explosive device (NED). 4 approaches that can reduce materials attractiveness with a very high degree of effectiveness are: -) diluting HEU with natural or depleted U to an enrichment of less than 10% U-235; -) storing Pu in nuclear fuel that is not man portable and with a dose rate greater or equal to 10 Gy/h at 1 m; -) storing Pu or HEU in heavy items, i.e. not transportable, provided the removal of the Pu or HEU from the item requires a purification/processing capability; and -) converting Pu and HEU to very dilute forms (such as wastes) that, without any security barriers, would require very long acquisition times to acquire a Category I quantity of Pu or of HEU. 2 approaches that can reduce materials attractiveness with a high degree of effectiveness are: -) converting HEU-fueled research reactors into LEU-fueled research reactors or dilute HEU with natural or depleted U to an enrichment of less than 20% U-235; -) converting U/Al reactor fuel into U/Si reactor fuel. Other approaches have been assessed as moderately or totally inefficient to reduce the attractiveness of nuclear materials.

  13. Determination of total and isotopic uranium by inductively coupled plasma-mass spectrometry at the Fernald Environmental Management Project

    International Nuclear Information System (INIS)

    Miller, F.L.; Bolin, R.N.; Feller, M.T.; Danahy, R.J.

    1995-01-01

    At the Fernald Environmental Management Project (FEMP) in southwestern Ohio, ICP-mass spectrometry (ICP-MS), with sample introduction by peristaltic pumping, is used to determine total and isotopic uranium (U-234, U-235, U-236 and U-238) in soil samples. These analyses are conducted in support of the environmental cleanup of the FEMP site. Various aspects of the sample preparation and instrumental analysis will be discussed. Initial sample preparation consists of oven drying to determine moisture content, and grinding and rolling to homogenize the sample. This is followed by a nitric/hydrofluoric acid digestion to bring the uranium in the sample into solution. Bismuth is added to the sample prior to digestion to monitor for losses. The total uranium (U-238) content of this solution and the U 235 /U 238 ratio are measured on the first pass through the ICP-MS. To determine the concentration of the less abundant U 234 and U 236 isotopes, the digestate is further concentrated by using Eichrom TRU-Spec extraction columns before the second pass through the ICP-MS. Quality controls for both the sample preparation and instrumental protocols will also be discussed. Finally, an explanation of the calculations used to report the data in either weight percent or activity units will be given

  14. Distribution of naturally occurring radioactivity and ¹³⁷Cs in the marine sediment of Farasan Island, southern Red Sea, Saudi Arabia.

    Science.gov (United States)

    Al-Zahrany, A A; Farouk, M A; Al-Yousef, A A

    2012-11-01

    The present work is a part of a project dedicated to measure the marine radioactivity near the Saudi Arabian coasts of the Red Sea and Arabian Gulf for establishing a marine radioactivity database, which includes necessary information on the background levels of both naturally occurring and man-made radionuclides in the marine environment. Farasan Islands is a group of 84 islands (archipelago), under the administration of the Kingdom of Saudi Arabia, in the Red Sea with its main island of Farasan, which is 50 km off the coast of Jazan City. The levels of natural radioactivity of (238)U, (235)U, (226)Ra, (232)Th and (40)K and man-made radionuclides such as (137)Cs in the grab sediment and water samples around Farasan Island have been measured using gamma-ray spectroscopy. The average activity concentrations of (238)U, (235)U, (226)Ra, (232)Th, (40)K and (137)Cs in the sediment samples were found to be 35.46, 1.75, 3.31, 0.92, 34.34 and 0.14 Bq kg(-1), respectively.

  15. Distribution of naturally occurring radioactivity and 137Cs in the marine sediment of Farasan island, southern red sea, Saudi Arabia

    International Nuclear Information System (INIS)

    Al-zahrany, A. A.; Farouk, M. A.; Al-yousef, A. A.

    2012-01-01

    The present work is a part of a project dedicated to measure the marine radioactivity near the Saudi Arabian coast of the Red Sea and Arabian Gulf for establishing a marine radioactivity database, which includes necessary information on the background levels of both naturally occurring and man-made radionuclides in the marine environment. Farasan Islands is a group of 84 islands (archipelago), under the administration of the Kingdom of Saudi Arabia, in the Red Sea with its main island of Farasan, which is 50 km off the coast of Jazan City. The levels of natural radioactivity of 238 U, 235 U, 226 Ra, 232 Thand 40 K and man-made radionuclides such as 137 Cs in the grab sediment and water samples around Farasan Island have been measured using gamma-ray spectroscopy. The average activity concentrations of 238 U, 235 U, 226 Ra, 232 Th, 40 K and 137C s in the sediment samples were found to be 35.46, 1.75, 3.31, 0.92, 34.34 and 0.14 Bq kg -1 , respectively. (authors)

  16. Comparison of Ruska and Rosemont pressure gages (U)

    International Nuclear Information System (INIS)

    Harvel, C.D.

    1991-01-01

    This paper reports that a 150,000 gallon tank was calibrated during the months of May and July of 1990. Six calibration runs were completed. Ruska and Rosemont pressure gages were installed to make in-tank liquid level measurements during the calibration process. A flow meter was used to measure the incremental volumes of water added to or removed from the tank. The Ruska and Rosemont gages were compared to determine the gage best suited for tank operation. One comparison criteria was the tolerance limits of error (LOE) for the predicted standardized in-tank volumes. For accountability purposes, the effects of the two gages on the LOE for the predicted inventory of U-235 were evaluated. The most important comparison criteria was the gage's contribution to the U-235 inventory LOE. The choice of which gage to use depends on the other measurement methods used for material accountability. The contributions to the inventory LOE were evaluated for two in-tank liquid level measurement methods, two concentration measurement methods, and one isotopic measurement method. The results indicate the Ruska pressure gage is best suited for tank operation only if the best concentration measurement method is used

  17. What the difference to use LEU and HEU fuel elements separately or together in a research reactor

    International Nuclear Information System (INIS)

    Kaya, S.; Uestuen, G.

    2005-01-01

    Concerning of nuclear material safety, most of the research reactors are advised to shift from HEU (high enriched-%93 U-235) to LEU (low enriched-%20 U-235) fuel elements. When LEU and HEU fuel elements are to be used together in a research reactor, some design and safety problems are encountered. According to use of the reactor, some research reactors such as MTR type may not show any considerable difference for HEU or LEU fuel elements, but the efficiency of radioisotope production generated by thermal neutron interaction may decrease about twenty-thirty percent when LEU fuel elements are used. Here, fine mesh-sized 3D neutronic analysis of TR-2 research reactor is presented to indicate the arising problem when LEU end HEU fuel elements are used together in a research reactor. Partial thermohydraulic analysis of the reactor is also given to show the betterness of the LEU fuel element design. However, there might be some points that should be noticed for safer operation of plate type fuelled research reactors. (author)

  18. A Field Test of the New Portable Gamma Spectrometry System

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Jung-Ki; Park, Uk Ryang; Park, Seunghoon; Chung, Heejun; Kwak, Sung-Woo [Korea Institute of Nuclear Non-proliferation and Control, Daejeon (Korea, Republic of); Kim, Yongkwn [NuCare Medical Systems, Inc., Incheon (Korea, Republic of)

    2015-10-15

    In order to perform a field test of the system, the measurement of U-235 enrichment for nuclear fuel pellets was conducted along with the IAEA Physical Inventory Verification (PIV) inspection at the KEPCO Nuclear Fuel (KNF). The enrichment value of U-235 was calculated based on the total counts of the 185.7 keV photopeak and compared with the reference line, drawn by certified sources. The goal of this study is to experimentally evaluate the system performance of the developed system. In this study, the new portable gamma spectrometry system showed a good linearity (R{sup 2}=1) but overestimated the enrichment values than IAEA inspection device. It could be caused by the stability of the new system since it found, right after this measurement, that the accuracy of the system gradually increases and becomes stable over time. Further steps will optimize the design parameter based on these results and repeat measurement with the same samples under the same environment.

  19. The University of Missouri Research Reactor, its fuel and productivity

    International Nuclear Information System (INIS)

    Brugger, R.M.; Schlapper, G.A.; Alger, D.M.

    1993-01-01

    This paper describes the University of Missouri Research Reactor (MURR) and presents a summary of contributions to education, research, and service. These efforts have helped offset the impact on the U.S. economy of research from other countries. Special emphasis is placed on fuel design developments that have allowed MuRR to keep the cost megawatt day (MWD) of fuel essentially constant. Also noted is the fact that the United States has missed some research opportunities because of a hold-the-line attitude. The slipping position of U.S. research reactors is compared with the rest of the world. As will be further outlined in the text, the MURR cannot (with available technology) decrease its U-235 enrichment level and maintain present research capabilities and fuel cycle costs. Data is presented to show how MURR, if permitted to use advanced fuel technology, could reduce fuel fabrication costs and onsite U-235 inventory. In addition it is shown that MURR could increase its capabilities provided that arbitrary institutional limits are removed

  20. The University of Missouri Research Reactor, its fuel and productivity

    International Nuclear Information System (INIS)

    Brugger, Robert M.; Schlapper, Gerald A.; Alger, Don M.

    1993-01-01

    This paper describes the University of Missouri Research Reactor (MURR) and presents a summary of contributions to education, research, and service. These efforts have helped offset the impact on the U. S. economy of research from other countries. Special emphasis is placed on fuel design developments that have allowed MURR to keep the cost megawatt day (MWD) of fuel essentially constant. Also noted is the fact that the United States has missed some research opportunities because of a hold-the-line attitude. The slipping position of U.S. research reactors is compared with the rest of the world. As will be further outlined in the text, the MURR cannot (with available technology) decrease its U-235 enrichment level and maintain present research capabilities and fuel cycle costs. Data is presented to show how MURR if permitted to use advanced fuel technology, could reduce fuel fabrication costs and onsite U-235 inventory. In addition it is shown that MURR could increase its capabilities provided that arbitrary institutional limits are removed

  1. HEU to LEU conversion and blending facility: Oxide blending alternative to produce LEU oxide for commercial use

    International Nuclear Information System (INIS)

    1995-09-01

    The United States Department of Energy (DOE) is examining options for the disposition of surplus weapons-usable fissile materials and storage of all weapons-usable fissile materials. Disposition is a process of use or disposal of material that results in the material being converted to a form that is substantially and inherently more proliferation-resistant than the original form. Examining options for increasing the proliferation resistance of highly enriched uranium (HEU) is part of this effort. This document provides data to be used in the environmental impact analysis for the oxide blending HEU disposition option. This option provides for a yearly HEU throughput of 1 0 metric tons (MT) of uranium metal with an average U235 assay of 50% blended with 165 MT of natural assay triuranium octoxide (U 3 O 8 ) per year to produce 177 MT of 4% U235 assay U 3 O 8 , for LWR fuel. Since HEU exists in a variety of forms and not necessarily in the form to be blended, worst case scenarios for preprocessing prior to blending will be assumed for HEU feed streams

  2. 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

  3. Technology development for nuclear material accountability

    International Nuclear Information System (INIS)

    Hong, Jong Sook; Lee, Byung Doo; Cha, Hong Ryul; Choi, Hyoung Nae; Park, Ho Jun.

    1990-01-01

    Neutron yields from 19 F(α,n) 22 Na reaction of uranium neutron interaction with uranium-bass materials, and the characteristics of shielded neutron assay probe have been studied. On the basis of the above examination, U-235 enrichment in UF 6 cylinders like model 30B and model 48Y was measured by the reaction and U-235 contents in the containers by non-destructive total passive neutron assay method. Total measurement efficiency as a result was found to be 6.44 x 10 -4 and 1.25 x 10 -4 for model 30B and model 40Y UF 6 cylinder, respectively. The uncertainty of measured enrichment as compared to Tag value obtained from chemical analysis approached about 5 % of relative error at 95 % confidence interval. In the follow-up action for the previously developed (1988) computer system of nuclear material accounting the error searching and treatment routine in accordance with code 10, of IAEA and respective facility attachment has been added to easing the burden of manual error correction by operator. In addition, the procedure for LEMUF calculation has been prepared to help bulk facility operators evaluating MUF in the period of material balance. (author)

  4. Low enrichment Mo-99 target development program at ANSTO

    International Nuclear Information System (INIS)

    Donlevy, Therese M.; Anderson, Peter J.; Beattie, David; Braddock, Ben; Fulton, Scott; Godfrey, Robert; Law, Russell; McNiven, Scott; Sirkka, Pertti; Storr, Greg; Wassink, David; Wong, Alan; Yeoh, Guan

    2002-01-01

    The Australian Nuclear Science and Technology Organisation (ANSTO, formerly AAEC) has been producing fission product Mo-99 in HIFAR, from the irradiation of Low Enrichment Uranium (LEU) UO 2 targets, for nearly thirty years. Over this period, the U-235 enrichment has been increased in stages, from natural to 1.8% to 2.2%. The decision to provide Australia with a replacement research reactor (RRR) for HIFAR has created an ideal opportunity to review and improve the current Mo-99 production process from target design through to chemical processing and waste management options. ANSTO has entered into a collaboration with Argonne National Laboratory (RERTR) to develop a target using uranium metal foil with U-235 enrichment of less than 20% The initial focus has been to demonstrate use of LEU foil targets in HIFAR, using existing irradiation methodology. The current effort focussed on designing a target assembly with optimised thermohydraulic characteristics to accommodate larger LEU foils to meet Mo-99 production needs. The ultimate goal is to produce an LEU target suitable for use in the Replacement Research Reactor when it is commissioned in 2005. This paper reports our activities on: - The regulatory approval processes required in order to undertake irradiation of this new target; -Supporting calculations (neutronics, computational fluid dynamics) for safety submission; - Design challenges and changes to prototype irradiation; - Trial irradiation of LEU foil target in HIFAR; - Future target and rig development program at ANSTO. (author)

  5. Temperature variation of criticality of thermal reactor lattices

    International Nuclear Information System (INIS)

    Velner, S.; Rothenstein, W.

    1975-01-01

    Departures from the asymptotic mode in the experimental setup have been examined in detail for two assemblies, one exponential, the other critical. It was found that the flux shape differed noticeably from the asymptotic mode in the core region especially for the exponential assemblies. On the other hand the departure from the fundamental mode has very little effect on the change of material buckling with temperature. Results of the calculations and their comparison with experiment are presented. The variation of material buckling with temperature is the same for ENDF/B-II and for ENDF/B-IV data, both for asymptotic reactor theory and for the buckling values derived from the flux calculated with the SN code. The results obtained with ENDF/B-IV data for both lattices are shown. In the small exponential assembly the results derived from S-4 calculations are compared with experiment. In the critical assembly the ratio of U-238 to U-235 fissions delta 28 and the relative conversion ratio - the ratio of U-238 captures to U-235 fissions in the lattice compared with the same quantity in a thermal column - are also shown. In both cases the experimental change of buckling with temperature is smaller than the calculated change. (B.G.)

  6. Critical and subcritical mass calculations of fissionable nuclides based on JENDL-3.2+

    International Nuclear Information System (INIS)

    Okuno, H.

    2002-01-01

    We calculated critical and subcritical masses of 10 fissionable actinides ( 233 U, 235 U, 238 Pu, 239 Pu, 241 Pu, 242m Am, 243 Cm, 244 Cm, 249 Cf and 251 Cf) in metal and in metal-water mixtures (except 238 Pu and 244 Cm). The calculation was made with a combination of a continuous energy Monte Carlo neutron transport code, MCNP-4B2, and the latest released version of the Japanese Evaluated Nuclear Data Library, JENDL-3.2. Other evaluated nuclear data files, ENDF/B-VI, JEF-2.2, and JENDL-3.3 in its preliminary version were also applied to find differences in results originated from different nuclear data files. For the so-called big three fissiles ( 233 U, 235 U and 239 Pu), analyzing the criticality experiments cited in ICSBEP Handbook validated the code-library combination, and calculation errors were consequently evaluated. Estimated critical and lower limit critical masses of the big three in a sphere with/without a water or SS-304 reflector were supplied, and they were compared with the subcritical mass limits of ANS-8.1. (author)

  7. Use of non-proliferation fuel cycles in the HTGR

    International Nuclear Information System (INIS)

    Baxter, A.M.; Merrill, M.H.; Dahlberg, R.C.

    1978-10-01

    All high-temperature gas-cooled reactors (HTGRs) built or designed to date utilize a uranium-thorium fuel cycle (HEU/Th) in which fully-enriched uranium (93% U-235) is the initial fuel and thorium is the fertile material. The U-233 produced from the thorium is recycled in subsequent loadings to reduce U-235 makeup requirements. However, the recent interest in proliferation-proof fuel cycles for fission reactors has prompted a review and evaluation of possible alternate cycles in the HTGR. This report discusses these alternate fuel cycles, defines those considered usable in an HTGR core, summarizes their advantages and disadvantages, and briefly describes the effect on core design of the most important cycles. Examples from design studies are also given. These studies show that the flexibility afforded by the HTGR coated-particle fuel design allows a variety of alternative cycles, each having special advantages and attractions under different circumstances. Moreover, these alternate cycles can all use the same fuel block, core layout, control scheme, and basic fuel zoning concept

  8. Development of a computer program of fast calculation for the pre design of advanced nuclear fuel 10 x 10 for BWR type reactors

    International Nuclear Information System (INIS)

    Perusquia, R.; Montes, J.L.; Ortiz, J.J.

    2005-01-01

    In the National Institute of Nuclear Research (ININ) a methodology is developed to optimize the design of cells 10x10 of assemble fuels for reactors of water in boil or BWR. It was proposed a lineal calculation formula based on a coefficients matrix (of the change reason of the relative power due to changes in the enrichment of U-235) for estimate the relative powers by pin of a cell. With this it was developed the computer program of fast calculation named PreDiCeldas. The one which by means of a simple search algorithm allows to minimize the relative power peak maximum of cell or LPPF. This is achieved varying the distribution of U-235 inside the cell, maintaining in turn fixed its average enrichment. The accuracy in the estimation of the relative powers for pin is of the order from 1.9% when comparing it with results of the 'best estimate' HELIOS code. With the PreDiCeldas it was possible, at one minimum time of calculation, to re-design a reference cell diminishing the LPPF, to the beginning of the life, of 1.44 to a value of 1.31. With the cell design with low LPPF is sought to even design cycles but extensive that those reached at the moment in the BWR of the Laguna Verde Central. (Author)

  9. Design of 50 MWe HTR-PBMR reactor core and nuclear power plant fuel using SRAC2006 programme

    International Nuclear Information System (INIS)

    Bima Caraka Putra; Yosaphat Sumardi; Yohannes Sardjono

    2014-01-01

    This research aims to assess the design of core and fuel of nuclear power plant type High Temperature Reactor-Pebble Bed Modular Reactor 50 MWe from the Beginning of Life (BOL) to Ending of life (EOL) with eight years operating life. The parameters that need to be analyzed in this research are the temperature distribution inside the core, quantity enrichment of U 235 , fuel composition, criticality, and temperature reactivity coefficient of the core. The research was conducted with a data set of core design parameters such as nuclides density, core and fuel dimensions, and the axial temperature distribution inside the core. Using SRAC2006 program package, the effective multiplication factor (k eff ) values obtained from the input data that has been prepared. The results show the value of the criticality of core is proportional to the addition of U 235 enrichment. The optimum enrichment obtained at 10.125% without the use of burnable poison with an excess reactivity of 3.1 2% at BOL. The addition Gd 2O3 obtained an optimum value of 12 ppm burnable poison with an excess reactivity 0.38 %. The use of Er 2O3 with an optimum value 290 ppm has an excess reactivity 1.24 % at BOL. The core temperature reactivity coefficient with and without the use of burnable poison has a negative values that indicates the nature of its inherent safety. (author)

  10. Irradiation program of slightly enriched fuel elements at the Atucha I nuclear power plant

    International Nuclear Information System (INIS)

    Casario, J.A.; Cesario, R.H.; Perez, R.A.; Sidelnik, J.I.

    1987-01-01

    An irradiation program of fuel elements with slightly enriched uranium is implemented, tending to the homogenization of core at Atucha I nuclear power plant. The main benefits of the enrichment program are: a) to extend the average discharge burnup of fuel elements, reducing the number of elements used to generate the same amount of energy. This implies a smaller annual consumption of elements and consequently the reduction of transport and replacement operations and of the storage pool systems as well as that of radioactive wastes; b) the saving of uranium and structural materials (Zircaloy and others). In the initial stage of program an homogeneous core enrichment of 0.85% by weight of U-235 is anticipated. The average discharge burnup of fuel elements, as estimated by previous studies, is approximately 11.6 MW d/kg U. The annual consumption of fuel elements is reduced from 396 of natural uranium to 205, with a load factor of 0.85. It is intended to reach the next equilibrium steps with an enrichment of 1.00 and 1.20% in U-235. (Author)

  11. Some aspects of the nuclear fission process

    International Nuclear Information System (INIS)

    Netter, F.

    1961-01-01

    In the following report one can find first a short general view on the present situation of our knowledge concerning the nuclear fission process, namely on the nucleus going through the saddle-point. Then there are some aspects connected with the excitation energy of the fissioning nucleus. The measurements made at Saclay on the fast neutron fission cross-section of U 233 , U 235 , Pu 239 , U 238 are described at the beginning of this work. It appears that for U 233 there is some characteristic shape modulation of the cross-section curve, in relation with the collective excited state of the deformed nucleus at the saddle-point. Good evidence of this is also given by the study of the relative fission rate with emission of long-range particles; it appears also that this ternary fission rate does not change substantially for neutron between thermal energy and 2 MeV, but that is very lower for the compound nucleus U 239 than for even-even compound nuclei. At the end there are some experiments on the strong 4,5 MeV gamma-ray originated by slow neutron absorption in U 235 . Time-of-flight device is used to establish that this 4,5 MeV gamma-ray seems mostly connected with radiative capture. (author) [fr

  12. In-reactor behaviour of centrifugally atomized U3Si dispersion fuel irradiated at high temperature in HANARO

    International Nuclear Information System (INIS)

    Kim, Ki Hwan; Park, Jong Man; Yoo, Byeong Ok; Park, Dae Kyu; Lee, Choong Sung; Kim, Chang Kyu

    2002-01-01

    The irradiation test on full-size U 3 Si dispersion fuel elements, prepared by centrifugal atomization and conventional comminution method, has been performed up to about 77 at.% U-235 in maximum burn-up at CT hole position having the highest power condition in the HANARO reactor, in order to examine the irradiation performance of the atomized U 3 Si for the driver fuels of HANARO. The in-reactor interaction of the atomized U 3 Si dispersion fuel meats is generally assumed to be acceptable with the range of 5-15 μm in average thickness. The atomized spherical particles have more uniform and thinner reaction layer than the comminuted irregular particles. The U 3 Si particles have relatively fine and uniform size distribution of fission gas bubbles, irrespective of the powdering method. The bubble population in the atomized particles appears to be finer and more homogeneous with the characteristics of narrower bubble size distribution than that of the comminuted fuel. The atomized U 3 Si dispersion fuel elements exhibit sound swelling behaviours of 5 % in ΔV/V m even at ∼77 at.% U-235 burn-up, which meets with the safety criterion of the fuel rod, 20vol.% for HANARO. The atomized U3Si dispersion fuel elements show smaller swelling than the comminuted fuel elements

  13. 1987 monitoring report for the defense waste lysimeters

    International Nuclear Information System (INIS)

    McIntyre, P.F.

    1987-01-01

    Low levels of radionuclides migrate through the soil to the sump. This report updates previous monitoring reports and discusses results obtained during the past year of operation. The effluents from the forty defense waste lysimeters continue to be analyzed on a monthly basis for gamma emitting radionuclides and quarterly for alpha emitting radionuclides and Sr-90. Cobalt-60, Sr-90, Sb-125, U-235, Pu-238, Pu-239 and Am-241 continue to be detected in sump effluent. Detectable levels of cobalt-60 and antimony-125 are each observed in only one lysimeter. Manganese-54, Ru-106 and Th-234 are no longer detected in effluent from any lysimeter. Significant levels of Sr-90 and Pu-238 are observed from several lysimeters, while others continue to show low levels of U-235, Pu-239 and Am-241. The release rates for transporting radionuclides through the soil to the sump indicate that migration is independent of whether a lysimeter is operated in a saturated or unsaturated mode. Pine trees continue to grow on the ten foot diameter lysimeters. No sampling of needles or woody stem portions was performed. The purpose of analyzing pine trees growing on lysimeters is to measure the amount of radionuclide uptake by the pine trees as their root systems come in contact with the waste material. 6 refs., 2 figs., 49 tabs

  14. Installation for the Mo-99 production from fission products

    International Nuclear Information System (INIS)

    Marques, R.O.; Cristini, P.R.; Marziale, D.P.; Furnari, E.S.; Fernandez, H.O.

    1988-01-01

    The installation to produce Mo-99 from nuclear fission started going on August 12th 1985 in Ezeiza Atomic Center. The characteristics of the process, the emplacement of the power plant, target, and irradiation conditions are presented. The targets are plates with a nucleus of Al/U alloy, with U-235 enriched to 90 % covered by Al plates. Each plate consists of about 1.10 -3 Kg of U-235 and 13.10 -3 Kg of Al. The plates are irradiated with a 3.10 13 n cm -2 s -1 flux during five days in the RA-3 nucleus. The Mo-99 separation method, is presented, where it is foreseen te I-131 separation. An account of the treatment of solid, liquid and gaseous waste is provided. An equipment to transfer the filter precipitate was designed in order to recover the U. The installation to continue the U recovery process, to separate I-131 and Xe-133 and to incorporate a Mo-99 purification stage for sublimation is being extended. (M.E.L.) [es

  15. Measurement of fast neutron induced fission cross section of minor-actinide

    International Nuclear Information System (INIS)

    Hirakawa, Naohiro

    1997-03-01

    In fuel cycles with recycled actinide, core characteristics are largely influenced by minor actinide (MA: Np, Am, Cm). Accurate nuclear data of MA such as fission cross section are required to estimate the effect of MA with high accuracy. In this study, fast neutron induced fission cross section of MA is measured using Dynamitron Accelerator in Tohoku University. The experimental method and the samples, which were developed or introduced during the last year, were improved in this fiscal year: (1) Development of a sealed fission chamber, (2) Intensification of Li neutron target, (3) Improvement of time-resolution of Time-of-Flight (TOF) electronic circuit, (4) Introduction of Np237 samples with large sample mass and (5) Introduction of a U235 sample with high purity. Using these improved tools and samples, the fission cross section ratio of Np237 relative to U235 was measured between 5 to 100 keV, and the fission cross section of Np237 was deduced. On the other hand, samples of Am241 and Am243 were obtained from Japan Atomic Energy Research Institute (JAERI) after investigating fission cross section of two americium isotopes (Am241 and Am 243) which are important for core physics calculation of fast reactors. (author)

  16. Report on the NAT-9 quality control exercise on uranium isotopes in two soil samples

    International Nuclear Information System (INIS)

    Bleise, Andreas

    2001-04-01

    The International Atomic Energy Agency (IAEA) section of Nutritional and Health related Environmental Studies (NAHRES) organized a quality control study for laboratories analysing samples from the UNEP field mission to Kosovo. Quality control was the major responsibility of the IAEA in the UN field assessment team. The NAT-9 quality control study consists of two soil materials from the IAEA Laboratories in Seibersdorf. The scope of this exercise was to determine the content of the uranium isotopes U-234, U-235 and U-238. The IAEA did not provide specific instructions, the participants were encouraged to apply their established analytical procedures to the samples. Five laboratories were invited to participate, four laboratories submitted results. For each soil sample 10 laboratory mean values were reported, using ICP-MS (3 laboratories) and α-spectrometry (1 laboratory). The participating laboratories were capable to distinguish the different uranium isotopes. All laboratories obtained the natural uranium ratio between U-235 and U-238. However, the results highlight a particular analytical weak spot. Although the methods of measuring the analytical signals are highly dependable, the sample preparation steps, in particular the sample dissolution procedure, appears to be lacking total quality control and has contributed to the deviations from the reported target values. One laboratory has documented evidence that extensive and well-controlled digestion methods can yield measurement results close to the target values. (author)

  17. IFBA credit in the Shearon Harris fuel racks with Vantage 5 fuel

    International Nuclear Information System (INIS)

    Boyd, W.A.; Schmidt, R.F.; Erwin, R.D.

    1989-01-01

    At the Shearon Harris nuclear plant, fuel management strategies are being considered which will result in feed fuel enrichments approaching 5.0 w/o U-235. These types of enrichments require a new criticality analysis to raise the existing fuel rack enrichment limit. It is receiving Westinghouse Vantage 5 fuel with integral fuel burnable absorber (IFBA) rods providing the depletable neutron absorber. An analysis was performed on the fuel racks which demonstrates that fuel enriched up to 5.0 w/o U-235 can be stored by taking credit for the IFBA rods present in the high enriched fuel assemblies. This is done by calculating the maximum Vantage 5 fuel assembly reactivity that can be placed in the fuel racks and meet the criticality K-eff limit. A methodology is also developed which conservatively calculates the minimum number of IFBA rods needed per assembly to meet the fuel rack storage limits. This eliminates the need for core designers to determine assembly K-inf terms for every different enrichment/IFBA combination

  18. Neutronics substantiation of possibility for conversion of the WWR-K reactor core to operation with low-enriched fuel

    International Nuclear Information System (INIS)

    Arinkin, F.M.; Gizatulin, Sh.H.; Zhantikin, T.M.; Koltochnik, S.N.; Takibaev, A.Zh.; Talanov, S.V.; Chakrov, P.V.; Chekushina, L.V.

    2002-01-01

    The studies are aimed to calculation and experimental justification of possibility for conversion of the WWR-R reactor core to low-enriched nuclear fuel (the 19.75-% enrichment in isotope U-235), resulting in reducing the risk of non-sanctioned proliferation of nuclear materials which can be used as weapons materials. The analysis of available published data, related to problem of reduction of enrichment in the fuel used in research thermal reactors, has been carried out. Basing on the analysis results, reference fuel compositions have been chosen, in particular, uranium dioxide (UO 2 ) in aluminum master form and the UA1 4 alloy. Preliminary calculations have shown that, with the WWR-K reactor core preserved existing critical characteristics (the fuel composition: UA1 4 ), the uranium concentration in the fuel element is to be increased by a factor of 2.0-2.2, being impossible technologically. The calculations have been performed by means of the Monte Carlo computational codes. The program of optimal conversion of the WWR-K reactor core to low-enriched fuel has been developed, including: development of calculation models of the reactor core, composed of various designs of fuel elements and fuel assemblies (FA), on a base of corresponding computational codes (diffusion, statistical, etc.); implementation of experiments in the zero-power reactor (critical assembly) with the WWR-C-type FA, in view of correction of the computational constants used in calculations; implementation of reactor core neutronics calculations, in view of selection of the U-235 optimal content in the low-enriched fuel elements and choice of FA reload strategy at the regime of reactor core after burning; determination of the fuel element specification; determination of the critical and operational loads for the reactor core composed of rod/tubular fuel elements; calculation of the efficiency of the protection control system effectors, optimization of its composition, number and locations in the

  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. In situ observation of thermal relaxation of interstitial-vacancy pair defects in a graphite gap

    International Nuclear Information System (INIS)

    Urita, Koki; Suenaga, Kazu; Iijima, Sumio; Sugai, Toshiki; Shinohara, Hisanori

    2005-01-01

    Direct observation of individual defects during formation and annihilation in the interlayer gap of double-wall carbon nanotubes (DWNT) is demonstrated by high-resolution transmission electron microscopy. The interlayer defects that bridge two adjacent graphen layers in DWNT are stable for a macroscopic time at the temperature below 450 K. These defects are assigned to a cluster of one or two interstitial-vacancy pairs (I-V pairs) and often disappear just after their formation at higher temperatures due to an instantaneous recombination of the interstitial atom with vacancy. Systematic observations performed at the elevated temperatures find a threshold for the defect annihilation at 450-500 K, which, indeed, corresponds to the known temperature for the Wigner energy release

  1. In situ observation of thermal relaxation of interstitial-vacancy pair defects in a graphite gap.

    Science.gov (United States)

    Urita, Koki; Suenaga, Kazu; Sugai, Toshiki; Shinohara, Hisanori; Iijima, Sumio

    2005-04-22

    Direct observation of individual defects during formation and annihilation in the interlayer gap of double-wall carbon nanotubes (DWNT) is demonstrated by high-resolution transmission electron microscopy. The interlayer defects that bridge two adjacent graphen layers in DWNT are stable for a macroscopic time at the temperature below 450 K. These defects are assigned to a cluster of one or two interstitial-vacancy pairs (I-V pairs) and often disappear just after their formation at higher temperatures due to an instantaneous recombination of the interstitial atom with vacancy. Systematic observations performed at the elevated temperatures find a threshold for the defect annihilation at 450-500 K, which, indeed, corresponds to the known temperature for the Wigner energy release.

  2. Influence of annealing temperature on the nanostructure TiO2-SnO2 prepared by electron gun method on the glass substrate and the aluminum/glass

    Directory of Open Access Journals (Sweden)

    N Beigmohammadi

    2013-09-01

    Full Text Available  TiO2-SnO2 thin films were coated on glass and Al / glass substrates by electron gun method. In coating process, the vacuum was 1.5×10-5 torr. Then, films were annealed at 450, 500 and 550 ˚ C. The crystallographic structure and film morphology were investigated by means of XRD and SEM. The electrical (I-V and optical properties were studied by the two point props system and UV/Vis/NIR spectrophotometer. The results showed the films under 550 ˚ C were crystalline. The thickness and grain size were 350 and 50 nm respectively. The electrical conductivity in the sample with Al / glass substrate under 550 ˚ C was better than the other samples. When temperature increased, the energy gap decreased from 4.05 to 4.03 eV for direct cases.

  3. Growth of optical grade germanium crystals

    International Nuclear Information System (INIS)

    Waris, M.; Akhtar, M.J.; Mehmood, N.; Ashraf, M.; Siddique, M.

    2011-01-01

    A novel design of Czochralski( CZ ) growth station in a low frequency induction furnace is described and growth of optical grade Ge crystal as a test material is performed achieving a flat solid-liquid interface shape. Grown Ge crystals are annealed in air at 450 -500 deg. C for 4 hrs and then characterized by determination of crystallographic orientation by Laue (back-reflection of X-rays) method, dislocation density studies by etch-pits formation, measuring electrical resistivity by 4-probe technique, conductivity type determination by hot probe method, measurement of hardness on Moh's scale and optical transmission measurement in IR region. The results obtained are compared to those reported in the literature. The use of this growth station for other materials is suggested. (author)

  4. Safflower (Carthamus tinctorius L. yield as affected by nitrogen fertilization and different water regimes

    Directory of Open Access Journals (Sweden)

    Reginaldo Ferreira-Santos

    2018-04-01

    Full Text Available Due to its origin and hardiness, safflower is usually cultivated in low-fertility soils with few inputs and no irrigation. In Brazil, little is known about its response to nitrogen (N and irrigation. This study was carried out near the city of Engenheiro Coelho, SP, Brazil, in 2014, in order to determine the effect of increasing nitrogen application rates (0, 50, 100, 150, 200, 250, 300, 350, 400, 450, 500, 550 and 600 kg ha-1 on safflower cultivation under irrigation and rainfed conditions. The use of irrigation during drought periods allowed stress reduction and significantly increased yield components and grain yield. Safflower yield was influenced by the interaction between water regimes and nitrogen rates. Grain yield may vary depending on several factors, however, maximum yield was achieved with rates of 208 and 214 kg N ha-1 under irrigation and rainfed conditions, respectively. For oil yield, 200 kg N ha-1 were sufficient, regardless of the water regime.

  5. Principle component analysis and linear discriminant analysis of multi-spectral autofluorescence imaging data for differentiating basal cell carcinoma and healthy skin

    Science.gov (United States)

    Chernomyrdin, Nikita V.; Zaytsev, Kirill I.; Lesnichaya, Anastasiya D.; Kudrin, Konstantin G.; Cherkasova, Olga P.; Kurlov, Vladimir N.; Shikunova, Irina A.; Perchik, Alexei V.; Yurchenko, Stanislav O.; Reshetov, Igor V.

    2016-09-01

    In present paper, an ability to differentiate basal cell carcinoma (BCC) and healthy skin by combining multi-spectral autofluorescence imaging, principle component analysis (PCA), and linear discriminant analysis (LDA) has been demonstrated. For this purpose, the experimental setup, which includes excitation and detection branches, has been assembled. The excitation branch utilizes a mercury arc lamp equipped with a 365-nm narrow-linewidth excitation filter, a beam homogenizer, and a mechanical chopper. The detection branch employs a set of bandpass filters with the central wavelength of spectral transparency of λ = 400, 450, 500, and 550 nm, and a digital camera. The setup has been used to study three samples of freshly excised BCC. PCA and LDA have been implemented to analyze the data of multi-spectral fluorescence imaging. Observed results of this pilot study highlight the advantages of proposed imaging technique for skin cancer diagnosis.

  6. Converting of oil shale and biomass into liquid hydrocarbons via pyrolysis

    International Nuclear Information System (INIS)

    Kılıç, Murat; Pütün, Ayşe Eren; Uzun, Başak Burcu; Pütün, Ersan

    2014-01-01

    Highlights: • Co-processing of oil shale with an arid land biomass for hydrocarbon production. • Co-pyrolysis in TGA and fixed-bed reactor. • Characterization of oil and char. - Abstract: In this study, co-pyrolytic behaviors of oil shale and Euphorbia rigida were investigated at different temperatures in a fixed bed reactor at 450, 500, and 550 °C with a heating rate of 10 °C/min in the presence of nitrogen atmosphere. The obtained solid product (char) and liquid product (tar) were analyzed by using different types of characterization techniques. Experimental results showed co-pyrolysis of oil shale and biomass could be an environmental friendly way for the transformation of these precursors into valuable products such as chemicals or fuels

  7. Effects of the addition of H2O and NH4OH in the electrical properties of thin films of Y2O3 deposited by pyrolytic spray

    International Nuclear Information System (INIS)

    Herrera S, H.J.; Alarcon F, G.; Aguilar F, M.; Falcony, C.; Garcia H, M.; Guzman M, J.; Araiza I, J.J.

    2005-01-01

    In this work we studied the electrical properties of yttrium oxide thin films obtained by spray pyrolysis from Y(acac) 3 and N,N-DMF. The films were deposited on Si(100) substrates at temperatures of 400, 450, 500 and 550 C. The electrical characteristic of the films was improved when a mist of H 2 O and/or NH 4 0H was simultaneously added to the deposition system. Current and capacitance versus voltage measurements were obtained when the Y 2 O 3 films were integrated in MOS (Metal-Oxide-Semiconductor) structures. Y 2 O 3 films with a dielectric constant up to 15 were obtained. The films can stand electric fields up to 2 MV/cm. An interface state density in the range of 10 10 -10 11 cm -2 eV -1 was measured at midgap from the high and low frequency capacitance measurements. (Author)

  8. Kinetic modeling of ethylbenzene dehydrogenation over hydrotalcite catalysts

    KAUST Repository

    Atanda, Luqman

    2011-07-01

    Kinetics of ethylbenzene dehydrogenation to styrene was investigated over a series of quaternary mixed oxides of Mg3Fe0.25Me0.25Al0.5 (Me=Co, Mn and Ni) catalysts prepared by calcination of hydrotalcite-like compounds and compared with commercial catalyst. The study was carried out in the absence of steam using a riser simulator at 400, 450, 500 and 550°C for reaction times of 5, 10, 15 and 20s. Mg3Fe0.25Mn0.25Al0.5 afforded the highest ethylbenzene conversion of 19.7% at 550°C. Kinetic parameters for the dehydrogenation process were determined using the catalyst deactivation function based on reactant conversion model. The apparent activation energies for styrene production were found to decrease as follows: E1-Ni>E1-Co>E1-Mn. © 2011 Elsevier B.V.

  9. The Jupiter program

    International Nuclear Information System (INIS)

    Ramirez, J.J.

    1995-01-01

    Jupiter is a Sandia initiative to develop the next generation of fast Z-pinch drivers for applications to high energy density physics, inertial confinement fusion, and radiation effects simulation. Jupiter will also provide unique capabilities for science research in a broad spectrum of areas involving ultra high magnetic fields, hot/dense plasmas, x-ray physics, intense neutron sources, etc. The program is based on the premise that a single facility using magnetically driven implosions can meet the needs in these multiple program areas. Jupiter requires a 450-500 TW, 8-10 MV, ∼ 100 ns pulsed power generator to impart - 15 MJ kinetic energy to an imploding plasma load. The baseline concept uses a highly modular, robust architecture with demonstrated performance reliability. The design also has the flexibility to drive longer implosion times. This paper describes the Jupiter accelerator concept, and the research underway to establish the technological readiness to proceed with construction of the facility

  10. Dynamic Diffraction Studies on the Crystallization, Phase Transformation, and Activation Energies in Anodized Titania Nanotubes.

    Science.gov (United States)

    Albetran, Hani; Vega, Victor; Prida, Victor M; Low, It-Meng

    2018-02-23

    The influence of calcination time on the phase transformation and crystallization kinetics of anodized titania nanotube arrays was studied using in-situ isothermal and non-isothermal synchrotron radiation diffraction from room temperature to 900 °C. Anatase first crystallized at 400 °C, while rutile crystallized at 550 °C. Isothermal heating of the anodized titania nanotubes by an increase in the calcination time at 400, 450, 500, 550, 600, and 650 °C resulted in a slight reduction in anatase abundance, but an increase in the abundance of rutile because of an anatase-to-rutile transformation. The Avrami equation was used to model the titania crystallization mechanism and the Arrhenius equation was used to estimate the activation energies of the titania phase transformation. Activation energies of 22 (10) kJ/mol for the titanium-to-anatase transformation, and 207 (17) kJ/mol for the anatase-to-rutile transformation were estimated.

  11. Performance of coils wound from long lengths of surface-coated, reacted, BSCCO-2212 conductor

    Energy Technology Data Exchange (ETDEWEB)

    Walker, M.S.; Hazelton, D.W.; Gardner, M.T. [Intermagnetics General Corp., Latham, NY (United States)] [and others

    1996-10-01

    React-before-wind surface-coated BSCCO-2212 is being established as a relatively low cost HTS conductor for practical applications. Quality tape is presently being manufactured in 450-500m lengths at a cost estimated to be 1/3-1/5 of the industry costs of BSCCO-2223 powder-in-tube tape. Robust, mechanically sound coils for applications ranging from NMR insert magnets to transformer windings are being made from this BSCCO-2212 tape. The coils have performed consistently through test and thermal cycling without degradation and as projected from short sample measurements. A hybrid approach, which uses mainly BSCCO- 2212 augmented by BSCCO-2223 conductor in the high radial field end regions, is expected to halve magnet system costs.

  12. Performance of the Antares large area cold cathode electron gun

    International Nuclear Information System (INIS)

    Rosocha, L.A.; Mansfield, C.R.

    1983-01-01

    The performance of the electron gun which supplies ionization for the Antares high power electron beam sustained CO 2 laser power amplifier is described. This electron gun is a coaxial cylindrical cold cathode vacuum triode having a total electron aperture area of approximately 9 m 2 . Electrons are extracted from the gun in pulses of 3-6 μs duration, average current densities of 40-60 ma/cm2, and electron energies of 450-500 keV. The main areas of discussion in this paper are the performance in terms of grid control, current density balance, and current runaway due to breakdown limitations. Comparison of the experimental results with the predictions of a theoretical model for the electron gun will also be presented

  13. Effects of the addition of H2O and NH4OH in the optical and structural properties of the thin films of Y2O3 deposited by pyrolytic spray

    International Nuclear Information System (INIS)

    Alarcon F, G.; Carvajal V, R.; Aguilar F, M.; Falcony, C.

    2005-01-01

    In this work we studied the optical and structural properties of yttrium oxide thin films deposited by spray pyrolysis. Yttrium acetylacetonate was used as raw material and N, N-DMF was used as solvent. The films were deposited on Si (100) and Si (111) substrates at temperatures of 400, 450, 500 and 550 C. The optical and structural characteristics of the films were dramatically improved when a mist of H 2 O and/or NH 4 0H was simultaneously added during deposition of the films. A refraction index up to 1.88, and deposition rates lower than 10 A/sec were obtained in the films. Infrared spectroscopy measurements indicate that the films resulted free from -OH bonds. X-ray diffraction patterns reveal that the films were polycrystalline. In addition, the relative chemical composition of the films was determined by Energy Dispersive Spectroscopy and the surface morphology was analyzed in the Atomic Force Microscope. (Author)

  14. Development of the white cast iron with niobium alloy, heat treating, to wear of the abrasive resistance; Desenvolvimento de uma liga de ferro fundido branco alto cromo com niobio, tratada termicamente, para resistencia ao desgaste abrasivo

    Energy Technology Data Exchange (ETDEWEB)

    Farah, Alessandro Fraga

    1997-07-01

    This work presents the heat treatment and abrasion tests results of a white cast iron with niobium alloy. The hardening heat treatment were made 950, 1000, 1050 e 110 deg C temperatures cooled by forced air. The tempering treatment were made at 450, 500 e 550 deg C temperatures. The heat treating alloy were compared, in the abrasive tests, with commercial alloys used as hardfacing by welding process in wear pieces. The abrasion tests was realized in pin on disk test. Additional tests were carried out for microstructural characterization to identify the different phases presents in the alloys. In a general way, the alloy studies showed the best wear rate for the heat treatments that results in higher hardness. It performance was superior than that of the commercial alloys. (author)

  15. Development of the white cast iron with niobium alloy, heat treating, to wear of the abrasive resistance

    International Nuclear Information System (INIS)

    Farah, Alessandro Fraga

    1997-01-01

    This work presents the heat treatment and abrasion tests results of a white cast iron with niobium alloy. The hardening heat treatment were made 950, 1000, 1050 e 110 deg C temperatures cooled by forced air. The tempering treatment were made at 450, 500 e 550 deg C temperatures. The heat treating alloy were compared, in the abrasive tests, with commercial alloys used as hardfacing by welding process in wear pieces. The abrasion tests was realized in pin on disk test. Additional tests were carried out for microstructural characterization to identify the different phases presents in the alloys. In a general way, the alloy studies showed the best wear rate for the heat treatments that results in higher hardness. It performance was superior than that of the commercial alloys. (author)

  16. Statistical Modeling for the Effect of Rotor Speed, Yarn Twist and Linear Density on Production and Quality Characteristics of Rotor Spun Yarn

    Directory of Open Access Journals (Sweden)

    Farooq Ahmed Arain

    2012-01-01

    Full Text Available The aim of this study was to develop a statistical model for the effect of RS (Rotor Speed, YT (Yarn Twist and YLD (Yarn Linear Density on production and quality characteristics of rotor spun yarn. Cotton yarns of 30, 35 and 40 tex were produced on rotor spinning machine at different rotor speeds (i.e. 70000, 80000, 90000 and 100000 rpm and with different twist levels (i.e. 450, 500, 550, 600 and 700 tpm. Yarn production (g/hr and quality characteristics were determined for all the experiments. Based on the results, models were developed using response surface regression on MINITAB�16 statistical tool. The developed models not only characterize the intricate relationships among the factors but may also be used to predict the yarn production and quality characteristics at any level of factors within the range of experimental values.

  17. Stokes shift spectroscopy for breast cancer diagnosis

    Science.gov (United States)

    Jeyasingh, Ebenezar; Prakashrao, Aruna; Singaravelu, Ganesan

    2010-02-01

    The objective of this study is to assess the diagnostic potential of stokes shift (SS) spectroscopy (SSS) for normal and different pathological breast tissues such as fibroadenoma and infiltrating ductal carcinoma. The SS spectra is measured by simultaneously scanning both the excitation and emission wavelengths while keeping a fixed wavelength interval Δλ=20 nm between them. Characteristic, highly resolved peaks and significant spectral differences between normal and different pathological breast tissues were observed. The SS spectra of normal and different pathological breast tissues shows the distinct peaks around 300, 350, 450, 500 and 600 nm may be attributed to tryptophan, collagen, NADH, flavin and porphyrin respectively. Using SSS technique one can obtain all the key fluorophores in a single scan and hence they can be targeted as a tumor markers in this study. In order to quantify the altered spectral differences between normal and different pathological breast tissues are verified by different ratio parameters.

  18. Study of 57Fe Mössbauer effect in RFe2Zn20 (R = Lu, Yb, Gd)

    Science.gov (United States)

    Bud'ko, Sergey L.; Kong, Tai; Ma, Xiaoming; Canfield, Paul C.

    2015-08-01

    We report measurements of 57Fe Mössbauer spectra for RFe2Zn20 (R = Lu, Yb, Gd) from  ˜4.5 K to room temperature. The obtained isomer shift values are very similar for all three compounds, their temperature dependence was analyzed within the Debye model and resulted in an estimate of the Debye temperatures of 450-500 K. The values of quadrupole splitting at room temperature change with the cubic lattice constant a in a linear fashion. For GdFe2Zn20, ferromagnetic order is seen as an appearance of a sextet in the spectra. The 57Fe site hyperfine field for T\\to 0 was evaluated to be  ˜2.4 T.

  19. Remanent and induced magnetization in the volcanites of Lipari and Vulcano (Aeolian Islands

    Directory of Open Access Journals (Sweden)

    R. Lanza

    1994-06-01

    Full Text Available The role of remanent and induced magnetization as sources of magnetic anomalies in the Lipari and Vulcano islands has been studied by systematic sampling. Remanent magnetization is higher than induced magnetization in almost all lithotypes. Its polarity is normal, and the mean directions are close to the present magnetic field. A slight thermal enhancement of the magnetic susceptibility occurs up to 450-500 °C, followed by a fall up to the Curie point, which is comprised in the range 550 ± 30 °C. This points to titanomagnetite as the main carrier of magnetization. The blocking temperature spectrum of the remanence ranges between the Curie point and 400 °C in most lithotypes, and falIs to 150-200 °C in the pyroclastic deposits. The results as a whole yield an outline of the areal distribution of the total magnetization intensity within the two islands.

  20. Thermodynamic and kinetic analysis of solid-phase interaction of alkali metal carbonates with arsenic pentoxide

    International Nuclear Information System (INIS)

    Pashinkin, A.S.; Buketov, E.A.; Isabaeva, S.M.; Kasenov, B.K.

    1985-01-01

    The thermodynamic analysis of solid-phase reactions of alkali metal carbonates with arsenic pentoxide showing the possibility of formation of all arsenates at a higher than the room temperature is performed. Energetically most advantageous is formation of meta-arsenates. It is shown that temperature increase favours the reaction process. By Gibbs standard energy decrease the reactions form the Li>Na>K>Rb>Cs series. On the base of calculation data linear dependence of Gibbs standard energy in reactions on the atomic number of alkali metalis established. By the continuous weighing method the kinetics of interaction of alkali metal carbonates with arsenic pentoxide under isothermal conditions in the 450-500 deg C range is studied. Studies is the dependence of apparent energy of interaction of carbonates wih As 2 0 5 an atomic parameters of al

  1. Qualite de vie apres un accident vasculaire cerebral au Senegal: a ...

    African Journals Online (AJOL)

    The accident was ischemic in 70% of cases, hemorrhagic in 30% of cases. 66% had a Barthel index between (60-100), 6% were between (0-20). The score was significantly better in young series (15-34), which all have a Barthel index between (60-100), more over 50% of patients in the portion above 75 years were in the ...

  2. "Living high - training low" vs. "living high - training high": erythropoietic responses and performance of adolescent cross-country skiers.

    Science.gov (United States)

    Christoulas, K; Karamouzis, M; Mandroukas, K

    2011-03-01

    To determine and compare the erythropoietic response and exercise performance of adolescent cross-country skiers, as a result of "living high-training high" (HH) and "living high-training low" (HL). Nine female and six male adolescent cross-country skiers volunteered to participate in separate trials. In the first trial (HH), the skiers lived and trained for 21 days at 1550-2050 m, while in the second trial (HL) they trained near sea level (450-500 m) but resided at 1550 m. All participants underwent maximal cycle ergometer tests for the determination of VO2max and cardiorespiratory parameters via an open circuit system at sea level before ascent to altitude, and 1-2 days after descent from altitude. Blood samples were drawn prior to and immediately after maximal cycle exercise testing, at sea level prior to ascent, on days 1 (D1) and 21 (D21) at altitude (1740 m), and 1-2 days post-altitude, for the determination of serum erythropoietin (EPO) concentration, haemoglobin (Hb), hematocrit (Ht), and red blood cell (RBC) volume. The results showed that both boys and girls cross-country skiers, significantly improved their sea level VO2max after 21 days of living at moderate altitude and training near sea level. The present study demonstrates that living at moderate altitude, 1550-2050 m and training low, near sea level (450-500 m) significantly increases VO2max and RBC mass for both boys and girls. Results indicate that applying the training concept "living high - training low" in adolescent athletes may improve their endurance performance.

  3. 238U, and its decay products, in grasses from an abandoned uranium mine

    Science.gov (United States)

    Childs, Edgar; Maskall, John; Millward, Geoffrey

    2016-04-01

    Bioaccumulation of radioactive contaminants by plants is of concern particularly where the sward is an essential part of the diet of ruminants. The abandoned South Terras uranium mine, south west England, had primary deposits of uraninite (UO2) and pitchblende (U3O8), which contained up to 30% uranium. When the mine was active uranium and radium were extracted but following closure it was abandoned without remediation. Waste rock and gangue, consisting of inefficiently processed minerals, were spread around the site, including a field where ruminants are grazed. Here we report the activity concentrations of 238U, 235U 214,210Pb, and the concentrations of selected metals in the soils, roots and leaves of grasses taken from the contaminated field. Soil samples were collected at the surface, and at 30 cm depth, using an auger along a 10-point transect in the field from the foot of a waste heap. Whole, individual grass plants were removed with a spade, ensuring that their roots were intact. The soils and roots and grass leaves were freeze-dried. Activity concentrations of the radionuclides were determined by gamma spectroscopy, following 30 days incubation for development of secular equilibrium. Dried soils, roots and grasses were also digested in aqua regia and the concentrations of elements determined by ICP techniques. Maximum activity concentrations of 238U, 235U, 214Pb and 210Pb surface soils were 63,300, 4,510, 23,300 and 49,400 Bq kg-1, respectively. The mean 238U:235U ratio was 11.8 ± 1.8, an order of magnitude lower than the natural value of 138, indicating disequilibrium within the decay chain due to mineral processing. Radionuclides in the roots had 5 times lower concentration and only grass leaves in the vicinity of the waste heap had measureable values. The mean soil to root transfer factor for 238U was 36%, the mean root to leaf was 3% and overall only 0.7% of 238U was transferred from the soil to the leaves. The roots contained 0.8% iron, possibly as

  4. Nuclear renaissance and thorium as a brighten of the next nuclear era

    International Nuclear Information System (INIS)

    Uenak, T.

    2001-01-01

    As is known well, the discovery of the fission of U-235 by Otto Hahn at the end of 1938 opened the gateway of nuclear era having a new and powerful energy source for our globe. The first applications of this new energy source were unfortunately shown as the atomic bombs on Japan in 1945. These first applications of nuclear energy were initiated by the use of fissile U-235 and then continued being added an artificial fissile material Pu-239. As a consequence of this initiative, today's nuclear technology has principally been based on the use of these fissile materials. Off course, the first application of nuclear energy as atomic bomb on Japan was very bad chance for the humanity to envisage really and first time the nuclear energy, and consequently, in the last half of 20th century, the majority of non-scientific community has not delighted even the peaceful applications of nuclear energy. This is also provoked by several polytized groups against to nuclear energy and related technologies. Nowadays, the nuclear technology has been interrogated also in the scientific platforms because of the proliferation of nuclear weapons, the long-lived wastes, the technical problems of excess Pu-239 either in dismantled nuclear weapons or in stocks. On the other hand, the nuclear energy cannot be abandoned because of the increasing global energy demand year by year, and so, sustainable nuclear energy production systems should be developed in long term. For this reason, the n uclear renaissance a s a new concept for sustainable nuclear energy technologies in the following years of 21st century was started to discuss in scientific platforms[1]. The three basic principles of nuclear renaissance can be summarized as follows: 1) nuclear energy production systems should eventually be non-proliferative of nuclear weapons, 2) minimized long-lived wastes resulting from fuel cycle, 3) good accordance with the environmentalist concepts, and so, easily accepted by the people. It is clearly

  5. SEU43 fuel bundles in CANDU 600

    International Nuclear Information System (INIS)

    Catana, Alexandru; Prodea, Iosif; Danila, Nicolae; Prisecaru, Ilie; Dupleac, Daniel

    2008-01-01

    Cernavoda Unit 1 and Unit 2 are pressure tube 650 MWe nuclear stations moderated and cooled with heavy water, of Canada design, located in Romania. Fuelling is on-power and the plant is currently fuelled with natural uranium dioxide. Fuel is encapsulated in a 37 fuel rod assembly having a specific standard geometry (STD37). In order to reduce fuel cycle costs programs were initiated in Canada, South Korea and at SCN Pitesti, Romania for design and build of a new, improved geometry fuel bundle and some fuel compositions. Among fuel compositions, which are considered, is the slightly enriched uranium (SEU) fuel (0.96 w% U-235) with an associated burn-up increase from ∼7900 MWd/tU up to ∼15000 MWd/tU. Neutron analysis showed that the Canadian-Korean fuel bundle geometry with 43 rods called SEU (SEU43) can be used in already operated reactors. A new fuel bundle resulted. Extended, comprehensive analysis must be conducted in order to assess the TH behavior of SEU43 besides the neutron, mechanical (drag force, etc) analyses. In this paper, using the sub-channel approach, main thermal-hydraulic parameters were analyzed: pressure drop; fuel, sheath and coolant temperatures; coolant density; critical heat flux. Some significant differences versus standard fuel are outlined in the paper and some conclusions are drawn. While, by using this new fuel, there are many benefits to be attained like: fuel costs reduction, spent fuel waste minimization, increase in competitiveness of nuclear power generation against other sources of generation, etc., the safety margins must be, at least, conserved. The introduction of a new fuel bundle type, different in geometry and fuel composition, requires a detailed preparation, a testing program and a series of neutron and thermal-hydraulic analysis. The results reported by this paper is part of this effort. The feasibility to increase the enrichment from 0.71% U-235 (NU) to 0.96% U-235, with an estimated burn-up increase up to 14000 MWd

  6. Analysis of fission ratio distribution in spherical lithium metal assembly with a graphite reflector

    International Nuclear Information System (INIS)

    Seki, Yasushi; Maekawa, Hiroshi; Hiraoka, Toru; Hirota, Jitsuya; Moriyama, Masatoshi.

    1975-08-01

    For the investigation of neutronics in the fusion reactor blanket, a spherical lithium assembly with a graphite reflector was prepared by piling up lithium and graphite blocks. The ratio of U-238 to U-235 fission rates was measured by micro fission chambers, and the result of the measurement was compared with that of the preliminary calculation. It has been shown that there is a large discrepancy between experiment and calculation, which is too large to be explained only by the experimental error. In this paper, the calculational procedure is reviewed and probable causes of the calculational error are listed. Further calculation of the fission ratio is carried out employing refined methods and with varied calculational models. As the result, it is concluded that the cause of the discrepancy is most likely the uncertainties of the nuclear data of the constituent elements of the assembly. (auth.)

  7. Burn-Up Determination by High Resolution Gamma Spectrometry: Spectra from Slightly-Irradiated Uranium and Plutonium between 400-830 keV

    Energy Technology Data Exchange (ETDEWEB)

    Forsyth, R S; Ronqvist, N

    1966-08-15

    Previously published studies of the short-cooled fission product spectra of irradiated uranium have been severely restricted by the poor energy resolution of the sodium iodide detectors used. In this report are presented fission product spectra of irradiated uranium and plutonium obtained by means of a lithium-drifted germanium detector. The resolved gamma peaks have been assigned to various fission products by correlation of measured energy and half-life values with published data. By simultaneous study of the spectra of two irradiated mixtures of plutonium and uranium, the possibility of using the activities of Ru-103 and Ru-106 as a measure of the relative fission rate in U-235 and Pu-239 has been briefly examined.

  8. Detection of tiny amounts of fissile materials in large-sized containers with radioactive waste

    Science.gov (United States)

    Batyaev, V. F.; Skliarov, S. V.

    2018-01-01

    The paper is devoted to non-destructive control of tiny amounts of fissile materials in large-sized containers filled with radioactive waste (RAW). The aim of this work is to model an active neutron interrogation facility for detection of fissile ma-terials inside NZK type containers with RAW and determine the minimal detectable mass of U-235 as a function of various param-eters: matrix type, nonuniformity of container filling, neutron gen-erator parameters (flux, pulse frequency, pulse duration), meas-urement time. As a result the dependence of minimal detectable mass on fissile materials location inside container is shown. Nonu-niformity of the thermal neutron flux inside a container is the main reason of the space-heterogeneity of minimal detectable mass in-side a large-sized container. Our experiments with tiny amounts of uranium-235 (<1 g) confirm the detection of fissile materials in NZK containers by using active neutron interrogation technique.

  9. Isotopic Determination of Nuclear Materials Using Nuclear Fission Track Registration Technique and Thermal Ionization Mass Spectrometric Technique

    International Nuclear Information System (INIS)

    Jeon, Young Sin; Pyo, Hyeong Yeol; Park, Yong Joon; Song, Kyu Seok; Kim, Won Ho; Jee, Kwang Yong

    2007-05-01

    It is very important to develope the technology for the determination of isotopic ratios of hot particles( 234 U, 235 U, 236 U etc.) detected from swipe samples of various nuclear facilities. This technology is highly competitive internationally and has to be established independently as long as our government maintains atomic energy and treats nuclear materials. In this text, sample pretreatment procedure, gamma-ray counting, alpha or fission track techniques, isotopic analysis of U and Pu, background problems and detection limits for mass determination, and their application to the real swipe sample were described with detailed procedure. This technology would contribute to the Korean economy's high growth rate as well as to superiority of government's leading research and development programs if successfully established

  10. Multigroup P8 - elastic scattering matrices of main reactor elements

    International Nuclear Information System (INIS)

    Garg, S.B.; Shukla, V.K.

    1979-01-01

    To study the effect of anisotropic scattering phenomenon on shielding and neutronics of nuclear reactors multigroup P8-elastic scattering matrices have been generated for H, D, He, 6 Li, 7 Li, 10 B, C, N, O, Na, Cr, Fe, Ni, 233 U, 235 U, 238 U, 239 Pu, 240 Pu, 241 Pu and 242 Pu using their angular distribution, Legendre coefficient and elastic scattering cross-section data from the basic ENDF/B library. Two computer codes HSCAT and TRANS have been developed to complete this task for BESM-6 and CDC-3600 computers. These scattering matrices can be directly used as input to the transport theory codes ANISN and DOT. (auth.)

  11. Obtention control bars patterns for a BWR using Tabo search; Obtencion de patrones de barras de control para un BWR usando busqueda Tabu

    Energy Technology Data Exchange (ETDEWEB)

    Castillo, A.; Ortiz, J.J.; Alonso, G. [Instituto Nacional de Investigaciones Nucleares, Km. 36.5 Carretera Mexico-Toluca, Ocoyoacac, Estado de Mexico 52045 (Mexico); Morales, L.B. [UNAM, IIMAS, Ciudad Universitaria, D. F. 04510 (Mexico); Valle, E. del [IPN, ESFM, Unidad Profesional ' Adolfo Lopez Mateos' , Col. Lindavista 07738, D. F. (Mexico)]. e-mail: jacm@nuclear.inin.mx

    2004-07-01

    The obtained results when implementing the technique of tabu search, for to optimize patterns of control bars in a BWR type reactor, using the CM-PRESTO code are presented. The patterns of control bars were obtained for the designs of fuel reloads obtained in a previous work, using the same technique. The obtained results correspond to a cycle of 18 months using 112 fresh fuels enriched at the 3.53 of U-235. The used technique of tabu search, prohibits recently visited movements, in the position that correspond to the axial positions of the control bars, additionally the tiempo{sub t}abu matrix is used for to manage a size of variable tabu list and the objective function is punished with the frequency of the forbidden movements. The obtained patterns of control bars improve the longitude of the cycle with regard to the reference values and they complete the restrictions of safety. (Author)

  12. Production of oxidants from α-radiolysis in the South Devon deposit

    International Nuclear Information System (INIS)

    Christensen, H.

    2001-08-01

    The maximum production of H 2 O 2 caused by α-radiolysis of water in the close vicinity of uranium-containing nodules found at the South Devon site has been estimated. The integrated α-doses from decay of U-238 and U-235 from the time of mineralization until today have been calculated. Calculations have been carried out for three different times: 140, 170, and 240 Myear. The maximum production from one nodule (diameter 4 cm) under 240 Myear was estimated to be 0.075 mole H 2 O 2 (= 2.5 g). For a number of causes, discussed in the report, the actual production will probably be much lower

  13. A review of prospects for an accelerator breeder

    International Nuclear Information System (INIS)

    Fraser, J.S.; Hoffman, C.R.; Schriber, S.O.; Garvey, P.M.; Townes, B.M.

    1981-12-01

    The scientific feasibility, engineering practicability and economic prospects for an Accelerator Breeder are reviewed. The scientific feasibiliity of high power accelerator components rests on a firm basis as a result of technical advances made in recent years but there is a need to combine all components in a demonstration model working under realistic conditions. The engineering practicability of Accelerator Breeder components should be tested in a staged development culminating in a full-scale demonstration plant. The economic assessment depends on calculations of allowed and estimated capital costs of an Accelerator Breeder for a CANDU system operating on the Th-U fuel cycle. The results indicate that the ratio of estimated to allowed capital cost is approximately 3.5 for a breeder with a 2% enriched uranium metal blanket and for separated U235 valued at 48 $/g

  14. Reprocessing yields and material throughput: HTGR recycle demonstration facility

    International Nuclear Information System (INIS)

    Holder, N.; Abraham, L.

    1977-08-01

    Recovery and reuse of residual U-235 and bred U-233 from the HTGR thorium-uranium fuel cycle will contribute significantly to HTGR fuel cycle economics and to uranium resource conservation. The Thorium Utilization National Program Plan for HTGR Fuel Recycle Development includes the demonstration, on a production scale, of reprocessing and refabrication processes in an HTGR Recycle Demonstration Facility (HRDF). This report addresses process yields and material throughput that may be typically expected in the reprocessing of highly enriched uranium fuels in the HRDF. Material flows will serve as guidance in conceptual design of the reprocessing portion of the HRDF. In addition, uranium loss projections, particle breakage limits, and decontamination factor requirements are identified to serve as guidance to the HTGR fuel reprocessing development program

  15. Occupational exposure to natural radioactivity in a zircon sand milling plant

    International Nuclear Information System (INIS)

    Ballesteros, Luisa; Zarza, Isidoro; Ortiz, Josefina; Serradell, Vicente

    2008-01-01

    Raw zirconium sand is one of the substances (naturally occurring radioactive material, NORM) which is widely used in the ceramic industry. This sand contains varying concentrations of natural radionuclides: mostly U-238 but also Th-232 and U-235, together with their daughters, and therefore may need to be regulated by Directive 96/29/EURATOM. This paper describes the method used to perform the radiological study on a zircon sand milling plant and presents the results obtained. Internal and external doses were evaluated using radioactivity readings from sand, airborne dust, intermediate materials and end products. The results on total effective dose show the need for this type of industry to be carefully controlled, since values near to 1 mSv were obtained

  16. Occupational exposure to natural radioactivity in a zircon sand milling plant

    Energy Technology Data Exchange (ETDEWEB)

    Ballesteros, Luisa [Laboratorio de Radioactividad Ambiental, Universidad Politecnica de Valencia, Apartado 22012, E-46071 Valencia (Spain)], E-mail: lballest@upvnet.upv.es; Zarza, Isidoro [Laboratorio de Radioactividad Ambiental, Universidad Politecnica de Valencia, Apartado 22012, E-46071 Valencia (Spain)], E-mail: iszarpe@upvnet.upv.es; Ortiz, Josefina [Laboratorio de Radioactividad Ambiental, Universidad Politecnica de Valencia, Apartado 22012, E-46071 Valencia (Spain)], E-mail: jortiz@iqn.upv.es; Serradell, Vicente [Laboratorio de Radioactividad Ambiental, Universidad Politecnica de Valencia, Apartado 22012, E-46071 Valencia (Spain)], E-mail: vserradell@iqn.upv.es

    2008-10-15

    Raw zirconium sand is one of the substances (naturally occurring radioactive material, NORM) which is widely used in the ceramic industry. This sand contains varying concentrations of natural radionuclides: mostly U-238 but also Th-232 and U-235, together with their daughters, and therefore may need to be regulated by Directive 96/29/EURATOM. This paper describes the method used to perform the radiological study on a zircon sand milling plant and presents the results obtained. Internal and external doses were evaluated using radioactivity readings from sand, airborne dust, intermediate materials and end products. The results on total effective dose show the need for this type of industry to be carefully controlled, since values near to 1 mSv were obtained.

  17. AIRFIL: a FORTRAN program for reduction of data obtained from alpha spectrometry of perimeter air filters

    International Nuclear Information System (INIS)

    Hinton, E.R. Jr.; Howell, R.L.

    1984-03-01

    Isotopic alpha spectrometry of 238 U, 235 U, and 234 U, along with gross alpha/beta counting of perimeter air filters, is performed by the Oak Ridge Y-12 Plant Laboratory in support of the Envrirnomental Monitoring Section of the Radiation Safety Department. Weekly samples are gross alpha/beta counted and the isotopic analysis performed on quarterly composites. Calculations and data management represents a major portion of the anlaysis time when performed manually, even with a desktop calculator. In order to reduce calculation time, perform orderly data manipulation and management, reduce errors due to redundant calculations, and eliminate report typing turnaround time, a computer program (AIRFIL) has been developed that performs these functions. The program accepts data through user prompts, then calculates and prints intermediate and final data, including detection limits

  18. Determination of Pb-210 and actinides by extraction chromatography and anion exchange chromatography

    International Nuclear Information System (INIS)

    Kalmykov, St.N.; Sapozhnikov, Yu.A.

    1997-01-01

    This work is devoted to the determination of Pb-210 and actinides (Pu-238, Pu-239, Am-241, U-235, U-238, Th-232) by means of highly selective chromatographic resins and anion exchangers. The special interest was paid to the analysis of large quantities of samples with high concentration of competitive ions like ocean sediments, bone ash and others.The commercially available TRU-Spec chromatographic resins was used for separation of actinides from the matrix. Then U, Th, Am, and Pu were separated from other using anion exchange chromatography with AG-1X4 anionite in Cl - form, electro-deposed and α-counted.Pb-21- and Bi-210 were determined by liquid scintillation counting. The developed procedure is rather express, effective and could be adopted for the determination of radionuclides like Ba-133, Ra, Np-239

  19. An analysis of ZEEP reaction rate measurements using the iterative five group scheme of METHUSELAH I

    Energy Technology Data Exchange (ETDEWEB)

    Allen, F R; Askew, J R [Technical Assessments and Services Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

    1964-02-15

    The five group iterative scheme of METHUSELAH I has been used to provide estimates of reaction rates for comparison with measurements in a number of loop experiments in ZEEP. Theory and experiment show good agreement for individual fission rate distributions and for the Pu-239/U-235 reaction rate ratios. Estimates of Lu/Mn ratios are in reasonable agreement with measurements in the H{sub 2}O cores, but become progressively less satisfactory as the D{sub 2}O content of the coolant is increased. It is concluded that for assessment calculations, the introduction of an iterative five group scheme improves the representation of thermal spectra in pressure tube reactor lattices. (author)

  20. The problem of reactivity and reaction-rate calculations for mixed graphite lattices

    International Nuclear Information System (INIS)

    Pitcher, H.H.W.

    1963-05-01

    The dependence of reactor physics quantities, such as η f and Pu239/U235 fission ratio, in a single cell on the environment of the cell, and the relationship of the reactivity of a mixed lattice to the reactivity of its components, in graphite-moderated reactors are investigated. In a particular case, a mixed lattice fuelled with uranium at 0 and 3000 MWD/Te showed at 8 cm. pitch a small but appreciable change (∼ 1%) in cell quantities, and at 25 cm. pitch a smaller change. It is found that the present method of calculating lattice reactivity, ignoring intercell effects, is probably adequate for standard-pitch metal-fuelled graphite-moderated systems. More general mixed-lattice systems, particularly if accurate values of cell quantities are required, may need special calculation techniques; these are discussed, and techniques adequate for most systems are presented. (author)

  1. Summarizing annual report 1989, on R and D work performed by the KfK Institute for Microstructural Engineering, IMT

    International Nuclear Information System (INIS)

    1990-03-01

    The main activities of the institute in the year under review continued the work with the separation nozzle method for U-235 enrichment, the testing of separation nozzle elements and the related methods for UF 6 production, and the design and testing of components for separation systems. In another working area, activities centered on microstructural techniques, as e.g. the production of masking blanks or structural masks for the deep-etch synchrotron radiation lithography, on the development of synchrotron radiation sources, or the design of irradiation facilities or wet chemical processes for microstructural techniques. The latter include molding of metals or plastics, surface etching and thin-film deposition. Experiments were carried out for microstructural process development, and product control and analyses for the purpose of quality assurance of microstructures. (HK) [de

  2. Environmental monitoring report for Pantex Plant covering 1986

    International Nuclear Information System (INIS)

    Laseter, W.A.

    1987-04-01

    This report summarizes the environmental monitoring program at Pantex Plant for 1986. It has been prepared in accordance with the US Department of Energy Order 5484.1. This report presents monitoring data for both radioactive and nonradioactive species in the local environment. Plant activities involve the handling of significant quantities of uranium, plutonium and tritium in the form of completed parts received from other DOE facilities, resulting in a very low potential for release of these radionuclides to the atmosphere. In 1986 only small releases of depleted uranium (depleted in the isotope U-235) and tritium occurred which could have affected the local environment. Monitoring data indicate that concentrations of these nuclides in the environment are below established criteria for air and water and therefore should not present a health hazard either to employees or to the public

  3. Optimization of neutronic characteristics of U3Si2 low enrichment fuel elements for a new design of IEA-R1 reactor core

    International Nuclear Information System (INIS)

    Mai, L.A.; Maiorino, J.R.; Gouvea, E.A.

    1989-01-01

    This work shows a study of neutronic optimization of U 3 Si 2 -Al low enrichment fuel element. This study has a goal to propose a optimized Core to be used in the research reactor IEA-R1. The external dimensions of the fuel element were maintained as constraints and the loss of reactivity along fuel life-time was defined as 'objective function', and it has been minimized by varying the fuel element dimensions. Cell calculations were made with HAMMER-TECH /3/ Code, for burnups up to 50% of U-235 initial mass. The Computer values of the objective function for several combinations of fuel element dimensions were fitted by a surface using the SAS system /9/, and it has been minimized by a Harwell subroutine /10/. (author) [pt

  4. Replacement of highly enriched uranium by medium or low-enriched uranium in fuels for research reactors

    International Nuclear Information System (INIS)

    Schwartz, J.P.

    To exclude the possibility of an explosive use of the uranium obtained from an elementary chemical process, one needs to use a fuel less enriched than 20 weight percent in U 235 . This goal can be reached by two ways: 1. The low density fuels, i.e. U or U 3 O 8 /Al fuels. One has to increase their U content from 1.3 g U/cm 3 presently qualified under normal operation conditions. Several manufacturers such as CERCA in France developed these fuels with a near-term objective of about 2 g U/cm 3 and a long-term objective of 3 g U/cm 3 . 2. The high density fuels. They are the UO 2 Caramel plate type fuels now under consideration, and U 3 Si and UMo as a long-term potential

  5. High-capacity neutron activation analysis facility

    International Nuclear Information System (INIS)

    Hochel, R.C.

    1979-01-01

    A high-capacity neutron activation analysis facility, the Reactor Activation Facility, was designed and built and has been in operation for about a year at one of the Savannah River Plant's production reactors. The facility determines uranium and about 19 other trace elements in hydrogeochemical samples collected in the National Uranium Resource Evaluation program. The facility has a demonstrated average analysis rate of over 10,000 samples per month, and a peak rate of over 16,000 samples per month. Uranium is determined by cyclic activation and delayed neutron counting of the U-235 fission products; other elements are determined from gamma-ray spectra recorded in subsequent irradiation, decay, and counting steps. The method relies on the absolute activation technique and is highly automated for round-the-clock unattended operation

  6. Approach to IAEA material-balance verification at the Portsmouth Gas Centrifuge Enrichment Plant

    International Nuclear Information System (INIS)

    Gordon, D.M.; Sanborn, J.B.; Younkin, J.M.; DeVito, V.J.

    1983-01-01

    This paper describes a potential approach by which the International Atomic Energy Agency (IAEA) might verify the nuclear-material balance at the Portsmouth Gas Centrifuge Enrichment Plant (GCEP). The strategy makes use of the attributes and variables measurement verification approach, whereby the IAEA would perform independent measurements on a randomly selected subset of the items comprising the U-235 flows and inventories at the plant. In addition, the MUF-D statistic is used as the test statistic for the detection of diversion. The paper includes descriptions of the potential verification activities, as well as calculations of: (1) attributes and variables sample sizes for the various strata, (2) standard deviations of the relevant test statistics, and (3) the detection sensitivity which the IAEA might achieve by this verification strategy at GCEP

  7. Device for thermonuclear reactor

    International Nuclear Information System (INIS)

    Yanagisawa, Yutaro; Kawarazaki, Yuki; Sugiyama, Yu.

    1996-01-01

    A member comprising hydrogen occluding materials is introduced to a reactor incorporated with U-235 as fuels in order to moderate and breed fast neutrons and to control the reactor. Since the amount of light hydrogen or heavy hydrogen is substantially the same as that of metal, etc. of hydrogen occluding material, a moderating efficiency substantially equal with that of a moderator comprising H 2 O can be obtained. In addition, since the member acting as a moderator has an effect of multiplying neutrons, use of only natural uranium 0.72% as nuclear fuels causes chain reaction to provide a function as a nuclear reactor. Further, the hydrogen occluding material can be used also as a control rod for controlling the reactor. The hydrogen occluding material may be Ti, Zr, Pd, proton conductor, Ag, Pt, Rh or oxides thereof or alloys thereof. The member comprising hydrogen occluding materials is preferably coated with a material not permeating hydrogen. (N.H.)

  8. Use of LEU in the aqueous homogeneous medical isotope production reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ball, R.M. [Babock & Wilcox, Lynchburg, VA (United States)

    1997-08-01

    The Medical Isotope Production Reactor (MIPR) is an aqueous solution of uranyl nitrate in water, contained in an aluminum cylinder immersed in a large pool of water which can provide both shielding and a medium for heat exchange. The control rods are inserted at the top through re-entrant thimbles. Provision is made to remove radiolytic gases and recombine emitted hydrogen and oxygen. Small quantities of the solution can be continuously extracted and replaced after passing through selective ion exchange columns, which are used to extract the desired products (fission products), e.g. molybdenum-99. This reactor type is known for its large negative temperature coefficient, the small amount of fuel required for criticality, and the ease of control. Calculation using TWODANT show that a 20% U-235 enriched system, water reflected can be critical with 73 liters of solution.

  9. Recent observations at the post-irradiation examination of low-enriched U-Mo miniplates irradiated to high burn-up

    International Nuclear Information System (INIS)

    Hofman, G.L.; Kim, Y.S.; Finlay, M.R.; Snelgrove, J.L.; Hayes, S.L.; Meyer, M.K.; Clark, C.R.

    2003-01-01

    High-density dispersion fuel experiment, RERTR-4, was removed from the Advanced Test Reactor (ATR) after reaching a peak U-235 burnup of ∼80% and is presently undergoing postirradiation examination at the ANL Alpha-Gamma Hot Cell Facility. This test consists of 32 mini fuel plates of which 27 were fabricated with nominally 6 and 8 g cm -3 atomized and machined uranium alloy powders containing 6.5 wt% to 10 wt% molybdenum. In addition, two miniplates contained solid U-10wt%Mo foils. Recent results of the postirradiation examination and analysis of RERTR-4 in conjunction with data from a companion test performed to 50% burnup, RERTR-5, are presented. (author)

  10. Progress in irradiation performance of experimental uranium - Molybdenum dispersion fuel

    International Nuclear Information System (INIS)

    Hofman, Gerard L.; Meyer, Mitchell K.

    2002-01-01

    High-density dispersion fuel experiment, RERTR-4, was removed from the Advanced Test Reactor (ATR) after reaching a peak U-235 burnup of ∼80% and is presently undergoing postirradiation examination at the ANL alpha-gamma hot cells. This test consists of 32 mini fuel plates of which 27 were fabricated with nominally 6 and 8 g cm -3 atomized and machined uranium alloy powders containing 7 wt% and 10 wt% molybdenum. In addition, two miniplates containing solid U-10 wt% Mo foils and three containing 6 g cm -3 U 3 Si 2 are part of the test. The results of the postirradiation examination and analysis of RERTR-4 in conjunction with data from previous tests performed to lower burnup will be presented. (author)

  11. Isotope separation method and apparatus

    International Nuclear Information System (INIS)

    Lyon, R.K.; Eisner, P.N.; Thomas, W.R.L.

    1980-01-01

    A method and apparatus are specified for separating a mixture of isotopes present in a compound, preferably a gaseous compound, into two or more parts in each of which the abundances of the isotopes differ from the natural abundances of the isotopes in the compound. The invention particularly relates to carrying out a laser induced, isotopically selective conversion of gaseous molecules in such a manner as to achieve more than one stage of isotope separation along the length of the laser beam. As an example, the invention is applied to the separation of the isotopes of uranium in UF 6 , in which either the U-235 or U-238 isotope is selectively excited by means of irradiation from an infrared laser, and the selectively excited isotope converted into a product that can be recovered from UF 6 by one of a variety of methods that are described. (U.K.)

  12. Release of fission products from irradiated aluminide fuel at high temperature

    International Nuclear Information System (INIS)

    Shibata, Toshikazu; Kanda, Keiji; Mishima, Kaichiro; Tamai, Tadaharu; Hayashi, Masatoshi; Snelgrove, James L.; Stahl, David; Matos, James E.; Travelli, Armando; Case, F. Neil; Posey, John C.

    1983-01-01

    Irradiated uranium aluminide fuel plates of 40% U-235 enrichment were heated for the determination of fission products released under flowing helium gas at temperatures up to and higher than the melting point of fuel cladding material. The release of fission products from the fuel plate at temperature below 500 deg. C was found negligible. The first rapid release of fission products was observed with the occurrence of blistering at 561±1 deg. C on the plates. The next release at 585. C might be caused by melting of the cladding material of 6061-Al alloy. The last release of fission product gases was occurred at the eutectic temperature of 640 deg. C of U-Al x . The released material was mostly xenon, but small amounts of iodine and cesium were observed. (author)

  13. Release of fission products from irradiated aluminide fuel at high temperature

    International Nuclear Information System (INIS)

    Shibata, T.; Kanda, K.; Mishima, K.

    1982-01-01

    Irradiated uranium aluminide fuel plates of 40% U-235 enrichment were heated for the determination of fission products released under flowing helium gas at temperatures up to and higher than the melting point of fuel-cladding material. The release of fission products from the fuel plate at temperature below 500 0 C was found negligible. The firist rapid release of fission products was observed with the occurrence of blistering at 561 +- 1 0 C on the plates. The next release at 585 0 C might be caused by melting of the cladding material of 6061-Al alloy. The last release of fission product gases was occurred at the eutectic temperature of 640 0 C of U-Al/sub x/. The released material was mostly xenon, but small amounts of iodine and cesium were observed

  14. MUICYCL and MUIFAP: models tracking minor uranium isotopes in the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Blum, S.R.; McLaren, R.A.

    1979-10-01

    Two computer programs have been written to provide information on the buildup of minor uranium isotopes in the nuclear fuel cycle. The Minor Uranium Isotope Cycle Program, MUICYCL, tracks fuel through a multiyear campaign cycle of enrichment, reactor burnup, reprocessing, enrichment, etc. MUICYCL facilities include preproduction stockpiles, U 235 escalation, and calculation of losses. The Minor Uranium Isotope Flowsheet Analyzer Program, MUIFAP, analyzes one minor isotope in one year of an enrichment operation. The formulation of the enrichment cascade, reactors, and reprocessing facility is presented. Input and output descriptions and sample cases are presented. The programs themselves are documented by short descriptions of each routine, flowcharts, definitions of common blocks and variables, and internal documentation. The programs are written in FORTRAN for use in batch mode

  15. Accident in science history. Hitler's atomic bomb; Stoerfall der Wissenschaftsgeschichte. Hitlers Atombombe

    Energy Technology Data Exchange (ETDEWEB)

    Popp, Manfred

    2016-07-01

    Up to now the historical explanation for the fact that NS Germany did not develop an atomic bomb was that there was enough scientific and technical knowledge but not enough resources. Recent historical research showed that the German scientists in the ''Uranverein'' did not know the cross section of U-235 and had no possibility to measure these cross sections. There exist only very few documents dealing with the bomb. The bomb was treated as a special case of a reactor. Obviously the German scientists did not know he bomb technology. It seems that the project was not success-oriented because the scientists expected to be sent to the front in case of a project stop.

  16. RERTR-7 Irradiation Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    D. M. Perez; M. A. Lillo; G. S. Chang; G. A. Roth; N. E. Woolstenhulme; D. M. Wachs

    2011-12-01

    The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-7A, was designed to test several modified fuel designs to target fission densities representative of a peak low enriched uranium (LEU) burnup in excess of 90% U-235 at peak experiment power sufficient to generate a peak surface heat flux of approximately 300 W/cm2. The RERTR-7B experiment was designed as a high power test of 'second generation' dispersion fuels at peak experiment power sufficient to generate a surface heat flux on the order of 230 W/cm2.1 The following report summarizes the life of the RERTR-7A and RERTR-7B experiments through end of irradiation, including as-run neutronic analyses, thermal analyses and hydraulic testing results.

  17. Standard test method for isotopic abundance analysis of uranium hexafluoride and uranyl nitrate solutions by multi-collector, inductively coupled plasma-mass spectrometry

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2014-01-01

    1.1 This test method covers the isotopic abundance analysis of 234U, 235U, 236U and 238U in samples of hydrolysed uranium hexafluoride (UF6) by inductively coupled plasma source, multicollector, mass spectrometry (ICP-MC-MS). The method applies to material with 235U abundance in the range of 0.2 to 6 % mass. This test method is also described in ASTM STP 1344. 1.2 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

  18. Initial conceptual design study of self-critical nuclear pumped laser systems

    Science.gov (United States)

    Rodgers, R. J.

    1979-01-01

    An analytical study of self-critical nuclear pumped laser system concepts was performed. Primary emphasis was placed on reactor concepts employing gaseous uranium hexafluoride (UF6) as the fissionable material. Relationships were developed between the key reactor design parameters including reactor power level, critical mass, neutron flux level, reactor size, operating pressure, and UF6 optical properties. The results were used to select a reference conceptual laser system configuration. In the reference configuration, the 3.2 m cubed lasing volume is surrounded by a graphite internal moderator and a region of heavy water. Results of neutronics calculations yield a critical mass of 4.9 U(235) in the form (235)UF6. The configuration appears capable of operating in a continuous steady-state mode. The average gas temperature in the core is 600 K and the UF6 partial pressure within the lasing volume is 0.34 atm.

  19. Use of LEU in the aqueous homogeneous medical isotope production reactor

    International Nuclear Information System (INIS)

    Ball, R.M.

    1997-01-01

    The Medical Isotope Production Reactor (MIPR) is an aqueous solution of uranyl nitrate in water, contained in an aluminum cylinder immersed in a large pool of water which can provide both shielding and a medium for heat exchange. The control rods are inserted at the top through re-entrant thimbles. Provision is made to remove radiolytic gases and recombine emitted hydrogen and oxygen. Small quantities of the solution can be continuously extracted and replaced after passing through selective ion exchange columns, which are used to extract the desired products (fission products), e.g. molybdenum-99. This reactor type is known for its large negative temperature coefficient, the small amount of fuel required for criticality, and the ease of control. Calculation using TWODANT show that a 20% U-235 enriched system, water reflected can be critical with 73 liters of solution

  20. 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

  1. Depleted uranium and the Gulf War syndrome

    International Nuclear Information System (INIS)

    1999-01-01

    Some military personnel involved in the 1991Gulf War have complained of continuing stress-like symptoms for which no obvious cause has been found. These symptoms have at times been attributed to the use of depleted uranium (DU) in shell casings which are believed to have caused toxic effects. Depleted uranium is natural uranium which is depleted in the rarer U-235 isotope. It is a heavy metal and in common with other heavy metals is chemically toxic. It is also slightly radioactive and could give rise to a radiological hazard if dispersed in finely divided form so that it was inhaled. In response to concerns, the possible effects of DU have been extensively studied along with other possible contributors to G ulf War sickness . This article looks at the results of some of the research that has been done on DU. (author)

  2. Chemical effects of fission recoils

    International Nuclear Information System (INIS)

    Meisels, G.G.; Freeman, J.P.; Gregory, J.P.; Richardson, W.C.; Sroka, G.J.

    1978-01-01

    The production of nitrogen from nitrous oxide at high density was employed to investigate the energy deposition efficiency of fission recoils produced from fission of U 235 in uranium-palladium foils clad with platinum. Nitrogen production varied linearly with fission recoil dose from 1.1 x 10 20 to 9.0 x 10 20 eV, and was independent of density between 12.5 and 127.5 g l -1 N 2 O. 16.2 +- 0.8% of the fission recoil energy was deposited external to the foil. Electron microprobe analysis showed some unevenness of new foil and polymer buildup on the surface after irradiation of ethylene-oxygen mixtures. Subsequent irradiation in the presence of nitrous oxide restored some of the original efficiency. This is ascribed to chemical oxidation of the polymer induced by reactive intermediates produced from nitrous oxide. (author)

  3. Outline of developing projects of atomic bomb in Japan and USA

    International Nuclear Information System (INIS)

    Fukui, Shuji

    2007-01-01

    The content of the title connecting with the World War II is described hoping that younger researchers of nuclear physics could know some of the facts that scientists and the military of Japan and USA, respectively, had have developing projects of atomic bomb by fission reaction, although there are no official documents of those in Japan, even if there are some unofficial documents that are uncertain partly in Japan. Described are a chronological table, the content of research and development of atomic bombs, Japan's experiments by Kikuchi Laboratory of Osaka Imperial University and Nishina Laboratory of RIKEN, as well as the USA's action such as production of fissile nuclide, Pu-239 and U-235, selection of the site to fabricate atomic bomb, investigation the state of research and development of atomic bombs in Germany, Italy and Japan. (S.Y.)

  4. Composition of lunar noble gases traped 2.5 AE and 3.5 AE ago

    International Nuclear Information System (INIS)

    Eugster, O.

    1986-01-01

    The times when the soils 74001 and 73261 were exposed on the lunar surface were determined by the U-235 - Xe-136 dating method. The isotopic composition of the trapped noble gases in these two soils is compared with that of the surface correlated noble gases in the young soils 12001 and in the present day solar wind. The surface correlated trapped gases are a mixture of implanted solar wind particles and retrapped lunar atmospheric gases. The observed changes are interpreted as a result of decreasing outgassing of radiogenic Ar-40 and perhaps He-4 and of fissiogenic Xe from the lunar crust. The old soils probably also contain surface correlated Kr-80 and Kr-82 produced by secondary cosmic ray neutron capture of adsorbed or retrapped bromine. To some extent the isotopic composition of the trapped gases in old lunar soil may also have been altered due to diffusion loss from material of low retentivity

  5. Determination of uranium isotopes in urine

    International Nuclear Information System (INIS)

    Lellis, I.R.; Silva, D.V.F.M. Rey; Taddei, M.H.T.

    2017-01-01

    Variable concentrations of uranium occur naturally in waters, plant products and soils. Small amounts of this element are routinely incorporated by man. Occupationally exposed individuals (IOEs) are subject to the incorporation of higher amounts of uranium into their work routines. The effects on human health resulting from the incorporation of uranium in environmental doses are not very well established and are currently recognized as of little relevance. The incorporation resulting from occupational activities, where higher doses can be found, represents a health risk resulting from chemical damages to the kidneys. Considering that uranium is eliminated from the human body through urine and feces, and that the concentration in the urine can be obtained by means of radiochemical analyzes, this can be considered an efficient indirect method to verify the incorporation of this element. In the work the isotopes of 234 U, 235 U and 238 U were analyzed in urine samples of IOEs and the rate of uranium present in them was verified

  6. Intercomparison of radionuclide measurements in marine sediment sample IAEA-135

    International Nuclear Information System (INIS)

    Ballestra, S.; Gastaud, J.; Lopez, J.J.; Parsi, P.; Vas, D.

    1993-08-01

    The results of an intercomparison exercise on a marine sediment from Irish Sea, IAEA-135, designed for the determination of artificial and natural radionuclides levels, are reported. The data from 151 laboratories representing 51 countries have been evaluated. The following are the recommended values, with confidence intervals, for 40 K, 60 Co, 134 Cs, 137 Cs, 154 Eu, 155 Eu, 226 Ra, 228 Ra, 232 Th, 238 Pu, 239+240 Pu (Reference date: 1 January 1992). Information values for 57 Co, 90 Sr, 106 Ru, 125 Sb, 210 Pb, 210 Po, 228 Th, 230 Th, 234 U, 235 U, 238 U and 241 Am are also reported. All values are expressed in Bq kg -1 dry weight. (author)

  7. Laser assisted jet nozzle isotope separation

    International Nuclear Information System (INIS)

    1981-01-01

    A process for separating fluent particles having different masses comprises the steps of: driving a fluent mixture of such particles around a curved passage toward a septum oriented to divide the mixture thereby accelerating such particles to impart a centrifugal force thereto; inducing type selective heating of a selected particle type in said mixture prior to termination of such acceleration; receiving the fraction of the mixture flowing past an outer surface of said septum in a first output conduit; and receiving the fraction of the mixture flowing past an inner surface of said septum in a second output conduit. The description of the process for isotope separation refers also to the use of infrared laser radiation to produce isotopically selective excitation of the U-235 isotope in UF 6 . (author)

  8. Development of a system of programs HAMMER - ORIGEN - 2 system

    International Nuclear Information System (INIS)

    Guimaraes, L.N.F.

    1984-07-01

    The development of a system of programs (Hamor-2) to calculate precision inventory of LWR reactors is presented. The Hamor-2 code generates with the Hammer-technion code the four (4) energy groupe effective cross section of the actinides: U-235, U-236, U-238, Pu-239, Pu-240, Pu-241, Pu-242. Hamor-2 collapses the four group energy effective cross section to one group energy effective cross section. Hamor-2 calculates with origen-2 the new concentrations for the actinides using the one group energy effective cross section. Two PWR reactors were analised. (HBR, KWO). The discrepancy between the final actinides conentration results calculated by Hamor-2 and the results of measures finded in the references were calculated. That discrepancies have been compared with the discrepancy between : the final actinide concentration results calculated by Origen-2 and some measured results finded in the references. (E.G.) [pt

  9. Effects of Radial Reflector Composition on Core Reactivity and Peak Power

    International Nuclear Information System (INIS)

    Park, Sang Yoon; Lee, Kyung Hoon; Song, Jae Seung

    2007-10-01

    The effects of radial SA-240 alloy shroud on core reactivity and peak power are evaluated. The existence of radial SA-240 alloy shroud makes reflector water volume decrease, so the thermal absorption cross section of radial reflector is lower than without SA-240 alloy shroud case. Finally, the cycle length is increased from 788 EFPD to 845 EFPD and the peak power is decreased from 1.66 to 1.49. In the case of without SA-240 alloy shroud, a new core loading pattern search has been performed. For the guarantee of the same equivalent cycle length of with SA-240 alloy shroud case, the enrichment of U-235 should be increased from 4.22 w/o to 4.68 w/o. The nuclear key safety parameters of new core loading pattern have been calculated and recorded for the future

  10. Effects of Radial Reflector Composition on Core Reactivity and Peak Power

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Yoon; Lee, Kyung Hoon; Song, Jae Seung

    2007-10-15

    The effects of radial SA-240 alloy shroud on core reactivity and peak power are evaluated. The existence of radial SA-240 alloy shroud makes reflector water volume decrease, so the thermal absorption cross section of radial reflector is lower than without SA-240 alloy shroud case. Finally, the cycle length is increased from 788 EFPD to 845 EFPD and the peak power is decreased from 1.66 to 1.49. In the case of without SA-240 alloy shroud, a new core loading pattern search has been performed. For the guarantee of the same equivalent cycle length of with SA-240 alloy shroud case, the enrichment of U-235 should be increased from 4.22 w/o to 4.68 w/o. The nuclear key safety parameters of new core loading pattern have been calculated and recorded for the future.

  11. A study of the material accountancy procedure at the uranium enrichment facility

    International Nuclear Information System (INIS)

    Shirahashi, J.; Akiba, M.; Omae, M.

    1984-01-01

    This paper describes an evaluation of material accountancy based on total uranium (U element MUF) to detect diversions of significant quantity in the uranium enrichment facility operating at a stated maximum enrichment level of 5%. Verification that material production is within the declared enrichment can be achieved by the inspection activities associated with limited - frequency unannounced access (LFUA) to cascade areas as treated extensively in HSP. According to the experience of the material accountancy at our facility, the reduction of the material accountancy capability by changing from U-235 isotope MUF to U element MUF is only about half. However, still the U element MUF approach can meet the current IAEA detection goals for the up to about 1000 tswu/a plant

  12. Molybdenum-UO2 cerment irradiation at 1145 K

    Science.gov (United States)

    Mcdonald, G.

    1971-01-01

    Two molybdenum-UO2 cermet fuel pins were fission heated in a helium-cooled loop at a temperature of 1145 K and to a total burnup of 5.3 % of the U-235. After irradiation the fuel pins were measured to check dimensional stability, punctured at the plenums to determine fission gas release, and examined metallographically to determine the effect of irradiation. Burnup was determined in several sections of the fuel pin. The results of the postirradiation examination indicated: (1) There was no visible change in the fuel pins on irradiation under the above conditions. (2) The maximum swelling of the fuel pins was less than 1%. (3) There was no migration of UO2 and no visible interaction between the molybdenum and the UO2. (4) Approximately 12% of the fission gas formed was released from the cermet cone into the gas plenum.

  13. Buckling and reaction rate measurements in graphite moderated lattices fuelled with plutonium-uranium oxide clusters at temperatures up to 400 deg. C

    International Nuclear Information System (INIS)

    Carter, D.H.; Gibson, M.; King, D.C.; Marshall, J.; Puckett, B.J.; Richards, A.E.; Wass, T.; Wilson, D.J.

    1965-07-01

    The Report describes a series of experiments carried out in SCORPIO I and II on sub-critical graphite moderated lattices fuelled with 21-rod clusters of PuO 2 /UO 2 fuel. Three fuel batches with nominal plutonium: uranium ratios of 0.25%, 0.8% and 1.2% were investigated at temperatures between 20 deg. C and 400 deg. C. Because of the limited amounts of the three fuels, exponential measurements were made in 2-zone stacks, the outer regions of which were loaded with suitably matched 'reference fuel'. Fine structure distributions in the lattice cell were obtained with manganese and indium foils. Pu239/U235 fission ratios were determined both by fission chambers and by fission-product counting techniques. (author)

  14. Calculation of Plutonium content in RSG-GAS spent fuel using IAFUEL computer code

    International Nuclear Information System (INIS)

    Mochamad-Imron

    2003-01-01

    It has been calculated the contain of isotopes Pu-239, Pu-240, Pu-241, and isotope Pu-242 in MTR reactor fuel types which have U-235 contain about 250 gram. The calculation was performed in three steps. The first step is to determine the library of calculation output of BOC (Beginning of Cycle). The second step is to determine the core isotope density, the weight of plutonium for one core, and one fuel isotope density. The third step is to calculate weight of plutonium in gram. All calculation is performed by IAFUEL computer code. The calculation was produced content of each Pu isotopes were Pu-239 is 6.7666 gr, Pu-240 is 1.4628 gr, Pu-241 is 0.52951 gr, and Pu-242 is 0.068952 gr

  15. The obtainment of highly concentrated uranium pellets for plate type (MTR) fuel by dispersion of uranium aluminides in aluminium

    International Nuclear Information System (INIS)

    Morando, R.A.; Raffaeli, H.A.; Balzaretti, D.E.

    1980-01-01

    The use of the intermetallic UAl 3 for manufacturing plate type MTR fuel with 20% U 235 enriched uranium and a density of about 20 kg/m 3 is analyzed. The technique used is the dispersion of UAl 3 particles in aluminium powder. The obtainment of the UAl 3 intermetallic was performed by fusion in an induction furnace in an atmosphere of argon at a pressure of 0.7 BAR (400 mm) using an alumina melting pot. To make the aluminide powder and attain the wished granulometry a cutting and a rotating crusher were used. Aluminide powders of different granulometries and different pressures of compactation were analyzed. In each case the densities were measured. The compacts were colaminated with the 'Picture Frame' technique at temperatures of 490 and 0 deg C with excellent results from the manufacturing view point. (M.E.L.) [es

  16. Reactor utilization; Eksploatacija reaktora

    Energy Technology Data Exchange (ETDEWEB)

    Zecevic, V [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Serbia and Montenegro)

    1963-02-15

    In 1962, the RA reactor was operated almost three times more than in 1961, producing total of 25 555 MWh. Diagram containing comparative data about reactor operation for 1960, 1961, and 1962, percent of fuel used and U-235 burnup shows increase in reactor operation. Number of samples irradiated was 659, number of experiments done was 16. mean powered level was 5.93 MW. Fuel was added into the core twice during the reporting year. In fact the core was increased from 56 to 68 fuel channels and later to 84 fuel channels. Fuel was added to the core when the reactivity worth decreased to the minimum operation level due to burnup. In addition to this 5 central fuel channels were exchanged with fresh fuel in february for the purpose of irradiation in the VISA-2 channel.

  17. Burn-Up Determination by High Resolution Gamma Spectrometry: Axial and Diametral Scanning Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Forsyth, R S; Blackadder, W H; Ronqvist, N

    1967-02-15

    In the gamma spectrometric determination of burn-up the use of a single fission product as a monitor of the specimen fission rate is subject to errors caused by activity saturation or, in certain cases, fission product migration. Results are presented of experiments in which all the resolvable gamma peaks in the fission product spectrum have been used to calculate the fission rate; these results form a pattern which reflect errors in the literature values of the gamma branching ratios, fission yields etc., and also represent a series of empirical correction factors. Axial and diametral scanning experiments on a long-irradiated low-enrichment fuel element are also described and demonstrate that it is possible to differentiate between fissions in U-235 and in Pu-239 respectively by means of the ratios of the Ru-106 activity to the activities of the other fission products.

  18. Actinide integral measurements in the CFRMF and integral tests for ENDF/B-V

    International Nuclear Information System (INIS)

    Anderl, R.A.

    1982-01-01

    Integral capture and/or fission rates have been reported earlier for several actinides irradiated in the fast neutron field of the Coupled Fast Reactivity Measurements Facility (CFRMF). These nuclides include 232 Th, 233 U, 235 U, 238 U, 237 Np, 239 Pu, 240 Pu, 242 Pu, 241 Am and 243 Am. This paper forucses on the utilization of these integral data for testing the respective cross sections on ENDF/B-V. Integral cross sections derived from the measured reaction rates are tabulated. Results are presented for cross-section data testing which includes integral testing based on a comparison of calculated and measured integral cross sections and testing based on least-squares-adjustment analyses

  19. Map of calculated radioactivity of fission product, (4)

    International Nuclear Information System (INIS)

    Takeda, Tsuneo

    1978-07-01

    The overall radioactivities of fission products depending on irradiation time and cooling time were calculated for 18 different neutron fluxes, which are presented in contour maps and tables. Irradiation condition etc. are the followings: neutron flux (n sub(th)) 1 x 10 12 - 6.8 x 10 14 n/cm 2 /sec, uranium quantity 1 mole (6 x 10 23 atoms, ca. 271 g UO 2 ), U-235 enrichment 2.7%, irradiation time 60. - 6 x 10 7 sec (1 min - 1.9 y), cooling time 0. and 60. - 6 x 10 7 sec (1 min - 1.9 y). The enrichment value represents those for LWRs. To calculate the overall radioactivities, 595 fission product nuclides were introduced. Overall radioactivities calculations were made for 68,000 combinations of irradiation time, cooling time and neutron flux. The many complex decay chains of fission products were treated with CODAC-No.6 computer code. (author)

  20. Performance Testing and Validation Plan for HMS4 Quantitative Gamma Measurements, K-25/K-27 D&D Project, East Tennessee Technology Park, Oak Ridge, Tennessee

    Energy Technology Data Exchange (ETDEWEB)

    Thiesing, J. W.; Donohoue, Tom; McCague, Jim; Martin, Ray; Royce, Ralph; Thomas, Troy

    2008-04-14

    The Holdup Measurement System 4 (HMS4) is a portable thallium activated sodium iodide (NaI[Tl]) gamma ray energy spectrometer that, when properly calibrated, is able to make quantifiable assessment of U-235 holdup in the presence of other uranium isotopes and prevailing background radiation. The use and calibration of the HMS4 is based upon the methodologies defined by Russo in La-14206, (Russo 2005), where detection efficiency determination protocols are defined (called Generalized Geometry Holdup [GGH]). The GGH methodology together with attenuation correction algorithms and other modeling parameters are combined in the HMS4 software package to provide a comprehensive tool for conducting in situ gamma-ray measurements. The fundamental principles of these capabilities are discussed.

  1. Analysis of the critical and first full power operating cores for PARR using leu oxide fuel

    International Nuclear Information System (INIS)

    Khan, L.A.; Qazi, M.K.; Bokhari, I.H.; Fazal, R.

    1989-10-01

    This paper explains the analysis for determining the first full power operating core for PARR using LEU oxide fuel. The core configuration selected for this first full power operation contains about 6.13 kg of U-235 distributed in 19 standard and five control fuel elements. The neutron flux level is doubled when core is shifted from 5MW to 10 MW. Total nuclear power peaking factor of the core is 2.03. The analysis shows that the core can be operated safely at 5 MW with a flow rate of 520 meter cube per hour and at 10 MW with a flow rate of 900 meter cube per hour. (A.B.). 10 figs

  2. Neutronic analysis of HEU to LEU conversion calculation for AEOI 5 MW pool-type MTR fuel research reactor core

    International Nuclear Information System (INIS)

    Pazirandeh, A.; Lutz, D.; Bartsch, G.

    1987-07-01

    The possibility of converting HEU(93%) fuel to LEU(20%) fuel without or with slight alteration to the fuel element geometry is discussed. The fuel density varies between 1.7 to 4.1 g U-235/cm. In cross section generation a unit cell with an extra zone to account for extra Al and water was considered. In burnup calculations a sequential shuffling pattern was assumed with fixed position control fuel elements. A cross section data set in 45 energy groups were generated using RSYST/CGM system using the cross section library JFET. Then for 2D-diffusion calculations homogenized and condensed 5 energy group cross sections were prepared. (orig./HP)

  3. Burnup credit applications in a high-capacity truck cask

    International Nuclear Information System (INIS)

    Boshoven, J.K.

    1992-09-01

    General Atomics (GA) has designed two legal weight truck (LWT) casks, the GA-4 and GA-9, to carry four pressurized-water-reactor (PWR) and nine boiling-water-reactor (BWR) fuel assemblies, respectively. GA plans to submit applications for certification to the US Nuclear Regulatory Commission (NRC) for the two casks in mid-1993. GA will include burnup credit analysis in the Safety Analysis Report for Packaging (SARP) for the GA-4 Cask. By including burnup credit in the criticality safety analysis for PWR fuels with initial enrichments above 3% U-235, public and occupation risks are reduced and cost savings are realized. The GA approach to burnup credit analysis incorporates the information produced in the US Department of Energy Burnup Credit Program. This paper describes the application of burnup credit to the criticality control design of the GA-4 Cask

  4. Analysis of the Daya Bay Reactor Antineutrino Flux Changes with Fuel Burnup

    Science.gov (United States)

    Hayes, A. C.; Jungman, Gerard; McCutchan, E. A.; Sonzogni, A. A.; Garvey, G. T.; Wang, X. B.

    2018-01-01

    We investigate the recent Daya Bay results on the changes in the antineutrino flux and spectrum with the burnup of the reactor fuel. We find that the discrepancy between current model predictions and the Daya Bay results can be traced to the original measured U 235 /Pu 239 ratio of the fission β spectra that were used as a base for the expected antineutrino fluxes. An analysis of the antineutrino spectra that is based on a summation over all fission fragment β decays, using nuclear database input, explains all of the features seen in the Daya Bay evolution data. However, this summation method still allows for an anomaly. We conclude that there is currently not enough information to use the antineutrino flux changes to rule out the possible existence of sterile neutrinos.

  5. Reentry safety for the Topaz II Space Reactor: Issues and analyses

    International Nuclear Information System (INIS)

    Connell, L.W.; Trost, L.C.

    1994-03-01

    This report documents the reentry safety analyses conducted for the TOPAZ II Nuclear Electric Propulsion Space Test Program (NEPSTP). Scoping calculations were performed on the reentry aerothermal breakup and ground footprint of reactor core debris. The calculations were used to assess the risks associated with radiologically cold reentry accidents and to determine if constraints should be placed on the core configuration for such accidents. Three risk factors were considered: inadvertent criticality upon reentry impact, atmospheric dispersal of U-235 fuel, and the Special Nuclear Material Safeguards risks. Results indicate that the risks associated with cold reentry are very low regardless of the core configuration. Core configuration constraints were therefore not established for radiologically cold reentry accidents

  6. Evaluation model for PWR irradiated fuel

    International Nuclear Information System (INIS)

    Gomes, I.C.

    1983-01-01

    The individual economic value of the plutonium isotopes for the recycle of the PWR reactor is investigated, assuming the existence of an market for this element. Two distinct market situations for the stages of the fuel cycle are analysed: one for the 1972 costs and the other for costs of 1982. Comparisons are made for each of the two market situations concerning enrichment of the U-235 in the uranium fuel that gives the minimum cost in the fuel cycle. The method adopted to establish the individual value of the plutonium isotopes consists on the economical analyses of the plutonium fuel cycle for four different isotopes mixtures refering to the uranium fuel cycle. (Author) [pt

  7. Development of long operating cycle simplified BWR

    International Nuclear Information System (INIS)

    Heki, H.; Nakamaru, M.; Maruya, T.; Hiraiwa, K.; Arai, K.; Narabayash, T.; Aritomi, M.

    2002-01-01

    This paper describes an innovative plant concept for long operating cycle simplified BWR (LSBWR) In this plant concept, 1) Long operating cycle ( 3 to 15 years), 2) Simplified systems and building, 3) Factory fabrication in module are discussed. Designing long operating core is based on medium enriched U-235 with burnable poison. Simplified systems and building are realized by using natural circulation with bottom located core, internal CRD and PCV with passive system and an integrated reactor and turbine building. This LSBWR concept will have make high degree of safety by IVR (In Vessel Retention) capability, large water inventory above the core region and no PCV vent to the environment due to PCCS (Passive Containment Cooling System) and internal vent tank. Integrated building concept could realize highly modular arrangement in hull structure (ship frame structure), ease of seismic isolation capability and high applicability of standardization and factory fabrication. (authors)

  8. Nondestructive Analysis of MET-5 Paint Can at TA35 Building 2 A-Wing Vault

    Energy Technology Data Exchange (ETDEWEB)

    Desimone, David J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Vo, Duc Ta [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-11-03

    In Building 2 A-wing vault MET-5 has some drums and other packages they wanted NEN-1 help identifying nondestructively. Measurements using a mechanically cooled portable high-purity germanium HPGe Ortec detective were taken of a paint can container labeled DU-2A to determine if any radioactive material was inside. The HPGe detector measures the gamma rays emitted by radioactive material and displays it as a spectrum. The spectrum is used to identify this radioactive material by using appropriate analysis software and identifying the gamma ray peaks. A paint can container, DU-2A, was analyzed with PeakEasy 4.84 and FRAM 5.2. The FRAM report is shown. The enrichment is 0.091% U235 and 99.907% U238. This material is depleted uranium. The measurement was performed in the near field, and to extract a mass a far field measurement will need to be taken.

  9. Important radionuclides and their sensitivity for groundwater pathway of a hypothetical near-surface disposal facility

    Energy Technology Data Exchange (ETDEWEB)

    Park, J. W.; Chang, K.; Kim, C. L. [Nuclear Enviroment Technology Institute, Taejon (Korea, Republic of)

    2001-04-01

    A radiological safety assessment was performed for a hypothetical near-surface radioactive waste repository as a simple screening calculation to identify important nuclides and to provide insights on the data needs for a successful demonstration of compliance. Individual effective doses were calculated for a conservative groundwater pathway scenario considering well drilling near the site boundary. Sensitivity of resulting ingestion dose to input parameter values was also analyzed using Monte Carlo sampling. Considering peak dose rate and assessment timescale, C-14 and I-129 were identified as important nuclides and U-235 and U-238 as potentially important nuclides. For C-14, the does was most sensitive to Darcy velocity in aquifer. The distribution coefficient showed high degree of sensitivity for I-129 release.

  10. Important radionuclides and their sensitivity for groundwater pathway of a hypothetical near-surface disposal facility

    International Nuclear Information System (INIS)

    Park, J. W.; Chang, K.; Kim, C. L.

    2001-01-01

    A radiological safety assessment was performed for a hypothetical near-surface radioactive waste repository as a simple screening calculation to identify important nuclides and to provide insights on the data needs for a successful demonstration of compliance. Individual effective doses were calculated for a conservative groundwater pathway scenario considering well drilling near the site boundary. Sensitivity of resulting ingestion dose to input parameter values was also analyzed using Monte Carlo sampling. Considering peak dose rate and assessment timescale, C-14 and I-129 were identified as important nuclides and U-235 and U-238 as potentially important nuclides. For C-14, the does was most sensitive to Darcy velocity in aquifer. The distribution coefficient showed high degree of sensitivity for I-129 release

  11. Calculations of nuclear data for the reactions of neutrons and protons with heavy nuclei at energy from 1 MeV up to 2 GeV

    International Nuclear Information System (INIS)

    Konshin, V.A.

    1995-01-01

    Several nuclear model codes were applied to calculations of nuclear data in the energy region from 1 MeV to 2 GeV. At energies from 1 to 20 MeV the statistical model code STAPRE was used for calculations of the neutron cross-sections for fission, (n,2n) and (n,3n) reaction cross-sections for 71 actinide isotopes. In the energy region from 10 to 100 MeV the nuclear theory code GNASH was used to calculate the neutron fission and (n,xn) cross-sections for 238 U, 235 U, 239 Pu, 232 Th, 237 Np, 238 Pu, 241 Am, 243 Am, 245 Cm and 246 Cm. At energies from 100 MeV to 2 GeV the intranuclear cascade-exciton model including the fission process was applied to calculations of the interactions of protons and neutrons with actinides and the calculated results are compared with experimental data. (author)

  12. General description and production lines of the Egyptian fuel manufacturing pilot plant

    International Nuclear Information System (INIS)

    Zidan, W.I.; Elseaidy, I.M.

    1999-01-01

    The Egyptian Fuel Manufacturing Pilot Plant, FMPP, is a new facility, producing an MTR-type fuel elements required for the Egyptian Second Research Reactor, ETRR-2, as well as other plates or elements for an external clients with the same type and enrichment percent or lower, (LEU). General description is presented. The production lines in FMPP, which begin from uranium hexaflouride (UF 6 , 19.7±0.2 % U 235 by wt), aluminum powder, and nuclear grade 6061 aluminium alloy in sheets, bars, and rods with the different heat treatments and dimensions as a raw materials, are processed through a series of the manufacturing, inspection, and quality control plan to produce the final specified MTR-type fuel elements. All these processes and the product control in each step are presented. The specifications of the final product are presented. (author)

  13. Benchmark calculations for critical experiments at FKBN-M facility with uranium-plutonium-polyethylene systems using JENDL-3.2 and MVP Monte-Carlo code

    International Nuclear Information System (INIS)

    Obara, Toru; Morozov, A.G.; Kevrolev, V.V.; Kuznetsov, V.V.; Treschalin, S.A.; Lukin, A.V.; Terekhin, V.A.; Sokolov, Yu.A.; Kravchenko, V.G.

    2000-01-01

    Benchmark calculations were performed for critical experiments at FKBN-M facility in RFNC-VNIITF, Russia using JENDL-3.2 nuclear data library and continuous energy Monte-Carlo code MVP. The fissile materials were high-enriched uranium and plutonium. Polyethylene was used as moderator. The neutron spectrum was changed by changing the geometry. Calculation results by MVP showed some errors. Discussion was made by reaction rates and η values obtained by MVP. It showed the possibility that cross sections of U-235 had different trend of error in fast and thermal energy region respectively. It also showed the possibility of some error of cross section of Pu-239 in high energy region. (author)

  14. Burn-Up Determination by High Resolution Gamma Spectrometry: Fission Product Migration Studies

    Energy Technology Data Exchange (ETDEWEB)

    Forsyth, R S; Blackadder, W H; Ronqvist, N

    1967-04-15

    The migration of solid fission products, in particular caesium and ruthenium, in high temperature oxide fuel can create a severe problem during the application of non-destructive burn-up methods employing gamma spectrometry, since caesium-137 is otherwise the most convenient long-lived burn-up monitor and ruthenium-106 can be used to distinguish between fissions in U-235 and Pu-239. As part of an experimental programme to develop burn-up methods, gamma scanning experiments have been performed on slices of irradiated UO{sub 2} pellets using a lithium-drifted germanium detector. The usefulness of the technique for migration studies has been demonstrated by comparing the fission product distribution curves across the specimen diameters with the microstructure of the specimens after polishing and etching.

  15. Separation and purification of Molybdenum-99 using the AG1x8 and Chelex-100 resins; Separacion y purificacion de Molibdeno-99 utilizando las resinas AG1x8 y Chelex-100

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez M, V; Lopez C, R; Millan S, S; Fucugauchi, L A

    1991-05-15

    In the separation of the fission products of the molybdenum-99 soluble in watery solution of sodium hydroxide its have been used advantageously ion exchange processes with resins, which present a great stability toward the oxidizers and reducers agents as well as toward the ionizing radiation. The main stages of the recovery process of Mo-99 starting from the fission products of U-235 enriched. The first step of the purification is the elimination of the alumina taken place in the alkaline breakup of the badges of alloy uranium-aluminum. For it is used it a resin of anion exchange denominated AG1X8. The second stage of the consistent process in the elimination of the remaining fission products that accompany the one Mo-99, by means of the use of the exchange resin Chelex-100. Lastly is carried out the a sublimation process to obtain the pure Mo-99. (Author)

  16. Resonance computations for cells with fuel annuli

    International Nuclear Information System (INIS)

    Hwang, R.N.; Gelbard, E.M.

    1990-01-01

    Two methods have been developed for the computation of resonance integrals in cells containing annular fuel regions. Both are based on rational approximations. One is a generalization of a one-term rational approximation method developed by Segev for a cell with a single fuel annulus. The second modifies the earlier Chen-Gelbard two-term method originally used for double-heterogeneity calculations. Both methods were tested, in cells with two fuel annuli, for various U 235 and U 238 resonances. Both gives resonance integrals accurate enough for practical purposes. The two-term fits are substantially more accurate in some NR cases, but are somewhat more difficult to correct for finite resonance widths. 8 refs., 4 tabs

  17. Conversion of highly enriched uranium in thorium-232 based oxide fuel for light water reactors: MOX-T fuel

    Energy Technology Data Exchange (ETDEWEB)

    Vapirev, E; Jordanov, T; Khristoskov, I [Sofia Univ. (Bulgaria). Fizicheski Fakultet

    1996-12-31

    The possibility of using highly enriched uranium available from military inventories for production of mixed oxide fuel (MOX) has been proposed. The fuel is based on U-235 dioxide as fissile isotope and Th-232 dioxide as a non-fissile isotope. It is shown that although the fuel conversion coefficient to U-233 is expected to be less than 1, the proposed fuel has several important advantages resulting in cost reduction of the nuclear fuel cycle. The expected properties of MOX fuel (cross-sections, generated chains, delayed neutrons) are estimated. Due to fuel generation the initial enrichment is expected to be 1% less for production of the same energy. In contrast to traditional fuel no long living actinides are generated which reduces the disposal and reprocessing cost. 7 refs.

  18. Certified reference material for radionuclides in fish flesh sample IAEA-414 (mixed fish from the Irish Sea and North Sea)

    International Nuclear Information System (INIS)

    Pham, M.K.; Sanchez-Cabeza, J.A.; Povinec, P.P.; Arnold, D.; Benmansour, M.; Bojanowski, R.; Carvalho, F.P.; Kim, C.K.; Esposito, M.; Gastaud, J.; Gasco, C.L.; Ham, G.J.; Hegde, A.G.; Holm, E.; Jaskierowicz, D.; Kanisch, G.; Llaurado, M.; La Rosa, J.; Lee, S.-H.; Liong Wee Kwong, L.; Le Petit, G.; Maruo, Y.; Nielsen, S.P.; Oh, J.-S.; Oregioni, B.; Palomares, J.; Pettersson, H.B.L.; Rulik, P.; Ryan, T.P.; Sato, K.; Schikowski, J.; Skwarzec, B.; Smedley, P.A.; Tarjan, S.; Vajda, N.; Wyse, E.

    2006-01-01

    A certified reference material (CRM) for radionuclides in fish sample IAEA-414 (mixed fish from the Irish Sea and North Seas) is described and the results of the certification process are presented. Nine radionuclides ( 4 K, 137 Cs, 232 Th, 234 U, 235 U, 238 U, 238 Pu, 239+24 Pu and 241 Am) were certified for this material. Information on massic activities with 95% confidence intervals is given for six other radionuclides ( 9 Sr, 21 Pb( 21 Po), 226 Ra, 239 Pu, 24 Pu 241 Pu). Less frequently reported radionuclides ( 99 Tc, 129 I, 228 Th, 23 Th and 237 Np) and information on some activity and mass ratios are also included. The CRM can be used for quality assurance/quality control of the analysis of radionuclides in fish sample, for the development and validation of analytical methods and for training purposes. The material is available from IAEA, Vienna, in 100 g units

  19. Main results on pilot operation during 5 years of the 3rd generation fuel in VVER-440 reactors of Kola NPP

    International Nuclear Information System (INIS)

    Saprykin, V.; Sumarokov, M.; Gagarinskiy, A.; Sumarokova, A.; Adeev, V.

    2015-01-01

    In the report the results of comparison of main neutron-physical data of exploitation of nuclear fuel are presented for the average enrichment (on U - 235) of 4.87 for the 2nd and 3rd (12 piece) generations with the results of calculations by the complex of the programs KASKAD for 5 fuel loadings of Kola NPP Unit 4 with the reactor VVER- 440. The basic feature of fuel of the 3rd generation as compared with the 2nd is a presence of ribs of inflexibility at corners instead of cover of the fuel assembly and also the increased amount of uranium. The arrangement of fuel rods with different enrichment in fuel assemblies of the 2nd and 3rd generations is chosen identical for the convenient comparison of neutronic and thermohydraulic characteristics of the fuel of different generations. The fuel of 3rd generation was situated in the core symmetrically to the fuel of 2nd one

  20. Evaluating the accuracy of uranium isotope amount ratio measurements performed by a quadrupole and a multi-collector magnetic sector inductively coupled plasma mass spectrometers for nuclear safeguards

    International Nuclear Information System (INIS)

    Pereira de Oliveira, O. Jr.; Sarkis, J.E.S.; Ponzevera, E.; Alonso, A.; De Bolle, W.; Quetel, C.

    2008-01-01

    The n(U 235 )/n(U 238 ) isotope amount ratio in a set of samples was measured using two modern analytical techniques: quadrupole inductively coupled plasma mass spectrometry (ICP-QMS) and multi-collector magnetic sector inductively coupled plasma mass spectrometry (MC-ICPMS). The measured ratios were compared to the certified ratios provided by the high accuracy gas source mass spectrometry (GSMS). The components of the uncertainty were identified and their contribution to the combined standard uncertainty was estimated using the recommendations of the ISO-GUM guide. The values of the measurement uncertainty and bias were determined and then compared to the International Target Values for Measurement Uncertainties in Safeguarding Nuclear Materials. It appears that only the measurements performed by MC-ICPMS can meet the stringent requirements of international nuclear safeguards. (authors)

  1. Method of measurement of cross sections of heavy nuclei fission induced by intermediate energy protons

    International Nuclear Information System (INIS)

    Kotov, Alexander; Chtchetkovski, Alexander; Fedorov, Oleg; Gavrikov, Yuri; Chestnov, Yuri; Poliakov, Vladimir; Vaishnene, Larissa; Vovchenko, Vil; Fukahori, Tokio

    2003-01-01

    The purpose of this work is experimental studies of the energy dependence of the fission cross sections of heavy nuclei, nat Pb, 209 Bi, 232 Th, 233 U, 235 U, 238 U, 237 Np and 239 Pu, by protons at the energies from 200 to 1000 MeV. At present experiment the method based on use of the gas parallel plate avalanche counters (PPACs) for registration of complementary fission fragments in coincidence and the telescope of scintillation counters for direct counting of the incident protons on the target has been used. First preliminary results of the energy dependences of proton induced fission cross sections for nat Pb, 209 Bi, 235 U and 238 U are reported. (author)

  2. Isotopic ratios of actinides used in British nuclear trials at Maralinga and Emu

    International Nuclear Information System (INIS)

    Johnston, P.N.; Burns, P.A.; Cooper, M.B.; Williams, G.A.

    1988-10-01

    Studies are underway to investigate the rehabilitation of the former nuclear weapons test sites at Maralinga and Emu in South Australia. Many of these studies are based on measurements of Am-241 which is a contaminant in the plutonium dispersed at these sites. Measurements of the ratio of activities of Pu-239 and Am-241 are presented for sites where suitable samples could be collected. Where possible, measurements were also made of Pu-240 and U-235 activities. Recommended values, current for mid 1988, for the Pu-239/Am-241 activity ratio for the major trial sites range between 20 and 40. At Taranaki values of 6 to 22 were reported, while for the minor sites, current values of the Pu-240/Am-241 activity ratio vary between 1 and 2. 6 refs., 6 tabs., 3 figs

  3. Burn-Up Determination by High Resolution Gamma Spectrometry: Spectra from Slightly-Irradiated Uranium and Plutonium between 400-830 keV

    International Nuclear Information System (INIS)

    Forsyth, R.S.; Ronqvist, N.

    1966-08-01

    Previously published studies of the short-cooled fission product spectra of irradiated uranium have been severely restricted by the poor energy resolution of the sodium iodide detectors used. In this report are presented fission product spectra of irradiated uranium and plutonium obtained by means of a lithium-drifted germanium detector. The resolved gamma peaks have been assigned to various fission products by correlation of measured energy and half-life values with published data. By simultaneous study of the spectra of two irradiated mixtures of plutonium and uranium, the possibility of using the activities of Ru-103 and Ru-106 as a measure of the relative fission rate in U-235 and Pu-239 has been briefly examined

  4. Behavior of uranium along Jucar River (Eastern Spain). Determination of 234U/238U and 235U/238U ratios

    International Nuclear Information System (INIS)

    Rodriguez-Alvarez, M.J.; Sanchez, F.

    1995-01-01

    The uranium concentration and the 234 U/ 238 U, 235 U/ 238 U activity ratios were studied in water samples from Jucar River, using low-level α-spectrometry. The effects of pH, temperature and salinity were considered and more detailed sampling was done in the neighbourhood of Cofrentes Nuclear Plant (Valencia, Spain). Changes were observed in the uranium concentration with the salinity and the 234 U/ 238 U activity ratio was found to vary with pH. Leaching and dilution, which depend on pH and salinity, are the probable mechanisms for these changes in the concentration of uranium and the activity ratios. (author) 25 refs.; 4 figs.; 1 tab

  5. ZZ TEMPEST/MUFT, Thermal Neutron and Fast Neutron Multigroup Cross-Section Library for Program LEOPARD

    International Nuclear Information System (INIS)

    Kim, Jung-Do; Lee, Jong Tai

    1986-01-01

    Description of problem or function: Format: TEMPEST and MUFT; Number of groups: 246 thermal groups in TEMPEST Format and 54 fast groups in MUFT Format. From this library, the program SPOTS4 generates a 172-54 group library as input to the code LEOPARD. Nuclides: H, O, Zr, C, Fe, Ni, Al, Cr, Mn, U, Pu, Th, Pa, Xe, Sm, B and D. Origin: ENDF/B-4; Weighting spectrum: 1/E + U 235 fission spectrum. Data library of thermal and fast neutron group Cross sections to generate input to the program LEOPARD. The data is based on ENDF/B-4 and consists of two parts: (1) 246 thermal groups in TEMPEST Format. (2) 54 fast groups in MUFT Format. From this library, the program SPOTS4 generates a 172-54 group library as input to the code LEOPARD (NESC0279)

  6. MIT LMFBR blanket physics project progress report No. 7, July 1, 1975--September 30, 1976

    International Nuclear Information System (INIS)

    Driscoll, M.J.

    1976-01-01

    Work during the period was devoted primarily to a range of analytical/numerical investigations, including evaluation of means to improve external blanket designs, beneficial attributes of the use of internal blankets, improved methods for the calculation of heterogeneous self-shielding and parametric studies of calculated spectral indices. Experimental work included measurements of the ratio of U-238 captures to U-235 fissions in a standard blanket mockup, and completion of development work on the radiophotoluminescent readout of LiF thermoluminescent detectors. The most significant findings were that there is very little prospect for substantial improvement in the breeding performance of external blankets, but internal blankets continue to show promise, particularly if they are used in such a way as to increase the volume fraction of fuel inside the core envelope. An improved equivalence theorem was developed which may allow use of fast reactor methods to calculate heterogeneously self-shielded cross sections in both fast and thermal reactors

  7. Influence of the particle discriminator for producing the microporous nuclear track etched membrane

    International Nuclear Information System (INIS)

    Thongphud, Apaporn; Ratanatongchai, Wichian; Supaphol, Pitt; Visal-athaphand, Pinpan

    2005-10-01

    The particle discriminator was used to focus the fission fragments from nuclear fission reaction between thermal neutron from the Thai Research Reactor and U-235 in uranium screen to strike almost normally to the polycarbonate (PC) film. The latent tracks in the thin 15 mm PC film were revealed after etching in 6N NaOH solution at 70 o C for 60 min. It was found that the tracks were porous. The porosity was more discrete and the pore shape was more circular as well. The track diameter was measured 3.73 +- 0.32 mm. It was also found that using particle discriminators with increasing thickness during exposure gives fewer pores in the PC film, after chemical etching under the same condition as above

  8. 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

  9. Dual isotope notch observer for isotope identification, assay and imaging with mono-energetic gamma-ray sources

    Science.gov (United States)

    Barty, Christopher P.J.

    2013-02-05

    A dual isotope notch observer for isotope identification, assay and imaging with mono-energetic gamma-ray sources includes a detector arrangement consists of three detectors downstream from the object under observation. The latter detector, which operates as a beam monitor, is an integrating detector that monitors the total beam power arriving at its surface. The first detector and the middle detector each include an integrating detector surrounding a foil. The foils of these two detectors are made of the same atomic material, but each foil is a different isotope, e.g., the first foil may comprise U235 and second foil may comprise U238. The integrating detectors surrounding these pieces of foil measure the total power scattered from the foil and can be similar in composition to the final beam monitor. Non-resonant photons will, after calibration, scatter equally from both foils.

  10. Statistical modelling of neural networks in γ-spectrometry applications

    International Nuclear Information System (INIS)

    Vigneron, V.; Martinez, J.M.; Morel, J.; Lepy, M.C.

    1995-01-01

    Layered Neural Networks, which are a class of models based on neural computation, are applied to the measurement of uranium enrichment, i.e. the isotope ratio 235 U/( 235 U + 236 U + 238 U). The usual method consider a limited number of Γ-ray and X-ray peaks, and require previously calibrated instrumentation for each sample. But, in practice, the source-detector ensemble geometry conditions are critically different, thus a means of improving the above convention methods is to reduce the region of interest: this is possible by focusing on the K α X region where the three elementary components are present. Real data are used to study the performance of neural networks. Training is done with a Maximum Likelihood method to measure uranium 235 U and 238 U quantities in infinitely thick samples. (authors). 18 refs., 6 figs., 3 tabs

  11. Chemistry of nuclear waste disposal

    International Nuclear Information System (INIS)

    Zimmer, E.

    1981-01-01

    In extractive purification of the low-enriched uranium fuel element (UO 2 -particle fuel element with SiC coating) no problems arise in the PUREX-process which have not already been solved when reprocessing LWR-type reactor and breeder fuel elements. Concerning the HTR-type reactor fuel elements containing thorium, there are two process cycles behind the head end; the pure U-235 is reprocessed in the same manner as the low-enriched uranium fuel, and the thorium, which is the bigger fraction, is reprocessed together with U-233 in the same manner as the mixed oxides. Only the CO 2 -off gas system, which contains krypton and carbon 14, leads to difficulties in nuclear waste disposal. (DG) [de

  12. High-capacity neutron activation analysis facility

    International Nuclear Information System (INIS)

    Hochel, R.C.; Bowman, W.W.; Zeh, C.W.

    1980-01-01

    A high-capacity neutron activation analysis facility, the Reactor Activation Facility, was designed and built and has been in operation for about a year at one of the Savannah River Plant's production reactors. The facility determines uranium and about 19 other elements in hydrogeochemical samples collected in the National Uranium Resource Evaluation program, which is sponsored and funded by the United States Department of Energy, Grand Junction Office. The facility has a demonstrated average analysis rate of over 10,000 samples per month, and a peak rate of over 16,000 samples per month. Uranium is determined by cyclic activation and delayed neutron counting of the U-235 fission products; other elements are determined from gamma-ray spectra recorded in subsequent irradiation, decay, and counting steps. The method relies on the absolute activation technique and is highly automated for round-the-clock unattended operation

  13. Assessment of the accident response of a light-water-moderated breeder-reactor system: AWBA development program

    International Nuclear Information System (INIS)

    High, H.M.

    1983-05-01

    The predicted accident response for a light water moderated, thorium/U-233 fueled, seed-blanket reactor concept was assessed. The first part of the assessment compared breeder accident response with that of a current commercial pressurized water reactor design for several different types of transients. Based on these comparisons and a review of the various parameter differences between the breeder and a U-235 fueled plant, the second part of the assessment studied the breeder accident behavior in more detail, particularly in areas of potential concern. Based on the two parts of the assessment, it was concluded that the breeder accident response would be very similar to that of present commercial pressurized water reactor plants. The large Doppler and moderator reactivity coefficients of the breeder would significantly reduce the severity of many of the accidents that must be considered. It is expected that the accident response of the breeder can be shown to meet regulatory criteria

  14. Ocean abandonment of radioactive waste. 2

    International Nuclear Information System (INIS)

    Kouyama, Hiroaki

    1994-01-01

    Now, the nuclear powered submarines armed with ballistic missiles have become the main strength of navy. In Russia, eight nuclear powered icebreakers are operated. Mainly PWRs are used for these nuclear ships. The fuel exchange for nuclear powered submarines is carried out after the use for nearly ten years, therefore, the degree of enrichment of U-235 in fuel seems considerably high. So far, the sinking accidents of five nuclear powered submarines were reported. Former USSR began the ocean abandonment of radioactive waste in 1959, and continued it up to recent date. The northern sea area where the abandonment was carried out and the abandoned amount of radioactivity are shown. Also those in Far East sea area are shown. The management system for radioactive waste in Russia, the course after the abandonment of liquid waste in Japan Sea by Russian navy, the response of Japan regarding the ocean abandonment of radioactive waste and so on are described. (K.I.)

  15. Description and user manual of the WIMSLIC, FIXER and COMPA programs; Descripcion y manual del usuario de los programas WIMSLIC, FIXER y COMPA

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar H, F [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    1992-08-15

    In this work the WIMS library and those WIMSLIC, FIXER and COMPA codes that are used to give him maintenance or to create a new one, the way to use them and its scopes are described. The objective of WIMSLIC, is the one of generating data nuclear with the WIMS code format, uses those results obtained with the NJOY system and the one POWR module, the FIXER function is to generate a new WIMS library or to already modify an existent, using the results of the one WIMSLIC code or those obtained with NJOY and WIMSR, while the COMPA function is to compare data groups with WIMS format. In the Appendix 1 the files of data are had that would use to generate a new library of WIMS that contain nuclear data for U-235, U-238, H-1, 0-16 and Al-27. In the Appendix 2 one has the listing of the programs before mentioned. (Author)

  16. Quantitative and quality test of cross section library ENDF/B-b2

    International Nuclear Information System (INIS)

    Zajac, R.; Necas, V.

    2006-01-01

    This article includes a test or in other words data verification of neutron ENDF/B-VIIb2 sub library. The first part consists from the process of preparation ACE files by NJOY 99.90. The starting point of data verification describes needed patches in NJOY 99.90, which are necessary to do for correctly production of ACE files. After the obtaining ACE files follow the test of all ACE files through GODIVA - input file for MCNP. GODIVA is high enrichment sphere of U-235, where every material is added as impurity. The aim of GODIVA test is to obtain a certainty if produced ACE files are able to run through MCNP. The second part of this article begins with choose of benchmarks from 'International Handbook of Evaluated Criticality Safety Benchmark Experiments, 2005'. From this source of criticality experiments were separated some benchmarks for quality verification of ACE files by MCNP (Authors)

  17. Separation and purification of Molybdenum-99 using the AG1x8 and Chelex-100 resins

    International Nuclear Information System (INIS)

    Sanchez M, V.; Lopez C, R.; Millan S, S.; Fucugauchi, L.A.

    1991-05-01

    In the separation of the fission products of the molybdenum-99 soluble in watery solution of sodium hydroxide its have been used advantageously ion exchange processes with resins, which present a great stability toward the oxidizers and reducers agents as well as toward the ionizing radiation. The main stages of the recovery process of Mo-99 starting from the fission products of U-235 enriched. The first step of the purification is the elimination of the alumina taken place in the alkaline breakup of the badges of alloy uranium-aluminum. For it is used it a resin of anion exchange denominated AG1X8. The second stage of the consistent process in the elimination of the remaining fission products that accompany the one Mo-99, by means of the use of the exchange resin Chelex-100. Lastly is carried out the a sublimation process to obtain the pure Mo-99. (Author)

  18. Inspection of domestic nuclear fuel rods using neutron radiography at the Tehran research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dastjerdi, Mohammad Hosein Choopan; Khalafi, Hossein; Kasesaz, Yaser [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of); Movafeghi, Amir

    2016-11-01

    Three unused domestic fuel rods were investigated qualitatively and quantitatively by means of thermal neutron radiography. The neutron radiography tests were performed by the image plate method at Tehran research reactor in order to check the fuel properties. The pellets of these three fuel rods contained three different U-235 enrichments and different sizes that were filled into a zircalloy tube. In the qualitative investigations, the difference in size and enrichment between the pellets and the gaps between them were obviously recognized in the image of the fuel rods. In the quantitative investigations, data of the pellets compositions, their sizes (lengths and diameters) and the gaps between them were extracted from obtained images. It was found that the measured data and the manufacturer's specifications are in good agreement.

  19. Nuclear design report for Ulchin nuclear power plant unit 2 cycle 5

    International Nuclear Information System (INIS)

    Park, Jin Ha; Park, Yong Soo; Cho, Byeong Ho; Zee, Sung Kyun; Lee, Sang Keun; Ahn, Dawk Hwan

    1993-09-01

    This report presents nuclear design calculations for cycle 5 of Ulchin unit it 2. Information is given on fuel loading, power density distributions, reactivity coefficients, control rod worths and operational limits. In addition, the report contains all necessary data for the startup tests including predicted values for the comparison with the measured data. The reload consists of 48 KOFA's enriched by nominally 3.50 w/o U 235 . Among the KOFA's, 20 fuel assemblies contain gadolinia rods. The fuel assemblies in the core are arranged in a low leakage loading pattern. The cycle length of cycle 5 amounts to 293 EFPD corresponding to a cycle burnup of 11780 MWD/MTU. (Author) 8 refs., 55 figs., 16 tabs

  20. Optimization of PWR fuel assembly radial enrichment and burnable poison location based on adaptive simulated annealing

    International Nuclear Information System (INIS)

    Rogers, Timothy; Ragusa, Jean; Schultz, Stephen; St Clair, Robert

    2009-01-01

    The focus of this paper is to present a concurrent optimization scheme for the radial pin enrichment and burnable poison location in PWR fuel assemblies. The methodology is based on the Adaptive Simulated Annealing (ASA) technique, coupled with a neutron lattice physics code to update the cost function values. In this work, the variations in the pin U-235 enrichment are variables to be optimized radially, i.e., pin by pin. We consider the optimization of two categories of fuel assemblies, with and without Gadolinium burnable poison pins. When burnable poisons are present, both the radial distribution of enrichment and the poison locations are variables in the optimization process. Results for 15 x 15 PWR fuel assembly designs are provided.